ML18064A888

Rev 0 to Palisades Nuclear Power Plant Third Interval ISI Program for Class 1,2 & 3 Components & Supports, Vols I & II

ML18064A888
Person / Time
Site:	Palisades
Issue date:	03/16/1995
From:	Van Wagner B CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18064A887	List: ML18064A886 ML18064A888 ML18065A098 ML18065A101 ML18065A102 ML18065A103 ML18065A105 ML18065A108 ML18065A109 ML18065A110 ML18065A111 ML18065A112 ML18065A113 ML18065A114 ML18065A115 ML18065A116 ML18066A740
References
PROC-950316, NUDOCS 9509120015
Download: ML18064A888 (469)

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Consumers Power Company 212 West Michigan Ave. Jackson, MI 49201 Third Ten-Year Inspection Interval Inservice Inspection Plan Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043

 **     Commercial Service Date:  12/31/71 Authorized Inspection Agency Factory Mutual Engineering Association 30150 Telegraph Road, Suite 141 Bingham Farms, MI 48025 I
                                             .,i

. )

PAL~SADES NUCLEAR PLANT INSERVICE INSPECTION PROGRAM Revision and-Approval S\jnimary TITLE: THIRD lO-YEAR INSPECTION INSERVl0:: INSPECTION PLAN SUBMiTTAL .. , . . Revision O Date

TABLE OF CONTENTS Section 1 Introduction 1 A. Historical Background 2 B. Upgrading Criteria 2-3 C. References 3 D. General 4-6 E. Inspection 6-7 Section 2 Outline of the Third 10 Year Program 8 A. Vessels 9 B. Piping 9-14 C. Extent of the Program 14-17 Section 3 Basis Statements 18 A. Exemptions .) 19-21 B. --Deferrals 22 C. iAugmentation 22-27 D. Additional Examinations 28 E. Consumers Powe~ Company - Palisades Nuclear Power Plant: Position 28 F. Repairs and Replacements 28-29 G. Containment Penetrations 29 Section 4 Palisades Plant Code Use 30-31 A. Code Cases for the Third Interva1 32 Section 5 Relief Requests 33-48 Section 6 Verification of Section XI Compliance 49 A. Introduction 50 B. Determination of Compliance 50 C. Number of Components 51 D. Interval Compliance 51 E. Midinterval Requirement Changes 51 F. Verification of Compliance - Third Interval by Category * - 52 G. Verification of Compliance - Third Interval by Category and Item Number 53 Section 7 Piping and Instrument Diagrams 54 Section 8 Third Interval, Section XI Category and Item Number - Scheduled Examinations by Period 55

TABLE OF CONTENTS (CONTINUED) PAGE(S) Section 9 Third Interval, Section XI Category and Item Number - Scheduled Examinations by Period and Outage Year 56 Section 10 ISI Isometrics for Scheduled Examinations during the Third Inservice Inspection Interval 57

• 1 SECTION 1 INTRODUCTION
• 1.

2 This document is a summary of the plan for Inservice Inspection (ISI) to be performed over the third 10-year interval on Class 1, Class 2 and Class 3 pressure retaining components and their supports of Consumers Power Company's (CPC) Palisades Nuclear Power Plant Unit 1. A. HISTORICAL BACKGROUND The Palisades Nuclear Power Plant was built in the late '60s and was placed in commercial service on December 31, 1971. During the first 40-month life of the plant, in order to comply with Paragraphs 4.3 and 4.12 of the Technical Specifications (dated September 1, 1972) of. the Provisional Operating License DPR-20 for the Palisades Nuclear Plant, which disc,usses ISI requirements of Class 1 components and systems, the nondestructive examinations were performed to satisfy the requirements of the ASME Section XI Code, 1971 Edition including the Winter 1972 Addenda. In February 1976, the NRC amended Paragraph 55a (g) of 10 CFR 50 to require nuclear plants to upgrade their Technical Specifications in the areas of the ISI requirements and the functional testing of pumps and valves. By amending Paragraph 55a (g) and by invoking Regulatory Guide 1.26, the NRC required nuclear plants to upgrade their- systems to include not only Class 1 systems, but also Class 2 and Class 3 systems in their ISI programs. B. UPGRADING CRITERIA The Construction of this Plan was based on the following documents:

1. Palisades Nuclear Plant's Piping and Instrument Diagrams and Plant Q-Li st.

• -*

_)

2. The 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Rressure Vessel Code, Section XI "Rules for

• Inservice Inspection of Nuclear Power Plant Components", Subsections IWA, IWB, IWC, IWD, and IWF, for Inspection Program B.

3

3. Requests for Relief From Provisions of Section XI, ASME B&PV Code, 1989 Edition.

4. USNRC "Rules and Regulations, Title 10, Chapter 1, Code of Federal Regulations-Energy", Part 50.55a.

5. Applicable sections of Paragraph 4.05, 4.3 and 4.12 of the Technical Specifications of the Provisional Operating License DPR-20 for the Pali sades Pl ant.

Components were scheduled for examination in accordance with the above stated rul,e-s and regulations. Examinations are conducted in accordance with the ASME Boiler and Pressure Vessel Code.

c. REFERENCES

1. 10CFR50.55a (g)

2. Provisional Operating License (DPR-20), Technical Specifications for the Palisades Plant, Docket No 50-255, Appendix A, Sections 4.3 and 4.12, per Change 9 dated October 9, 1973 and Amendment 53 dated October 15, 1979.

3. ASME Boiler and Pressure Vessel Code, Section XI, 1989 Edition.

4. Consumers Power Quality Assurance Program Description for Operational Nuclear Power Plants, CPC-2A.

5. Palisades Nuclear Plant Administrative Procedures .

 )    6. Palisades Plant.Engineering Manual Procedure.

D. GENERAL

1. This Inservice Inspection Plan for the four 10-year Inservice Intervals (see E below) has been developed, reviewed and approved by Consumers Power Company for use at the Palisades Nuclear Power Plant Unit 1. This plan incorporates all applicable relief requests and periodic surveillance requirements of References C-2, C-3 and C-4 for the 40-year service lifetime.

The start of the first 10-year interval coincides with the date of first commercial operation, December 31, 1971. The length of the first three and one-third year period-was extended to October 30, 1976 by adding eighteen months cumulative shutdown time between August 1973 and April 1975 in accordance with IS-241, Section XI, 71/W72~--- The second period of the first 10-year interval was scheduled to end on February 28, 1980. The beginning of the third period was delayed until June 1, 1980 due to the 1979/1980 extended refueling outage. The Palisades Plant was out of serv1ce from September 1979 through May 1980. The interval was extended to November 9, 1983, per

                             -- .

IWA-2400(c) - 77S78 Addenda, to coincide with outage inspection completion. The second interval (first period), therefore began November 10, 1983. The Palisades Plant was out of service from August 12, 1983 through July 30, 1984 and from May 19, 1986 through April 3, 1987 due to Extended Maintenance Outages. Therefore, the second interval has been extended to May 11, 1995, per IWA-2400(c), ASME Boiler and Pressure Vessel Code, Section XI, 83S83.

2. Responsibility for the maintenance of this plan and the development of subsequent plans rests with the Palisades Inservice Inspection Section.

3. In view of the fact.that the design of the Palisades Nuclear Plant was completed prior to the issuance of Section XI of the ASME B&PV **

Code, the inspection access requirements of IS-142, '71 Edition were* not available to impact the plant design parameters. The Technical Specifications/Relief Request of this plan detail specific Code requirements which cannot be met.

4. Examination methods performed are intended to be representative of past ISi practice or of preservice methods utilized. In either case, it should be recognized that either UT or RT are acceptable volumetric exams and either PT or MT are acceptable surface exams.

Unique weld joint parameters may, of course, dictate more restrictive selection criteria; eg, high background radiation will preclude RT, austenitic stainless materials will preclude MT, ~tc. It is intended that tne process which selects exam method for inspections under this plan treat UT and RT as interchangeable and PT and MT as interchangeable with consideration given to past practice in light of the reproducibility of results.

5. The .Preservice Inspection (PSI) conducted at the Palisades Plant was of limited scope and included all or portions of only the following components/lines:

a. Reactor Vessel

b. Primary Side of the Steam Generators

c. Pressurizer

d. Regenerative Heat Exchangers

e. Main Recirculation Loops

f. Primary Coolant Pump Flywheels

g. 12" Safety Injection Lines (Class 1)

h. 6" Safety Injection Lines (Class 1)

i. Pressurizer Surge Line

j. 12" Shutdown Cooling Line (Class 1) k

• Pressurizer Relief Lines

)     l. Pressurizer Spray Lines

m. Auxiliary Spray Lines

6 Specific information concerning the exact scope and the results of the PSI are contained in the "Palisades Nuclear Power Plant Preoperational Inspection Report" (SWRI Project 17-2249), dated July 25, 1971.

6. This plan does not account for the pump and valve, snubber or pressure testing programs of Section XI. This plan does not administer the steam generator eddy current programs which is required by Section XI and the Palisades Plant Technical Specifications. These programs are described in the following Engineering Manual Procedures.

a. EM-09-02, "Inservice Testing of Plant Valves"

b. EM-09-04, Inservice Testing of Selected Safety Related Pumps 11 11

c. EM-09-05, Steam Generator Inservice Inspection 11 11

.) d. EM-=p-9--11, "Palisades Eddy Current Test Procedure for Data Management for Replacement Steam Generators

e. EM-09-13, "Inservice Inspection Pressure Testing Program

f. EM-09-07, Testing of Plant Snubbers" 11 11 11 E. INSPECTION INTERVALS The following table delineates the inspection intervals for the Palisades Plant:

Table II 1st Interval From 12/31/1971 to 09/30/1983 1st Period 12/31/1971 to 10/31/1976 *See D Above 2nd Period 11/01/1976 to 05/31/1980 *See D Above 3rd Period 06/01/1980 to 11/09/1983 *See D Above 2nd Interval From 11/10/1983 to 05/11/1995 *See D Above 1st Period 11/10/1983 to 09/10/1988 2nd Period 09/11/1988 to 01/11/1992 3rd Period 01/12/1992 to 05/11/1995 3rd Interval From 05/12/1995 to 05/11/2005 1st Period 05/12/1995 to 09/11/1998 2nd Period 09/12/1998 to 01/11/2002 3rd Per'i od 01/12/2002 to 05/11/2005

7

• 4th Interval 1st Period From 05/12/2005 to 05/11/2015 2nd Period 3rd Period

8 SECTION 2 OUTLINE OF THE THIRD 10-YEAR PLAN

• /

  )

2. OUTLINE OF THE THIRD 10-YEAR PLAN

. A. VESSELS The examination areas in the reactor pressure vessel and closure head, pressurizer, steam generators, reactor coolant pumps and regen~rative heat exchanger are identified either by name or by the Combustion Engineering component identi.fication number shown on.the appropriate Consumers Power Company drawings. - B. PIPING The following detail descriptions provide the systematic process and logic for the ass.ignment of weld identification.

1. The first character set in each code consi~ts of alphabetic codes which designate the main piping system under consid~ration. Below is a table showing the codes and their respective systems.

CCS - Component Cool*ing System CVC - Chemical and Volume Control System -

       - - DMW - Demi nera l i zer Water System ESS - Engineered Safeguards System FWS ~ Feedwater System MSS - Main Steam System PCS - Primary C.ool ant System PMW - Primary Makeup Water System RWS - Radwaste System .

SFP Spent Fuel Pool System SWS - Service Water System VAS Vent and Air tonditioning System

2. The second character set consists of a number designating nciminal diameter of the pipe under consideration.

3. The third character set consists of alphabetic ~odes with indicate subsystems or functions of the line under consideration. Below is a table showing subsystem codes.

ARH Auxiliary Return Header ASH - Auxiliary Supply Header AWS - Auxiliary Feedwater System CCS - Component Cooling System CHL - Charging Line CHP - Charging High Pressure CHX - Cooling Heat Exchanger CMU - Condensate Makeup CPU - Component Cooling Pump CPL Cavity Pool Line CRS ~ohtainment Air Cooling

• CSH - Containment Supply Header CSS - Core Spray System CST - Condensate Storage Tank CSW - Critical Service Water CWR - Clean Waste Receiver DRL -. Drain Line EPS - Emergency Power Supply FPF - Fuel Pool Fill FPP - Fuel Pool Pump FWL - Feedwater Line IRS - Iodine Removal System LDD - Letdown Drain LDL - Letdown Line LTC - Long Term Cooling MSL - Main Steam Line MSV - Main Steam Valve Line PRS - Pressure Relief System PSL - Pressurizer Surge Line PSS - Pressurizer Spray System PTO - Sample Tap
• RtL RE RHC

          -
          -
          -

Main Reactor Coolant Loop Recirculaticin*Ltne Return Header Combined 11 RWS - Radwaste System RVR - Main Steam Relief Valve Riser SCH - Shutdown Cooling Heat Exchanger SCS - Shutdown Cooling System SOC - Shutdown Cooling SFP - Spent Fuel Pool SFX - Spent Fuel Heat Exchanger SIS - Safety Injection System SRT - Safety Refµeling Tank SWP - Service Water Pump SX - Sample Line ., 4~ The fourth character set consists of the following: a .. A number-letter combin~tion ~esignating, in the case of the Main Reactor Coolant Loop, the line under consideration. Iri the case of associated auxiliary or core cooling lines, this code designates the line in the Main Reactor Coolant Loop which the line under consideration ultimately ties into. The codes and their respective designations are as follows: IA 30" Cold Leg "A" From Steam Generator "A" to Reactor Vessel 18 30 11 Cold Leg "B" From Steam Generator "A" to Reactor Vessel* 1H Loop* "A" Hot Leg - 42 11 2A 30" Cold Leg "C" From Steam Generator "B" to* Reactor Vessel 28 30" Cold Leg "D" From Steam Generator "B" to Reactor Vessel 2H Loop 8 Hot 11 11 Leg - 42 11 IP Pressurizer

b. A number directly following the loop designation distinguishes between two or more lines of the same subsystem relating to the

12 same loop. In the case of the same subsystem relating to a given 1oop, the number l11 wi 11 be used for consistency of the code. 11

c. A line will be identified with a loop, .if possible; if not, it will be identified with a major component.

5. The fifth character set represents the weld number respective to the direction of flow or direction of longitudinal weld from a circumferential weld and a symbolic indication, if appropriate, according to the legend of symbols. Some examples are as follows:

1, 2, 3, 4 and 5 are circumferential weld numbers in the direction of flow, 4LU is longitudinal weld upstream from circumferential weld 411 11 and 6LD is longitudinal weld downstream from circumferential weld 116 *II In the case of two or more longitudinal welds in a pipe, the Number 1 weld would be at 11 Lo 11 as chosen from "Method for Determining Zero Reference Location of Pipe Welds" (see Section XI, Appendix III ASME Code) or first weld clockwise from Lo facing direction of flow. 11 11 The Number 2 weld would then be the next.weld clockwise, etc. In an elbow, Number 1 would be inside and 2 outside. 6LD1 is 11 11 11 11 longitudinal weld downstream from circumferential weld 6 when there 11 11 is more than* one longitudinal weld and.this was the first one CW of

     .

11 Lo." 7PR is a pipe lug downstream from circumferential weld 7; 3PS is pipe support downstream from circumferential weld 3. The Class 2 piping, the weld numbers begin with weld number 201. This number system is used to designate only the Class 2 weld-numbers and their associated pipe hanger, restrairits, supports and bolting.

6. If ~here is a sixth character set, it describes the successive welds of the fifth character set. 3PS6 is sixth pipe support downstream

~) from circumferential weld 3.

13 Letters refer to sequential branch connections. Numbers refer to hanger lugs, pipe supports and restraint lugs. If there is more than one hanger lug at a particular position on a pipe, they are numbered clockwise with respect to direction of flow (similar to the L dimension) .

.) 14

7. For branch conn~ctions larger than one inch, the following notation is used. The nctation for these branch connections is in three parts:

(ESS-12-SIS-lAl-l) I (ESS-2-SIS-lAl) (*) (0) (*) . Contains the systems, loop and weld number upstream.

         /      Indicates the b~anch ~onnection.

(0) Contains the branch line identifier.

8. Some examples of weld numbering are as follows:

ESS-6-SIS-lAl-6 The sixth circumferential weld from the designated system boundary in direction of flow in the 6" Safety Injection Line connecting the Loop "A" Cold Leg. PCS-30-RCL-18-4/ 12" Safety Injection Line lB Branch Connection ESS-12-SIS-181 weld into 30" Cold Leg Loop "B" downstream from circumferential weld number "4". PCS-12-PSL-1H-3PL Pipe Restraint Lug toward 42" Hot Leg "A" from Number 3 circumferential weld in the Pressurizer Surge Line .

        . PCS-30-RCL-2A-3LD-2    The Second longitudinal weld CW in direction of flow from "Lo" in a pipe. This weld in downstream from circumferential weld Number "3."

c. EXTENT OF THE PROGRAM This submittal covers the third 10-year inspection interval. The Palisades Plant utilizes Program Bas set forth in IWA-2430, Section XI.

15 Selected portions of the major components and/or systems to be examined in accordance with Section XI are as follows:

1. Cl ass 1

a. Reactor Pressure Vessel

b. Reactor Pressure Vessel Closure Head

c. Steam Generator - Primary Side

d. Pressurizer

e. Regenerative Heat Exchanger - Primary Side

f. Piping

           - Primary Coolant System
           - Engineering _Safeguards System
           - Chemical and Volume Control System

g. Primary Coolant Pumps

.I 2.

h. Valves Class 2

a. Pressure Vessels

            - Concentrated Boric Acid Tanks
            - Safety Injection Tanks
            - Boric Acid Filter
            - Steam Generators - Secondary Side
            - Regenerative Heat Exchangers - Secondary Side
            - Shutdown Cooling Heat Exchangers
            - SIRW Tank
            - Iodine Removal Tanks

b. Piping

            - Primary Coolant System
            - Main Steam System
            - Feedwater System
            - Engineered Safeguards System
            - Chemical and Volume Control System
            - Radwaste System

• 16

        - Vent and Air Conditioning System
        - Component Cooling System
        - Spent Fuel Pool System

c. Pumps

        - Containment Spray Pumps
        - Charging Pumps
        - Concentrated Boric Acid Pumps
        - High-Pressure Safety Injection Pumps
        - Low-Pressure Safety Injection Pumps
        - SIRW Tank Recirculation Pump

3. Class 3

a. Pressure Vessels

        - Condensate Storage Tank
        - Component Cooling Surge Tank
        - Spent Fuel Pool Heat Exchangers
        - Component Cooling Heat Exchangers

• - Letdown Heat Exchanger .

        - Shutdown Cooling Heat Exchangers
        - Shield Cooling Heat Exchangers
        - Engineered Safeguards Room Coolers
        - Control Room Air Conditioning Units
        - Containment Air Coolers

b. Piping

        - Primary Coolant System
        - Main Steam System
        - Feedwater System
        - Engineered Safeguards System
        - Chemical and Volume Control System
        - Service Water System Component Cooling System
        - Chemical Addition System
        - Spent Fuel Pool System
        - Makeup and Demineralizer Water System
        - Primary Makeup Water System

17

c. Pumps

     - Service Water Pumps
     - Auxiliary Feedwater Pumps
     - Fuel Pool Cooling Pumps
     - Component Cooling Pumps
     - Auxiliary Feedwater Pump Turbine Driver

.)

• 18 SECTION 3 BASIS STATEMENTS
• 19

3. BASIS STATEMENTS The following seven sections delineate the basis used by Consumers Power Company for exemptions, exclusions, or deferral of examinations or other modifications of the requirements of Section XI.

A. EXEMPTIONS All Classes 1, 2 and 3 pressure retaining components (and their supports) are subject to examination. However, Section XI provides rules for exempting components from volumetric and surface examinations (ie, IWB-1220, IWB-2500, IWC-1220, IWC-2500, IWD-1220) and Federal law allows the regulatory authority (NRC) to grant relief from specific portions of the code upon demonstrated need.

                      -          -
        .Tables Al and A2 list all the Class 2 components exempted except those

• exempted due to pipe/nozzle size. Class 1 piping less than or equal to l" nominal diameter is exempt by IWB-1220(b)(l). All Class 1 valve bodies (B-M-2) not exceeding 4" nominal pipe size are excluded by Table IWB-2500.

TABLE Al CLASS 2 OPERATIONAL EXEMPTIONS IWC-122l(e) 1989 Edition Component Function T-82A SI Bottle T-828 SI Bottle T-82C . SI Bottle T-820 SI Bottle

• • )
• BASIS: IWC-122l(e), 1989 Edition exempts components from NOT requirements.

when a vessel, piping, pumps, valves, other components and component connections of any size in statically pressurized, passive (i.e., no 20 pumps) safety injection system of pressurized water reactor plants. The SI bottles are filled to capacity (with borated water and nitrogen) and as such can be considered statically pressurized. The bottles are maintained at the pressure required to operate .

• TABLE A2 c:.ASS 2 PRESSURE TEMPERATURE EXEMPTION IWC-1222(C) 1989 Edition 21 Maximum Maximum Operating Operating Component Function Pressure Temperature T-53A Boric Acid Storage Tank Atm 200°F T-538 Boric Acid Storage Tank Atm 200°F T-102 NaOH Tank 16 Psig 150°F T-103 NaOH Makeup Tank 16 Psig 150°F BASIC: IWC-1222(c), 1989 Edition allows exemptions from NOT requirements for components where the maximum operating pressure and temperature do not exceed 275 psig and 200°F, respectively. Data obtained from M-250 or manufacturer's drawings .
• B. . DEFERRALS 22 Section XI of the code provides a degree of latitude in the scheduling of major component examination in that certain category examinations may be deferred to the end of the inspection interval. Examples of major component examinations are mechanized UT of the reactor vessel, pump and valve teardowns and core internal examinations.

COMPONENTS DEFERRABLE TO THE END OF INTERVAL Section XI Category Component* Comments B-L-1 Pump Casing Welds Code Case N-481 B-L-2 Pump Casing Code Case N-481

      *B-M-1           Valve Body Welds B-M-2           Valve Bodies

• C.

B-N-3 B-0 AUGMENTATION Core Support Structures Welds in CROM Housing Plant Technical Specifications may at times require more frequent examination scheduling than does Section XI as is the case with the regenerative heat exchanger and high-energy piping. Section XI requirements are superseded by Technical Specifications and these examinations are not subject to Paragraphs IWB-2400 and IWC-2400, Section XI. The following Palisades Technical Specifications sections apply to the Augmented Program and will be performed throughout the life of the plant.

• TABLE 4.3.2 Miscellaneous Surveillance Items 23 Eguigment Method Freguenc}'.

1. Regenerative Heat Exchanger

a. Primary Side Shell Volumetric 5-Year Maximum Interval (100%)

to Tube Sheet Welds

b. Primary Heat Volumetric 5-Year Maximum Interval (100%)

2. Primary Coolant Pump Volumetric 100% Upper Flywheel Each Flywheels Refueling

• AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH-ENERGY LINES OUTSIDE OF CONTAINMENT 24 APPLICABILITY Applies to welds in piping system or portions of systems located outside of containment where protection from the consequences of postulated ruptures is not provided by a system of pipe whip restraints, jet impingement barriers, protective enclosures, and/or other measures designed specifically to cope with such ruptures.

For the Palisades Plant, this specification applies to welds in the main steam and main feedwater lines located inside the Main Steam and Feedwater Penetration Rooms.

...:

OBJECTIVE

• To provide assurance of the continued integrity of the piping systems over their service lifetime.

SPECIFICATION 4.12.1 For welds identified in Figure 4.12.A (Main Steam Lines) and Figure 4.12.B (Feedwater Lines): A. At the first o~tage period, such as refueling,. a volumetric ex~mination shall be performed with 100 percent inspection of welds in accordance with the requirements of ASME Section Xl,Code, "Inservice Inspection of Nuclear Power Plant Components. 11 B. The inservice inspection at each weld shall be performed in accordance with the requirements of ASME Section XI Code, "Inservice Inspection of Nuclear Power Plant Components," with the following schedule (the inspection intervals identified below sequeniially follow the baseline examination of 4.12.1.A*above):

• First Inspection Interval

1. First 3-1/3 years (or nearest 100% Volumetric inspection 25 refueling outage). of all welds.

2. Second 3-1/3 years (or nearest* 100% volumetric inspection refueling outage). of all welds.

3. Third 3-1/3 years (or nearest 100% volumetric inspection refueling outage). of all welds.

Successive Inspection Intervals Every 10 years thereafter (or Volumetric inspection of 1/3 nearest refueling outage) of the welds at the expiration of each 1/3 of the inspection

*  '

interval with a cumulative 100% coverage of all welds. The welds selected during each inspection period shall be distributed among the total number to be examined to provide a* representative sampling of the conditions of all welds. C. Examinations that reveal unacceptable structural defects i~ a weld during

• an inspection should be extended to require an additional inspection of another 1/3 of the welds. If further unacceptable defects are detected in the second sampling, the reminder of the welds shall be inspected.

D. In the event repairs of any welds are required following any examination during successive inspection intervals, the inspection schedule for the repaired welds will revert back to the first inspection interval. 4.12.2 For other welds {excluding those identified in Figure 4.12.A and

• • -)

Figure 4.12.B):

• A. Welds in the main steam lines including the safety valve attachment 26 welds and in the feedwater lines shall be examined in accordance.with the requirements of Subsections ISC-100 through 600 of the 1972 Winter Addendum of ASME Section XI Code.

4.12.3 For all welds in the main steam lines and main feedwater lines located inside the Main Steam and Feedwater Penetration Rooms: A. A visual inspection of the surface of the insulation at all weld locations shall be performed on a weekly basis for detection of leaks. Any detected leaks shall be investigated and evaluated. If the leakage is caused by a through-wall flaw, either the plant shall be shut down or the leaking piping isolated. Repairs shall be performed prior to return of this line to service.

                  *-*

B. Repairs, reexamination and* p1p1ng pressure tests shall be conducted

• in accordance with the rules of ASME Section XI Code .

Under normal plant operating conditions, the piping materials operate under ductile conditions and within the stress limits considerably below the ultimate strength properties of the materials. Flaws which could grow under such conditions are generally associated with cyclic loads that fatigue the metal, and lead to leakage cracks. The inservice examination and the frequency of inspection will provide a means for timely detection even before the flaw penetrates the wall of the piping.

• Miscellaneous Augmented Examinations Palisades Plan for examining the "Structural Integrity for the Auxiliary 27 Feedwater System Piping Associated with the Steam Generators, reference letters RJB 34-88 dated May 18, 1988, BVV 88-032 dated July 14, 1988 and THF 88-001 dated January 28, 1988, which shows evidence of examinations already performed.

A synopsis of those letters mentioned above consists of the examinations listed below:

1. Pipe- to -Elbow weld - Perform volumetric and surface examinations

2. Elbow- to -Pipe weld - Perform volumetric and surface examinations

3. Pipe- to -Nozzle weld - Perform volumetric and surface examinations

4. Peiform ultrasonic wall thinning examinations beginning at the Elbow- to -

Pipe weld downstream of the Stea~ Generators.

5. Perform visual examinatio~s of internal knuckle region, provided the.Steam

• Generators are open for secondary side inspections .

The above examinations are to be done once each 3-1/3 years (equivalent to once each ISI period). These examinations are to apply to both Steam* Generators and are included in the 40-Year Master Inservice Inspection Plan .

• 28 D. ADDITIONAL EXAMINATIONS Examinations performed during any inspection that reveal indications exceeding the allowable indication standards of Section XI IWB-3000 shall be subject to Palisades ISI evaluation per Section XI IWB-IWC-IWF-2400.

Subsection 2430. If weld processing discontinuities (ie. porosity, slag, incomplete fusion or penetration) are detected during any inspection that exceed the allowable indication standards of Section XI IWB-3000 they shall be subject to Palisades ISI evaluation but no expansion of examination scope will be required. The additional examinations as required by Section XI IWB-IWC-IWF-2400. Subsection 2430 will only be performed if service induced discontinuities are detected. E. CONSUMERS POWER PALISADES NUCLEAR POWER PLANT: POSITION The transition pieces between the carbon steel nozzles and the carbon steel piping are also carbon steel and thus not dissimilar metal safe ends. The nozzle to safe end welds are therefore classed as B9.ll. B-J welds rather than B5.10 or BS.30, B-F welds. (See CE Dwg E-232-119-11.) These welds will be included and inspected to Category B-J requirements with the restriction that the inspection be expanded to include 100% of each weld during each inspection interval. F. REPAIRS AND REPLACEMENTS Repairs and replacements at Palisades Nuclear Power Plant are performed in accordance with the ASME Section XI Repair and Replacement Program. As required by Articles IWA-4000 and IWA-7000. this program delineates the essential requirements of the complete repair cycle including weld repairs. procurement and installation of replacements. The program consists of administrative procedures 0hich describe overall departmental responsibilities and interfaces. the Authorized Nuclear Inspector's involvement and documentation requirements. Also.

• 29 Maintenance and Quality Assurance departmental procedures implement controls for special ~roc2sses essential to the repair program such as flaw removal, weld repair, post weld heat treatment and non-destructive examination.

The Repair and Replacement Program complies with the requirements if IWB, IWC, IWD and IWF-4000 and 7000 of ASME Section XI. G. CONTAINMENT PENETRATIONS Palisades was constructed to the 1955 Edition of ASA 831.1 and is not subject to the rules of ASME Section III. However, Palisades has optionally upgraded all containment penetrations to ASME Class 1 or ASME Class 2. Con ta i nmen,Cpenetrat ions at Pali sades are tested in accordance with

• Palisades Technical Specification program to satisfy the requirements of IQ CFR 50 Appendix J. Containment penetrations which are part of* ASME

    *Class l; 2 or 3 system will also be tested under Palisades Section XI program for these systems (ie, hydrostatic testing, period leakage tests and NOE).

Containment penetrations which are not part of an ASME Class 1, 2 or 3 system will only be tested under Palisades Appendix J program as appropriate. Repairs of containment penetrations are conducted in accordance with Palisades Section XI Repair/Replacement Program .

• 30 SECTION 4 PALISADES PLANT .

Code Usage

• 31 INSERVICE INSPECTION PROGRAM - CODE USE APPLICABLE EDITIONS & ADDENDA OF .A..~ME BOILER & PRESSURE VESSEL CODE SECTION XI Pursuant to Paragraph 50.55a(g) of 10 CFR Part 50, the inservice examination requirements applicable to nondestructive examination at the Consumers Power Company, Palisades Plant, are based upon the rules set forth in the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code, for Class 1, 2, 3 piping and component supports .
• A. CONSUMERS POWER COMPANY - PALISADES NUCLEAR POWER PLANT INSERVICE INSPECTION PROGRAM CODE CASE UTILIZATION Consumers Power Company elects per 10 CFR 10.55a(a)(3) and Footnote 6 to 32 utilize the following ASME Section XI Code Cases as an integral part of the Third 10-Year Interval Inservice Inspection Plan for the Palisades Nuclear Power Plant.

Code Case N-311 Alternate Examination of Outlet Nozzle on Secondary Side of Steam Generators Code Case N-457 Qualification Specimen Notch Location for Ultrasonic Examination of Bolts and Studs Code Case N-460 Alternate Examination Coverage for Class 1 and 2 Weld Code Case N-461 Alternate Rules for Piping Calibration Block Thickness Code Case N-463-1 Evaluation Procedures and Acceptance Criteria for Flaws in Class 1 Ferritic Piping that Exceed the Acceptance Standards of IWB-3514.2

• Code Case N-481 Code Case N-489 Alternate Examination requirements for Cast Austenitic Pump Casings (see attachment 1 to this submittal)

Alternate Rules for Level Ill NOE Qualification Examinations Code Case N-491 Alternate Rules for Examination of Class 1, 2, 3 and MC Component Supports of Light-Water Cooler Power Plants Code Case N-494 Pipe Specific Evaluation Procedures and Acceptance Criteria for Flaws in Class 1 Ferritic Piping that Exceed the Acceptance standards of IWB-3514.2 Other Code Cases not approved by Regulatory Guide 1.147 Revision 10 are included as Relief Requests for NRC approval.

• COMBUSTION ENGINEERING OWNERS GROUP CEN-412 Revision 2 RELAXATION OF REACTOR COOLANT PUMP CASING INSPECTION REQUJREMENTS CEOG TASK 678 Prepared for the C-E OWNERS GROUP APRIL 1993 jl 1111 ABB Combustion Engineering Nuclear Power ,.,,,1, ASEA BROWN BOVERI

LEGAL NOTICE

                                                      **-

THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION INGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF: A. MAKES ANY WARRANTY OR ~ATION, EXPRFSS OR IMPUm INCLUDING THE WARllANTm OF~ FOR A PARTICULAR. PURPOSE OR MERCHANTABILITY, WITH ~TO THE ACCURACY, CO~, OR~ OF THE INFORMATION CONTAINm IN THIS REPORT, OR mAT THE USE OF ANY INFORMATION, APPARATIJS, ME'IHOD, OR PR~ DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS; OR B. ASmJMIS ANY UABII.JTm WITH ~ TO THE USE OF OR FOR DAMAGD DmJLTING FROM THE USE OF ANY INFORMATION, APPAR.ATIJS, METHOD ORPROOSS DISCLOSED IN THIS REPORT.

• This report shows that all of the casings evaluated have the necessary material characteristics to justify a relaxation from the ASME 10-year inspection intervals.

3

ABSTRACT The ASME Boiler and Pressure Vessel Code requirement for inservice inspection of reactor coolant pump (RCP) casings

• consists of a very difficult and costly volumetric examination of welds and visual inspection of internal surf aces at 10-year intervals. An extension of the inspection interval would reduce personnel radiological exposure significantly and would help to reduce plant operating and maintenance costs.

Five utilities have funded material and structural evaluations by ABB Combustion Engineering Nuclear Power to justify increasing the inservice inspection intervals for their RCP casings. Justification for increasing the inspection interval is dependent upon demonstrating that flaws that might have been present initially in the casing will not grow to an unacceptable size during the proposed relaxed inspection interval. An initial crack size is postulated based upon the established cqde acceptance criteria for the baseline material pre-service examination. An empirical correlation for cyclic crack growth is then used to predict growth histories of hypothetical cracks over extended periods of time. Original material certification records and stress analyses for each pump casing provide plant-specific details for this analysis. Ligament stren9th is evaluated and assessments of stress intensity factors associated with the growing crack are compared to the reduction of material fracture toughness with time, to

                                        ~

determine the number of years needed to reach an end-point crack size. An engineering evaluation of the resulting time period provides a means for establishing inspection intervals which ensure -that RCP casing integrity will be maintained during the inspection interval. 2

TABLE OF CONTENTS SECTION TITLE ABSTRACT 2 LIST OF TABLES 6 LIST OF FIGURES 7 1.0

SUMMARY

13

2.0 INTRODUCTION

16 2.1 Purpose 16 2.2 Scope and Applicability 16

3.0 BACKGROUND

17 3.1 Goals 17 3.2 RCP Casing Descriptions 17 3.2.1 General 17 3.2.2 Casing Design 18 3.2.3 Fabrication 18 3.2.4 Inservice Inspections 19 3.2.5 Codes and Standards 19 3.3 RCP Casing Material Degradation Mechanisms 27 3.3.1 Fatigue and Crack Growth Rates 27 3.3.2 Thermal Embrittlement 29 Section 3 References 33 4

SECTION TITLE 4.0 CURRENT INSPECTION REQUIREMENTS FOR RCP CASINGS 35 4.1 NRC Requirements 35 4.2 ASME Requirements 35 4.3 Applicability of ASME Code Case N-481 36 Alternative 4.4 ASME Code Case Reference Flaw 37 4.5 Postulated Initial Crack Depth 38 Section 4 References 38 5.0 METHODOLOGY FOR EVALUATING PUMP CASINGS 39 e 5.1 crack Growth Analysis 39 5.1.1 Selection of Locations for Most C~nservative Analysis 39 5.1.2 Initial Flaw Size 40 5.1.3 Plant Operatinq History 41 _5.1.4 Calculation of Crack Growth Rates 42 5.1.5 Plant-specific Results 48 5.2 Thermal Embrittlement Analysis 56 5.2.1 Material Heat Data 57 5.2.2 Estimate of Minimum Charpy Impact Enerqy 59 5.2.3 Conversion of Charpy Impact Enerqies to J-inteqrals 60 5.2.4 Conversion of J-inteqrals to Plane 5.2.5 Strain Fracture Touqhness Plant-specific Results 68 69 e 5

.. SECTION TITLE PAGE 5.3 End Point crack Size Determination 70 5.3.1 Crack Size for Non-ductile Crack Propagation 71 5.3.2 Crack Size for Unstable Ductile Tearing 73 5.3.3 Crack Size for Ultimate Strength Limit 74 5.3.4 Crack Depth for Emergency Condition* and Faulted Condition Loads 74 5.3.5 Plant Specific Results 91 6.0 INSPECTION INTERVAL DETERMINATION 6.1 Safety Margins 94 6.2 Inspection Alternatives 96 6.2.1 Volumetric Examination 96 6.2.2 Visual Examination 97 6~3 Inspection Interval 98 6.4 Plant-specific Inspection Intervals 99

7.0 CONCLUSION

S 100 TABLE NUMBER LIST OF TABLES 3.2-1 Reactor Coolant Pump Fabrication History 20 5.1-1 Coefficients for Calculating KI 45 5.3-1 Comparison of KI(applied) and KJc Touqhness 72 7.0-1 End-point Crack Sizes as a Percent of Wall Thickness 101 6

LIST OF FIGURES e

                                                               '

FIGURE NUMBER TITLE 3.2-1 Typical Pump Vertical Cross-section 21 3.2-2 Typical Casing Vertical Cross-section 22 3 .* 2-3 Typical Casing Horizontal Cross-section 23 3.2-4 Hub Section After Casting, Before Welding 24 3.2-5 Volute Section After Casting, Before Welding 25 3.2-6 Welded Casing Assembly Before Machining 26 3.3-1 Time-Temperature Curve 34 5.1-1 Growth Curves for Palisades RCPs 51 5.1-2 Growth Curves for Fort Calhoun RCPs 52 5.1-3 Growth curves for Calvert Cliffs RCPs 53 5.1-4 Growth Curves for San Onofre RCPs 54 5.1-5 Growth Curves for St. Lucie RCPs 55 5.2-1 5.2-2 Flow Diagram for Fracture Toughness Curves Comparison of J-R Curves, 63 64 e 5.2-3 Comparison of Ferrite Content Predictions 65 5.2-4 curves to Define Jic 66 5.2-5 Flow Stress Values for Aged and Unaged Cast SS 67 5.3-1 Palisades curves With Design Condition Limits 76 5.3-2 Fort Calhoun Curves With Design Condition Limits 77 .5. 3-3 Calvert Cliffs Curves With Design Condition Limits 78 5.3-4 San Onofre Curves With Design Condition Limits 79 5.3-5 St. Lucie Curves With Design Condition Limits 80 5.3-6 Palisades Curves With Emergency Condition Limits 81 5.3-7 Ft. Calhoun Curves With Emergency Condition Limits 82 5.3-8 Calvert Cliffs Curves With Emergency Condition Limits 83 5.3-9 San Onofre Curves With Emergency Condition Limits 84 7

LIST OF FIGURES (Cont'd) FIGURE NUMBER TITLE PAGE 5.3-10 St. Lucie Curves With Emergency Condition Limits 85 5~3-11 Palisades Curves With Faulted Condition Limits 86

5. 3-12 Fort Calhoun curves With Faulted Condition Limits 87 5.3-13 Calvert Cliffs Curves .With Faulted Condition Limits 88 5.3-14 San Onofre curves With Faulted condition Limits 89 5.3-15 St. Lucie Curves With Faulted Condition Limits 90 APPENDIX TITLE A. APPLICATION OF METHODOLOGY TO PALISAJ;>ES 1.0 Purpose 5 2.0 Pre-service Inspection Data Evaluation 6 3.0 Operatinq History 8 4.0 Thermal Embrittlement 10 10 10 11

                                                            . 11 12 26 26 5.2  Reference Stress Reports                  26 5.3  Selection of High Stress Locations        26 5.4  Stresses and Wall Thicknesses at Limiting Locations                        27 5.5  Calculation of Crack Growth Rates         28 8

APPENDIX TITLE PAGE e 5.6 Stresses Under Emerg_ency and Faulted Conditions 32 5.7 Results 33 6.0 Inspection Interval 34 Appendix A References 35 B. APPLICATION OF METHODOLOGY TO FORT CALHOUN 1.0 Purpose 5 2.0 Pre-service Inspection Data Evaluation 6 3.0 *Operating History 8 4.0 Thermal Embrittlement 10 4.1 Material Identification and Chemical Properties 10 4.2 Material Specifications and Mechanical Properties 10 4.3 Thermal Aging Behavior 11 4.4 Toughness Properties of Aged Materials 11 4.5 Limiting Values 12 5.0 crack Growth Analysis 30 5.1 Scope 30 5.2 Reference stress Reports 30 5.3 Selection of High Stress Locations 30 5.4* Stresses and Wall Thicknesses at Limiting Locations 31 5.5 Calculation of. Crack Growth Rates 32 5.6 Stresses Under Emergency and Faulted Conditions 37 5.7 Results 39 6.0 Inspection Interval 40 Appendix B References 41 9

APPENDIX TITLE C. APPLICATION OF METHODOLOGY TO CALVERT CLIFFS 1&2 1.0 Purpose 6 2.0 Pre-service Inspection Data Evaluation 7 3.0 Operating History 9 4.0 Thermal Embrittlement 12 4.1 Material Identification and Chemical Properties 12 4.2 Material Specifications and Mechanical Properties 12 4.3 Thermal Aging Behavior 13 4.4 Toughness Properties of Aged Materials 13 4.5 Limiting Values 14 5.0 Crack Growth Analysis 39 5.1 Scope 39 5.2 Reference Stress Reports 39 5.3 Selection of High Stress Locations 39 5.4 Stresses and Wall Thicknesses at Limi~ing Locations 40 5.5 Calculation of Crack Growth Rates 42

                                             '

5.6 Stresses Under Emergency and Faulted Conditions 48 5.7 Results 51 6.0 Inspection Interval 52 Appendix c References 53 D. APPLICATION OF METHODOLOGY TO SAN ONOFRE 2&3 1.0 Purpose 6 2.0 Pre-service Inspection Data Evaluation 7 10

APPENDIX TITLE PAGE e 3.0 Operating History 9 4.0 Thermal Embrittlement 12 4.1 Material Identification and Chemical Properties 12 4.2 Material Specifications and Mechanical Properties 12 4.3 Thermai Aging Behavior 13 4.4 Toughness Properties of Aged Materials 13 4.5 Limiting Values 14 5.0 Crack Growth Analysis 43 5.1 Scope 43 5.2 Reference Stress Reports 43 5.3 Selection of High Stress Locations 43 5.4 Stresses and Wall Thicknesses at Limiting Locations 44 5.5 Calculation of Crack Growth Rates 46 e 5.6 Stresses Under Emergency and Faulted Conditions 47 5.7 Results 56 6.0 Inspection Interval 57 Appendix D References 58 E.* APPLICATION OF METHODOLOGY TO ST. -LUCIE 1&2 1.0 Purpose 6 2.0 Pre-service Inspection Data Evaluation 7 3.0 Operating History 9 4.0 Thermal Embrittlement 12 4.1 Material Identification and Chemical Properties 12 4.2 Material Specifications and Mechanical Properties 12 4.3 Thermal Aging Behavior 13 11

APPENDIX TITLE PAGE 4.4 Toughness Properties of Aged Materials 13 4.5 Limiting Values 14 5.0 Crack Growth Analysis 39 5.1 Scope 39 5.2 Reference Stress Reports 39 5.3 Selection of High Stress Locations 39 5.4 Stresses and Wall Thicknesses at* Limiting Locations 40 5.5 Calculation of Crack Growth Rates 42 5.6 Stresses Under Emergency and Faulted Conditions 48 5.7 Results 51 6.0 Inspection Interval 52 Appendix E References 53 F. COMPUTER CODE LISTINGS Description of Listings 2 Database File Structure 4 Computer dBase Program CASEINFO.PRG 6 computer dBase Pro9ram FRACTOUG.PRG 7 Computer dBase Program TEARMOD.PRG 9 12

1.0

SUMMARY

This report provides the results of an evaluation performed by ABB Combustion Engineering Nuclear Power for five participating members of the CE Owners Group to justify relaxation of several current requirements for reactor coolant pump (RCP) casing inspections. The five utility participants are Consumers Power Company, Omaha Public Power District, Baltimore Gas and Electric Company, the Southern California Edison Company, and the Florida Power and Light Company. The purpose, scope, and applicability of this evaluation are provided in Section 2.0. Section 3.0 provides a description of the RCP casings and a discussion of the material degradation mechanisms upon which the inspections are based. During the operating life of RCPs, periodic casing inspections are required by the NRC. These are stipulated in 10CFR50.55a to tt be in accordance with ASME code requirements and are expected to take place as part of the scheduled ten-year inservice inspection activities. Inservice inspection requirements are referred to in the plant Technical Specifications, which invoke ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components." ASME Section XI Rules and 10CFRS0.55a specify that a volumetric inspection of pump casing welds and a visual inspection of pump casing internal surfaces be performed on one reactor coolant pump of a group within each ten-year interval. Such inspections require access to the inside surfaces of the pump casing and, as such, are extremely difficult to perform once the pump is welded to the primary piping and is operated in a radioactive environment. Radiographic inspection requires access to the inside surfaces of the pump casing; ultrasonic examination is not acceptable because of the unfavorable acoustic characteristics associated with the grain structure of the cast stainless steel casing material. 13

Because of these difficulties, and the inherent high degree of toughness of cast austenitic stainless steel, the ASME has approved Code Case N-481 which provides alternate examination procedures specifically for cast austenitic pump casings. In lieu of performing a volumetric examination of the casing welds, the Code Case specifies visual examinations of the external surfaces, a visual examination of internal surfaces when a pump is disassembled for maintenance, and an analytical evaluation of the pump casing integrity with an assumed flaw extending 25% (l/4t} into a highly stressed region of the casing pressure boundary. This report documents the analytical evaluations performed to demonstrate RCP casing integrity for RCPs operated by the five participating utilities. The evaluations include analyses requirements of ASME Code Case N-481 in that they include the analysis of casings having postulated l/4t cracks and they evaluate thermal embrittlement. The evaluations are based on the use of original stress analyses reports to locate regions of high stress concentrations. Conservative crack growth rates are calculated for locations of high stress to determine the time required (in years of service) for an assumed initial crack to propagate into the casing wall as a result of repeated applications of stress. Crack growth curves (crack size vs time) are shown for the most limiting stress locations in each casing. Also shown for each curve is the applied stress intensity factor associated with the crack size as the crack grows. Three failure modes are investigated. Estimated values of RCP casing material toughness, based on casing material certifications, provide a basis for calculating the end-point crack size limits for the two failure modes related to thermal embrittlement: non-ductile propagation and ductile tearing. The third criterion for establishing end-point crack size is based on the flow stress of the material. To the extent that the end-point crack depth determined by the above criteria is greater 14

than l/4t (one-quarter casing wall thickness), the Code Case requirement for the postulated crack to be evaluated is satisfied. In all of the cases analyzed, the end point crack depth is limited by the flow stress, not the fracture toughness criteria. In addition, in all cases the limiting end-point crack size is significantly greater than l/4t. The time required to reach the limiting* end-point crack size is 46 years or longer. Plant specific results are as follows: Limiting Minimum End-point Time to Reach Plant .

• crack Size . Crack Size End-point . **

Palisades 0.36t 46 years Fort Calhoun 0.32t 165 years Calvert Cliffs 1&2 0. 38t 130 years San Onofre 2&3 0. 43t 77 years st. Lucie 1&2 0.38t 130 years

• t = thickness of pump case wall at stress point analyzed

      **   Based on conservative use of design values of annual stress cycles.

Revision 2 of CEN-412 is identical to Revision 1 except for added or modified material on Pages 88, 90, 100, and 101 of the generic portion and Pages so of Appendicies C and E. These modifications reflect a removal of excessive conservatism from the stress values extracted from the original stress reports for the discharge nozzle at Calvert Cliffs 1&2 and St. Lucie 1&2 under faulted conditions. A table and further discussion have been added to show all of the ~ end poinnt crack sizes and to further support the conclusion that the l/4t stability criterion of ASME Code Case N-481 is met. 15

2.0 INTRODUCTION

2.1 Purpose ASME Section XI inspection requirements specify visual examination of the inside surface of reactor coolant pump casings and volumetric examination of the casing welds of one of the RCPs during each 10-year inservice inspection interval. ASME Code Case N-481, approved by the ASME in March 1990, provides a method by which Section XI requirements can be relaxed for cast austenitic stainless steel RCP casings. The purpose of this report is to present justification for relaxing the Section XI inservice inspection requirements and extending inspection intervals for the reactor coolant pumps operated at the Palisades, Fort Calhoun, Calvert Cliffs 1&2, San Onofre 2&3, and St. Lucie 1&2 plants. 2.2 Scope and Applicability

                                                                .

The scope of this report is limited to the reactor coolant pumps now in service at eight nuclear power units: Fort Calhoun, Palisades, Calvert Cliffs 1&2, San Onofre 2&3, and st. Lucie 1&2. The analys~s are limited to the casing materials and stress values applicable to these 32 pumps (4 per unit). Because the methodology is generic, similar results are possible for similar pumps at other plants but plant-specific calculations need to be performed in each case to incorporate plant-specific materials, stress analyses and design data. 16

3.0 BACKGROUND

3.1 Goals The goals of this report are: (1) - To describe the RCP casings being evaluated. (2) - To describe the methodology used to justify relaxed inspection intervals *

        . (3) - To apply the methodology to plant-specific RCP casings.

(4) - To present the results obtained in applying the methodology to all of the pump casings under consideration. 3.2 RCP Casing Descriptions 3.2.1 General The thirty-two reactor coolant pumps {RCPs) being evaluated {4 per unit) circulate coolant through the reactor coolant systems so that heat can be transferred from the reactor cores to the steam generators during power operation. Each unit is designed for operation with a 2-loop, 2-pumps per loop configuration. The pumps are all Type E {as defined in ASME Code Sect III, NB3400), vertical, constant speed, single stage,* diffuser-type centrifugal pumps with bottom suction and tangential horizontal discharge. All of the pumps under consideration were manufactured by the Byron Jackson Pump Division of the Borg-Warner Corporation (now known as BW/IP International, Inc.). A typical pump vertical cross-section is shown in Figure 3.2-1. 17

3.2.2 casing Design A typical casing vertical cross-section is shown in Figure 3.2-2. The circular casings were formed from castings of austenitic stainless steel in a heavy walled, symmetrical configuration which allows all support loads to be taken through the casing wall. Each casing consists of a volute section and a hub with integrally cast diffuser vanes. A horizontal cross-section of this component is shown in Figure 3.2-3. This arrangement of components performs two functions: the suction nozzle, volute and discharge nozzle guide the fluid from the suction pipe through the impeller to the discharge pipe; the diffuser vanes and volute convert the velocity energy in the fluid imparted by the impeller into pressure energy to recirculate the primary cooling through the reactor core and the steam generators. The RCP casing.is designed to withstand the full range of structural loads and those cyclic loads associated with temperature and pressure changes in the reactor coolant system. The cyclic loads are introduced by normal plant transients, reactor trips, and startup and shutdown operations. The numbers of these event cycles which were taken into consideration in the design for each of the plants are listed in the plant-specific appendices to this report. 3.2.3 Fabrication Fabrication considerations required that the volute and hub sections be cast independently and subsequently 18

welded together. This work was done for the various 4i1 plants as indicated in Table 3.2-1. Figures 3.2-4 and 3.2-5 show typical hub and volute sections after casting, and Figure 3.2-6 shows a welded casing assembly. The fabricating process required that the castings be inspected for voids, cracks or any other defects resulting from the process. If indications were found that exceeded the established acceptable limits, they were repaired. The hub and volute sections were welded together and inspected in accordance with ASME Section III requirements. Following the welding and inspection, the casing assemblies were annealed and machined. Materials certification records, inspection results, and casing stress analyses were documented in accordance with quality assurance requirements. 3.2.4 Inservice Inspections Plant Technical Specifications define the requirements for inservice inspection of RCP casings. These inspections are performed in accordance with ASME Section XI requirements, as specified in 10 CFR 50.55a. Any relief from these requirements must be provided in writing from the Nuclear Regulatory Commission. 3.2.5 Codes and Standards In 1968, a Draft ASME Code for Pump6-_ and Valves for Nuclear Power was issued for trial use and comment, and to provide specific reactor coolant pump requirements in ASME Section III. These specific requirements for pumps and valves in nuclear power plants were added to Section III of the ASME Boiler and Pressure Vessel Code in the 1971 edition and Section III was renamed "Nuclear Power Plant Components". 19

Table 3.2-1 Reactor Co9lant Pump Fabrication History Year of Delivery Plant Casing Configuration Foundry From Foundry Palisades 2 pieces GE Foundry 1967 (hub and volute) Division Schenectady, NY Fort Calhoun 5 pieces (hub Armco, National 1968 and 4-piece Supply Division volute) Torrence, CA Calvert 2 pieces ESCO 1971 Cliffs 1&2 (hub and volute) Portland, OR San Onofre 2 pieces ESCO 1976 2&3 (hub and volute) Portland, OR and 1977 N 0 st. Lucie 1&2 2 pieces ESCO 1971 (hub and volute) Portland, OR and 1977

Coupling Spacer i-:t---1~~~--_.~_,Hydrostati c Bearing FIGURE 3.2-1 TYPICAL REACTOR COOLANT PUMP - CROSS-SECTIONAL VIEW 21

HUB

                           '

I FIGURE 3.2-2 TYPICAL PUMP CASING - VERTICAL CROSS-SECTIONAL VIEW 22

CROTCH VANE 18

• FIGURE 3.2-3 TYPICAL PUMP CASING - HORIZONTAL CROSS SECTIONAL VIEW 23

FIGURE 3.2-4 TYPICAL RCP CASING HUB SECTION BEFORE WELDING TO VOLUTE 24

                                                       .,.
                                                         ~ *-;,
                                                          . :-
                                                            ~!
                                                                .e FIGURE 3.2-5 TYPICAL RCP CASING VOLUTE SECTION BEFORE WELDING TO HUB 25

l!t-~llU.~C..:t~@~

                                                                ~..

\ ..

    * :O::.;...
                ~~:;.rj~       .. -: 1** -
                . =-::*-* ",\, *;.-<!..;.:-

c~

                ~:..~-

FIGURE 3.2-6 WELDED PUMP CASING ASSEMBLY BEFORE MACHINING 26

• In 1971, the ASME Boiler and Pressure Vessel Code Committee added Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components."

ASME Section XI specifies the inservice inspection requirements for nuclear components, including RCPs. The Byron Jackson Type DFSS (Diffuser Single suction) pump is classified as a Type E pump in section III of the latest version. of the ASME Code. The purpose of the ASME classification is to provide both a pictorial and verbal descr~ption of the basic pump body geometries. A Type E pump is one having a volute radially split casing with multi-vane diffusers that provide a structural function to the casing. 3.3 RCP Casing Material Degradation Mechanisms 3.3.1 Fatigue and crack Growth Rates Fatigue is an ~ge-related degradation mechanism in which a material is subjected to a large number of repeated load cycles. The completion of a load cycle may be a return to a lower, or zero, stress or a reversal from a peak tensile to a compressive stress. When stresses are sufficiently large, an incremental amount of local plasticity can occur at grain boundaries with each cycle, which produces fatigue damage. such local damage is associated with peak stresses which would be in excess of material yield stress. Initially, material ductility can accommodate a large number of cycles without failure~ Eventually, however, as dislocations are immobilized within the 27

material, its ductility diminishes locally and non-ductile failure may occur. The maximum acceptable number of cyclic loadings, which depends on changes in stress intensity, can be determined through use of a cyclic fatigue curve as published in the ASME Section III Appendix 1, using methods defined in Subsection NB-3000. Compliance with this Subsection ensures that primary plus secondary plus peak stresses at a highly stressed region are within acceptable limits for the design number of load cycles applied to the structure, or, alternatively, that the number of load cycles is sufficiently low to accommodate the local change in stress intensity. A fatigue analysis considers all loading conditions which are categorized as Service Level A (Normal Operation) or Service Level B (Upset Conditions) as specified by the designer. Loadings associated with Service Levels c (Emergency) or D (Faulted) are exempt from inclusion in a fatigue analysis due to their infrequency. By definition, any event which is expected to occur 25 or more times must be classified at or below Service Level B. As stipulated in Paragraph NB-3222.4(d) a fatigue analysis is not required by Subsection NB-3000 under certain conditions. The RCP casing stress reports prepared initially by the Byron Jackson Company for each of the five plants being evaluated justified application of this exemption by demonstrating compliance with the conditions. The data used in this demonstration contain the number of atmospheric-to-operating pressure cycles for which the pumps were designed by the Byron Jackson Company. This number was 28

shown to be lower than the allowable number of cycles from the fatigue curve thereby justifying exemption from the fatigue analysis. The crack growth analysis of this report was performed to snow the annual growth rates of hypothetical cracks in highly stressed regions based on the number of stress cycles originally used in the design. Such cycles are assumed to occur at a uniform frequency, referred to in this report as the design rate, during the forty-year license period of each plant. 3.3.2 Thermal Embrittlement Thermal embrittlement refers to the time-dependent effect of exposure to elevated temperatures on the ductility and toughness properties of a material. Austenitic stainless steels are subject to thermal embrittlement due to a variety of metallurgical reactions which are dependent upon the composition, microstructure, and time and temperature. Combinations of these variables can result in thermal embrittlement of both austenitic and martensitic wrought stainless steels, cast stainless steels, and stainless steel weld metals. A comprehensive review and evaluation of the effects of long-term elevated temperature exposure on these types of materials has been performed by Yukawa (Ref. 3~1). The following discussion will focus on the mechanisms of thermal embrittlement as they relate to the cast stainless steel and weld metal of the RCP casings. Cast austenitic stainless steels and austenitic stainless steel weld metal are actually dual phase materials. The microstructures exhibit a duplex 29

structure consisting of an austenitic matrix with a distribution of a second phase called ferrite. The volume fraction and morphology of the ferrite phase is controlled by the chemical composition of the material and the casting or welding process used to produce the material. Typically the ferrite distribution in large castings or heavy weldments is not uniform due to variations in cooling rates and segregation of alloying elements. For these reasons, a nominal ferrite content is either calculated from the chemical composition or measured indirectly by detecting the relative magnetic response of the material. The presence of ferrite is beneficial for improving strength, castability and weldability, and resistance to stress corrosion cracking. However, the ferrite phase in stainless steel castings and weld metal is responsible for the susceptibility of these materials to thermal embrittlement. Various carbide phases, intermetallic compounds such as sigma and chi phases, and the chromium-rich bee phase alpha-prime (a!)_ can precipitate in the ferrite during service at elevated temperatures and* lead to substantial degradation in mechanical properties (Ref. 3-2}. - Figure 3.3-1 shows the time-temperature curves for formation of vari~us phases and the change in impact strength of thermally aged cast stainless steel. At the operating temperature of light-water reactors, 280° to 320°C (535° to 610°F), thermal embrittlement is primarily caused by a' precipitation. The embrittlement of cast stainless steel results in a brittle fracture associated with either the cleavage of the ferrite phase or separation of the ferrite/austenite phase boundary. The austenitic 30

matrix retains ductility and toughness. The degree of embrittlement, and, hence, the toughness of the material is controlled by the amount of brittle fracture. The extent of thermal embrittlement of a given cast stainless steel is, therefore, highly dependent on the amount, size, and distribution of the ferrite phase. Other metallurgical reactions occurring in this temperature range also contribute to the overall extent of thermal embrittlement. The presence of phase boundary carbides at the austenite/ferrite interfaces is another important parameter in controlling the fracture mode (Ref. 3-2) and the resultant loss of toughness. The higher carbon grades of cast stainless steels have a higher charpy impact transition temperature due to the presence of phase boundary carbides. The carbides weaken the austenite/ferrite boundaries and lead to phase boundary separation, an additional low toughness fracture mode which acts in conjunction with the cleavag*e of the embrittled ferrite phase. One other factor that influences the low temperature thermal aging response is the precipitation of an additional second-phase particle in the ferrite called the G phase. The kinetics of the precipitation of this phase are dependent upon the chemical composition of the cast material. In general, precipitation of G phase is faster in the molybdenum-containing CFSM steels (i.e. Type 316) (Refs. 3-2, 3 ... 3). The CFSM castings exhibit larger decreases in toughness than CFS grades with similar ferrite contents and thermal exposure histories. 31

Austenitic stainless steel weld metal and castings are similar in that both contain ferrite which can transform by the metallurgical reactions described above. This feature results in a general similarity in the initial kinetics of the decrease in toughness with exposure time. Yukawa (Ref. 3-1) observed that for longer thermal exposures the weld metals tend to stabilize at a higher retained value of toughness while castings continue to decrease to very low values before saturating. The long term result is a greater degradation in toughness for cast material even though short term aging data indicate similar kinetics. Based on this observation by Yukawa (Ref. 3-1), the prediction of thermal embrittlement for the RCP casings will assume that the castings and weld metals behave similarly. This assumption is conservative since the actual embrittlement of the weld metal may be less than predicted. The method for predicting the fracture toughness of thermally embrittled cast stainless steels is discussed in Section 5.2 of this report. Thermal aging of cast stainless steels results in increases in hardness and tensile strength and decreases in ductility, Charpy impact toughness, and fracture toughness of the material. However, it has been observed that the low-cycle fatigue behavior and fatigue crack propagation rates are not significantly altered by thermal aging (Refs. 3-4, 3-5). This factor was considered in selecting the fatigue crack growth rate relationship for the crack growth analysis portion of this report. 32

SECTION 3 REFERENCES 3-1 s. Yukawa, "Review and Evaluation of the Toughness of Austenitic Steels and Nickel Alloys After Long-Term Elevated Temperature Exposures," Prepared for Pressure Vessel Research Committee, Aug. 1990. (DRAFT) 3-2 o. K. Chopra and A. Sather, "Initial Assessment of the Mechanisms and Significance of Low-Temperature Embrittlement of Cast Stainless Steels in LWR Systems," NUREG/CR-5385 (ANL-89/17), U. s. Nuclear Regulatory Commission, Washington, D.C., August 1990. 3-3 M. Vrinat, R. Cozar and Y. Meyzaud, "Precipitated Phases in the Ferrite of Aged Cast Duplex Stainless steel," Sci. Metall., 20, 1101 (1986). 3-4 E. I Landerman and W. H. Bamford, "Fracture Toughness and Fatigue Characteristics of Centrifugally Cast Type 316 Stainless Steel Pipe After Simulated Thermal Service Conditions," Ductility and Toughness Considerations in Elevated Temperature Service, ASME MPC-8, New York, 1978, pp. 99-125. 3-5 G. Slama, P. Petrequin and T. Mager, "Effect of Aging on Mechanical Properties of Austenitic Stainless Steel Castings and Welds," SMIRT Post-Conference Seminar 6, Assuring Structural Integrity of Steel Reactor Pressure Boundary Components, August 29-30, 1983, Monterey, CA. 33

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w I- 4'00 300 200 0.1 1 *10 100 1000 TIME (hours) FIGURE J.J-1 TIME-TEMPERATURE CURVE FOR (TOP} FORMATION OF VARIOUS PHASES AND (BOTTOM} DECREASE IN ROOM TEMPERATURE IMPACT ENERGY IN CAST STAINLESS STEEL (REF-3.2} 34

4.0 CURRENT INSPECTION REQUIREMENTS FOR RCP CASINGS 4.1 NRC Requirements NRC requirements for RCP inservice inspections are established by 10 CFR 50.55a(g) and are referenced in plant Technical Specifications and plant FSARs. For example, Section 5.4.1 of the Calvert Cliffs Unit 2 Technical Specifications requires that the reactor coolant system be maintained in accordance with the code requirements specified in Section 4 of the FSAR. Section 4.1.5.6 of the FSAR, "In-service Inspection," refers to inservice inspections being made in accordance with the ASME Boiler and Pressure Vessel Code, Section XI. Title 10 CFR 50.55a(g) (6) (i) however, allows the Commission to grant relief from the ASME Code if the licensee has determined that conformance with certain Code requirements is impractical for his facility. 4.2 ASME Requirements ASME Section XI, Subsection IWB-2412 establishes 10 years as a permissible inservice inspection interval. Article IWB-2500 states that components shall be examined and tested as specified in Table IWB-2500-1, which requires the following for examination category B-L-1 (Pump Casing Welds) and B-L-2 (Pump Casing): B-L-1 Volumetric examination of all welds B-L-2 Visual (VT-3) examination of all internal surf aces to the extent practicable but only when the pump is disassembled for maintenance, repair, or volumetric examination. 35

4.3 Applicability of ASME Code Case N-481 Alternative Code case N-481, approved by the ASME on March 5, 1990, provides "Alternative Examination Requirements for Cast Austenitic Pump Casings - Section XI, Division 1. 11 In lieu of requiring the volumetric examination specified in Table IWB-2500-1, Examination Category B-L-1, Item B12.10, the Code Case requires: (a) VT-2 visual examination of pump exterior during the hydrostatic pressure test required by Table IWB-2500-1, Category B-P. (b) VT-1 visual examination of the external surf aces of the weld of one pump casing. (c) VT-3 visual examination of the internal surf aces whenever a pump is disassembled for maintenance. (d) Perform an evaluation to demonstrate the safety and serviceability of the pump casing. The evaluation shall include the following: (1) evaluating material properties, including fracture toughness values; (2) performing a stress analysis of the pump casing; (3) reviewing the operating history of the pump; (4) selecting locations for postulating cracks; (5) postulating 1/4 thickness reference crack with a length six times its depth; (6) establishing the stability of the selected crack under the governing stress conditions; (7) considering thermal aging embrittlement and any other process that may degrade the properties of the pump casing during service. (e) A report of this evaluation shall be submitted to the regulatory and enforcement .authorities having jurisdiction at the plant site for review. 36

As of the time of issuance of Revision 8 of NRC Regulatory ~ Guide 1.14 7, "Inservice Inspection Code Case Acceptability" (November, 1990), Code Case N-481 was not yet on the approved l.ist. Because the numerical listing of Code Cases in Revision 8 lists Code Case N-472 as the most recently approved, it is apparent that N-481 has not yet completed being reviewed and approved by the NRC. 4.4 ASME Code Case Reference Flaw The one-quarter thickness reference flaw with a length six times its depth, as postulated in Code Case N-481, was selected to be ten times an estimate of the largest flaw that could be missed during pre-service inspections (Reference 4-1). Using the assumptions of the Code case, the following evaluation can be made of the largest flaw that could be missed: Initial Flaw Size = Reference Flaw 10

            = Reference crack de:Eth x Reference crack length 10
                  = ll4t x 6l4t 10
                  =     t x    6t 4./10   4./10
                  = .oat x .48t As given in the above derivation, the reference flaw consists of a l/4t crack depth with an aspect ratio of 1/6.

This reference flaw is ten times the size of an initial flaw having a crack depth of 8%t. 37

An initial crack depth of 8%t would have been easily detectable by pre-service inspection, as it is four times the depth of the 2% crack required by the ASME Code to be detectable by radiographic detection techniques. 4.5 Postulated Initial Crack Depth Based on the foregoing derivation, an initial crack depth of 8%t was conservatively postulated for all of the crack growth curves developed in Section 5 of this report. SECTION 4 REFERENCES 4-1 ABB-CE internal memorandum, o. F. Hedden to A. G. Schoenbrunn, dated 9/20/91, commenting on work by the PVRC Ad Hoc Group on Toughness Requirements. This group prepared Welding Research Council Bulletin 175, August 1972, "PVRC Recommendations for Ferritic Materials", in which the l/4t reference flaw criterion originally appeared. 38

5.0 METHODOLOGY FOR EVALUATING PUMP CASINGS 5.1 Crack Growth Analysis In this section, the methodology for selecting the most critical locations for hypothetical crack growth, and estimating the growth histories of such cracks, starting from a conservative initial condition, are discussed. 5.1.1 Selection of Locations for Most Conservative Analysis The object of the selection process is to locate those high-stress regions within the pump casing at which a hypothetical crack would propagate most quickly. The growth rate in the critical region serves to "envelope" (i.e. conservatively bound) the growth rate of any other hypothetical crack elsewhere in the pump casing. tt Crack histories are initialized and developed on a dimensionless (i.e. percent-of-thickness) basis. The selection of limiting locations is, in general, consistent with those highly-stressed regions which had been previously identified in the original RCP casing Stress Reports authored by the Byron-Jackson Company. The original RCP casing Stress Reports provided the appropriate Design Certification in accordance with criteria set forth in ASME, Division 1, Section III. The quantification of stresses at the limiting locations, as provided in the Byron-Jackson Company stress reports, verifies that the highly-stressed regions satisfy ASME Section III requirements for Design Conditions. These stresses are summarized in the individual stress reports. Design condition loads ~ 39

consist of the summation of stresses due to: design pressure (2500 psia), forces and moments on the pump casing in normal operation, closure bolt preload effects, and Operating Basis Earthquake (O.B.E.) loads. As discussed below, Design Condition loads are combined with transient thermal stresses from plant heatup or cooldown (whichever is worse) to define the peak of the stress cycles on which the crack growths are based. In general, the following regions in the RCP casings were either analyzed or shown to be enveloped by a worse region and, thus, exempted: (1) suction nozzle, (2) volute ("scroll")/lower flange junction, (3) volute, (4) volute/hanger bracket vicinity, (5)- volute/upper flange junction, (6) diffuser vane, (7) crotch region, and (8) discharge nozzle. The membrane (section-average) and bending (through-wall variation) components of local stress in the*design condition were then established for each potentially limiting location. 5.1. 2 Initial Flaw Size As discussed in Section 4.5, the growth histories of hypothetical cracks were calculated using an assumed initial depth of 8 percent of the section thickness for all cases. Since any detectable cracks in the pump casings would have been required to be repaired as part of preoperational inspection, and the radiographic standard for detection sensitivity was 2 p~rcent, the assumed initial depth reflects a conservative estimate of the largest undetectable crack. The initial crack length was postulated to be 48 percent of section thickness based on an aspect ratio (depth:length) of 1:6. This initial crack size postulation is supported 40

by Bulletin WRC-175 of August, 1972. which states that, ~ with the combination of examination requirements of ASME Sections III and XI (radiography & ultrasonic mapping, respectively), the probability of a crack four times greater than the 2 percent radiography standard escaping detection is evaluated as being "very low". Accordingly, an 8% crack is judged to be a conservative starting point for this analysis, which predicts the growth histories of any hypothetical crack that would have been initially undetectable. In accordance with the recommendation of ASME Section III, Article A-5000, the growth histories of the hypothetical cracks in this analysis are based on the assumption of self-similar enlargement, in which aspect ratios are assumed to remain constant at {1:6). 5 .1. 3 Plant Operating History The growth histories of hypothetical cracks in the RCP casings were calculated based on a conservative estimate of the significant stress cycles which the pumps would be expected to experience. Engineering judgment was used to combine knowledge of the pump duty cycle; as given in the Design Specifications, with the mathematical expression of the incremental crack growth correlation {discussed below). This would (1) account for those stress cycles which would produce meaningful crack growth, and {2) eliminate those stress cycles which would produce negligible crack growth. As discussed below, the incremental crack growth correlation used for this analysis is based on changes in the applied stress intensity factor, KI' {from fracture mechanics) raised to the fourth power. Thus, minor changes in the primary coolant pressure or 41

temperature would produce negligible crack growth. Because of this fourth-power effect, the only stress cycles which would meaningfully contribute to crack growth histories are those in which the primary coolant pressure changed from atmospheric to operating and back to atmospheric during heat up and cool down. As indicated in Section 3.3.1 the expected number of such cycles was given in the Byron-Jackson company stress reports in the process of justifying exemptions from fatigue analyses. Thus, conservative estimates of the rate at~which atmospheric -to-operating pressure cycles occur were used to estimate growth histories. 5.1.4 Calculation of Crack Growth Rates The objective of this portion of the analysis was to combine previous stress analyses with fracture mechanics methodology and an empirical crack-growth relationship, and thereby to predict the growth history (depth versus time) of the fastest-growing hypothetical crack(s) in the RCP casings. A selection of limiting locations was made for each RCP casing, based on design condition stresses as published in the respective Byron-Jackson Company stress reports. At each such region, the through-wall stress variation was obtained. This provided the membrane (section-average) and bending (surface minus membrane) components of linearized stress. Transient thermal bending stress (based on conservative assumptions of through-wall temperature gradients, if not explicitly stated) were incorporated into the bending stress component. Wall thicknesses were obtained from the stress reports and from available as-built drawings. 42

It can be mathematically shown that, for a given ~ membrane and bending stress, a crack which is initially at a given percent of thickness would subsequently propagate more quickly into a thicker section. Specifically, the time needed for a crack to transit a given percent-of-thickness is inversely proportional to section thickness. For example, the time needed for a crack to grow from 8 % to 30 % of depth in a 3 11 thick section would be only two-thirds as long as the time needed to go from 8% to 30% in a 2 11 section. This is a direct consequence of; (1) the square-root-of-length dependency in the applied stress intensity factor (KI) and (2) the fourth-power dependency of crack growth on changes in applied stress intensity factor. Incremental distances are proportional to thickness to the first power; propagation velocities are proportional to thickness squared. Hence, estimates of section thicknesses, where necessary, were conservatively biased towards larger values. Thicker-than-average sections were not analyzed if the published stress reports indicated lower-than-average stresses there. Also, membrane stress has a more significant effect than bending stress when applied stress intensity factor is calculated. Hence, if peak surface stress is known, any uncertainty about membrane and bending stress can be conservatively treated by assigning a somewhat larger value to membrane stress while correspondingly adjusting bending stress downward. Based on the methodology given in ASME Section XI, Article A-3000, the applied stress .intensity factor

    '

applicable to a fracture mechanics evaluation was calculated using equation A-3300 (1): 43

(Eqn. 5-1) Here am and ab are the membrane and bending components of stress. Q is the "flaw shape parameter" from Figure A-3300-1. Mm and Mb are the "membrane stress correction factor" and "bending stress correction factor" for surface cracks from Figures A-3300-3 and A-3300-5, respectively. The determination of Q, Mm, and Mb was based on a surface crack with an aspect ratio a/L of 1:6. The value of Q obtained from Figure A-3300-1 was conservatively based on a total stress (membrane plus bending) equal to yield stress. The linear depth of the crack, "a", is the only independent variable in the calculation of KI from equation A-3300 (1) using the above methodology and assumptions. Therefore, the functional dependence of "K" upon "a" was determined, with "a" being expressed as a fraction of "t", the applicable wall thickness~* By algebraic manipulation, equation A-3300 (1) is made into a function of dimensionless crack depth, a/t, as follows: j!. (Eqn. 5-2) t Mm and Mb' are both functions of a/t in their applicable figures (since surface crack aspect ratio, a/L, was fixed at 1/6). Q, which depends only on a/L, is a constant. The dependence of these three coefficients on a/t is given in Table 5.1-1. 44

Table 5.1-1 COEFFICIENTS Mm' Mb & Q for CALCULATING KI with a/L = 1/6 = constant KI = (am Mm + aB MB) ;7rt /~ Q t M  !:!b ~ -m _Q_ 0.1 1.10 0.95 1.05 0.15 1.12 0.89 1. 05 0.20 1.15 0.94 " 0.25 1.185 0.80 " 0.30 1. 225 0.76 " 0.35 1.27 0.73 " 0.40 1.325 0.70 " 0.45 1.395 0.685 " 0.50 1.48 0.675 1.05 from from from Fig. A-3300-3* A-3300-5* A-3300-1* am + uB

                                              = 1.0 **

O' ys

• From ASME Section XI Article A-3000
  • Conservative Value 45

The functional dependence of KI on "a" is readily converted to its dependence on "a/t". The end result is that the applied KI was tabulated versus a/t for each pump case region being evaluated, from the initial depth of the hypothesized crack (a/t = 0.08) through an arbitrary end-point (a/t = 0.50) sufficient to complete the crack growth history. Per-cycle crack growth was then calculated using an empirical relationship for cast stainless steel in water, hereafter referred to as the Bernard & Slama equation (Reference 5-9): in. (Eqn. 5-3) cycle Here the quantity "R" refers to the ratio of applied stress intensity factors, KI (minimum)/ KI (maximum) through the cycle. A simplifying and conservative* assumption of KI (minimum) = o was applied, rendering R = o. K1 {maximum) is associated with the Applied Stress Intensity Factor from the Design Condition {plus heatup/cooldown thermal gradients), and hence, ~K =KI (maximum). The Bernard & Slama equation then simplifies to: in (Eqn. 5-4) cycle Multiplying the per-cycle crack growth, above, by the design number of cycles per year (a figure which varies from 12.55 to 17.88, depending on the plant) then yields the annual crack growth rate as_ a function of KI (maximum). This is an application of the "Chain Rule"; in this case: da/dT = (da/dN) (dN/dT) {Eqn. 5-5) 46

where "T" is time in years, "N" is the number of atmospheric-to-operating pressure cycles, and dN/dT is e the design annual rate of occurrence. KI (maximum) i.s a function of either crack depth, "a" , or of dimensionless crack depth "a/t". Consequently da/dT becomes a function of a/t, i.e. the rate-of-deepening (inches/year) of a hypothetical crack can be expressed as a function of the dimensionless crack depth itself. This constitutes a non-linear differential equation which has no closed form solution but can be treated numerically. Substituting for the expression KI' the general form of the differential equation is: d(a/t) .

       = constant x f {u, a/t} x {a/t) 2     (Eqn. 5-6) dT where f {u, a/t} indicates the inclusion in KI of membrane and bending stress correction factors Mm and Mb.

Using this differential relationship and beginning at a/t = o.os, the subsequent crack growth history can be determined by integrating the variable growth rate through the section thickness. The integration process yields accumulated Time in years {"T") as a function of a/t. To get this result, incremental transit times, ~T, are calculated using Time = Distance/Velocity where incremental distances correspond to a change in a/t {i.e~ a percent of thickness interval) and velocity, da/dT. A conservative integration procedure expedites the process considerably. Crack growth rate increases with~ increasing KI' which assumes greater values with deeper 47

cracks (larger a/t values). The crack growth rate accelerates as the crack deepens. The integration procedure conservatively calculates incremental transit times by using the end-point rate (i.e., the fastest rate) of each depth interval. In order to conservatively calculate the time it would take to propagate from a/t = 0.08 to 0.10, the fastest growth rate, at a/t=0.10, would be used for the entire interval. This procedure yields the shortest incremental transit time for each interval of crack depth, and the most rapid growth histories. The final step in calculating the predicted growth history of a hypothetical crack is a summation of the incremental transit times, from the initial condition of a/t = 0.08 at Time= O. 5.1.5 Plant Specific Results The predicted growth histories of hypothetical cracks in the limiting locations are shown in Figures ~.1-1 through 5.1-5 for Palisades, Fort Calhoun, Calvert Cliffs 1&2, San Onofre 2&3, and St. Lucie 1&2 respectively. Two or more additional growth curves are plotted for each plant to show the sequentially less-limiting growth histories. The choice of locations plotted reflects the most highly-stressed regions originally identified in the Byron Jackson Company stress analyses of the pump casings. The stress components and local thicknesses for the plotted regions are listed in the plant-specific appendices to this report. Figures 5.1-1 to 5.1-5 indicate the following: 48

(1) cyclic crack growth rates based on the Bernard & Slama equation, and constant local stresses, increase with both time and crack depth. The time for a crack to grow from a depth of 8% to a depth of 15 % of the wall thickness is roughly equal to the time to grow from 15% depth to 50%. This is due to the square root-of-crack length term in the definition of KI' the applied stress intensity factor. Secondarily, the increase in Membrane Stress Correction Factor, Mm' outweighs the decrease in Bending Stress Correction Factor, Mb, in the definition of KI. (2) In the cases of Calvert Cliffs 1 & 2, St. Lucie 1 & 2 and Fort Calhoun, the limiting region is a Diffuser Vane located near the discharge no~zle. In the cases of* San Onofre 2 & 3, and Palisades, the limiting region is the Crotch vicinity, which is also immediately adjacent to the discharge nozzle. (3) The next most critical regions are: hanger bracket

1. 1 vicinity (Calvert Cliffs and St. Lucie); Junction of Scroll to suction Nozzle Flange (Fort Calhoun and Palisades); and the last Diffuser Vane (San Onofre).

(4) The predicted times for a crack in the limiting region to grow from 8% to 25% of local thickness are: 40 years (Palisades); 59 years (San Onofre); 110 years (Calvert Cliffs and St. Lucie); 155 years (Fort Calhoun). (5) The result for Palisades is characterized by the fastest growth history. This is attributed to published surface stresses of 55 Ksi in the crotch region, which substantially exceeds the results for the other three plants. The design condition stresses for the Palisades plant included Safe Shutdown Earthquake

(SSE) effects ("design basis earthquake" in the original terminology) which exceed the Operating Basis Earthquake (OBE) stresses used in the other plants, and this is thought to have contributed to the faster growth history for Palisades. In particular, the magnitude of the SSE is twice that of the OBE. It is also believed that finite-element capabilities at the time of the Palisades stress analysis were still in a developmental stage (1969) and as such would have tended to overpredict local secondary stress effects, relative to the other finite-element analyses which were performed later. Nevertheless, the result for the Palisades plant is considered acceptable in terms of the intent of this report. The known conservatisms in the analysis methodologies reinforce this viewpoint. For example, the inclusion of secondary stress effects in the definition of design stresses on which the predicted crack growths are based. (6) The 25%-of-depth cracks are found to be stable against Design Condition, Emergency Condition, and Faulted Condition loads. A substantial period of time would be required after a crack had reached 25 % of depth, before it would reach an end-point crack size (more than 5 additional years in the case of Palisades; 10 or more years for the other plants). (7) Design crack growth histories used a large and conservative rate of stress cycling based on the original Design Specifications. Actual stress cycling, at least to the present, occurs at a substantially lower frequency. 50

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CJ) Ro 0 N 0 20 40 60 80 100 120 140 160 180 200 220 240 Years (1) Vane 8 (4) Volute @ Junction with Lower Flange (2) Discharge Nozzle, Crotch Vicinity (5) Hanger Bracket #1 Vicinity (3) Suction Nozzle

• Bernard & Slama Growth Equation @ 700 cy/40 yr; a/L = 1/6 (3) (4) (1 )(5) (2)

                                                                                                                                          -
                    .50
                    .45                                                         I                           fj'                        I Ci) CJ>
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                    .05
 ~Ro                  0 CJ) (A) 0     10     20    30     40          50  60     70     80              90   100  110     120 130  140 150        160 Years (1) Suction Nozzle                          (4) Vane (2) Junction Lower Flange with Volute       (5) Volute in Vane 5 Vicinity (3) Crotch

Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1/6 (1) (5) (2)

                        .50
                        .45 j
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   ~I\)                   0 0     20       40      60       80       100    120        140   160       180      200  220     240 Years (1) Vane 8                             (4) Volute @ Junction with Lower Flange (2) Discharge Nozzle, Crotch Vicinity  (5) Hanger Bracket #1 Vicinity (3) Suction Nozzle 01 01

5.2

• Thermal Embrittlernent Analysis A procedure and'correlations have been developed for predicting the change in fracture toughness of cast stainless steel components due to thermal aging during service in LWR's at 280°-330°C (535°-625°F). The flow diagram for estimating fracture toughness J-R Curves of cast stainless steels exposed to LWR environments is shown in Figure 5.2-1 (Ref. 5-1). The method outlined in Section A defines "lower-bound" fracture toughness J-R curves for cast stainless steels of unknown chemical composition. Different "lower-bound" curves are defined depending on whether or not the ferrite content of the casting has been measured.

Section B (Figure 5.2-1) outlines the approach for estimating the fracture toughness of cast stainless steels when the Certified Material Test Report (CMTR) is available for the material. The CMTR includes the chemical composition of the casting, as well as the tensile properties and often includes the ferrite content for the material. The approach outlined in Section B describes the estimation of "saturation" J-R Curves, i.e. the lowest toughness that is anticipated for the material after long-term service. The lower bound J-R curve for the unaged cast stainless steels is applied when the J-R curve estimated for the aged material results in a higher predicted toughness than is defined by the lower-bound curve for the unaged material. The approach outlined in Section B of Figure 5.2-1 is the method that has been employed in this . evaluation of the RCP casings. This method has been applied to both the castings and weld metal used in fabricating the pump casings. As noted in Section 3.3.2, applying this same method to the weld metal is a conservative assumption, since the empirical data suggest that stainless steel weld metal may saturate at higher levels o.f toughness than is predicted for castings (Ref. 3-1). 56

Section C of Figure 5.2-1 represents a further refinement the approach for predicting the aged fracture toughness properties of a casting. This method employs the actual service temperature and time history of the component to provide a better estimate of the toughness of a casting for some intermediate aged condition. Figure 5.2-2 shows a comparison of the results of these three methods. The method in Section A results in the J-R Curve labeled lower bound in Figure 5.2-2. The Section B approach defines the curve labeled "saturation" and Section C results in the curves for different times of exposure defined in Effective Full Power Years (EFPY) in Figure s.2-2. The curve labeled "unaged" in Figure 5.2-2 is the lower bound curve for toughness of unaged cast stainless steels. For the purposes of the current evaluation, the more conservative approach of using the saturation J-R curves described by Section B was utilized. The refinement of using actual service history (Section C) could have been used for any RCP casings where the estimated aged toughness was found to be limiting. such further evaluation was not found to be necessary at this time. The following discussion provides a detailed description of the methodology for predicting the saturated fracture toughness properties for the thermally aged RCP casing cast stainless steels and weld metals. 5.2.1 Material Heat Data The CMTR's for the casting or weld metal to be evaluated are used to obtain the chemical composition for the material. The weight percentages of the various alloy elements are used to calculate the ferrite content. There are several methods for 57

estimating the ferrite content of a casting based on chemical composition (Refs. 5-2, 5-3). Chopra in Reference 5-1 recommends calculating ferrite content in terms of Hull's equivalent factors, as follows: Creq = Cr + 1.21 (Mo) + 0.48 (Si) - 4.99 (Eqn. 5-7) Ni eq =.Ni + 0.11 (Mn) - 0.0086 (Mn) 2 + 18.4 (N) + 24.5 (C) + 2.77 (Eqn. 5-8) Since nitrogen (N) is not a specified alloying element in the material specification, it is not normally reported in the chemical analysis provided in the CMTR's. When no value is available for the nitrogen ., content, Chopra (Reference 5-1) recommends using a nominal value of 0.04% (See Figure 5.2-1). The ferrite content is determined from the chromium and nickel equivalents by the following equation: 2 %F = 100.3 (Creq/Nieq> - 170.72 (Creq/Nieq> (Eqn. 5-9) ,.- + 74.22 Determination of ferrite content for the ASTM A 351 castings, if required, would ordinarily be estimated by one of the methods in ASTM A 800 (Reference 5-2). The estimation of ferrite in castings by both the Hull's equations and the Schaefer diagram of ASTM A 800 were compared by Aubrey et al. (Reference 5-3). It was found that both methods predicted very similar ferrite contents (Figure 5.2-3) and the statistical correlation coefficients for both methods were essentially the same. Ferrite content of weld metal is normally estimated in terms of ferrite numbers from the Schaefler diagram (Reference 5-4). However, there is also good agreement between the measured ferrite 58

content of the weld metals and the values predicted by ~ Hull's equations. Therefore, the calculated ferrite values using Hull's equations were used for all of the materials to calculate the material parameter and to estimate the minimum Charpy impact toughness. s.2.2 Estimate of Minimum Charpy Impact Energy The "saturation" fracture toughness pr~perties of stainless steel castings are estimated from the degree of thermal embrittlement where further changes in toughness properties no longer occur with longer exposure time. The "saturation" toughness is therefore the minimum expected toughness value for a material. Since much of the experimental work on thermal aging of cast stainless steels utilized Charpy impact testing following aging, the degree of thermal embrittlement is characterized in terms of the "normalized" Charpy impact energy (i.e. the absorbed energy per unit area of the notched Charpy impact specimen). Two different correlations were developed for estimating the minimum Charpy impact energy, depending on the type of casting. One correlation is for the equivalent Type 304 grades, both low-carbon CF3 and sta~dard CFS. The second correlation is for molybdenum-containing Type CFSM castings, equivalent to Type 316 stainless steel. The reason for the.two correlations is because of the increased susceptibility to thermal embrittlement due to formation of G phase in molybdenum-containing steels. Since all of the pump casings are Type CFSM with Type 316 weld filler metals, the second correlation has been used in this evaluation. The material parameter for CFSM castings is defined by the equation: 59

  ~ = [%F x Cr x (C + 0.4N) x (Ni+Si) 2 ]/100          (Eqn. 5-10) e* When no reported value for nitrogen content is available, the same value of 0.04% nitrogen is assumed that was used to calculate the ferrite content. The saturation value of room-temperature Charpy impact energy is given by:

log 10 ~sat= 1.15 + 1.532 exp(-0.0467 ~) (Eqn. 5-11) 2 This equation gives a value for ~sat in joules/cm . Multiplying this value by a factor of 0.588 converts c to Charpy impact energy in foot/pounds. Vsat 5.2.3 Conversion of Charpy Impact Energies to J-Integrals The saturation fracture toughness J-R curves are estimated from the minimum room-temperature Charpy impact energies based on correlations of actual J-R curve test data at room-temperature and elevated-temperature with Charpy impact test data. For Type CFSM castings the J-R curve at room temperature is estimated by the equation: J d

              = {91  (25 4)n (C

• Vs at

                                      ) 0

• 67 }[~a]n (Eqn. 5-12) where the exponent n is defined as:

0 67 n = 0.35 .+ 0.0025(C.._

                              -vsa t] *                (Eqn. 5-13)

At elevated-temperatures in the range of 290°-320°C (550°-610°F),_ the J-R curve for thermally aged Type CFSM castings is estimated by the equation: 60

(Eqn. 5-14) e where the exponent n is defined as: n = 0.24 + 0.0063[CVs at ]0. 49 (Eqn. 5-15) For these equations, the room temperature impact energy 2 <=vsat is in (J/cm ) and Jd and ~a are expressed in (in-lb/in 2 ) and (in.), respectively. According to the method in Figure s.2-1, these J-R curves are then compared with the lower-bound unaged J-R curve for static-cast stainless steels given by: (Eqn. 5-16) This equation is applicable to static-cast stainless steels from room temperature up to 320°C (610°F). The tt purpose of this comparison is to check that the saturation J-R curve for thermally embrittled castings does not predict higher

                    .

toughness than the lower-bound

                           .

for unaged material. If the initial toughness properties are not known, which is usually the case since no impact tests are required for these castings, the lower-bound curve for unaged material is used when the J-R curve estimated from the chemical composition is higher. The relation between the saturation curve and the unaged lower bound is shown in Figure 5.2-2 for a typical Type CFSM casting. Once the power law equations for the J-R curves at room and elevated temperature are obtained, a value for the elastic-plastic fracture toughness, Jic' can be calculated. The Jic values are determined from the power law J-R curve equation according to the methods 61

of ASTM E 813 (Reference 5-5). Jic is defined as the intersection of the 0.2 mm offset line with the power law equation of the J-R Curve as shown in Figure 5.2-4. Note that ASTM E 813 uses a slope of two times the flow stress for the blunting, data exclusion and offset lines in Figure 5.2-4. The flow stress is defined as the average of the yield strength and tensile strength. For the evaluation of the elastic-plastic fracture toughness properties of stainless steels, it has been found that a slope of four times the flow stress for the blunting line and exclusion line provides a better representation of the material behavior (References 5-1, 5-6). Room temperature tensile properties are available for the RCP casing castings; however, there are no elevated temperature tensile data and there are no tensile properties for the weld metals. In order to calculate Jic values, flow stress values had to be estimated. Tabulated values of flow stress for unaged and aged cast stainless steels were plotted as shown in Figure 5.2-5 (Ref. 3-2). Approximate upper bound curves were fit to the aged flow stress as a function of the unaged room temperature flow stress (Figure 5.2-5). These relationships were used to estimate flow stress values for the castings at elevated temperatures based on the room temperature tensile properties. For weld metals, where no room temperature tensile data was available, a normal distribution was fit to the flow stress for the castings to establish a mean value. Flow stress for the weld metal was assumed to be equal to 78.6 ksi at room temperature and 58 ksi at 550°F, these values represent the mean value of the castings plus one standard deviation. This approach is conservative for estimating the Jic values, since the determination is 62

A Section B Input compositio

                                                                         .Cr,Mo,Si.Mn,Ni,&C N known Input l> 111 N c-0.04 UI  11111111111 Lower bound       Lower bound J-R curves         J-R curves Eqs. 2.1-2.4     Eqs.        ~2.12 Calculate '
                                                                                 .3.4 Calculate Eq.3.5 CVm Section C Service time oughness desired e known T8 (*C)        e                       Lower-Bound
                             <280  3.3                         Unaged Input* e     280-330        2.9                        J-R curve
                    ~             2.5                     Eqs. 3.16, 3.1 Calculate Q, P Eq. 4.1, 4.2 .

Input C Vint CVint

• 200 Calculate Calculate a,p,OI a,p,OI Eqs. 4.3-4.5 Eqs. 4.3-4.5 Yes WIU~I Lower-Bound UU Unaged J-R curve qs. 4.6-4.13 s. 3.16, 3.,17 I

                                                                                            . !

FIGURE 5.2-1 FLOW DIAGRAM FOR ESTIMATING FRACTURE TOUGHNESS J-R CURVES OF CAST STAINLESS STEEL IN LWR SYSTEMS (REF. 5-1) 63

Crack Extension ~a (In.) 0.0 0.1 0.2 0.3

        - aoo                                       - - - -
                                               ---

5000 C'll-efpy at 290°C c

        --

C'I E

         ""')
                                                             ------------ 16
                                                            --------- 32        4000

c

                                                                                      - I.
        ~                                                   ------ <48
        ""') 600                                                                       c c                                                                    3000   ..,

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        .52 ca 400 E                                                                           ;

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        -

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                                                                                      -

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                                                                                        ....

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                                                                                               ".

Crack Extension, da (mm) Crack Extension d8 (In.) 0.1 0.2 0.3

        --

C'll E 800 Heat L CF-SM 290-320°C

                                      ,, ,,,

efpy at 320"C

                                                             -*---*--** 16 5000 C'll-
                                                                                      ]5 c
        -" "')
         ~
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                                                                                       -

Q E 0 CD 0 0 2 4 6 8 10 Crack Extension, l1.a (mm) FIGURE 5.2-2 COMPARISON BETWEEN LOWER-BOUND J-R CURVE AND J-R CURVES AFTER 16, 32 AND 48 EFPY AT 290 AND 320°C FOR STATIC-CAST PLATE OF CF-SM STEEL (REF. 5-1) 64

It! !S FIGURE 5.2-3 COMPARISON OF FERRITE CONTENT PREDICTION SCHOEFER EQN. (A 800) VS. HULL'S FACTORS 65

• 48 RELIEF REQUEST NUMBER - RR-10 {cont'd)

PROPOSED ALTERNATE EXAMINATION Requirements* shall be met for both transverse and parallel flaws at the intersection of the welds and for that length of longitudinal weld within the

• circumferential weld examination volume. Therefore, surface examinations and volumetric examinations extending beyond the volume and length aforementioned

      *will not be performed for the third inspection interval .

• )

--* '

    - - - - -         .J-
                       --.,.- - - - - - - -                       -.,

• POINTS useo FOR ReGRESSfON ANAl.YSIS l

600 I ( 500 300 1.5 mm El$CUIStON UNE I 100 v~max I G.25 GJiO Cl.JS . 1.G 1.50 CRACK EXTENSION Cmml 1.75 ' curves to Define JIC FIGURE 5.2-4 66

a 2SC UNAGED

                                                                                                                                          +       25C AGED I     ! I J I I ! I ' I ' ~ I I I I ! *I '

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      ~.=
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I 1. 1 ' ' f' I', I I I ! Ii

• I! I ; ' I ' 'j ' ' ' I i' ' I ' I
• I ' , ; ' ** , t' ,t ' I I' , , , I , It t
• Ii , , 1 I' .. I ' I I t UNAGED R()(Jif nMPERA1'URE FLOW STRESS, HPa FIGURE 5.2-5 FLOW STRESS OF AGED AND UNAGED CAST STAINLESS STEEL AT 25C & 290C 67

not highly sensitive to the value of flow stress and the higher than average estimated flow stress (one standard deviation) will result in a slightly lower estimated Jic value. The other parameter to be determined from the J-R curve is the Tearing Modulus. Since the power law form of the J-R curve is non-linear, an average value of the Tearing Modulus must be estimated over the range of the power law between the two exclusion lines shown in Figure 5.2-4. The Tearing Modulus is defined as: (Eqn. 5-17) where E is the modulus of elasticity, af is the flow stress and dJ/da is the slope of the J-R curve. The least squares linear fit of the average J-R curve slope is performed by the method defined in Reference 5-6. 5.2.4 conversion of J-Integrals to Plane strain Fracture Toughness The elastic-plastic fracture toughness, Jic can be converted to an equivalent linear-elastic fracture toughness, Kic" When plane strain conditions predominate, this relationship between J and K is given

• by the equation:

(Eqn. 5-18) where E is the elastic modulus and v is Poisson's ratio. At the critical point of crack initiation, Jic is equivalent to Kic through this relationship. The equivalent Kic value will be referred to as KJc to denote that it* is a derived fracture toughness value from the J-R curve. 68

5.2.5 Plant Specific Results Quality Assurance Document Packages were searched to locate the CMTR's for all four pumps at each of the eight units. The thirty-two CMTR packages were reviewed to compile a DBase III Plus database of the chemical compositions of the RCP casing scroll and diffuser/hub castings and the weld filler metals used in joining the casing halves. Weld filler metals used for weld repairs of the castings or of the casing welds were not included in the database. Measured ferrite contents and tensile properties reported on the CMTR's were also included in the database. Appendix F provides a detailed listing of all of the information contained in the database with the actual variable names and a definition of the information contained in each variable. The information in the database was utilized in the approach detailed above f ~r estimating the aged fracture toughness for each pump casing. The first step was to calculate the ferrite content, material parameter and minimum Charpy impact energy for each ~aterial. This was done with a short DBase III Plus program called CASEINFO.PRG. The listing of this program is shown in Appendix F. This program calculates the ferrite content using Hull's factors and then proceeds to calculate the material parameter, minimum Charpy impact energy and the coefficients and exponents for the respective power law J-R curve equations. The resultant values for all of these values are then written into the database.

                                             . 69

The Jic and Kic values are calculated by another DBase t9 III Plus program called FRACTOUG.PRG, which is also listed in Appendix F. This program calculates the fracture toughness values from the J-R curve at room-and elevated-temperature by iteratively solving for the intersection of the power law equation and the 0.2 mm offset line according to ASTM E 813. Once again all of the calculated values are written into the database. The third DBase III Plus program called TEARMOD.PRG is also shown in Appendix F. This program performs the. linear regression fit of a linear curve from the power law equation in the range of acceptable data as defined by ASTM E 813. The calculated average tearing modulus values are written into the database. Detailed results for each heat of material for the ~ thirty-two pumps included in this study are provided in Appendices A, B, c, D and E. The Tables included in the Appendices list the complete information on the chemical composition for each heat of material, the

                       -     --    .   -

tensile properties {when available), the calculation of ferrite content and the determination of both room- and elevated-temperature fracture toughness parameters. For each plant the information is grouped for the individual pumps. The information for each pump is listed in order of increasing fracture toughness, such that the first entry represents the heat of material with the minimum estimated fracture toughness for aged material in a given pump. 5.3 End-Point Crack size Determination In this section the growth of a hypothetical crack is evaluated to establish at what point in its history it first 70

reaches an unacceptable size. An unacceptable crack size ~ would be indicated as soon as any one of the following three criteria is met:

1. The crack is unstable against non-ductile propagation.

2. The crack is unstable against ductile tearing.

3. The remaining ligament cannot carry its original loading, based on its flow stress.

Compliance with the above criteria are to be demonstrated for Design Conditions, Emergency Conditions, and Faulted Conditions as defined in the stress analysis reports. 5.3.1 crack size for Non-ductile Crack Propagation For this case, the criterion for acceptability is that ~ the applied stress intensity factor, previously calculated as described in Section 5.1.4, should be less than the aged material toughness, KJc" That this condition is met throughout the plotted growth histories from Figures 5.1-1 through 5.1-5, is demonstrated by the results in Table 5.3-1. In Table 5.3-1 the largest applied KI obtained from the fracture mechanics calculation is compared with the minimum and median measured material toughness from all heats of material. The material toughness at the point of maximum K , most probably the median toughness, is 1 seen to be substantially qreater than the minimum calculated KJc* In turn, the minimum calculated toughness, KJc' substantially exceeds the maximum applied KI for all of the casings evaluated. 71

Table 5.3-1 Applied KI From Design Condition vs Actual Toughness at 550 F Minimum Median KI KJc KJc Time, yrs Applied Toughness Toughness Plant/Limiting Location & Catt) CKsi-yinch) (Ksi-yinch) CKsi-vinch) Calvert Cliffs/Vane 8 133 (0.45) 108. 151. 0 186.7 Palisades/ Crotch 49 (0.45) 130. 152.3 223.5 Ft Calhoun/Vane 7 173 (0.45) 109. 141. 5 223.8 San Onofre/Crotch 77 (0.45) 101. 142.8 188.4 st . Lucie/Vane 8 133 (0.45) 108. 142.1 184.9 ...... N

                                                -- -- ---- --- - -:~t* -~

Since Applied Stress Intensity Factor is a strictly increasing function of dimensionless crack depth, a/t, it is sufficient to note that after the indicated Time in years none of the limiting cracks would have reached an unacceptable size as defined by this criterion. 5.3.2 Crack Size for Unstable Ductile Tearing It is possible for a crack to begin to tear under an applied load and then to self-arrest after a certain increase in its length. This effect is attributed to local work-hardening in the vicinity of the crack tip. It is numerically characterized by a material Tearing Modulus, which is the slope of the material "J-integral" (Jd) versus crack extension. The J-inteqral required for the onset of crack extension (the "intercept" of the Tearing Modulus slope) can be e related to the material Toughness, KJc' through linear fracture mechanics under plane-strain conditions, as discussed in Section 5.2.4. Ordinarily, stability against ductile tearing is demonstrated by comparing applied J-integrals for a series of crack depths to the Tearing Modulus slope, and demonstrating that after a given crack extension the Jd eventually exceeds the applied J. This is unnecessary if the applied J never reaches the J-integral required to initiate ductile tearing. That condition is satisfied provided the KI (applied) can be shown not to exceed material toughness KJc as in the above criterion. 73

Hence, stability against ductile tearing is established, since the onset of further crack propagation (other than per-cycle crack growth) would

                                                      --

not occur. 5.3.3 Crack Size for Flow stress Limit The next verification that the hypothetical RCP casing cracks have not reached an unacceptable size is based on flow stress considerations. The remaining ligament (uncracked portion) in a cracked section must remain capable of carrying the applied force and moment. Accordingly, a conservative, two-dimensional approximation method was used to establish the limiting crack depth for which this would no longer be possible. Once the flow-stress limited crack depth has been reached, any subsequent crack growth renders the remaining ligament incapable of supporting the load application. These conservative limiting crack depths are indicated in Figures 5.3-1 through 5.3-5 based on the Design condition. 5.3.4 Crack Depth for Emergency Condition and Faulted Condition Loads To completely determine the lower limit of unacceptable crack size, it is necessary to ascertain the stability of each analysis region under Emergency Condition and Faulted Condition loads. Accordingly, these were extracted from the Byron Jackson Company stress reports, the details of which are given in the 74

plant-specific Ap~endices A through E. As before, the - - membrane and bending components of stress were obtained for either condition. Some important differences relative to the design condition stresses are: (1) secondary* stresses are not considered, since they represent a surface effect, (2) thermal stresses are not added on, and (3) SSE (Safe Shutdown Earthquake) stresses replace OBE. The elimination of surface and thermal secondary stresses is consistent with the evaluation of Emergency Condition and Faulted Condition loads, since in those extreme cases only loadings to the gross section are considered (i.e., net membrane stresses and bending caused by sustained mechanical loads). Under Emergency Condition or Faulted Condition loads, local surface yielding due to secondary effects is acceptable. Therefore, it is reasonable that Emergency Condition and Faulted Condition stresses do not exceed the Design Condition ti (plus thermal) stresses used in the crack growth analysis, in many cases. Accordingly, applied stress intensity factors were calculated for Emergency Condition and Faulted Condition loads at large crack depths (e.g., 0.45 and 0.50). In all cases the Applied KI does not exceed the established minimum toughness (KJc) for the pump casings. This establishes an acceptable result for the dept~s indicated, and below. Flow stress limits were also checked under Emergency conditions and Faulted Condition. These are indicated in Figures 5.3-6 to 5.3-15. 75

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5.3.5 Plant Specific Results When each of the criteria for unacceptable crack size was applied to the plant-specific cases that were evaluated, the limiting condition was found to be ligament flow stress in all cases. The number of years of cyclic loadings preceding the end-point crack size based on ligament flow stress are as follows: Palisades 46 years Fort Calhoun 165 years Calvert Cliffs 1&2 130 years San Onofre 2&3 77 years st. Lucie 1&2 130 years 91

SECTION 5 REFERENCES 5-1 o. K. Chopra, "Estimation of Fracture Toughness of Cast stainless Steels During Thermal Aging in LWR Systems," NUREG/CR-4513 (ANL-90/42), U.S. Nuclear Regulatory Commission, Washington, D.C., June 1991. 5-2 ASTM A 800, "Standard Practice for Steel Casting, Austenitic Alloy, Estimating Ferrite Content Thereof," American Society for Testing and Materials, Philadelphia, PA. 5-3 L. s. Aubrey, et al., "Ferrite Measurement and Control in Cast Duplex Stainless Steels," Stainless Steel Castings, ASTM STP 756, V. G. Behal and A. S. Melilli, Eds., American Society for Testing and Materials, 1982, pp. 126-164. 5-4 NB-2433 "Delta Ferrite Determination," ASME Boiler and Pressure Vessel Code, Division 1 - Subsection NB, Class 1 Components, American Society of Mechanical Engineers, New York, NY. 5-5 ASTM E 813, "Standard Test Method for Jic' A Measure of Fracture Toughness," American Society for Testing and Materials, Philadelphia, Pa. 5-6 A. L. Hiser, "Tensile and J-R Curve Characterization of Thermally Aged cast Stainless Steels," NUREG/CR-5024 (MEA-2229), U.S. Nuclear Regulatory Commission, Washington, o.c., September 1988. 5-7 A. L. Hiser, F. J. Loss, and B. H. Menke, "J-R Curve Characterization of Irradiated Low Upper Shelf Welds," Appendix H, NUREG/CR-3506 (MEA-2028), U.S. Nuclear Regulatory Commission, Washington, D.C., April 1984. 92

5-8 B. A. Pistolese, "Limiting Crack Depth Based on Material Ultimate Strength", MISC-ME-C-112 (ABB Combustion Engineering), June 21, 1991. 5-9 c. L. Hoffma.n, "Fatigue Crack Growth Rate of Cast Stainless Steel in PWR Water", MCC-91-285 (ABB Combustion Enginee~ipg), July 30, 1991. 93

6.0 INSPECTION INTERVAL DETERMINATION 6.1 Safety Margins The results obtained in assessing th~ s.tabili ty of RCP casings subjected to postulated cracks.in .high stress regions support a relaxation of the req~ired io~year inspection intervals. For all plants reviewed, the end-point crack size is not reached until after the 40-year license period of the host plants. Also, for all plants reviewed, the RCP casing evaluations show stability throughout the 40-year license period despite a postulated 1/4t crack. In developing these results, many conservative assumptions and input data have been used. Certainly, refinement of these data can provide support for considerably longer time periods for reaching material instabilities. Because these results were obtained while using very conservative input data, the use of an explicit overall numerical safety factor is not applied to the final results. The following are some of the implicit (i.e., inherent) conservatisms that appear in the calculations: 1- Assumed initial crack size - The.assumed 8%t initial crack size is a factor of 4 larger than the 2%t detection sensitivity required by Section III. No detectable cracks are permitted by Section III during fabrication. Any that were found upon inspection were repaired. The required detection sensitivity for this pre-service inspection was 2%t. Any cracks too small to be detected would therefore be less than or equal to 2%t in depth. 2- Number of stress cycles - Use of 505 to 715 stress cycles in 40 years is based on pump ~esign specifications which establish conservatively the 94

numbe~ of cyclic events to be considered in performing the casing design. The number of such events actually ti experienced per year during plant operation to-date is considerably lower, as discussed in each plant-specific appendix. Based on individual plant operations from initial start-up to the present, this represents a safety factor of from 2 to 9. 3- Events considered in stress analysis - The original stress analysis includes the stress represented by an Operating Basis Earthquake. It also includes thermal stresses from heat ups and cool-downs at rates greater than those normally experienced by the pump casings. In calculating stress from system pressure, the calculation uses design pressure rather than operating pressure. These abnormal stresses are applied to the crack growth rate during each of the cycles assumed to occur per year, even though a normal cycle would not experience them. 4 - Inclusion of Secondary Stresses - The. Design Stresses which were specifically identified as secondary in the Byron-Jackson Company reports. In many cases, secondary stresses were a significant portion of _the local bending component of linearized stress. Secondary stresses are often caused by strain-compatibility effects between portions of a structure having differing thicknesses, e.g., the volute-to~flange junction, or the crotch region. By definition, they are inherently self-limiting since localizeq yielding would cause them to subside. They are not "load-controlled" as is the case for primary stresses, e.g., as pressurization would cause a primary membrane stress. This aspect of the crack growth 95

analysis is thus conservative, since the secondary stresses are assumed not to gradually subside with successive cycles. 5 - Estimated Fracture Toughness Properties - The estimated fracture toughness properties of the casing materials were evaluated from the saturation toughness J-R curves. Actual service time-temperature histories J of the pump casings were not used to evaluate the decrease in toughness due to thermal embrittlement. In addition, conservative estimates for flow stresses were used to calculate JIC values from the J-R curves. 6 - Bias Toward Assumptions of Thick Sections - As discussed in Section 5.1.4, wherever a choice of section thickness was open because of lack of dimensional data, the thicker value was chosen to develop a conservative (faster) crack growth rate. 7 - Use of End-of-Interval Growth Rates - When calculating crack growth, the growth rate at the end of the crack growth interval was applied to the entire interval. As discussed in Section 5wl.4, this results in faster calculated crack growths than would be the case if average crack growth rates for the interval had been used. 6.2 Inspection Alternatives 6.2.l Volumetric Examination The volumetric examination required by the ASME Code was performed by radiography during.fabrication. The scroll and hub/diffuser sections were radiographed 96

separately before being welded together and the welds joining the two were radiographed afterward, access to the internal surf aces being gained through the discharge nozzle. This same extent of weld volumetric examination would be extremely difficult to perform during 10-year inservice inspections required by Section XI because of limited access to the internals of the installed pump and because of the high level of attendant radiation. Consequently, conformance with the Section XI requirement for volumetric examination of the casing welds during inservice inspection intervals is considered to be impractical for the Type E RCPs being evaluated. Even if radiography were possible, it would be difficult to characterize the three dimensional structure of the crack without costly and radiation intensive effort. An alternative to the volumetric examination is application of the methodology discussed in Section 5.0 to demonstrate analytically that RCP casing stability is assured during at least the 40 year license period for each plant and to show that this stability bounds the 1/4t reference crack postulated in ASME Code Case N-481. 6.2.2 Visual Examination Section XI also requires a visual examination (VT-3) of the internal surfaces of pump casings at each 10-year inservice inspection interval. For the same reasons as described above, conformance with this requirement is considered to be impractical for the Type E RCPs being evaluated. It is important to note that visual inspections of the inside surface are of little usefulness~ In particular, visual exams give no indications of the depth of the crack. Furthermore, visual inspection is complicated by the long-term operation in a hot-water environment. Additionally, 97

any cracks would tend to close tight during the visual ~ exam, which is performed without pressure and thermal loads present. Application of the methodology presented in Section 5.0 to demonstrate that the end-point crack size bounds the postulated reference crack of ASME Code case N-481 is a reasonable alternative to the Section XI visual examination requirements. The VT-1, and VT-2 visual examinations of external surfaces called for in Code case N-481 are not considered to be effective from a risk/benefit standpoint. The removal and replacement of pump insulation for access to the external surf aces and the examination itself increase personnel exposure with limited benefit over the analytical demonstration of stability described in Section 5.0. It is proposed, however, that a VT-3 visual examination as specified in Code Case N-481 be required as an additional alternative to the VT-1 examination specified by Section XI. The VT-3 visual of the internal surfaces would be performed.to the extent practical whenever a pump is disassembled for maintenance. This visual exam would be performed to detect gross damage, loose parts and abnormal conditions. 6.3 Inspection Interval Based on the foregoing considerations, this report proposes that requirements for RCP inspection be relaxed from the currently specified volumetric and internal visual examinations during each 10-year inservice inspection 98

interval, and that an alternate program based largely on ASME Code Case N-481 be considered. The alternate program would require that an evaluation be performed for each pump casing to show that the end-point crack size bounds the postulated reference crack of Code Case N-481, and would require a VT-3 visual examination of internal surfaces, to the extent practical within limitations of design and geometry, whenever a pump is disassembled for maintenance. Under this program the inspection interval is dependent on the time periods between disassembly of -the RCP for maintenance or on the number of years of operation at which the calculated end-point crack size is reached, whichever occurs first. Further operation of the pump upon reaching the calculated end-point crack size condition would require either additional evaluation to justify continued operation or performance of technologically available non-destructive examination techniques to characterize the integrity of the RCP casing. 6.4 Plant Specific Inspection Intervals For the eight units evaluated in this report, the RCP casing volumetric inspection interval can safely be extended in each case to at least the initial 40 year license period. If desirable in the future for the Palisades plant, a further review of the conservatisms embodied in this evaluation will most probably extend the volumetric inspection interval beyond the presently indicated 46 years. For Fort Calhoun, Calvert Cliffs 1&2, San Onofre 2&3, and St. Lucie 1&2, the conservatively calculated volumetric inspection interval of greater than 70 years is ample to preclude the need for such inspections for the life of the plant. In all eight cases it is considered prudent and sufficient to perform a VT-3 visual examination of internal surfaces to the extent practical whenever a pump is disassembled for maintenance. 99

7.0 CONCLUSION

S Based on the plant-specific evaluations described in the appendices of this report, it is apparent that RCP casing materials used in each of the reactor coolant pumps operating at* Palisades, Fort Calhoun, Calvert Cliffs 1&2, San Onofre 2&3, and St. Lucie 1&2 can withstand 'the effects of conservatively established cyclic stress loadings.over at least the initial 40-year license periods without reaching an end-point crack depth condition at which rapid crack growth would occur. All of the casings were also found to meet th~ materials criteria in ASME Code Case N-481 for waiving volumetric examinations of cast austenitic pump casings. As shown in Table 7.0-1, a postulated ~one-quarter thickness reference flaw will remain stable under governing design, emergency and faulted condition stresses. For each casing evaluated, the end-point crack depth at which instability occurs was found to be greater than the one-quarter thickness reference flaw postulated in Code Case N-481. -- Based on this evaluation, it is concluded that inservice volumetric examinations of these RCP casings are unnecessary for the 40-year license periods of the plants evaluated, but visual (VT-3) examinations of casing inside surfaces, to the extent practical, are prudent whenever an RCP is disassembled for maintenance. Casing integrity is shown to be retained for each of the pump casings for at least the following total periods, starting from initial operation: Palisades 46 years Fort Calhoun 165 years Calvert Cliffs 1&2 130 years San Onofre 1&2 77 years St. Lucie 1&2 130 years 100

Table 7.0-1 END-POINT CRACK SIZES AS A PERCENT OF WALL THICKNESS Location as Defined in Crack Growth History Graphs

                       -1        ~         ....J.    ......!  _s_

Palisades Design Condition 36 40 42 Emerg. Condition 36 40 42 Faulted Condition 36 40 42 Fort Calhoun Design Condition 42 43 so 32 *~

                                                                    *=*

Emerg. Condition S6 so 48 so Faulted Condition 40 46 49 53 Calvert Cliffs 1&2

• Design Condition Ernerg. Condition Faulted Condition San Onofre 2&3 Design Condition 44 57 39 42 49 42 33 49 47 42 38 46 S4 46 34 41 46 46 34 37 Emerg. Condition S9 S4 52 4S S5 Faulted Condition 35 39 44 41 68 St. Lucie 1&2 Design Condition 44 49 47 54 46 Emerg. Condition S7 42 42 46 46 Faulted Condition 39 33 38 34 34 101

APPENDIX A APPLICATION OF GENERIC METHODOLOGY FOR RELAXATION OF THE PALISADES REACTOR COOLANT PUMP CASING INSPECTION INTERVAL

• Appendix A 1

ABSTRACT Appendix A was prepared to demonstrate the amount of inspection interval relaxation appropriate for the reactor coolant pump casings at the Palisades plant, based on application of the generic methodology presented in the main body of this report. Appendix A 2

APPENDIX A TABLE OF CONTENTS Section Title Page 1.0 PURPOSE 5

2. 0 . PRE-SERVICE INSPECTION DATA EVALUATION 6 3.0 OPERATING HISTORY 8 3.1 Design Specifications 8 3.2 Stress Cycles Used in Evaluation 9 3.3 Stress Cycles at Palisades To-date 9 4.0 THERMAL EMBRITTLEMENT 10 4.1 Material Identification and Chemical Properties 10 4.2 Material Specifications and Mechanical Properties 10 e

4.3 Thermal Aging Behavior 11 4.4 Toughness Properties of Aged Materials 11 4.5 Limiting Values 12 5.0 CRACK GROWTH ANALYSIS 26 5.1 ~cope 26 5.2 Reference Stress Reports 26 5.3 Selection of High Stress Locations 26 5.4 Stresses and Wall Thicknesses at Limiting Locations 27 5.5 Calculation of Crack Growth Rates 28 5.6 Stresses Under Emergency and Faulted Conditions 32 5.7 Results 33

6. 0 . INSPECTION INTERVAL 34 APPENDIX A REFERENCES 35 e

I i Appendix A 3 I L

LIST OF TABLES TABLE 4-1 Material Identification and Chemical Compositions 13 TABLE 4-2 Material Specifications and Tensile Properties 16 TABLE 4-3 Predicted Thermal Aging Behavior 18 TABLE 4-4 Predicted Toughness Properties of Aged Materials (70°F) 20 TABLE 4-5 Predicted Toughness Properties of Aged Materials (550°F) 22 TABLE 4-6 Limiting and Controlling Values of Jic and KJc at 70°F 24 TABLE 4-7 Limiting and Controlling Values of Jic and KJc at 550°F 25 TABLE 5-1 Crack Growth Rate at Suction Nozzle Flange 29 TABLE 5-2 Crack Growth Rate at Crotch Region 30 TABLE 5-3 Crack Growth Rate at Discharge Nozzle 31 Appendix A 4

1.0 PURPOSE The purpose of Appendix A is to document application of the methodology presented in the generic portion of this report to plant-specific data for the reactor coolant pump casings at the Palisades plant, and to quantify the resulting extent of inspection-interval relaxation available. Appendix A 5 I I b:::=~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

2.0 PRE-SERVICE INSPECTION DATA EVALUATION Pre-service inspection data for the Palisades reactor coolant pumps numbered 661-N-0764 through 661-N-0767 was collected from QA data packages originally prepared by the Byron Jackson Company and stored in archives by ABB Combustion Engineering Nuclear Power. Information in these data packages concerning welding procedures, radiographic inspections, non-destructive testing and dye penetrant testing were examined. The testing and inspection procedures that were followed for all reactor coolant pumps at Palisades were found to be the same in all significant aspects. The most relevant information obtained from this review of the QA data packages* were the reports on radiographic examination of the RCP casing castings, pressure retaining welds, and repair welds. Radiographic examination requirements invoked ASME Section III rules for examination procedures and sensitivity. The required radiograph sensitivity was 2-2T according to applicable ASTM Standard Refere~ce Radiograph Procedure requirements (i.e. ASTM E71, El86, E280) as determined by the casting thickness. The 2-2T sensitivity is consistent with a 2% initial flaw size, because the requisite image quality indicator (IQI) for this level of examination is specified as a penetrameter with a minimum hole-size diameter equal to 2% of the casting thickness. The acceptance criteria for interpretation of the radiographs was severity Level 2 for sand, porosity or shrinkage indications. Linear indications such as cracks, hot tears, and unfused chaplets or chills were unacceptable at any level. Any such discernable indications required rejection of weld repair and a repeated radiographic examination of the affected casting or weldment. Appendix A 6

The results of this review of pre-service RCP casing examinations confirm the assumed detectable flaw of 2% thickness described in Section 4.4 of the generic report. Appendix A 7

                                               \

3.0 OPERATING HISTORY 3.1 Design Specifications The Palisades RCPs were delivered to the site in 1967 and were first placed in cormnercial operation in 1971. Reactor coolant system design pressure and temperature are 2485 psig and 650°F respectively. Each pump is designed to deliver 83,000 gpm of coolant at a head of 260 feet. These pumps have 30 inch diameter suction and discharge piping. The design specification (Reference 3-1) calls for the pumps to be capable of withstanding the following transient conditions during their 40-year license period: Assumed Occurrences Transient During 40 Year Condition License Period

• Heat-Up 500 Cool-Down 500 Hydrostatic Test (3110 psia) 10 Leak Test In Conjunction with Heatup (2485 psig) 320 Loss of Secondary Pressure 2 Reactor Trip 500 Appendix A 8

3.2 Stress Cycles Used in Evaluation As indicated in Section 5.1.4 of the generic portion of this report, crack growth was evaluated on the basis of an assumed number of stress cycles between atmospheric and operating pressures during heatup and cooldown over the nominal 40-year life of the plant. The number of such

   -cycles used in the stress analyses performed by the Byron Jackson company is 502, as given on page 20 of Reference 5-1. This total is slightly greater than the 500 heat-up cooldown cycles specified in the RCP design specifications.

On an annual basis, the average number of stress cycles, based on 502 per 40 years, is 12.55 per year, and the hypothetical crack growth calculations and curves were prepared accordingly. 3.3 Stress Cycles at Palisades To-date Details of the actual operating history of the Palisades plant from 1971 to 1991 were furnished in Reference 3-2 and are as follows:

• Heatup/Cooldown - 106 Hydrostatic Test - 3 Leak Test - 67 Loss of Secondary Pressure - o Reactor Trip - 112 Because heatup-plus-cooldown, taken together, constitute one cycle, and the remaining events represent relatively minor stresses, the number of stress cycles to-date is seen to be 106 over a 20 year time period. This time period is equivalent to only 8.4 years at the design rate of pressure cycling. The actual rate of cycle accrual is thus seen to be only 42% of the design rate: a significant conservatism ~

in the evaluation. Appendix A 9

                                                \

4.0 THERMAL EMBRITTLEMENT Thermal embrittlement evaluation of the Palisades casings is discussed and plant specific data are presented in the five following reports. All equations referenced below are found in the main body of this report, which is also referred to as the generic report. 4.1 Material Identification and Chemical Properties The chemical compositions provided in Report #1 (Table 4-1) for each RCP casing at Palisades were obtained from Quality Assurance documents originally supplied by the Byron Jackson Company and stored at ABB Combustion Engineering Nuclear-- Power. For each individual pump casing, chemical compositions are given for specific casing welds as well as for individual castings. 4.2 Material Specifications and Mechanical Properties The material specifications and mechanical properties found in Report #2 (Table ~-2) for each RCP casing were obtained from the same data source as in Section 4.1 above. For each individual pump casing the material specification, material_ type_and heat number are given for specific casing welds as well as for individual castings. It is evident from the report that data obtained for mechanical properties (i.e. yield strength, tensile strength, total elongation and reduction in area) for each material was only available for the castings, and was not available for the casing welds. The unaged flow stress at 70"F and the aged flow stress at 70"F and SSO"F were calculated as discussed in Section 5.2.3 of the generic report. Appendix A- 10

4.3 Thermal Aging Behavior Report #3. (Table 4-3) contains predicted thermal aging behavior data for all of the Palisades RCP casings. The measured ferrite contents listed for specific casing welds and individual castings were supplied by the Byron Jackson Company in the same QA package as referenced in Section 4.1 above. In most cases a value was obtainable for the measured ferrite content. In cases where a value was not given, a zero was recorded. The chromium and nickel equivalents for the castings and weld metal, as well as the chromium/nickel ratio for the castings, were calculated using equations 5-7 and 5-8 respectively, as discussed in Section 5.2.1 of the generic report. Values for ferrite content of the castings were computed using two methods: for ferrite*content #1, the values were computed using the method which follows ASTM ASOO/ASOOM (Reference 5-2); for ferrite content #2, the values were computed using equation 41t 5-9 as discussed in Section 5.2.l of the generic report. The latter method follows work performed by O.K. Chopra (Reference 4-1). 4.4 Toughness Properties of Aged Materials The predicted toughness properties of aged material at 70°F and 55o*F are given respectively in Report #4 (Table 4-4) and Report #5 (Table 4-5). The measured ferrite contents listed for all heat numbers are the same as the values given in Report #3. The material aging parameter was calculated using equation 5-10 of the generic report. The room-temperature charpy impact energy, Cvsat, of the various materials was calculated using equation s-11. The Jic values were determined in accordance with the methods of ASTM E813 as discussed in Section 5.2.3. The plane strain fracture toughness, KJc' and minimum tearing modulus, T, at 41!t L i Appendix A

                                                    \

11

70°F and 550°F were calculated using equations 5-17 and 5-18. The values listed for the material constants N and c at 70°F and 550°F were calculated using equations 5-12 through 5-15. These constants were needed in computing the values for Jic' KJc and T. All equations used in Report #4 are found in Section 5.2 of the generic report. 4.5 Limiting Values The limiting and controlling values for Jic and KJc at 70°F and 550°F for each individual pump at Palisades are given in Tables 4-6 and 4-7. Appendix A 12

Table 4-1 ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL

)>

"'O REACTOR COOLANT PUMP CASINGS "'O CD

I a.

REPORT fl* MATERIAL.IDENTIFICATION &CHEMICAL COMPOSITIONS I

-'*

x

)>

MATERIAL HEAT NO. c Mn St s p Cr Ht Mo N Cb

         ** PLANT I.D. PALISADES

• RCP PUMP CASING 661*N*0764 CASE ASSEM. 4847 0.07 0.66 1.20 0.008 0.020 19.30 9.30 2.16 0.04 o.oo CASE ASSEM. 4851 0.04 1.10 1.15 0.010 0.020 19.00 9.40 2.10 0.04 o.oo CASING WELD 03033A 0.06 1.91 0.49 0.019 0.012 18.88 12.25 2.33 0.04 o.oo*

CASING WELD 03067A 0.08 1.88 0.43 0.021 0.011 18.84 12.40 *2.43 0.04 .* o.oo CASING WELD 0214C 0.05 1.73 0.46" 0.024 0.010 19.00 12.65 2.35 0.04 0.00 CASING WELD 03104 0.07 1.80 0.38 0.016 0.011 20.14 9.80 o.oo 0~04 o.oo CASING WELD 02362A 0.03 1.72 0.48 0.018 0.022 19.10 11.55 2.29 0.04 o.oo CASING WELD 02475C 0.03 1.68 0.53 0.022 0&010 18.92 11.49 2.29 0.04 o.oo*

• RCP PUMP CASING 661*N*0765 CASING WELD A7B24R 0.06 1.42 o.so 0.012 0.011 20.52 .J0.56 2.66 0.04 o.oo CASING WELD 6A58A 0.08 2.43 0.23 0.014 0.022 19.14 11.92 2.47 0.04 0.00 CASING WELD 87Hl3R 0.05 1.12 0.45 0.025 0.025 1'9.88 10.87 2.60 0.04 o.oo CASE ASSEM. 4849 0.06. 1.20 1.20 0.015 0.020 18.80 9.70 2.14 0.04 o.oo CASING WELD 6A57A 0.05 2.29 0.21 0.016 0.016 19.30 11.51 2.52 0.04 o.oo CASING WELD 6A59A 0.06 2.48 0.23 0.016 0.022 19.49 11.72 2.52 0.04 o.oo CASE ASSEM. 4853 0.06 0.79 0.67 0.015 0.020 18.70 9.40 2.12 0;04 o.oo CASING WELD B7Hl4R 0.03 1.15. 0.40 0.025 0.025 19.62 10.82 2.87 0.04 o.oo CASING WELD 7E34B 0.05 1.84 0.31 0.006 0.010 19.50 12.71 2.05. o.*04 0.00 CASING WELD 03104 0.07 1.80 0.38 0.016 0.011 20.14 9.80 o.oo 0.04 o.oo CASING WELD 03252 0.04 1.46 0.40 0.015 0.011 18.99 *11.80 2.24 0.04 0.00

    /     CASING WELD  78143A    0.05   1.33  .* 0.51  0.010   0.031 19.53  "11.57   2.24  0.04     o.oo CASING WELD  02362A    0.03  .1. 72   ~0.48 0.018    0.022 19.10    11.55  2.29  0.04     o.oo CASING WELD  03100     0.03   1.66*   'c>.40 0.018 . 0.010 19.06    12.20  2.25  0.04     0.00 CASING WELD  02475C    0.03   1.68    *o.53 0.022    0.010 18.92    11.49  2.29  0.04     o.oo CASING WELD  7H32A     0.03   1. 73    0.34 0.008    0.013 19.12    11.88  2.02  0.04   . o.oo
 .......

w e

Table 4 . (Continued) /- ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL

~

"'C REACTOR COOLANT PUMP CASINGS "'C ct>

s

a. REPORT 11

• MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS

-'*

x

~

MATERIAL HEAT NO. c Mn S1 s p Cr N1 Mo N Cb CASING WELD 7H37A 0.03 2.08 0.27 0.009 0.012 18.68 12.28 2.08 0.04 o.oo

• RCP PUMP CASING 661-N-0766 CASING WELD 87Hl5R 0.07 1.35 0.48 0.022 0.013 20.16 10.98 2.73 0.04 0.00 CASING WELD 87H13R 0.05 1.12 0.45 0.025 0.025 19.88 10.87 2.60 0.04 o.oo CASING WELD 87J1R 0.06 1.25 0.49 0.030 0.013 20.27 10.75 2.65 0.04 0.00 CASING WELD 03033A 0.06 1.91 0.49 0.019 0.012 18.88 12.25 2.33 0.04 0.00 CASE ASSEM. 4850 0.06 0.83 0.97 o*.01s 0.020 19.00 9.80 2.18 0.04 o.oo CASE ASSEM. 4854 0.04 1.01 1.00 0.013 0.020 18.90 9.40 2.09 0.04 0.00 CASING WELD 87Hl4R 0.03 1.15 0.40 0.025 0.025 19.62 10.82 2.87 0.04 o.oo CASING WELD 03067A 0.08 1.88 0.43 0.021 . 0*.011 18.84 12.40 2.43 0.04 0.00 CASING WELD 03378A 0.04 1.35 0.29 0.011 0.009 21.54 10.04 o.oo 0.04 0.00 CASING WELD 033788 0.04 1.35 0.29 0.011 0.009 21.54 10.04 o.oo 0.04 0.00 CASING WELD 7F43A 0.06 1.30 0.37 0.011 0.015 18.79 11.22 2.20 0.04 o.oo CASING. WELD 0214C 0.05 1. 73 0.46 0.024 0.010 19.00 12.65 2.35 0.04 o.oo CASING WELD 03252 0.04 1.46 0.44 0.015 0.011 18.99 11.80 2.24 0.04 o.oo CASING WELD 03100 0.03 1.66 0.40 0.018 0.010 19.06 12.20 2.25 *0.04 0.00
• RCP PUMP CASING 661-N-0767

          . CASING WELD  87Hl5R   0.07 1.35 0.48 0.022 0.013 20.16 10.98          2.73
                                                                                          .

0.04 o.oo CASING WELD B7H13R 0.05 1.12 0.45 0.025 0.025 19.88 10.87 2.60 0.04 o.oo CASING WELD 87J1R 0.06 1.25 0.49 0.030 0.013 20.27 10.75 2.65 0.04 o.oo CASE ASSEM. 4848 0.06 1.03 0.99 0.023 0.030 19.00 9.80 2.18 0.04 o.oo . CASE ASSEM. 4852 0.05 0.88 ,0. 76 0.014 0.020 19.30 .. 9.80 2.09 0.04 0.00 CASING WELD B7Hl4R 0.03 1.15 *,'.0.40 0.025 0.025 19.62 . 10.82 2.87 0.04 0.00 CASING WELD 03252 0.04 1.46 ;p.44 0.015 0.018 18.99 '11.80 2.24 0.04 o.oo CASING WELD 7F43A. 0.06 1.30 *o.37 0.011 0.015 18.79 11.22 2.20 0.04 o.oo* CASING WELD 02362A 0.03 l. 72 0.48 0.018 0.022 19.10 11.55 2.29 0.04 o. 00 *'

 ....... CASING WELD   03100    0.03 1.66 0.40 0.018 0.010 19.06 12.20          2.25  0.04  o.oo
 .i::-

Table 4-1 (Continued) ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL

)::o "O

REACTOR COOLANT PUMP CASINGS "O tD

s 0..

REPORT 11

• MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS

-'*

x

)::o MATERIAL HEAT NO. c   Mn     S1      s      p      Cr    N1    Mo     N     Cb CASING WELD 02475C   0.03 1.68   Q.53 0.022 0.010 18.92 11.49 2.29 0.04 0.00 CASING WELD 10184    0.02 1.52   0.54 0.015 0.* 014 18.56 12.81 2.30 0.04 0.00

e Tao.ie 4-29 e

                                                                                                          ~'

ANALYSIS OF THERMAL AGING OF CAST STAIN~ESS STEEL

):>

"'O REACTOR COOLANT PUMP CASINGS "'O CD

I

a. REPORT 12

• MATERIAL SPECIFICATION &TENSILE PROPERTIES

--'*

x

):>     MATERIAL     MATERIAL MATERIAL                 YIELD TENSILE       TOTAL RED. IN   UNAGED    AGED       AGED OR        SPEC. TYPE

• STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (k~i) (ks1) (%) . (%) STRESS STRESS STRESS

                                                                                            @ 70F   @ 70F    @ 550F
                                                            ..
                                                            .i
       ** PLANT I.D. PALISADES

• RCP PUMP CASING 661-N-0764 CASE ASSEM. A 351 CFSM 4847 38.5 80.0 57.0 o.o 59 69566 48972 CASE ASSEM. A 351 CF8M 4851 39.3 81.8 58.0' o.o 61 71624 51030 CASING WELD 1T3454 E316-16 03033A , o.o o~o o.o o.o 0 78600 58000 CASING WELD 1T3454 . E316-16 03067A o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316*16 0214C o.o o.o o.o o.o 0 78600 58000 CASING WELD A 298 E308*16 03104 o.o o*.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316*16 02362A o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316*16 02475C o.o o.o o.o o.o 0 78600 58000
• RCP PUMP CASING 661-N-0765 CASING WELD A 351 CF8M A7B24R o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-15 6A58A o.o o.o o.o . o.o 0 78600 58000 CASING WELD A 351 CR8H B7Hl3R o.o o.o o.o o.o 0 78600 58000 CASE ASSEM. A 351 CF8M 4849 36.5 81.0 54.0 o.o 59 68774 48180 CASING WELD E316-15 6A57A o.o o.o. o.o o.o 0 78600 58000 .

CASING WELD E316*15 6A59A o.o 0.0* o.o o.o* 0 78600 58000 CASE ASSEM. A 351 CF8M 4853 34.5 77.5 50.5 o.o 56 64420 43826 CASING WELD A 351 CF8H B7H14R o.o o.o o.o o.o* 0 78600 58000 CASING WELD E316-15 7E34B o.o .o.o 0.0 o.o 0 78600 .:58000 CASING WELD A 298 E308-16 03104 o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316-16 0325~'. o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-15 7814~ o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316*16 0236 A o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316-16 03100 o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 £316-16 02475C o.o o.o o.o 0.0 0 78600 58000 ....... O'I CASING WELD E316-15 7H32A o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-15 7H37A o.o o.o o.o o.o 0 78600 58000

Table 4-2 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

)>

-0 REACTOR COOLANT PUMP CASINGS -0 Ill

s

a. REPORT 12

• MATERIAL SPECIFICATION &TENSILE PROPERTIES

~.
)(

MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED

)>

OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                                     @70F       @ 70F    @550F

• RCP PUMP CASING 661-N-0766 o.o.

                                                                            \

o.o o.o CASING WELD A 351 CF8M B7Hl5R o.o 0 78600 58000 CASING WELD A 351 CF8M B7Hl3R OiO o.o o.o o.o . 0 78600 58000 CASING WELD A 351 CF8M B7J1R o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316-16 03033A o.o o.o o.o . 0.0 0 78600 58000 CASE ASSEM. A 351 CF8M 4850 '38.0 77.5 52.0 o.o 58 67191 46597 CASE ASSEM. A 351 CF8M 4854 39.3 77 .s . 51.0 o.o ** 58 68220 47626 CASING WELD *A 351 CF8M 87H14R o.o o.o o.o o.o 0 0. 78600 58000 CASING WELD 1T3454 E316-16 03067A o.o o.o o.o o.o 78600 58000 CASING WELD E308 03378A o.o .. o.o o.o o.o 0 78600 58000 CASING WELD E308 033788 o.o o.o o.o o.o 0 78600 58000 CASING WELD E316*15 7F43A o.o o.o o.o 0.0 0 78600 58000 CASING WELD 1T3454 E316*16 0214C o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316-16 03252 o.o o.o o.o o.o 0 78600 . 58000 CASING WELD 1T3454 E316-16 03100 o.o o.o o.o o.o 0 78600 58000

          * *RCP PUMP CASING 66l*N*0767 CASING WELD    A 351    CF8M       B7H15R          o.o       o.o        o.o      o..o          0      78600    58000 CASING WELD    A 351    CF8H       B7Hl3R          o.o       o.o        o.o . o.o           0      78600    58000 CASING WELD    A 351    CF8M       B7J1R           o.o       o.o        o.o      o.o           0      78600    58000 CASE ASSEM. A 351    CF8M       4848           37.0      75.0       48.5      o.o          56      64420    43826 CASE ASSEM. A 351    CF8M      *4852           38.5    . 79.3       43.0      o.o          59      69011 .: 48417 CASING WELD    A 351    CF8H       B7H\4R          o.o       o.o        o.o      0.0           0      78600    58000
       ,*

CASING WELD 1T3454 E316-16 03252 o.o o.o o.o o.o 0

• 1 78600 58000 CASING WELD E316-15 7F4~ o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316-16 023 2A o.o o.o 0.0 o.o 0 78'600 58000 CASING WELD 1T3454 E316-16 03100 o.o o.o o.o o.o 0 78600 58000 CASING WELD 1T3454 E316-16 02475C o.o o.o o.o o.o 0 78600 58000

 .......
 -...i     CASING WELD              E316-16   10184           o.o       o.o        o.o      o.o           0      78600    58000 e
                                                                       -                                                      e

         .***

I - ' Table 4 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

                                                                                                                             --
)>

"O REACTOR COOLANT PUMP.CASINGS "O 11>

s 0..

_,, REPORT 13

• PREDICTED THERMAL AGING BEHAVIOR x

)>

HEAT MEASURED CHROMIUM NICKEL Cre/Nte CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR cbNTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS* n (t:) 12 (%) WELDS WELDS (%)

      ** PLANT I.O. PALISADES

• RCP PUMP CASING 661-N-0764 4847 o.o 17.500 14.590 1.20 13.7 13.7 23.260 12.930 11.0 4851 o.o 17.103 13.997 1.22 14.7 15.4 22.825 12.350 12.0 03033A 7.5 16.944 17.454 0.97 ' 2.5 3.0 21.945 16.265 5.0 03067A s.o 16.997 17.944 0.95 1.5 2.5 21.915 16.820 3.8 0214C 5.0 17.074 17.619 0.97 *2.s 3.0 22.040 16.305 s.o

     . 03104           6.0   15.332   15.069     1.02            2.7               4.4     20.110   13.850          6.2 02362A          o.o   17.111   15.955     1.07       I :  6.2               6.5     22.110   14.510          9.9 02475C*         OiO   16.955   15.892     1.07            6.2               6.2     22.005   14.430         10.0

• RCP PUMP CASING 661-N-0765 A7B24R 12.0 18.989 15.626 1.22 11.8 14.9 23.930 14.210 9.8 6A58A 7.0 17.249 17.554 0.98 1.8 3.3 21.955 16.675 4.0 B7H13R 11.0 18.252 15.713 1.16 9.7 11.2 23.155 14.130 14.2 4849 o.o 16.975 . 14.796 1.15 10.5 10.4 22.740 13.300 15.3 6A57A 9.0 17.460 16.448 1.06 4.5 6.0 22.135 15.355 7.7 6AS9A 7.0 17.660 16.965 1.04 3.8 5.2 22.355 16.020 7.5 4853 . o.o 16.597 14.458 1.15 7.3 10.4 21.825 12.795 13.4 B7H14R 13.0 18.295 15.152 1.21 12.3 14.3. 23.090 13.465
• 15.S 7E34B 6.0 17.139 17~663 0.97 2.1 3.0 22.015 16.390 4*.0 ..

03104 6.0 15.332 15.069 1.02 2.7 *4.4 20.710 13.850 . 6.2 03252 6.0 16.902 16.551-: 1.02 4.1 4.5 21.830 15.080 1I 716 78143A 02362A 5.0 o.o 17.495 16.481 17.111 15.955 : i 1.06 1.07 5.9 6.2 6.0 6.5 22.535 22.110 - 14.995 14.510 10.0 9*, 9 03100 5.5 16.984 16.649 ' 1.02 4.1 4.4 21.910 15.190 9.7 02475C 0.0 16.955 15.892 1.07 6.2 6.2 22.005 14.430 10.0 I-' co 7H32A o.o 16.737 16.237 1.03 4.5 4.8 21.650 14.785 7.6 7H37A o.o 16.336 16.762 0.97 2.1 3.1 21.165 15.480 4.0

                                                                   -- - - - - - * - _ . __h.

Table 4-3 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

)::a

"'C REACTOR COOLANT PUMP CASINGS "'C Cb

s a.. REPORT 13

• PREDICTED THERMAL AGING BEHAVIOR

......

x HEAT MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED

)::a NO.       FERRITE EQUIV. EQUIV.             RATIO FERRRITE        FERRITE   EQUIV.      EQUIV.      FERRITE CONTENT        FOR      FOR         FOR CONTENT CONTENT              FOR           FOR WELD METAL

(%) CASTINGS CASTINGS CASTINGS fl (%) 12 (%) WELDS . WELDS (%)

• RCP PUMP CASING 661-N-0766 B7H15R 8.0 18.704 16.383 1.l4 8.5 10.0 23.610 15.015 7.0 B7H13R 11.0 18.252 15.713 1.16 9.7 11.2 23.155 14.130 . 14.2.

B7JlR 10.0 18.722 15.850 1.18 10.5 12.5 23.655 14.375 9.0 03033A 7.5 16.944 17.454 0.97 2.5 3.0 21.945 16.265 5.0 4850 o.o 17 .113 14.861 1.15 11.0 10.6 22.635 13.215 15.2 4854 o.o 16.919 13.988 1.21 13.7 14.5 22.490 1'2 .305 10.l B7H14R 13.0 18.295 15.152 1.21 . 12.3 14.3 23.090 13.465 15.5 03Q67A 5.0 16.997 17.944 0.95 1.5 2.5 21.915 16.820 3.8 03378A o.o 16.689 14.659 1.14 .. 7.3 9.9 21.975 13.115 13.0 033788 0.0 16.689 14.659 1.14 7.3 9.9. 21.975 13.115 13.0 7F43A 6.0 16.640 16.324 1.02 4.1 4.4 21.545. 14.870 7.4 0214C 5.0 17 .074 17.619 0.97 2.5 3.0 22.040:. 16.305 5.0 03252 6.0 16.922 16.551 1.02 4.5 4.5 21.890 15.080 7.9 03100 5.5 16.984 16.649 1.02 4.1 4.4 21.910 15.190 7.7

• RCP PUMP CASiNG 661-N-0767 B7Hl5R 8.0 18.704 16.383 1.14 8.5 10.0 23.610 15.015 7.0 B7Hl3R 11.0 18.252 15.713 1.16 *9. 7 .. 11.2 23.155 14.130 14.2 B7J1R ' 10.0 18.722 15.850 1.18 10.5 12.5 23.655 14.375 9.0 4848 0.0 17.123 14.880 1.15 10.5 10.6 22.665 13.315 9.0 4852 0.0 17.204 14.621 1.18 11.4 12.2 22.530 12.940 10.0 ..

B7Hl4R 13.0 18.295 15.152

• 1.21 12.3 14.3 23.090 13.465 15.5 03252 6.0 16.922 16. 551 * *I 1.02 4.5 4.5 21.890 15.080 .'7 .9 7F43A o.o 16.640 16.324 '. 1.02 4.1 4.4 21.545 14.870 7.4 02362A o.o 17.111 15.955 1.07 6.2 6.5 22.110 14.510 9,:9 03100 5.5 16.984 16.649 1.02 4.1 4.4 21.910 15.190 7.1

...... 02475C o.o 16.955 15.892 1.07 612 6.2 22.005 14.430 10.0 \.0 10184 o.o 16.612 17. 051 0.97 3.0 3.1 21.670 15.490 1.0 e

                                                                        -                                                -

e e Table 4-4 \

)>

"'C ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR "'C tt> COOLANT PUMP CASINGS AT 70F

I

....x 0.

REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL

)>

HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT . Jlc KJc T MOO. CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (%) @ 70F @ 70F (ft-lbs) (lb/in) (ksiVin)

      ** PLANT I. D. PALISADES

• RCP PUMP CASING 661-N-0764 4847 0.0 25.16 3828.0 0.38 24.69 675.2 168.9 71.4 4851 0.0 18.20 5229.0 0.39 37.50 905.4 195.6 89.3 02475C 0.0 7.86 '10979.0 0.43 95.67 1722.0 269.7 141.3 02362A 0.0 8.26 10576.0 0.42 91.46 1734.0 270.6 137.5 03104 6.0 7.36 11520.0 0.43 101.30 1826.0 277.8 147.2 03033A 7.5 7.20 11700.0 0.43 103. 20 1861.0 280.3 149.l 03067A 5.0 7.14 11777 .0 0.43 104.00 1876.0 281.5 150.0 0214C 5.0 6.70 12313.0 0.43 109.60 1982.0 289.4 155.7
• RCP PUMP CASING 661-N-0765 A7824R 12.0 27.63 3508.0 0.38 21.93 607.3 160.2 52.0 B7Hl3R 11.0 18.91 5040.0 0.39 35. 72 858.8 190.5 72.3 4849 0.0 17 .62 5398.0 0.39 39.10 942.8 199.6 99.3 B7Hl4R 13.0 15.91 5956.0 0.40 44.48 984.2 203.9 83.7 4853 0.0 15.01 6295.0 0.40 47.79 1079. 0 213.4 128.2 78143A 5.0 11.66 7941.0 0.41 64.27 1302.0 . 234.5 107.7 6A59A 7.0 11.27 8184.0 0.41 66.75 1346.0 238.5 110. 7 6A57A 9.0 10.53 8680.0 0.41 71.82 1444.0 247.0 116.5 6A58A 7.0 8.80 10066.0 0.42 86.15 1635.0 262.8 131.8 02475C 0.0 7.86 10979.0 0.43 95.67 1722 .0 269.7 141. 3 02362A 0.0 8.26 10576.0 0.42 91.46 1734.0 270.6 137.5 03252 6.0 7.72 11120.0 0.43 97.15 1749.0 271.8 142.8 03104 6.0 7.36 11520.0 0.43 101. 30 1826.0 277 .8 147.2 N

7E34B 6.0 6.75 12249.0 0.43 109.00 1969.0 288.4 155.0 0 03100 5.5 6.45 12626.0 0.43 112. 90 2045.0 293.9 159.0 7H32A 0.0 6.05 13172.0 0.44 118.60 2045.0 293.9 164.1 7H37A 0.0 4.39 15895.0 . 0.45 147.10 2458~0 322.2 190.7

                                                                             .... '-::'...*

Table 4-4 (Continued) ANALYSIS OF. THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS AT 70F REPORT 14

• PREDICTED TOUGHNESS PROPERTIES OF AGED.MATERIAL HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD.

CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F {%) 9 70F @70F (ft-lbs) (1 b/1n) (ksfVf n)

• RCP PUMP CASING 661-N-0766 B7J1R 10.0 24~34 3949.0 0.38 25.76 691.4 170.9 58.1 B7Hl5R 8.0 23.41 4099.0 0.38 27.09 720.4 174.4 60.1 B7H13R 11.0 18.91 5040.0 0.39 35.72 858.8 190.5 72.3 4850 0.0 17.81 5343.0 0.39 38.58 934.8 198.7 102.9 B7H14R 13.0 15.91 5956.0 0.40 44.48 984.2 203.9 83.7 4854 o.o 16.56 5732.0 0.39 42.31 1011.0 206.7 106.4 0337BA o.o 12.70 7358.0 0.40 58.37 1254 .o . 230.2 101.2 033788 0.0 12.70 7358.0 0.40 58.37 1254.0 230.2 101. 2 7F43A 6.0 8.47 10371.0 0.42 89.33 1695.0 267.6 135.2 03252 6.0 7.84 10994.0 0.43 95*.83 1727 .o 270.l 141. 5 03033A 7.5 7.20 11700.0 0.43 103.20 1861.0 280.3 149.1 03067A 5.0 7.14 11777 .o 0.43 104.00 1876.0 281.5 150.0 0214C 5.0 6.70 12313.0 0.43 109.60 1982.0 289.4 155.7 03100 5.5 6.45 12626.0 0.43 112.90 2045.0 "293.9 159.0
• RCP PUMP CASING 661-N-0767 B7J1R 10.0 24.34 3949.0 0.38 25.76 691.4 170.9 58 .1 B7H15R 8.0 23.41 4099.0 0.38 27.09 720.4 174.4 60.1 B7H13R 11.0 18.91 5040.0 0.39 35.72 858.8 190.5 72.3 4848 0.0 *17,79 . 5347 .o 0.39 38.62 941.8 199:5 111. 5 4852 o.o 17.34 5482.0 0.39 39.91 959.1 201.3 100.1 B7H14R 13.0 15."9.l 5956.0 0.40 44.48 984.2 203.9 83.7 7F43A 0.0 8.4~ 10371.0 0.42 89.33 1695.0 267.6 135.2 02475C 0.0 7.86 10979.0 0.43 95.67 1722.0 '269. 7 141.3 03252 6.0 7.84 10994.0 0.43 95.83 1727. 0 270. l 141. 5 N

02362A 0.0 8.26 10576.0 0.42 91.46 1734.0 270.6 137.5 ...... 03100 5.5 6.45 12626.0 0.43 112. 90 2045.0 293.9 159.0 10184 0.0 4.10 16459.0 0.45 153.00 2571.0 "329. 5 196.2

         .e                                                       .e                                                  e

e Table 4-5

)::o                                         ANALYSIS OF THERMAL AGING OF S~AINLESS STEEL

"'C "'C REACTOR COOLANT PUMP CASINGS AT SSOF Cl)

l 0.

-'*

REPORT 15 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x

)::o  HEAT           MEASURED      MATERIAL     MATERIAL     MATERIAL       MINIMUM      MINIMUM     MINIMUM  MINIMUM NO.             FERRITE         AGING     CONSTANT     CONSTANT          IMPACT         Jlc         KJc  T MOD.

CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @550F @550F ( ft-*1 ~s )1 {lb/in) (ks1/1n)

     ** PLANT 1.0. PALISADES

• RCP PUMP CASING 661-N-0764 4847 0.0 25.16 3082.0 0.28 24.69 905.6 159.3 103.6 4851 0.0 18.20 3893.0 0.29 37.50 1116.0 176.8 117.9 0247SC 0.0 7.86 6746.0 0.32 95.67 1784.0 223.6 149.1 02362A 0.0 8.26 6564.0 0.31 91.46 1814.0 225.5 145.1 03104 6.0 7.36 6988.0 0.32 101.30 1861.0 228.3 153.3 03033A 7.5 7.20 7068.0 0.32 103.20 1886. 0 . 229.9 154.7 03067A 5.0 7.14 7102.0 0.32 104.00 1897.0 230.5 155.3 0214C 5.0 6.70 7336.0 0.32 109.60 1973.0 235.1 159.3
• RCP PUMP CASING 661-N-0765 A7B24R 12.0 27.63 2887.0 0.28 21.93 827.9 152.3 71. 2 B7H13R 11.0 18.91 3787.0 0.29 35.72 1064.0 172. 7 90.8 4849 0.0 17.62 3987.0 0.29 39.10 1155.0 179.9 133.6 B7Hl4R 13.0 15.91 4290.0 0.29 44.48 1224.0 185.2 101. 0 4853 0.0 15.01 4471.0 0.29 47~79 1338.0 193.6 174.1 78143A 5.0 11.66 5314.0 Q.30 64.27 1489.0 204.3 121. 5 6A59A 7.0 11.27 5434.0 0.30 66.75 1527.0 206.9 123.7 6A57A 9.0 10.53 5675.0 0.31 71.82 1529.0 207.0 129.0 6A58A 1.0 8.80 6330.0 0.31 86.15 1738.0 220.7 140.9 02475C 0.0 7.86 6746.0 0.3Z 95.67 1784.0 223.6 149.1 02362A 0.0 8.26 6564.0 0.31 91.46 1814.0 225.5 145.1 03252 6.0 ~.7~ 6810.0 0.32 97 .15 1804.0 224.8 150.3 03104 6.0 7.3 6988.0 0.32 lOi.30 1861. 0 228.3 153.3 7E34B 6.0 6.75 7308.0 0.32 109.00 1963.0 234.5 158.9 N

03100 5.5 6.45 7472.0 0.32 112. 90 2015.0 237.6 161. 6 N 7H32A 0.0 6.05 7705.0 0.32 118.60 2092.0 242.I 165.5 7H37A 0.0 4.39 8826.0 0.33 147.10 2349.0 256.5 184.8

Table 4-5 (Continued*) I J:> ANALYSIS OF THERMAL AGING OF STAINLESS STEEL REACTOR ~OOLANT PUMP CASINGS AT 550F Cl)

I a.

 -'*

REPORT 15 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x

 )::>  HEAT           MEASURED      MATERIAL      MATERIAL     MATERIAL      MINIMUM      MINIMUM       MINIMUM*       MINIMUM NO.             FERRITE          AGING     CONSTANT     CONSTANT       IMPACT           Jlc            KJc       T MOD.

CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @550F @ 550F (ft-lbs) (1 b/1 n) (ks1/1n)

• RCP PUMP CASING 661-N-0766 B7J1R 10.0 24.34 3155.0 0.28 25.76 912.3 159.9 77 .0 B7H15R 8.0 23.41 3244.0 0.28 27.09 941.4 162.4 7B.9 B7Hl3R 11.0 18.91 3787.0 0.29 35.72 1064.0 172. 7 90.B 4850 0.0 17 .81 3956.0 0.29 38.58 1151.0 179.6 141.0 B7Hl4R 13.0 15.91 4290.0 0.29* 44.48 1224.0 185.2 101.0 4854 o.o 16.56 4170.0 0.29 42.31 1218.0 184.7 141. 6 03378A o.o 12.70 5022.0 0.30 58.37 1395.0 . 197.7 116.0 033788 o.o 12.70 5022.0 0.30 58.37 1395.0 197.7 116.0 7F43A 6.0 8.47 6470.0 0.31 89.33 1783.0 223.5 143.3 03252 6.0 7.84 6753.0 0.32 95.83 1787.0 223.8 149.2 03033A 7.5 7.20 7068.0* 0.32 103.20 1886.0 229.9 154.7 03067A 5.0 7.14 7102.0 0.32 104.00 1897.0 230.5 155.3 0214C 5.0 6.70 7336.0 0.32 109.60 1973.0 235.1 159.3 03100 . 5.5 6.45 7472.0 0.32 112.90 2015.0 237.6 161. 6
• RCP PUMP CASING 661-N-0767 B7JlR 10.0 24.34 3155.0 0.28 25.76 912.3 159. 9 . 77 .0 B7Hl5R 8.0 23.41 3244.0 0.28 27.09 941.4 162.4 7B.9 B7Hl3R 11.0 18.91 3787.0 0.29 35.72 1064.0 *172. 7 90.8 4848 0.0 . 17. 79 3959.0 0.29 38.62 1162. 0 180.5 157.9 4852 o.o 17 .34 4033.0 0.29 39.91 1170.0 181.0 133.7 B7H14R 13.0 15.*i1 4290.0 0.29 44.48 1224.0 I 185.2 101. 0 7F43A o.o .8. 7 6470.0 0.31 89.33 . 1783.0 .223. 5 143.3 02475C o.o 7.86 6746.0 0.32 95.67 1784.0 '223.6 149 .1 03252 6.0 .7.84 6753.0 0.32 95.83 . 1787.0 223.8 . 149.2 N 02362A 0.0 8.26 6564.0 0.31 91.46 1814.0 225.5 145 .1 w 03100 5.5 6.45 7472.0 0.32 112. 90 2015.0 .237.6 161. 6 9050.0 0.34 153.00 2305.0 254.l 189.5 10184 e.

0.0

                                   .      4.10 e                                                      e

Table 4-6 Limiting and Controlling Values of Jic and KJc at 70°F Jlc KJc RCP PUMP CASING HEAT  ! (lb/in) (ksi/ in) 661-N-0764 4847 675.2 168.9 661-N-0765 A7B24R 607.3 160.2 661-N-0766 B'?JlR 691. 4 170.9 661-N-0767 B7J1R 691. 4 170.9

Table 4-7 Limiting and Controlling Values of J 1 c and KJ c at 550°F Jic KJc RCP PUMP CASING HEAT i (lb/in) (ksi/in) 661-N-0764 4847 905.6 159.3 661-N-0765 A7B24R 827.9 152.3 661-N-0766 B7J1R 912.3 159.9 661-N-0767 B7J1R 912.3 159.9

-- 5.0 CRACK GROWTH ANALYSIS In this section, the methodologies discussed in Section 5.1 of the generic report are applied to plant specific conditions at the Palisades plant. The growth history of a worst-case hypothetical crack is conservatively developed based on information contained in the vendor's (Byron-Jackson Company) stress analysis reports (References 5-1 to 5-4). 5.1 Scope The analysis which follows pertains to the 36x36x38 DFSS Reactor Coolant Pump casings, Serial Numbers 661-N-0764 to 661-N-0767, inclusive, at the Palisades Plant. 5.2 Reference Stress Reports Stress values used in the crack growth analyses were obtained from the original stress analysis reports prepared by the Byron Jackson Company and retrieved from storage at ABB Combustion Engineering Nuclear Power (References 5-1 through 5-4). The Revision 1 documents (References 5-2 and 5-4) were prepared as a result of revised seismic input conditions. Predicted membrane and surface stresses in the limiting regions of the pump casing were revised accordingly, as summarized in Table 5.0 of Reference 5-3. 5.3 Selection of High Stress Locations The methodology described in Section 5.1.1 of the generic report was applied to identify three regions as potentially limiting: (1) Junction, Volute with suction Nozzle Flange (2) crotch Region (3) Discharge Nozzle Appendix A 26

All other regions in the stress summary were considered and ~ were found to have lower stresses than the above regions. 5.4 Stresses and Wall Thicknesses at Limiting Locations Membrane and Through-Wall Bending components of the limiting regions were obtained from References 5-1 through 5-4 as follows: (1) For the Junction of Volute with Suction Nozzle Flange: Key Elements = # 319 & 320 in Finite-Element Model Membrane stress = 22.2 Ksi (p. xi, Table 5.0, Rev. 1) Bending stress = 24.6 Ksi (p. xi, Table 5.0, Rev. 1) Thickness = 4. 75 in. (RS 21343, "Pump Case Shell Measurements"). This is the largest tabulated thickness from cuts 1--9, Azimuthal Locations A--G (2) For the Crotch Region: Key Elements = # 69 & 70 in Finite-Element Model Membrane stress = 21.85 Ksi (p. xi, Table 5.0, Rev. 1) Bending stress = 32.8 Ksi (p. xi, Table 5.0, Rev. 1) Thickness = 4.5 in. (RC 21343, p. 7) from Position S7-8 (3) For the Discharge Nozzle Key Elements = #83, 84, 85, and 86 in finite element model Membrane stress = 26.34 Ksi (p. xi, Table 5.0, Rev. 1) Bending stress = 8.64 Ksi (p. xi, Table 5. 0-, Rev. 1) Thickness = 3.5 in. (estimate) e Appendix.A 27

Surface stresses, from which bending is derived, were provided both with and without the effect of thermal gradients associated with heatup/cooldown. For a conservative analysis procedure, the larger values of associated bending stress were used. 5.5 Calculation of Crack Growth Rates The methodology described in Sections 5.1.4 of the generic report was applied to the above plant-specific conditions, using the annual rate of stress-cycling given in Section 3.2 of this Appendix. An integration procedure was used to predict dimensionless crack depth, a/t, as a function of time, T. Results are summarized in Tables 5-1 through 5-3. For each region the calculated entries are listed against crack depth, a/t, as follows: (1) Applied Stress Intensity Factor, KI' was calculated using the ASME Section XI procedure, as further described in Section 5.1.4 of the generic report. Units for KI are Ksi--squareroot inch. (2) Crack growth rate, da/dT, was calculated using the Bernard & Slama equation (with R=O), multiplied by the annual_ rate of stress cycling (12.55 cycles/year). The final equation is:

                                  -10    4.0 da/dT =  5.403 x 10     KI Units for da/dT are inches/year.

Appendix A 28

Table 5-1 Palisades RCP Casing Crack Growth Rates At suction Nozzle Flange (Junction with Volute) a/t Interval KI da/dT ATime {fraction} (KSI /IN} (IN/YEAR) {YEARS} 0.08 -- 0.10 56.97 5.69 x 10- 3 16.7 0.10 -- 0.15 68.27 1.17 x 10- 2 20.2 (a) 0.15 -- 0.20 77.88 1. 99 x 10- 2 11. 9 (b) 0.20 -- 0.25 86.68 3.05 x 10- 2 7.8 0.25 -- 0.30 94.76 4.36 x 10- 2 5.4 0.30 -- 0.35 102.9 6.06 x 10- 2 3.9 0.35 -- 0.40 111.2 8.26 x 10- 2 2.8 0.40 -- 0.45 120.9 0.116 2.0 0.45 -- 0.50 131.8 0.163 1.4 (a) for Ata = 0.01 Ll°.1 = 27 .3 (b) for Ata = 0.01 +/-T.1 = 14.8 Appendix A 29 Ii.ii;!..=======~:;;,;,;_;;;=;;;;;;.;,,,;:= -------------- - - - - - - - - - - - - - - - -

Table 5-2 Palisades RCP Casing Crack Growth Rates At Crotch Region a/t Interval KI da/dT t:,.Time {fraction {KSI /IN) {IN/YEAR) {YEARS} 0.08 -- 0.10 64.05 9.09 x I0- 3 9.9 a .10 -- a .15 76.26 1.83 x io- 2 12.3 (a) 0.15 -- 0.20 86.45 3.02 x Ia- 2 7.4 (b) 0.20 -- 0.25 95.65 4.52 x io- 2 5.0 0.25 -- 0.30 103.9 6.30 x io- 2 3.5 0.30 -- 0.35 112.2 8.57 x io- 2 2.6 0.35 -- 0.40 120.5 0.114 2.0 0.40 -- 0.45 130.3 0.156 1.4 0.45 -- 0.50 141.4 0.216 1.0 (a) for t:,.ta = 0.01 I:T.1 = 16. 4 (b) for t:,.ta = 0.01 I:T; = 9.1 Appendix A 30

Table 5-3 Palisades RCP Casing Crack Growth Rates At Discharge Nozzle a/t Interval KI da/dT t:..Time {fraction} {KSI Im} {IN/YEAR} {YEARS} 0.08 -- 0.10 38.05 1.13 x 10- 3 61.8 0 .10 -- 0 .15 46.61 2.55 x 10- 3 68.6 0 .15 -- 0. 20 54.34 4.71 x 10- 3 37.1 0.20 -- 0.25 61.60 7.82 x 10- 2 22.3 0.25 -- 0.30 68.83 1.21 x 10- 2 14.4 0.30 -- 0.35 76.11 1.81 x 10- 2 9.6 0.35 -- 0.40 83.8 2.67 x 10- 2 6.5 10- 2 0.40 0.45

     --
     --

0.45 0.50 92.61 102.5 3.97 5.98 x x 10- 2 4.4 2.9 e Appendix A 31

(3) Incremental time, dT, in which the crack will grow through an indicated interval of dimensionless crack depth values, a/t, was calculated as described in Section 5.1.4 of the generic report. Units for dT are years. The summation of time increments yields the total Time for a crack to grow to a given a/t value. The predicted growth curves for hypothetical cracks show the functional relationship between a/t and total Time, using the initial condition of a/t = o.os at Time

          = o. The first incremental time listed is based on a change in a/t in the amount 0.02 (i.e. 2% of thickness), to indicate the time needed for the crack to grow from a/t = 0.08 to a/t = 0.10. Subsequent incremental times are based on the time needed for the crack to grow through changes in a/t of 0.05 (i.e. 5%

of thickness) . The first such incremental time is given for the range a/t = 0.10 to 0.15. The final incremental time is given for the range a/t = 0.45 to a.so, whereupon the analysis is terminated. For the purpose of improving the crack growth histories, the integration process was refined further from a/t = 0.10 to 0.20. In ~his interval, incremental times were re-calculated with 1% steps of a/t, using intermediate values of KI. (Since the crack growth is inherently accelerating, using smaller a/t increments in its initial growth phase avoids an unnecessary over-conservatism). 5.6 Stresses Under Emergency and Faulted Conditions In order to verify that limiting sections containing hypothetical cracks could withstand Emergency Condition and Faulted Conditions Loads the methodology described in Section 5.1.4 of the generic discussion was again applied. Applied stress intensity factors were calculated at the Appendix A 32

limiting locations. The Design Condition stresses were t9 used. Stresses published specifically for Emergency Condition and Faulted Condition Loads were not contained in References 5-1 to 5-4, the available Byron Jackson Company stress reports. The Design Condition stresses are nevertheless considered appropriate for use in this part of the analysis because they (1) include SSE ("Design Basis Earthquake") effects (2) do not include thermal bending stresses, (3) are based on Design Pressure in excess of that which would correspond to an Emergency Condition, and (4) include local surface stresses which would be categorized as peak (Q) values and ordinarily not part of this evaluation. Accordingly, the judgement was made that these stresses could be conservatively utilized in the Emergency Condition and Faulted Condition Loads analysis of applied stress intensity factor and end-point crack size. 5.7 Results Results of the crack growth analysis for the Palisades RCP casings are shown in Figure 5.3-1 of the generic portions of this report. Figur~ 5.3-1 shows that the postulated 8%t initial crack at the crotch region will grow to 25%t in slightly over 40 years under the influence of the conservatively defined stress cycles in the design specification. Calculations also indicate that the hypothesized crack will then grow until it reaches an end-point crack size of 36%t, limited by flow stress, in about 46 years. Appendix A 33

6.0 INSPECTION INTERVAL Results reported in this Appendix support the position that the 10-year volumetric examination interval required by Section XI is not necessary to ensure safe operation during the 40-year licensed life of the plant. The conservatively calculated end-point crack size is not reached until 46 years after initial operation. The demonstrated casing integrity also supports a relaxation of the 10-year interval for visual examinations, as currently required by ASME Section XI and Code Case N-481. Such examinations add unnecessarily to personnel exposure with no significant benefit to plant safety. The ASME Section XI requirement for VT-3 examination of internal surfaces when the pump is disassembled is an appropriate low-marginal-exposure monitoring activity to the extent practical, but only when the pump is disassembled for maintenance or _repair. Based upon the results contained in this evaluation, relaxation of the casing inspection interval for the Palisades RCPs from 10. years to 40 years is considered to be justified. Appendix A 34

APPENDIX A REFERENCES 3-1 Engineering Specification for a Primary Coolant Pump for Consumers Power, 70P-005, Rev. 3. 3-2 Letter B. Kubacki, Consumers Power Company to P. Richardson, ABB C-E Nuclear Power, dated 7/26/91 4-1 o. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in the LWR Systems," NUREG/CR-4513 (ANL-90/42), U.S. Nuclear Regulatory Commission, Washington, D.c., June 1991. 5-1 Byron Jackson, Stress Report Summary, October 1969 5-2 Byron Jackson, Stress Report Summary, Revision 1, May 1970 5-3 Byron Jackson, Stress Report Volume IIA, Pump Case, May 1969 5-4 Byron Jackson, Stress Report Volume IIA, Pump Case, Revision 1, November 1969 Appendix A 35

APPENDIX B APPLICATION OF GENERIC METHODOLOGY FOR RELAXATION OF THE FORT CALHOUN REACTOR COOLANT PUMP CASING INSPECTION INTERVAL Appendix B 1

ABSTRACT Appendix B was prepared to demonstrate the amount of inspection interval relaxation appropriate for the reactor coolant pump casings at the Fort Calhoun plant, based on application of the generic methodology presented in the main body of this report. Appendix B 2

APPENDIX B TABLE OF CONTENTS Section Title Page

1. 0 PURPOSE 5 2.0 PRE-SERVICE INSPECTION DATA EVALUATION 6 3.0 OPERATING HISTORY 8 3.1 Design Specifications 8 3.2 Stress Cycles Used In Evaluation 9 3.3 Stress Cycles at Fort Calhoun To-date 9 4.0 THERMAL EMBRITTLEMENT 10 4.1 Material Identification and Chemical Properties 10 4.2 Material Specifications and Mechanical Properties 10 4.3 Thermal Aging Behavior 11 4.4 Toughness Properties of Aged Materials 11

4. 5 Limiting V~_lues 12 5.0 CRACK GROWTH ANALYSIS 30 5.1 sc_ope 30 5.2 Reference Stress Reports 30 5.3 Selection of High Stress Locations 30 5.4 Stresses and Wall Thicknesses at Limiting Locations 31 5.5 Calculation of crack Growth Rates 32 5.6 Stresses Under Emergency and Faulted Conditions 37 5.7 Results 39 6.0 INSPECTION INTERVAL 40 APPENDIX B REFERENCES 41 Appendix B 3

APPENDIX B LIST OF TABLES TABLE 4-1 Material Identification and Chemical Compositions 13 TABLE 4-2 Material Specifications and Tensile Pl;'operties 16 TABLE 4-3 Predicted Thermal Aging Behavior 19 TABLE 4-4 Predicted Toughness Properties of Aged Materials (70°F) 22 TABLE 4-5 Predicted Toughness Properties of Aged Materials (550°F) 25 TABLE 4-6 Limiting and Controlling Values of Jic and KJc at 70°F 28 TABLE 4-7 Limiting and Controlling Values of Jic and KJc at 550°F 29 TABLE 5-1 Crack Growth Rate at Junction of Scroll and suction Nozzle Flange 33 TABLE 5-2 Crack Growth Rate at Scroll 34 TABLE 5-3 Crack Growth Rate at Junction of Scroll and Bolt Circle Flange 35 TABLE 5-4 Crack Growth Rate at Diffuser Vane Number 7 36 Appendix B 4

1.0 PURPOSE The purpose of Appendix B is to document application of the methodology presented in_ the generic portion of this report to the plant-specific data for the reactor coolant pump casings at the Fort Calhoun plant, and to quantify the resulting extent of inspection-interval relaxation available. Appendix B 5

2.0 PRE-SERVICE INSPECTION DATA EVALUATION Pre-service inspection data for the Fort Calhoun reactor coolant pumps numbered 671-N-0029 through 671-N-0032 was collected from QA data packages originally* prepared by the Byron Jackson Company and stored in archives by ABB Combustion Engineering Nuclear Power. Information in these data packages concerning welding procedures, radiographic inspections, non-destructive testing and dye penetrant testing were examined. The testing and inspection procedures that were followed for all reactor coolant pumps at Fort Calhoun were found to be the same in all significant aspects. The most relevant information obtained from this review of the QA data packages were the reports on radiographic examination of the RCP casing castings, pressure retaining welds, and repair welds. Radiographic examination requirements invoked ASME Section III rules for examination procedures and sensitivity. The required radiograph sensitivity was 2-2T according to applicable ASTM Standard Reference Radiograph Procedure requirements {i.e. ASTM E71, El86, E280) as determined by the casting thickness. The 2-2T sensitivity is consistent with a 2% initial flaw size, because the requisite image quality indicator {IQI) for this level of examination is specified as a penetrameter with a minimum hole-size diameter equal to 2% of the casting thickness. The acceptance criteria for interpretation of the radiographs was Severity Level 2 for sand, porosity or shrinkage indications. Linear indications such as cracks, hot tears, and unfused chaplets or chills were unacceptable at any level. Any such discernable indications required rejection of weld repair and a repeated radiographic examination of the affected casting or weldment. Appendix B 6

The results of this review of pre-service RCP casing examinations confirm the assumed detectc. *.le flaw of 2% thickness described in Section 4.4 of the generic report. Appendix B 7

3.0 OPERATING HISTORY 3.1 Design Specifications The Fort Calhoun RCPs were delivered to the site in 1968 and were first placed in commercial operation in 1973. Reactor coolant system design pressure and temperature are 2500 psia and 650"F respectively. Each pump is designed to deliver 47,500 gpm of coolant at a head of 260 feet. These pumps have 24 inch diameter suction and discharge piping. The design specification (Reference 3-1) calls for the pumps to be capable of withstanding the following transient conditions during their 40-year license period: Assumed Occurrences Transient During 40 Year Condition License Period Heat-Up (lOO"F/hr) 500 Cool-Down (lOO"F/hr) 500 Hydrostatic Test (3125 psia 100-400.F) 10 Leak Test - In Conjunction with Heatup (2100 psia 100-400.F) 200 Loss of Secondary Pressure 5 Reactor Trip or Loss of Load 400 Appendix B 8

3.2 Stress Cycles Used in Evaluation As indicated in Section 5.1.4 of the generic portion of this report, crack growth was evaluated on the basis of an assumed number of stress cycles between atmospheric and operating pressures during heatup and cooldown over the nominal 40-year life of the plant. The number of such cycles used in the stress analyses performed by the Byron Jackson Company is 715, as given on page 285 of Reference 3-2. This total is slightly greater than the 700 heat-up cooldown and leak test cycles specified in the RCP design specifications. on an annual basis, the average number of stress cycles, based on 715 per 40 years, is 17.875 per year, and the hypothetical crack growth calculations and. curves were prepared accordingly. 3.3 Stress cycles at Fort Calhoun To-date 9 Details of the actual operating history of the Fort Calhoun e. plant from 1973 to 1991 were furnished in Reference 3-3 and are as follows: Heatup - 44 Cooldown - 43 Leak Test - 36 Loss of Secondary Pressure - O Heatup-plus-cooldown, taken together, constitute one cycle. When these are added to the leak test cycles a total of 80 is obtained over the 18 year operating period ending in mid 1991 (4.44 cycles per year). This time period is equivalent to 4.5 years at the assumed rate of stress cycling (17.875 per year). The actual rate of cycle accrual is seen to be only 25% of the design rate; a significant conservatism in the evaluation. Appendix B 9

4.0 THERMAL EMBRITTLEMENT Thermal embrittlement evaluation of the Fort Calhoun casings is discussed and plant specific data are presented in the five following reports. All equations referenced below are found in the main body of this report, which is also referred to as the generic report. 4.1 Material Identification and Chemical Properties The chemical compositions provided in Report #1 (Table 4-1) for each RCP casing at Fort Calhoun were obtained from Quality Assurance documents originally supplied by the Byron Jackson company and stored at ABB Combustion Engineering Nuclear Power. For each individual pump casing, chemical compositions are given for specific casing welds as well as for individual castings. 4.2 Material Specifications and Mechanical Properties The material specifications and mechanical properties found in Report #2 (Table _4-2) for each RCP casing were obtained from the same data source as in Section 4.1 above. For each individual pump.casing the material specification, material type and heat number are given for specific casing welds as well as for individual castings. It is evident from the report that data obtained for mechanical properties (i.e. yield strength, tensile strength, total elongation and reduction in area) for each material was only available for the castings, and was not available for the casing welds. The unaged flow stress at 70°F and the aged flow stress at 70°F and 550°F were calculated ~s discussed in Section 5.2.3 of the generic report. Appendix B 10

4.3 Thermal Aging Behavior Report #3 (Table 4-3) contains predicted therrr.~l aging behavior data for all of the Fort Calhoun RCP casings. The measured ferrite contents listed for specific casing welds and individual castings were supplied by the Byron Jackson Company in the same QA package as referenced in Section 4.1 above. In most cases a value was obtainable for the measured ferrite content. In cases where a value was not given, a zero was recorded. -The chromium and nickel equivalents for the castings and weld metal, as well as the chromium/nickel ratio for the castings, were calculated using equations 5-7 and 5-8 respectively, as discussed in Section 5.2.l of the generic report. Values for ferrite content of the castings were computed using two methods: for ferrite content #1, the values were computed using the method which follows ASTM ASOO/ASOOM (Reference 5-2 of generic report); for ferrite content #2, the values were computed using equation 5-9 as discussed in Section 5.2.1 of the generic report. The latter method follows work performed by O.K. Chopra (Reference 4-1). 4.4 Toughness Properties of Aged Materials The predicted toughness properties of aged material at 70°F and 55o*F are given respectively in Report #4 (Table 4-4) and Report #5 (Table 4-5). The measured ferrite contents listed for all heat numbers are the same as the values given in Report #3. The material aging parameter was calculated using equation 5-10 of the generic report. The room-temperature charpy impact energy, Cvsat, of the various materials was calculated using equation 5-11. The Jic values were determined in accordance with the methods of ASTM E813 as discussed in Section 5.2.3. The plane strain fracture toughness, KJc' and minimum tearing modulus, T, at Appendix B 11

70°F and 550°F were calculated using equations 5-17 and tt 5-18. The values listed for the material constants N and c at 70°F and 550°F were calculated using equations 5-12 through 5-15. These constants were needed in computing the values for Jic' KJc and T. All equations used in Report #4 are found in Section 5.2. 4.5 Limiting Values The limiting and controlling values for Jic and KJc at 70°F and 550°F for each individual pump at Fort Calhoun are given in Tables 4-6 and 4-7. Appendix B 12

                       ---------------- - - - - - -   -

e e ANALVIS OF THERMAL AGING OF CAST STAINLESS STEEL

i

:.

  "'O      Table 4-1                                                     REACTOR COOLANT PUMP CASINGS
  "'O CD

I

a. REPORT 11 G MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS

   ..4, x

OJ MATERIAL HEAT NO. c Mn Si s p Cr Ni Mo N Cb

                     ** PLANT I.D. FT. CALHOUN

• RCP PUMP CASING 671-N-0029 CASE ASSEM. 516942 0.06 1.00 1.04 0.017 0.031 20.54 9.66 2.42 0.04 0.00 CASING WELD 03233 0.06 1.65 0.43 . 0.021 0.011 20.00 11. 76 2.23 0.04 0.00 CASING WELD 03314 0.06 1.66 0.45 0.020 0.009 19.80 11.86 2.35 0.04 0.00 CASE ASSEM. 714649 0.06 1.21 0.59 0.018 0.027 18.88 10.82 2.24 0.04 0.00 CASING WELD 02318A 0.06 ' 1.67 0.53 0.016 0.010 18.86 11.52 2.30 0.04 ' 0.oo CASING WELD 03107 0.06 1.72 0.44 0.015 0.010 19.12 12.20 2.10 0.04 0.00 CASING WELD 03336 0.06 1. 74 0.58 0.020 0.010 19.65 12.38 2.39 0.04 0.00 CASING WELD 02370A 0.06 1.65 0.49 0.020 0.010 19.02 11.54 2.29 0.04 0.00 CASING WELD 024758 0.04 1.66 0.56 0.019 0.009 19.33 11.52 2.41 0.04 0.00 CASING WELD A7719 0.04 1.80 0.31 0.012 0.005 19.11*12.19 2.29 0.04 0.00 CASING WELD 02362A 0.03 1. 72 0.48 0.018 0.022 19.10 11.55 2.29 0.04 o.oo CASING WELD 03252 0.04 1.46 0.44 0.015 0.011 18.99 11.80 2.24 0.04 0.00 CASING WELD 02475A 0.03 1.82 0.59 0.022 0.011 18.93 11.88 2.48 0.04 0.00 CASE ASSEM. 714637 0.04 1.32 0.59 0.020 0.031 18.26 11.10 2.40 0.04 0.00 CASING WELD 9208F 0.02 2.21 0.42 0.011 0.008 19.32 12.34 2.45 0.04 0.00 CASING WELD 9358F 0.03 1.67 0.38 0.010 0.010 18.80 12.70 2.40 0.04 0.00 CASING WELD A9663F 0.02 1.94 0.39 0.008 0.012 19.*.05 12.45 2.32 0.04 0.00 CASING WELD 8878 0.02 1. 72 0.32 0.013 0.005 18..96 12.19 2.30 0.04 0.00 CASING WELD D8122 0.02 1.86 0.41 0.009 0.019 18.95 12.37 2.27 0.04 0.00 CASING WELD A7490 0.02 1.85 0.35 0.009 0.005 18.93 12.30 2.16 0.04 0.00 CASING WELD A9872F 0.02 1.95 0.38 0.013 0.011 191.33 *12.22 2.33 0.04 0.00 CASING WELD 022918 0.03 1.85 0.57 0.038 0.009 18.69 12.48 2.23 0.04 o.oo CASING WELD A8025.'. 0.02 1. 75 0.31 0.011 0.005 18.74 12.15 2.38 0.04 0.00 CASING WELD 02~12p 0.03 1.93 0.51 0.015 0.010 18.25 11.53 2.19 0.04 0.00 CASING WELD A8626' 0.02 1.92 0.33 0.009 0.007 19.16 12.55 2.29 0.0.4 o.oo CASING WELD 9599F 0.02 1.87 0.39 0.009 0.011 rn.79 12.44 2.11 0. 0'4 0.00

..... w

fi '/ e i ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL* REACTOR COOLANT PUMP CASINGS I ):::> Table 4-1 (Continued)

     "'O
     "'O m

REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS

I i a.

      -'*

x. MATERIAL i I I O:J HEAT NO. c Mn Si s p Cr Ni Mo N Cb CASING WELD A9310F 0.02 2.01 0.32 0.008 0.008 19.22 12.67 2.84 0.04 0.00

• RCP PUMP CASING 671-N-0030 CASING WELD 03314A 0.06 1.56 0.44 0.023 0.010 19.82 11.80 2.40 0.04 o.oo CASE ASSEM. 714737 0.07 1.04 0.67 0.023 0.033 18.63 11.09 2.40 0.04 0.00 CASING WELD 01765 0.03 2.32 0.36 0.028 0.010 23.44 12.84 0.00 0.04 o.oo CASE ASSEM. 517116. 0.05 1.06 0.35 0.020 0.038 19.19 9.73 2.23 0.04 0,.00 CASE ASSEM. 714334 0.04 0.66 0.38 0.020 0.032 19.29 9.73 2.29 0.04 o.*oo CASING WELD 03314 0.06 1.66 0.45 0.020 0.009 19.80 11.86 2.35 0.04 o.oo CASE ASSEM. 714649 0.06 1.21 0.59 0.018 0.027 18.88 10.82 2.24 0.04 0.00 CASING WELD 03107 0.06 1. 72 0.44 0.015 0.010 19.12 12.20 2.10 0.04 0.00 CASING WELD L933 0.04 0.77 0.95 0.012 0.011 19.80 8.30 0.00 0.04 0.00 CASING WELD 03252 0.04 1.46 0.44 0.015 0.011 18.99 11.80 2.24 0.04 0.00 CASING WELD 01746 0.02 1.69 0.42 0.012 0.011 20.99 9.47 0.00 0.04 o.oo CASE ASSEM.
• 714637 0~04 1.32 0.59 0.020 0.031 18.26 11.10 2.40 0.04 0.00 CASING WELD 03681 0.03 1.80 0.47 0.016 0.010 19. 95* 9.90 o.oo 0.04 0.55 CASING WELD 02964E 0.04 1. 73 0.52 0.016 0.012 19.22 9.40 0.00 0.04 0.66 CASING WELD 01124K 0.03 1.90 0.47 0.026 0.009 19.56 10. 71 0.00 0.04 0.51
• RCP PUMP CASING 671-N-0031 CASE ASSEM. 714737 0.07 1.04 0.67 0.023 0.033 18.63 11.09 2.40 0.04 0.00 CASE ASSEM. 517312 0.07 1.30 0.75 0.023 0.032 18.98 10.92 2.21 0.04 0.00 CASE ASSEM. 517246 0.06 1.28 0.80 0.023 0.028 18.54 10.63 2.40 0.04 0.00 CASING WELD L933 . . 0.04 0.77 0.95 0.012 0.011 19.80 8.30 o.oo 0.04 0.00 CASING WELD 752627 0.02 1.66 0.39 0.010 0.009 19.50 13.00 2.28 0.04 o.oo
• RCP PUMP CASING 671-N~0032 CASE ASSEM. 517571 * ~0.06 1.30 0.69 0.017 0.029 19.79 10.80 2.12 0.04 o. oo CASE ASSEM. 714737 0.07 1.04 0.67 0.023 0.033 18.63 11.09 2.40 0.04 o:oo

   ......
   .,i::.

  ):> Table 4-1 (Continued)
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l a.

                 . .                             ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL x
  -'*

REACTOR COOLANT PUMP CASINGS OJ REPORT #1.

• MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS MATERIAL HEAT NO. c Mn Si s p Cr Ni Mo N Cb CASING WELD 01765 0.03 2.32 0.36 0.028 0.010 23.44 12.84 o.oo 0.04 o.oo CASE ASSEM. 714334 0.04 0.66 0.38 0.020 0.032 19.29 9.73 2.29 0.04 0.00 CASE ASSEM. 517312 0.07 1.30 0.75 0.023 0.043 18.98 10.*92 2.21 0.04 0.00 CASE ASSEM. 517246 0.06 1.28, 0.80 0.023 0.028 18.54 10.63 2.40 0.04 0.00 CASING WELD L933 0.04 0.77 0.95 0.012 0.011 19.80 8.30 0.00 0.04 0.00 CASING WELD 01746 0.02 1.69 0.42 0.012 0.011 20.99 9.47 o.oo 0.04 0.00 CASING WELD 03100 0.03 1.66 0.40 0.018 0.010 19.06 12.20 2.25 0.04 0.00 CASING WELD 03681 0.03 1. 73 0.52 0.016 0.010 19.95 9.90 o.oo 0.04 0',55 CASING WELD 02964E 0.03- 1. 73 0.52 0.016 0.012 19.22 9.40 0.00 0.04 0.66 CASING WELD 01124K 0.03 1.90 0.47 0.026. 0.009 19.56 10.71 0.00 0.04 0.51

...... U1

e e ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS

)::o "t:J Table 4-2

• REPORT 12
• MATERIAL SPECIFICATION & TENSILE PROPERTIES

-c m

I MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED a.

.....                                                                                                FLOW     FLOW x

OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW PART HEAT NO. (ksi) (ks1) (%) (%) STRESS STRESS STRESS C:J

                                                                                            @ 70F   @ 70F  @ 550F
      ** PLANT I.D. FT. CALHOUN

• RCP PUMP CASING 671-N-0029 CASE ASSEM. A 351 CF8M 516942 45.4 87.9 30.5 38.7 67 81282 60688 CASING WELD E316-16 03233 0.0 o.o o.o 0.0 0 78600 58000 CASING WELD E316-16 03314 o.o o.o . o.o o.o 0 78600 58000 CASE ASSEM. A 351 CF8M 714649 . 37 .3 71 ~9 52.5 58.9 55 62203 41609 CASING WELD E316-16 02318A o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-16 03107 o.o 0.0 o.o o.o 0 78600 58000 .

CASING WELD E316-16 03336 o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-16 02370A o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-16 024758 o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-16 A7719 o.o 0.0 o.o o.o 0 78600 58000 CASING \ilELD E316-16 02362A o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-16 03252 o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-16 02475A o.o o.o o.o 0.0 0 78600 58000 CASE ASSEM. A 351 CF8M . 714637 32.8 71.2 49.6 49.2 52 58087 37493 CASING WELD E316-16 9208F 0.0 o.o o.o o.o 0 7860.0 58000 CASING WELD E316-16 9358F o.o o.o. o.o o.o 0 78600 58000 CASING. WELD. E316-16 A9663F 0.0 0.0 o.o o.o 0 78600 58000 CASING WELD E316-16 8878 . o.o 0.0 o.o o.o 0 78600 58000 CASING WELD E316-16 D8122 0.0 0.0 0.0 o.o 0 78600 .:58000 CASING WELD E316-16 A7490 o.o o.o 0.0 0.0 0 78600 58000 CASING WELD E316-16 A9872F o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-16 022918 o.o 0.0 0.0 o.o 0 78600 58000 CASING WELD E316-16 A8025 0.0 o.o o.o o.o 0 78600 58000 CASING WELD E316-16 029128~ o.o o.o 0.0 o.o 0 78600 58000 CASING WELD E316-16 A8626 ~ o.o o.o o.o o.o 0 78600. 58000 CASING WELD E316-16 9599F o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-16 A9310F o.o o.o o.o 0.0 0 78600 58000 ...... OI

                                '*'

e ..* ___ . Table 4-2 (Continued)

):o "t:J ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL "t:J CD REACTOR COOLANT PUMP CASINGS

I 0..

-'*

x REPORT 12

• MATERIAL SPECIFICATION &TENSILE PROPERTIES co MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                               @70F    @ 70F   @ 550F

• RCP PUMP CASING 671-N-0030.

CASING WELD E316*16 03314A o.o o.o o.o o.o 0 78600 58000 CASE ASSEM. A 351 CF8M 714737 37.9 72.9 50.0 74.0 55 63470 42876 CASING WELD X5620S 01765 o.o o.o o.o o.o 0 78600 58000 CASE ASSEM. A 351 CF8M 517116 38.0 78.2 48.0 75.0 58 67745 47151 CASE ASSEM. A 351 CF8M 714334 39. l 76.7 49.8 68.7 58 67428 46834 CASING WELD E316*16 03314 o.o o.o 0.0 o.o 0 78600 58000 CASE ASSEM. A 351 CFSM 714649 36.0 73.2 52.0 77.0 55 62203 41609 CASING WELD E316*16 03107 o.o o.o o.o 0.0 0 78600 58000 CASING WELD A.362 CF8M l933 o.o. o.o o.o o.o 0 78600 58000 CASING WELD E316-16 03252 o.o o.o o.o o.o .o 78600 58000 CASING WELD E347 01746 o.o o.o o.o o.o 0 78600 58000 CASE ASSEM. A 351 CF8M 714637 31.6 71. 7 50.0 75.0 52 57532 36938 CASING WELD E347-16 03681 o.o o.o o.o o.o 0 78600 58000 CASING WELD E347-16 02964E o.o o.o o.o o.o 0 78600 58000 CASING WELD E347-16 01124K o.o 0.0 o.o o.o 0 78600 58000

• RCP PUMP CASING 671-N-0031 CASE ASSEM. A 351 CF8M 714737 35.0 72.-9 56.0 74.5 54 61174 40580 CASE *ASSEM. A 351 CF8M 517312 38.0 71.7 57.0 74.9 55 62599 42005 CASE ASSEM. A 351 CF8M 517246 34.0. 73.0 58.0 68.0 54 60462 39868 CASING WELD A 362 CF8M l933 o.o o.o 0.0 o.o 0 78600 .: 58000 CASING WELD E316-16 752627 o.o 0.0 o.o o.o 0 78600 58000

                                                 ~

• RCP PUMP CASING 671-N-0032 ~  :

CASE ASSEM. A 351 CF8M 517571 39.0 75.0 46.9 69.0 57 66003 45409 CASE ASSEM. A 351 CF8M 714737 36.0 71.0 55.0 66.3 54 60462 39868

 ......
 ......

CASING WELD XS620S 01765 o.o o.o o.o o.o 0 78600 58000 CASE ASSEM. A 351 CF8M 714334 36.4. 72.7 60.0 75.0 55 62124 41530 CASE ASSEM. A 351 CF8M 517312 38.0 71.0 55.5 65.8 54 62045 41451 CASE-EM. I\ 151 CFSM . 517246 42.0 74- 49.9 70.7 58 67586 46392 e

                                                                                 ~  -- .

e .e Table 4-2 (Continued)

)>

"O ANALYSIS OF THERMAL AGING OF CAST STAINL~SS STEEL "O rt> REACTOR COOLANT PUMP CASINGS

I 0..

-lo
><

REPORT 12 - MATERIAL SPECIFICATION &TENSILE PROPERTIES OJ MATERIAL MATERIAL MATERIAL YIELD TENSILE . TOTAL RED. IN UNAGED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA . FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                         @ 70F   @ 70F    @ SSOF CASING WELD A 362    CF8M      L933           0.0        0.0      o.o      o.o        0   78600     58000 CASING WELD          E347      01746          0.0        0.0      0.0      o.o        0   78600     58000 CASING WELD          E316*16   03100          o.o        o.o      o.o      0.0       *O   78600     58000 CASING WELD          E347-16   03681          o.o        0.0      o.o      o.o        0   78600     58000 CASING WELD          E347*16   02964E         o.o        o.o      o.o      o.o        0   78600     58000 CASING WELD          E347-16   01124K         o.o        0.0      o.o      0.0        0 . 78600     58000
                                                                                                      ' .
 .....

00

              . I

e e e

  )::>
 "'C ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL
 "'C Cl>               Table 4-3                                REACTOR COOLANT PUMP CASINGS

s 0..

I x' REPORT 13 - PREDICTED THERMAL AGING BEHAVIOR OJ . HEAT MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS 11 (%) 12 (%) WELDS WELDS (%}

          ** PLANT I.D. FT. CALHOUN

• RCP PUMP CASING 671-N-0029 516942 20.0 18.977 14.639 1.30 18.4 21.5 24.520 13.040 17.0 03233 7.0 17. 915 16.968 1.06 5.1 5.8 22.875 15.675 6.2 03314 6.5 17.870 16.995 1.05 5.1 5.6 22.825 15.690 6.0 714649 8.0 16.884 15.868 1.06 6.2 6.1 22.005 I 14 .365 10.0 0231BA 6.0 16.907 16.729 1.01 3.8  : 4.1 21.955 15.445 6.8 03107 5.5 16;882 17 .340 . 0.97 2.5 3.1 21.880 16.060 . s.0
• 03336 5.0 17.830 17.521 1.02 4.1 4.4 22.910 16.250 5.2 02370A 6.0 17.036 16.601 1.03 4.5 4.7 22.045 15.275 7.6 024758 8.0 17: 525 16.042 1.09 7.0 7.4 22.580 14.600 11.5 A7719 7.0 17 .040 16.797 1.01 3.7 4.3 21.865 15.430 7.4 02362A 8.0 17.111 15.955 1.07 6.2 6.5 22 .110 14.510 9.9 03252 6.0 16.922 16.551 1.02 4.5 4.5 21.890 15.080 7.9 02475A 7.0 17.224 16.391 1.05 5.5 5.6 22.295 15.010 9.2 714637 r 70 o 16.457 15.790 1.04 5.5 5.2 21.545 14.250 9.2 9208F 0.0 17. 496 16.537 1.06 5. l .. : 5.9 22.400 15.245 9.2 9358F 6.0 16.896 16.978 1.00 3.4 *. 3.7 21. 770 15.485 6.8 A9663F 6.0 17 .054 16.701 1.02 4.1 4.5 21.955 15.310 7.6 8878 7.0 16.907 16.301 1.04 4.8 . 5.0 21. 740 14.790 8.0 D8122 6.0 16.904 16.590 1.02 4.1 4.4 21.835 15.160 7.6 A7490 6.0 16.722 16.519 1.01 3.8 *4.2 21.615 15.085 7.o .

A9872F 6.0 17 .342 16.398 1.06 5.1 5.9 22.230 14.995 9.2 022918 5.0 16.672 16.7.73 0.99 3.4 3.6 21. 775 15.355 7:.4 A8025 6.0 16. 779 16.312 1.03 4.5 4.7 21.585 14.825 8.2 029128 5.0 16.155 15.853 1.02 4.1 4.4 21.205 14.475 6.6

  ...... AB626           5.0   17 .099  16.750       1.02        4.1           4.5 21.945       15.340         7.6 l.D 9599F           5.0   16.540   16.636       0.99        3.1           3.6 21.485       15.205         6.2 A9310F          5.0   17.820   16.828       1.06        5.5           5.9 22.540       15.445        8.8 e                                                         e                                           e

                                  --- ------

e Table 4-3 (Continued)

)>

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL "'C -0 REACTOR COOLANT PUMP CASINGS fl)

I 0.

-'*

REPORT 13 - PREDICTED THERMAL AGING BEHAVIOR x aJ HEAT MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS 11 (%) 12 (%) WELDS WELDS (%)

• RCP PUMP CASING 671-N-0030 03314A 8.0 17 .945 16.927 1.06 5.5 6.0. 22.880 15.580 10.8 714737 8.0 16.866 16.416 1.03 5.1 4.7 22.035 14.910 9.0 01765 9.0 18.623 17.216 1.08 4.8 6.9 23.980 16.010 12.5 517116 12.0 17 .066 14.617 1.17 9.7 11.6 21.945 13.020 13.6 714334 13.0 17.253 14.407 1.20 11.8 13.6 22.150 12.610 14.5 03314 6.5 17.870 16.995 1.05 5.1 5.6 22.825 I 15.690 6.0 714649 8.0 16.884 15.868 1.06 6.2 . : 6.1 22.005 14.365 . 10.0 03107 5.5 16.882 17.340 0.97 2.5 3.1 21.880 16.060 5.8 l933 o.o 15.266 12.866 1.19 11.8 12.9 21.225 11.085 15.3 03252 . 6.0 16.922 16.551 1.02 4.5 4.5 . 21.890 15.080 7.9 01746 0.0 16.202 13.652 1.19 9.3 12.9 21.620 12.145 14.1 714637 7.0 16.457 15.790 1.04 5.9 5.2 21.545 14.250 9.2 03681 0.0 15.186 14.311 1.06 5.9 6.0 20.930 12.900 10.0 02964E o.o 14.480 13.977 1.04 5.5 5.0 20.330 12.575 8.* 4 Oll24K o.o 14.796 15.227 0.97 2.7 3.0 20.520 13.880 5.9
• RCP PUMP CASING 671-N-0031 714737 8.0 16.866 16.416 1.03. 5.1 4.7 22.035 14.910 9.0 517312 7.0 17 .024 16.220 1.05 5.9 5.5 22.315 14.810 6.1 517246 8.0 16.838 15.684 1.07 . 7.0 . 6. 5 22.140 14.210 10.7 L933 0.0 15.266 12.866 1.19 11.8 12.9 21.225 11. 085 15.3 ..

752627 0.0 17 .456 17.228 1.01 4.1 *4,2 22.365 15.720 6.6

• RCP PUMP CASING 671-N-0032 517571 10.0 17.696 15.904 1.11 8.1 8.4 22.945 14.450 13:0 714737 a.a 16.866 16.416 1.03 5.1 4.7 22.035 14.910 9.0 01765 9.0 18.623 17.216 1.08 4.8 6.9 23.980 16.010 12.5 N

0 714334 13.0 17.253 14.407 1.20 11.8 13.6 22.150 12.610 14.5 517312 7.0 17.024 16.220 1.05 5.9 5.5 22.315 14.810 6.1 5172- 8.0 16.838 15.684 1.07 7.0 e 6~5 22.140 14.210 10.7

)>

""C Table 4-3 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL ""C CD REACTOR COOLANT PUMP CASINGS

s 0..

......                                           . REPORT 13

• PREDICTED THERMAL AGING BEHAVIOR

><

OJ HEAT MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS 11 (%) 12 (%) WELDS WELDS (%) L933 o.o 15.266 . 12 .866 1.19 11.8 12.9 21.225 11.085 15.3 01746 0.0 16.202 13.652 1.19 9.3 12.9 21.620 12.145 14.1 03100 5.5 16.984 16.649 1.02 4.1 4.4 21.910 15.190 7.7 03681 o.o 15.210 14.306 1.06 6.2 6.1 21.005 12.865 10.6 02964E 0.0 14.480 13.806 1.05: 5.9 5.5 20.330 12.365 8.4 01124K 0.0 14.796 15.227 0.97 2.7 3.0 20.520 13.880 5.9 N

.......

I '

                                                                                                                            ---------

e Table.4-4

  )::>>
 -0 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

. -0 fl) COOLANT PUMP CASINGS AT 70F

I 0..

  ......                                     REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x

a:I HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOO. CONTENT PARAMETER c N ENERGY @70F @70F @70F (%) 9 70F @70F (ft- lbs) ( 1b/1n) (ks 1V1n")

         ** PLANT I. D. FT. CALHOUN

• RCP PUMP CASING 671-N-0029 516942 20.0 36.34 2773.0 0.37 15.85 495.1 144.6 39.2 03233 7.0 13.57 6920.0 0.40 53.98 1168.0 222.1 95.8 03314 6.5 12.78 .7315.0 0.40 57.93 1245.0 229.3 100.7 03336 5.0. 10.93 8405.0 0.41 69.00 1389.0 242.3 113.2 024758 8.0 10.67 8582.0 0.41 70.81 1423.0 245.2 115.3 714649 8.0 11.14 8268.0 0.41 67.60 1424.0 245.2 172.7 02370A 6.0 9.35 9585.0 0.42 81.17 1542.0 255.2 126.3 02318A 6.-0 8.93 9943.0 0.42 84.87 1610.0 260.8 130.4 03252 6.0 7.84 10994.0 0.43 95.83 1727.0 270.1 141.5 02362A. 8.0 8.26 10576.0 0.42 91.46 1734.0 270.6 137.5 02475A 7.0 8.21 10621.0 0.42 91.93 1744.0 271.4 138.0 03107 5.5 7.15 11758.0 0.43 103.80 1873.0 281.3 149.8 714637 7.0 1.12* 11127 .o 0.43 97.22 1885.0 282.2 248.6 A7719 7.0 6.86 12114.0 0.43 107.60 1941.0 286.4 153.6 A9310F 5.0 6. 71 12298.0 0.43 109.50 *1980.0 289.2 155.5 9208F 8.0 6.65 12376.0 0.43 110.30 1995.0 290.3 156.3 A9872F 6.0 6.47 12610.0 0.43 112. 70 2042.0 293.7 158.9 A9663F 6.0 . 5.49 14009.0 0.44 127.40 2207.0 305.4 172.6 A8025 6.0 4.96 14863.0 0.45 136.30 2256.0 308.7 180.5 029128 5.0 4.89 14978.0 0.45 137.50 2280.0 310.3 181. 7 A8626 5.0 5.29 14377.0 0.44 131.20 2282.0 310.5 176.3 9358F 6.0 4.8f 15098.0 0.45 138.80 2302.0 311.8 182.8 A7490 6.0 4.81 15118.0 0.45 139.00 2305.0 312.0 183.0 08122 6.0 5.18 14503.0 0.44 -132.60 2309.0 312.3 177. 7 N

N 022918 5.0 4.73 15265.0 0.45 140.50 2334.0 314.0 184.6 8878 7.0 5.09 14639.0 0.44 134.00 2334.0 314.0 179.0 9599*.. 5.0 4.16 16344.0 *

• 45 151.80 2549.0 328.1 195.1 e

Table 4~4 (Continued)

                                                                                                                    --

t>

"'C ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR "'C m . COOLANT PUMP CASINGS AT 70F

I a.

-'*
><

REPORT 14 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL o:J HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACl Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @70F @70F (%) @70F @70F (ft-lbs) ( l b/1n) (ks1V1n)

• RCP PUMP CASING 671-N-0030 714334 13.0 16.38 5793.0 0.39 42.90 1026.0 208.1 109.9 517116 12.0 15.41 6140.0 0.40 46.27 1041.0 *209. 7 114.1 03314A 8.0 13.45 6977.0 0.40 54.54 1179.0 223.2 96.6 L933 o.o 12.21 7624.0 0.41 61.06 1241.0 229.0 103.9 03314 6.5 12.78 7315.0 0.40 57.93 1245.0 229.3 100.7 Oi765 9.0 12.14 7664.0 0.41 61.46 1248.0 229.6 104.4 714649 8.0 11.14 8268.0 0.41 67.60 1424.0 245.2 172.7 01746 0.0 9.78 9235.0 0.42 77.54 1476.0 249.7 122.3 714737 8.0 10.40 8773.0 0.41 72.78 1524.0 253.7 175.0 03252 6.0 7.84 10994.0 0.43 95.83 1727 .0 270.1 141. 5 03107 5.5 7.15 11758.0 0.43 103.80 1873.0 281.3 149.8 714637 7.0 7.72 11127 .0 0.43 97.22 1890.0 282.5 252.9 03681 0.0 5.92 13358.0 0.44 120.60 2080.0, *296.4 166.0 02964E 0.0 5.02 14769.0 0.45 135.30 2238.0 307.5 179.6 01124K 0.0 3.71 17268.0 0.46 161.30 2596.0 331.1 203.2
• RCP PUMP CASING 671-N-0031 L933 0.0 12.21 7624.0 0.41 61.0G 1241.0 229.0 103.9 517312 7.0 . 12.00 7742.0 0.41 62.25 1315.0 235.7 161. 4 517246 8.0 11.73 7903.0 0.41 63.88 1357.0 239.4 175.1 714737 8.0 10.40 8773.0 0.41 72. 78* 1535.0 254.7 187.1 752627 0.0 5.7;4 13618.0 0.44 123.30 2130.0 299.9 168.6
• RCP PUMP CASING 671-N-0032 517571 10.0 16.76 5666.0 0.39 41.67 1003.0 205.8 112 .2 N

w 714334 13.0 16.38 5793.0 0.39 42.90 1037.0 209.3 128.3 L933 0.0 12.21 7624.0 0.41 61.06 1241.0 *229.0 103.9 01765 9.0 12.14 7664.0 0.41 61.46 1248.0 229.6 104 . *

e. e

                                                                                                                  ~

e Table 4-4 (Continued) I )> ANALYSIS OF THERMAL AGING OF CAST ,*STAINLESS STEEL REACTOR "O

 "O
                                                      .COOLANT PUMP CASINGS AT 70F CD

J 0..

  ......                               REPORT 14 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x

o:J HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @70F @70F @ 70F (%) @70F @ 70F (ft-lbs) (lb/in) (ks1V1n) 517312 7.0 12.00 7742.0 0.41 62.25 1317 .0 235.9 164.0 517246 8.0 11.73 7903.0 0.41 63.88 1328.0 236.8 142.4 01746 0.0 9.78 9235.0 0.42 77.54 1476.0 249.7 122.3 714737 8.0 10.40 8773.0 0.41 72.78 1540.0 .255.0 191.3 03681 0.0 6.07 13151.0 0.44 118.40 2042.0 293.7 163.8 03100 5.5 6.45 12626.0 *o.43 112.90 2045.0 293.9 159.0 02964E 0.0 4.78 15170.0 0.45 139.50 2316.0 312.8 183.6 01124K 0.0 3.71 17268.0 0.46 161.30 2596.0 331.1 203.2 e.

e - e I Table 4-5 I -6"

 "'C ANALYSIS OF THERMAL AGING OF STAINLESS STEEL fl)

s REACTOR COOLANT PUMP CASINGS AT 550F 0...

  -'*

x REPORT 15 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL OJ HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jic KJc T MOD. CONTENT PARAMETER c N ENERGY @550F @ 550F @ 550F (%) @550F @ 550F (ft-lbs} (1 b/in) (ks1/1n)

       ** PLANT T.D. FT. CALHOUN

• RCP PUMP CASING 671-N-0029 516942 20.0 36.34 2419.0 0.27 15.85 715.0 141.5 55.5 03233 7.0 13.57 4798.0 0.30 53.98 1325.0 192.7 111.6 03314 6.5 12.78 5000.0 0.30 57.93 1388.0 197.2 115.6 03336 5.0 10.93 5542.0 0.31 69.00 1489.0 204.2 126.5 024758 8.0 10.67 5628.0 0.31 70.81 1515.0 206.0 128.l 714649 8.0 11.14 5475.0 0.30 67.60 1635.0 214.0 225.4 02370A 6.0 9.35 6106.0 0.31 81.17 1666.0 216.l 136.9 02318A 6.0 8.93 6273.0 0.31 84.87 1720.0 219.6 139.9 03252 6.0 7.84 6753.0 0.32 95.83 1787.0 223.8 149.2 02362A 8.0 8.26 6564.0 0.31 91.46 1814.0 225.5 145.1 02475A 7.0 8.21 6584.0 0.31 91.93 1822.0 225.9 145.4 03107 5.5 7.15 7093.0 0.32 103.80 1894.0 230.4 155.1 714637 7.0 7.72 6813.0 0.32 97.22 1985.0 235.8 322.8 A7719 7.0 6.86 7249.0 0.32 107.60 1944.0 233.4 157.8 A9310F 5.0 6.71 7329.0 0.32 109.50 1969.0 234.9 159.2 9208F 8.0 6.65 7363.0 0.32 110.30 1980.0 235.5 159.8 A9872F 6.0 6.47 7465.0 0.32 112. 70 2013.0 237.5 161. 5 A9663F 6.0 . 5.49 8057.0 0.33 127.40 2102.0 242.7 172.3 A8025 6.0 4.96 8409.0 0.33 136.30 2213.0 249.0 178.l 029128 5.0 4.89 8456.0 0.33 137.50 2229.0 249.9 178.9 A8626 5.0 5.2~ 8210.0 0.33 131.20 2150.0 245.5 174.8 9358F 6.0 4.8 8505.0 0.33 138.80 2244.0 250.8 179.6 A7490 6.0 4.81 8513.0 0.33 139.00 2247.0 '250.9 179.8 08122 6.0 5.18 8262.0 0.33 132.60 2166.0 246.4 175.7 N

U'1 022918 5.0 4.73 8573.0 0.33 140.50 2267.0 252.0 180.8 8878 7.0 5.09 8317 .0 0.33 134.00 2184.0 247.4 176.6 9599. 5.0 4.16 9004.0 ti.34 151.80 2291. 0 253.4 * .8

e ... e Table 4-5 (Continued)

l>

"O ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "O n> REACTOR COOLANT PUMP CASINGS AT 550F

,

c..

.......

x REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL IXI HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @550F @ 550F @ 550F (%) @550F @550F (ft-lbs) (1 b/1n) (ks1/1n)

• RCP PUMP CASING 671-N-0030 714334 13.0 16.38 4203.0 0.29 . 42.90 1233.0 185.8 147.0 517116 12.0 15.41 4389.0 0.29 46.27 1295.0 190.5 150.3 03314A 8.0 13.45 4827.0 0.30 54.54 1334.0 193.3 112.3 L933 0.0 12.21 5156.0 0.30 61.06 1439.0 200.8 118.5 03314 6.5 12.78 5000.0 0.30 57.93 1388.0 197.2 115.6 01765 9.0 12.14 5176.0 0.30 61.46 1444.0 201.2 118.8 714649 8.0 11.14 5475.0 0.30 67.60 1635.0 . 214.0 225.4 01746 0.0 9.78 5941.0 0.31 77.54 1614.0 212.7 133.9 714737 8.0 10.40 5720.0 0.31 . 72.78 1631.0 213.8 223.0 03252 6.0 7.84 6753.0 0.32 95.83 1787 .o 223.8 149.2 03107 5.5 7.15 7093.0 0.32 103.80 1894.0 230.4 155.1 714637 7.0 7.72 6813.0 0.32 97.22 1992.0 236.3 331.0 03681 0.0 5.92 7784.0 0.33 120.60 2015.0 237.6 167.9 02964E 0.0 5.02 8371.0 0.33 135.30 2201.0 248.3 177 .5 01124K 0.0 3.71 9367.0 0.34 161.30 2406.0 259.6 194.4
• RCP PUMP CASING 671-N-0031 L933 o.o 12.21 5156.0 0.30 61.06 1439.0 200.8 118.5 517312 7.0 12.00 5215.0 0.30 62.25 1539.0 207.6 213.6 517246 8.0 11. 73 5295.0 0.30 63.88 1584.0 210.7 236.9 714737 8.0 10.40 5720.0 0.31 72.78 1649.0 215.0 245.5 752627 0.0 5.74 7893.0 0.33 123.30 2049.0 239.6 169.7
• RCP PUMP CASING 671-N-0032 517571 10.0 16.76 4133.0 0.29 41.67 1214.0 184.5 153 .3 N

O'I 714334 13.0 . 16.38 4203.0 0.29 42.90 1255.0 187.6 182.6 L933 0.0 12.21 5156.0 0.30 61.06 1439.0 200.8 118. 5

                                                                                                                          -

9.0 12.14 5176.0 0.30 61.46 1444.0 201.2 118.8 0176 * . e

e -e Table 4-5 (Continued)

)>

-0 ANALYSIS OF THERMAL AGING OF STAINLESS STEEL -0 11> REACTOR COOLANT PUMP CASINGS AT 550F

I a.

-'*

x REPORT #5

• PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL o:J HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINI~UM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jic KJc T MOD.

CONTENT PARAMETER c N ENERGY @550F @ 550F @ 550F (%) @ 550F @550F (ft-lbs) {lb/in) (ks1/1n) 517312 7.0 12.00 5215.0 0.30 62.25 1543.0 207.9 218. 7 517246 8.0 11. 73 5295.0 0.30 63.88 1536.0 207.4 177 .1 01746 0.0 9.78 . 5941.0 0.31 77 .54 1614.0 212.7 133.9 714737 8.0 10.40 5720.0 0.31 72.78 1656.0 215.4 253.3 03681 0.0 6.07 7~96.0 0.32 118.40 2089.0 242.0 165.3 03100 5.5 6.45 7472.0 0.32 112. 90 2015.0 237.6 161. 6 02964E 0.0 4.78 . 8534.0 0.33 139.50 2254.0 251.3 180.2 01124K 0.0 3. 71 9367.0 0.34 161.30 2406.0 259.6 194.4 e*

I i I I I I I I I~"CJ Table 4-~ - Fort Calhoun I ~0, I Limiting and Controlling Values of I ~- Jic and KJc at 70°F I ~ I Jic KJc RCP PUMP CASING HEAT Jt (lb/in) (ksi/in) 671-N-0029 516942 495.1 144.6 671-N-0030 714334 1026.0 208.1 671-N-0031 L933 1241. 0 229.0 671-N-0032 517571 1003.0 205.8

Table 4-7 Limiting and Controlling Values of Jic and KJc at 550"F Jic KJc RCP PUMP CASING HEAT .i (lb/in) (ksi/ in) 671-N-0029 516942 715.0 141.5 671-N-0030 714334 1233.0 185.8 671-N-0031 L933 1439.o* 200.8 671-N.;.,0032 517471 1214.0 184.5

5.0 CRACK GROWTH ANALYSIS In this section, the methodologies discussed in Section 5.1 of the generic report are applied to plant specific conditions at the Fort Calhoun plant. The growth history of a worst-case hypothetical crack is conservatively developed based on information contained in the vendor's (Byron-Jackson Company) stress analysis report (Reference 3-1). 5.1 Scope The analysis which follows pertains to the 28x28x34 DFSS Reactor Coolant Pump casings, Serial Numbers 671-N-0029 to 671-N-0032, inclusive, at the Fort Calhoun Plant. 5.2 Reference Stress Reports Stress values used in the crack growth analyses were obtained from the original stress analysis reports prepared by the Byron Jackson Company in January, 1973, and retrieved from storage at ABB Combustion Engineering Nuclear Power (Reference 3-2). 5.3 Selection of High Stress Locations The methodology described in Section s.1.1 of the generic report was applied to identify four regions as potentially limiting: (1) Junction, Scroll to Suction Nozzle Flange (2) Scroll (3) Junction, Scroll to.Bolt-Circle Flange (4) Diffuser Vane Number 7 All other regions in the stress summary were considered and -- were found to have lower stresses than the above regions. Appendix B 30

5.4 Stresses and Wall Thicknesses at Limiting Locations tt Membrane and through-wall bending components of the limiting regions were obtained from Reference 3.2 under the Design Condition basis, .. ~s follow~: (1) For the Junction, Scroll to Suction Nozzle Flange (Region 3): Design Condition = #25 Key Element = #445 in Finite-Element Model Membrane Stress = 24.47 Ksi (pp. 201, 202) Bending Stress = 13.82 Ksi (pp. 201,202) Thickness = 2.06 in.(Reference 3-4). (2) For the Scroll (Region 4) Design Condition = #25 Key Element = #444 in Finite Element Model Membrane stress = 27.7 Ksi (p. 208) Bending stress 2.2 Ksi (p. 208) Thickness = 2.06" (by estimation, the same as adjacent element 445 above) (3) For the Junction, Scroll to Bolt-Circle Flanqe (Reqion

5) :

Design Condition = # 15 Key Element = # 438 in Finite-Element Model Membrane stress = 21.9 Ksi (p. 212) Bending stress = 8.0 Ksi (p. 212) Thickness = 3.0 11 (estimated) Appendix B 31

(4) For Diffuser Vane Number 7 (Region 7) Design Condition = #13 Key Element = # 264 in Finite-Element Model Membrane stress = 28.33 Ksi (p. 220) Bending stress = 20.5 Ksi (p. 220) Thickness = 5.15 in* (Reference 3-2) 5.5 Calculation of Crack Growth Rates The methodology described in 5.1.4 of the generic report was applied to the above plant-specific conditions, using the annual rate of stress-cycling given in Section 3.2 of this appendix. An integration procedure was used to predict dimensionless crack depth, a/t, as a function of time, T. Results are swnmarized in Tables 5-1 through 5-4. For each region the calculated entries are listed against crack depth, a/t, as follows: (1) Applied Stress Intensity Factor, KI' was calculated using the ASME Section XI procedure, as further described in 5.1.4 of the generic report. Units for KI are Ksi--squareroot inch. (2) Crack growth rate, da/dT, was calculated using the Bernard & Slama equation (with R=O), multiplied by the Expected annual rate of stress cycling. The final equation is:

                                           -10    4.0 da/dT =  7.695 x 10     KI Units for da/dT are inches/year.

Appendix B 32

• Table 5-1.

Fort Calhoun RCP Casing crack Growth Rates At

                ~unction of Scroll and Suction Nozzle Flange (Region 3) a/t Interval                  KI                 da/dT                fl.Time (fraction}               (KSI /TN)            (IN/YEAR)            (YEARS}

0.08 -- 0.10 31.44 7.52 x 10"" 4 54.8 0.10 -- 0.15 38.18 1.63 x 10- 3 63.0 0 ..15 -- 0.20 44.13 2.92 x 10- 3 35.3 0.20 -- 0.25 49.72 4.70 x 10- 3 21.9 0.25 -- 0.30 55.04 7.06 x 10- 3 14.6 0.30 -- 0.35 60.46 1. 03 x 10- 2 10.0 0.35 -- 0.40 66.10 1.47 x 10- 2 7.0 0.40 -- 0.45 72.62 2.14 x 10- 2 4.8 0.45 -- 0.50 79.95 3.14 x 10- 2 3.3 (am= 24.47, ub = 13.82, t = 2.06")

                                                               /lit = 2. 544 /IN Appendix B                                                      33

Table 5-2 Fort Calhoun RCP Casing crack Growth Rate At Scroll (Region 4) a/t Interval KI da/dT ~Time (fraction} (KSI /IN) (IN/YEAR) (YEARS} 0.08 -- 0.10 25.56 3.29 x 10- 4 125. 0.10 -- 0.15 31. 71 7. 78 x 10 -4. 132. 0.15 -- 0.20 37.42 1. 51 x 10- 3 68.2 0.20 -- 0.25 42.93 2.61 x 10- 3 39.4 0.25 -- 0.30 48.41 4.23 x 10- 3 24.3 0.30 -- 0.35 54.03 6.56 x 10- 2 15.7 0.35 -- 0.40 60.05 1.00 x 10- 2 10.3 0.40 -- 0.45 66.86 1.54 x 10- 2 6.7 0.45 -- 0.50 74.58 2.38 x 10- 2 4.3 e (um= 27.7, ub = 2.2, t = 2.06") Appendix B 34

Table 5-3 Fort Calhoun RCP Casing Crack Growth Rate At Junction of Scroll and Bolt tircle Flange (Region 5) a/t Interval KI da/dT t.T {fraction} {KSI JIN) {IN/YEAR) {YEARS} 0.08 -- 0.10 30.02 6.25 x 10- 4 96. 0 .10 -- 0 .15 36.72 1. 40 x 10- 3 107. 0.15 -- 0.20 42.75 2.57 x i0- 3 58.4 0.20 -- 0.25 48.46 4.24 x 10- 3 35.3 0.25 -- 0.30 54.0 6.54 x 10- 3 22.9 0.30 -- 0.35 59.65 9.74 x 10- 3 15.4 0.35 -- 0.40 65.59 1. 42 x 10- 2 10.5 10- 2 0.40 0.45

     --
     --

0.45 0.50 72.41 80.10 2.12 3.17 x x 10- 2 7.1 4.7 e (am= 21.9, ab = 8.0, t = 3.0") Appendix B 35

Table 5-4 Fort Calhoun RCP Casing crack Growth Rate At Diffuser Vane Number 7 a/t Interval KI da/dT ~T {fraction} (KSI Im} {IN/YEAR} {YEARS} 0.08 -- 0.10 41.95 2.38 x 10- 3 43.2 0.10 -- 0.15 51.99 5.62 x 10- 3 45.8 (a) 0.15 -- 0.20 61.28 1. 08 x 10- 2 23.7 (b) 0.20 -- 0.25 70.24 1.87 x 10- 2 13.7 0.25 -- 0.30 79.13 3.02 x 10- 2 8.5 0.30 -- 0.35 88.24 4.67 x 10- 2 5.5 0.35 -- 0.40 98.0 7 .1. x 1.0- 2 3.6 0.40 -- 0.45 109.05 0.109 2.3 0.45 -- 0.50 121.5 0.168 1.5 {am= 28.33, ab= 2.77, t = 5.15")

                -

1% a/t steps: E~Ti = 65.4 (a) E~Ti = 31.1 (b) Appendix B 36

(3) Incremental time, dT, in which the crack will grow through the indicated interval of dimensionless crack depth values, a/t, was calculated as described in 5.1.4 of the generic portion of the report. Units for dT are years. The summation of time increments yields the total Time for a cf'ack to grow to a given a/t value. The predicted growth curves for hypothetical cracks show the functional relationship between a/t and total Time, using the initial condition of a/t = o.oa at Time = o. The first incremental time listed is based on a change in a/t in the amount 0.02 (i.e. 2% of thickness), to indicate the time needed for the crack to grow from a/t = 0.08 to a/t = 0.10. Subsequent incremental times are based on the time needed for the crack to grow through changes in a/t of 0.05 (i.e. 5% of thickness). The first such incremental time is given for the range a/t = 0.10 to 0.15. The final incremental time is given for the range a/t = 0.45 to 0.50, whereupon the analysis is terminated. 5.6 Stresses Under Emergency and Faulted Conditions In order to verify that limiting sections containing hypothetical cracks can withstand Emergency Condition Loads and Faulted Conditions Loads, as discussed in Section 5.3.4 of the generic portion of this report, the methodology described in Section 5.1 was again applied. Applied stress intensity factors were calculated at the limiting locations, based on the following data from Reference 3-2. 5.6.1 Emergency Condition Stresses (1) Junction, Scroll to suction Nozzle Flange ("Region 3 11 ) Appendix B 37

APPENDIX C APPLICATION OF GENERIC METHODOLOGY FOR RELAXATION OF THE CALVERT CLIFFS 1&2 REACTOR COOLANT PUMP CASING INSPECTION INTERVAL

'*

Appendix c 1 _j

ABSTRACT Appendix c was prepared to demonstrate the amount of inspection interval relaxation appropriate for the reactor ~ coolant pump casings at the Calvert Cliffs 1&2 plants, based an application of the generic methodology presented in the main body of this report. Appendix c 2

APPENDIX C TABLE OF CONTENTS Section Title Page

1. 0 PURPOSE 6 2.0 PRE-SERVICE INSPECTION DATA EVALUATION 7 3.0 OPERATING HISTORY 9 3.1 Design Specifications 9 3.2 Stress Cycles Used In Evaluation 10 3.3 Stress Cycles at Calvert Cliffs To-date 10 4.0 THERMAL EMBRITTLEMENT 12 4.1 Material Identification and Chemical Properties 12 4.2 Material Specifications and Mechanical Properties 12 4.3 Thermal Aging Behavior 13 4.4 Toughness Properties of Aged Materials 13 4.5 Limiting Values 14 5.0 CRACK GROWTH ANALYSIS 39 5.1 Scope 39 5.2 Reference Stress Reports 39 5.3 Selection of High Stress Locations 39 5.4 Stresses and Wall Thicknesses at Limiting Locations 40 5.5 Calculation of Crack Growth Rates 42 5.6 Stresses Under Emergency and Faulted Conditions 48 5.7 Results 51 6.0 INSPECTION INTERVAL 52 APPENDIX C REFERENCES 53 Appendix C 3

LIST OF TABLES TABLE 4-1 Material Identification and Chemical Compositions - Unit 1 15 TABLE 4-2 Material Specifications and Tensile Properties - Unit 1 17 TABLE 4-3 Predicted Thermal Aging Behavior - Unit 1 19 TABLE 4-4 Predicted Toughness Properties of Aged Materials (70°F) - Unit 1 21 TABLE 4-5 Predicted Toughness Properties of Aged Materials (550°F) - Unit 1 23 TABLE 4-6 Limiting and Controlling Values of Jic and KJc at 70°F - Unit 1 25 TABLE 4-7 Limiting and Controlling Values of Jic and KJc at 550°F - Unit 1 26 TABLE 4-8 Material Identification and Chemical Compositions - Unit 2 27 TABLE 4-9 Material Specifications and Tensile Properties - Unit 2 29 TABLE 4-10 Predicted Thermal Aging Behavior - Unit 2 31 TABLE 4-11 Predicted Toughness Properties of Aged Materials (70°F) - Unit 2 33 Appendix C 4

LIST OF TABLES (Continued) TABLE 4-12 Predicted Toughness Properties of Aged Materials (550°F) - Unit 2 35 TABLE 4-13 Limiting and Controlling Values of Jic and KJc at 70°F - Unit 2 37 TABLE 4-14 Limiting and Controlling Values of Jic and KJc at 550°F - Unit 2 38 TABLE 5-1 Crack Growth Rates at Vane Number 8 43 TABLE 5-2 Crack Growth Rates at Discharge Nozzle-Crotch Vicinity 44 TABLE 5-3 Crack Growth Rates at Suction Nozzle-Level c 45 TABLE 5-4 Crack Growth Rates at Volute Junction With Lower Flange 46 TABLE 5-5 Crack Growth Rates at Hanger Bracket #1 Vicinity 47 Appendix c 5

1.0 PURPOSE The purpose of Appendix C is to document the application of methodology presented in the main body of this report to the plant-specific data for the reactor coolant pump casings at the Calvert Cliffs 1&2 plants, and to quantify the extent of inspection interval relaxation available. Appendix c 6

2.0 PRE-SERVICE INSPECTION DATA EVALUATION Pre-service inspection data for the Calvert Cliffs reactor coolant pumps numbered 681-N-0437 through 681-N-0444 was collected from QA data packages originally prepared by the Byron Jackson Company and stored in archives by ABB Combustion Engineering Nuclear Power. Information in these data packages concerning welding procedures, radiographic inspections, non-destructive testing and dye penetrant testing were examined. The testing and inspection procedures that were followed for all reactor coolant pumps at Calvert Cliffs were found to be the same in all significant aspects. The most relevant information obtained from this review of the QA data packages were the reports on radiographic examination of the RCP casing castings, pressure retaining welds, and repair welds. Radiographic examination requirements invoked ASME Section III rules for examination procedures and sensitivity. The required radiograph sensitivity was 2-2T according to applicable ASTM standard Reference Radiograph Procedure requirements (i.e. ASTM E71, El86, E280) as determined by the casting thickness. The 2-2T sensitivity is consistent with a 2% initial flaw size, because the requisite image quality indicator (IQI) for this level of examination is specified as a penetrameter with a minimum hole-size diameter equal to 2% of the casting thickness. The acceptance criteria for interpretation of the radiographs was severity Level 2 for sand, porosity or shrinkage indications. Linear indications such as cracks, hot tears, and unfused chaplets or chills were unacceptable at any level. Any such discernable indications required rejection of weld repair and a repeated radiographic examination of the affected casting or weldment. Appendix c 7

The results of this review of pre-service RCP casing examinations confirm that cracks of 2% described in Section 4.4 of the generic report were detectable but none were left unrepaired. Appendix c 8

3.0 Operating History 3.1 Design Specifications The Calvert Cliffs Unit 1 and 2 RCPs were delivered to the site in 1971 and were first placed in commercial operation in 1975 and 1977 respectively. Reactor coolant system design pressure and temperature are 2500 psia and 650°F respectively. Each pump is designed to deliver 81,200 gpm of coolant at a head of 300 feet. These pumps have 30 inch diameter suction and discharge piping. The design specification (Reference 3-1) calls for the pumps to be capable of withstanding the following transient conditions events during the 40-year license period: Assumed Occurrences Transient During 40 Year Condition License Period Heat-Up (100.F/hr) 500 Cool-Down {l00°F/hr) 500 Hydrostatic Test (3125 psia 100-400°F) 10 Leak Test-In Conjunction With Heatup (2250 psia 100-400°F) 320 Loss of Secondary Pressure 5 Reactor Trip or Loss of Load 400 Appendix c 9

3.2 Stress Cycles Used in Evaluation As indicated in Section 5.1.4 of the generic portion of this report, crack growth was evaluated on the basis of an design number of stress cycles between atmospheric and operating pressures during heatup and cooldowri over the nominal 40-year life of the plant*. The number of such cycles used in the stress analyses performed by the Byron Jackson Company is soo, as given on page 94 of Reference 3-2. This total was increased to sos to include Loss of Secondary Pressure. The total is therefore S transient events greater than the soo heat-up cooldown cycles specified in the RCP design specifications. On an annual basis, the average number of stress cycles, based on sos per 40 years, is 12.62S per year, and the hypothetical crack growth calculations and curves were prepared accordingly. 3.3 Stress Cycles at Calvert Cliffs To-date Details of the actual operating history of the Calvert Cliffs RCPs from 197S (Unit 1) and 1977 (Unit 2) to 1991 were furnished in Reference 3-3 and are as follows: Unit 1. Unit 2 Heatup/Cooldown 84 57 Reactor Trip -* 112 86 Heatup-plus-cooldown, taken together, constitute one cycle. The average number of stress cycles per year over the 16 operating years for Unit 1 and 14 years for Unit 2, ending in mid 1991 is 5.25 and 4.07 cycles per year respectively. This time period is equivalent to 6.7 years for Unit 1 and 4.S years for Unit 2 at the design rate of stress cycling Appendix c 10

(12.625 per year). The actual rate of cycle accrual for each plant is seen to be only 42% and 32% of the design rate for Units 1 and 2, respectively, a significant conservatism. Appendix c 11

4.0 THERMAL EMBRITTLEMENT Thermal embrittlement evaluation of the Calvert Cliffs casings is discussed and plant specific data are presented in the five following reports. All equations referenced below are found in the main body of this report, which is also referred to as the generic report. 4.1 Material Identification and Chemical Properties The chemical compositions provided in Report #1 (Table 4-1) for each RCP casing at Calvert Cliffs were obtained from Quality Assurance documents originally supplied by the Byron Jackson Company and stored at ABB Com.Pustion Engineering Nuclear Power. For each individual pump casing, chemical compositions are given for specific casing welds as well as for individual castings. 4.2 Material Specifications and Mechanical Properties The material specifications and mechanical properties found in Report #2 (Table 4-2) for each RCP casing were obtained from the same data source as in Section 4.1 above. For each individual pump casing the material specification, material type and heat number are given for specific casing welds as well as for individual castings. It is evident from the report that data obtained for mechanical properties (i.e. yield strength, tensile strength, total elongation and reduction in area) for each material was only available for the castings, and was not available for the casing welds. The unaged flow stress at 70°F and the aged flow stress at 10°F and 550°F were calculated as discussed in Section 5.2.3 of the generic report.

Appendix c 12

4.3 Thennal Aging Behavior Report #3 (Table 4-3) contains predicted thermal aging behavior data for all of the Calvert Cliffs RCP casings. The measured ferrite contents listed for specific casing welds and individual castings were supplied by the Byron Jackson Company in the same QA package as referenced in Section 4.1 above. In most cases a value was obtainable for the measured ferrite content. In cases where a value was not given, a zero was recorded. The chromium and nickel equivalents for the castings and weld metal, as well as the chromium/nickel ratio for the castings, were calculated using equations 5-7 and 5-8 respectively, as discussed in Section 5.2.1 of the generic report. Values for ferrite content of the castings were computed using two methods: for ferrite content #1, the values were computed using the method which follows ASTM ASOO/ASOOM (Reference 5-2); for ferrite content #2, the values were computed using equation 5-9 as discussed in Section 5.2.1 of the generic report. The latter method follows. work performed by O.K. Chopra (Reference 4-1). 4.4 Toughness Properties of Aged Materials The predicted toughness properties of aged material at 70°F and 550°F are given respectively in Report #4 (Table 4-4) and Report #5 (Table 4-5). The measured ferrite contents listed for all heat numbers are the same as the values given in Report #J. The material aging parameter was calculated using equation 5-10 of the generic report. The room-temperature charpy impact energy, CVsat' of the various materials was calculated using equation 5-11. The Jic values were determined in accordance with the methods of Appendix C 13

ASTM E813 as discussed in Section 5.2.3 of the generic report. The plane strain fracture toughness, KJc' and minimum tearing modulus, T, at 70°F and 550°F were calculated using equations 5-17 and 5-18. The values listed for the material constants N and C at 70°F and 550°F were calculated using equations 5-12 through 5-15. These constants were needed in computing the values for Jic' KJc and T. All equations used in Report #4 are found in Section 5.2 of the generic report. 4.5 Limiting Values The limiting and controlling values for Jic and KJc at 70°F and 550°F for each individual pump at Calvert Cliffs Units 1&2 are given in Tables 4-6 and 4-7 for Unit 1 and 4-13 and 4-14 for Unit 2. Appendix C 14

C

>         Table 4-1 C

l> . ANALVIS OF THERMAL AGING OF CAST STAINLESS STEEL s

...
 ..                                                   REACTOR COOLANT PUMP CASINGS n                                     REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS MATERIAL HEAT NO. c     Mn     S1      s       p     Cr   . N1   Mo     N     Cb
      ** PLANT l.D. CALVERT CLIFFS UNIT 1

• RCP PUMP CASING 681-N-0437 CASING WELD 04146 0.02 1.84 i.47 0.014 0.014 20.48 9.96 2.70 0;04 0.00 HUB/DIFFUSER 41588 0.05 0.65 1.10 0.006 0.034 19.35 9.60 2.22 0.04 0.00 CASING WELD 03063 0.04 1.70 0.53 0.016 0.011 19.89 10.29 2.81 0 .04 ' 0.00 CASING WELD 03165 0.04 1.73 0.61 0.014 0.013 19.66 9.76 2.65 0.04 0.00 CASING WELD 03036 0.03 1.53 0.47 0.016 0.013 19.01 9.70 2.81 0.04 0.00 CASING WELD 03003 0.04 1.54 0.44 0.013 0.013 19.11 9.89 2.48 0.04 0.00 CASING WELD 04313 0.02 0.91 0.52 0.016 0.015 19.59 9.93 2.59 0.04 0.00 CASE SCROLL 41141 0.07 0.60 0.85 0.006 0.036 18.69 9.23 2.15 0.04 0.00 CASING WELD 04195 0.04 1.98 0.52 0.017 0.019 19.45 11.70 2.32 0.04 0.00 CASING WELD 03493 0.02 1. 74 0.44 0.019 0.*016 18.82 9.36 2.76 0.04 0.00 CASING WELD 04286 0.02 0.90 0.52 0.015 0.015 18.83 10.12 2.41 0.04 0.00 CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 19.90 9.10 1.39 0.04 0.00 CASING WELD 03793 0.02 1.66 0.34 0.018 0.008 19 .18 9.34 2.78 0.04 0.00
• RCP PUMP CASING 681-N-0438 CASE SCROLL 43393 0.07 0.72 1.34 0.003 0.032 18.87 9.27 2.17 0.04 0.00 HUB/DIFFUSER 43836 0.08 0.72 1.34 0.006 0.035 18.91 9.36 2.2S 0.04 0.00 CASING WELD 03063 0.04 1. 70 0.53 0.016 0.011 19.89 10.29 2.81 0.04 0.00 CASING WELD 03165 0.04 1. 73 0.61 0.014 0.013 19.66 9.76 2.65 0.04 0.00 CASING WELD 03036A 0.04 1.52 ;0.47 0.015 0.014 18.69 9.89 2.84 0.04 0.00 CASING WELD 03003 0.04 1.54. b.44 0.013 0.013 19.11 9.89 2.48 0.04 0.00 CASING WELD 03036 0.03 1.53 b.47 0.016 0.013 19.01 9.70 2.81 .0.04 0.00 CASING WELD 04313 0.02 0.91 0.52 0.016 0.015 19.59 9.93 2.59 0.04 0.00 I

I-' Ul CASING WELD 04286 0.02 o..90 0.52 0.015 0.015 18.83 10.12 2.41 0.04 0.00 i CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 '19. 90 9.10 2.31 0.04 0.00 CASING WELD 03493 0.02 1. 74 0.44 0.019 0.016 18, 9.36 2.76 0.04 0.00 e

e Table 4-1 (Continued) ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL l> -a REACTOR COOLANT PUMP CASINGS -a fl)

s 0..

......                                     REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS
><

("") MATERIAL HEAT NO. c Mn Si s p Cr Ni Mo N Cb CASING WELD 03793 0.02 1.66 0.34 0.018 0.008 19 .18 9.34 2.78 0.04 0.00

• RCP PUMP CASING 681-N-0439 CASE SCROLL 44581 0.07 0.72 1.34 0.003 0.035 19.27 9.29 2.23 0.04 0.00 HUB/DIFFUSER 44936 0.06 0.72 1.48 0.001 0.034 18.98 9.24 2.10 0.04 0.00 CASING WELD 03063 0.04 1.70 0.53 0.016 0.011 19.89 10.29 2.81 0.04 0.00 CASING W~LD 03036 0.03 1.53 0.47 0.016 0.013 19.01 9.70 2.81 0.04 0.00

         *cASING WELD   03003      0.04  1. 54     0.44  0.013   0.013    19.11 9.89 2.48 0.04' 0.00 CASING WELD   03036A     0.04  1.52      0.47  0.015   0.014    18.69 9.89 2.84 0.04 0.00 CASING WELD   04313      0.02  0.91      0.52  0.016   0.015    19.59 9.93 2.59 0.04 0.00 CASING WELD   03493      0.02  1. 74     0.44  0.019   0.016    18.82 9.36 2.76 0.04 0.00 CASING WELD   04455      0.02  0.95      0.51  0.017   (). 013  18.87 9.86 2.46 0.04 0.00 CASING WELD   04286      0.02  0.90      0.52  0.015   0.015    18.83 10.12 2.4X 0.04 0.00 CASING WELD   X43439     0.03  1.39      0.36  0.017   0.016    19.90 9.10 2.31 0.04 0.00
                        !
                        '

• RCP PUMP CASING 681-N-0440 CASING WELD 04146 0.02 1.84 1.47 0.014 0.014 20.48 9.96 2.70 0.04 0.00 CASE SCROLL 45164 0.05 0.88 0.97 0.007 0.034 19.50 9.26 2.15 0.04 0.00 HUB/DIFFUSER 44734 0.05 0.72 1.34 0.003 0.034 19 .16 9.58 2.27 0.04 0.00 CASING WELD 03063 0.04 1. 70 0.53 0.016 0.011 19.89 10.29 2.81 0.04 0.00 CASING WELD 03036 0.03 1.53 0.47 0.016 0.013 19.01 9.70 2.81 0.04 0.00 CASING WELD 03003 0.04 1. 54 . 0.44 0.013 0.013 19 .11 9.89 2.48 0.04 0.00 CASING WELD 03036A 0.04 1.52 0.47 0.015 0.014 18.69 9.89 2.84 0.04 0.00 CASING WELD 04313 0.02 0.91 0.52 0.016 0.015 19.59 9.93 2.59 0.04 0.00 CASING WELD 04455 0.02 0.95 ~0.51 0.017 0.013 18.87 9.86 2.46 0.04 0.00 CASING WELD CASING WELD 04286 X43439 0.02 0.03 0.90 .o. 52 0.015 0.015 18.83 10.12 2.41 1.39 0.36 0.017 0.016 19.90 9.10 2.31 1

0.04 0.00 0.04 0.00

......

O"I

Table 4-2 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

)::>>

-0 REACTOR COOLANT PUMP CASINGS -0 Cl>

J a..

-'*

REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES x CJ MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNA GED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (ksi) (ksi) (%) {%) STRESS STRESS STRESS

                                                                                                     @ 70F   @ 70F @ 550F
              ** PLANT I.D. CALVERT CLIFFS UNIT 1

• RCP PUMP CASING 681-N-0437 58000 CASING WELD 04146 0.0 0.0 0.0 0.0 0 78600 HUB/DIFFUSER A 351 CFBM 41588 45.3 84.5 45.0 72.0 65 78511 57917 CASING WELD A 298 E316-16 03063 0.0 0.0 . O.Q. 0.0 0 78600 58000 CASING WELD A 298 E316-16 03165 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03036 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03003 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04313 o. 0 . 0.0 0.0 0.0 0 78600 58000

             . CASE SCROLL A 351     CFSM       41141          40.7      85.5      63.0     69.0         63  75661   55067 CASING WELD A 298     E316-16    04195           0.0       0.0      . 0.0     0.0          0  78600   58000 CASING WELD A 371     ER-316     03493           0.0       0.0        0.0     0.0          0  78600   58000 CASING WELD A 371     ER-316     04286           o.o       0.0        0.0     0.0          0  78600   58000 CASING WELD A 298     E316-16    X43439          0.0       0.0        0.0     0.0          0  78600   58000 CASING WELD A 371     ER-316     03793           o.o       0.0        0.0     o.o          0  78600   58000

• RCP PUMP CASING 681-N-0438 63 75345 54751 CASE SCROLL A 351 CF8M 43393 42.1 83 .7 63.0 67.0 HUB/DIFFUSER A 351 CF8M 43836 42.8 85.3 55.0 76.0 64 77165 56571 CASING WELD A 298 E316-16 03063 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03165 0.0 0.0 0.0 0.0 0 78600 . 58000 CASING WELD A 298 E316-16 03036A 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03003 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03036 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04313 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04286 0.0 0.0 0.0 0.0 0 78600 58000

     ...... CASING WELD  A 298    E316-16    X43439          0.0       0.0        0.0     o.o          0  78600   58000
     .......

CASING WELD A 371 ER-316 03493 0.0 0.0 o.o 0.0 0 78600 58000 CASING WELD A 371 ER.:.316 03793 0.0 0, o.o 0.0 0 78600 58000-e

)>                                                 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

"'O "'O Table 4-3 REACTOR COOLANT PUMP CASINGS CD

3 0.

-'*                                                   REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR
)(

(""') HEAT MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FER.RITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS #1 (%) #2 (%) WELDS WELDS (%}

            **PLANT I.D. CALVERT CLIFFS UNIT 1

• RCP PUMP CASING 681-N-0437 04146 23.0 19.463 14.129 1.38 24.0 29.4 25.385 12.680 19.. 0 41588 14.0 17.574 14.399 1.22 15.2 15.3 23.220 12.625 12.0 03063 15.0 18.555 14.840 1.25 13.7 17.6 23.495 13.420 10.0 03165 14.*0 18.169 14.411 1.26 14.2 18.4 23.225 13.025 12.0 03036. 16.0 17.646 14.138 1.25 13.7 17.4 22.525 12.625 10.0 03003 12.0 17 .332 14.525 1.19 11.0 13.3 22.250 13.060 14.2 04313 18.0 17.983 14.019 1.28
• 16. 7 20.3 . 22.960 12.185 13.0 41141 8.0 16.709 14.514 1.15 11.0 10.6 22 .115 12.830 14.0 04195 8.0 17.517 16.321 1.07 6.2 6.5 22.550 15.030 10.7 03493 17.0 17.381 13.619 1.28 14.7 19.7 . 22.240 12.150 11. 0 04286 15.0 17 .006 14.208 1.20 12.3 13.6 22.020 12.370 15.3 X43439 14.0 16.765 13 .477 1.24 12.3 17 .1 21.830 11. 895 lLO 03793 16.0 17.717 13.495 1.31 16.2 23.0 22.470 11. 970 14.0
• RCP PUMP CASING 681-N-0438 43393 14.0 17.149 14.566 1.18 13.2 12.3 23.050 12.930 13.0 43836 12.0 17 .286 14.901 1.16 12.3 11.2 23 .170 13.320 1.0 03063 15.0 18.555 14.840 1.25 13.7 17 .6 23.495 13.420 10.0 03165 14.0 18.169 14.411 1.26 14.2 18.4 23.225 13.025 12.0 03036A 13.0 17.362 14.523 1.20 11.0 13.5 22.235
• 13.050 f3 .8 03003 12.0 17 .332 14.525 1.19 11.0 13.3 22.250 13 .060 14.2 03036 15.0 17.646 14.089 1.25 13.7 17.7 22.525 12.565 1010 04313 18.0 17.983 14.019 1.28 16.7 20.3 22.960 12.185 13.0 04286 15.0 17 .006 14.208 1.20 12.3 13.6 22.020 12.370 15.3

   ........

X43439 14.0 17.878 13.477 1.33 17 .2 24.3 22.750 11. 895 11. 0

   "° 03493                                13.521       1.29         15.2     20.5    22.240       12.030        11. 0 03793 17,0 17 .381 16.0 17.717     13.495       1.31         16.2  e  23.0    22.470       11. 970       14.0 e

e

)>

Table 4-3 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL "'C "'C REACTOR COOLANT PUMP CASINGS CD

I a..

......                                           REPORT 13 - PREDICTED THERMAL AGING BEHAVIOR x

n HEAT MEASURED CHROMIUM NICKEL Cre/Nie CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS #1 (%) 12 (%) WELDS WELDS (%)

• RCP PUMP CASING 681-N-0439 44581 13.0 17.621 14.586 1. 21 14.7 14.4 23.510 12.950 13.0 44936 14.0 17 .241 14.291 1.21 15.2 14.2 23.300 12.600 13.0 03063 15.0 18.555 14.840 1. 25 13.7 17.6 23.495 13.420 10.0 03036 16.0 17.646 14.138 1.25 13.7 17 .4 22.525 12.625 lo:o 03003 12.0 17.332 14.525 1.19 11.0 13.3 22.250 13.060 14.2 03036A 13.0 17.362 14.474 1. 20 11. 4 13.8 22.235 1'2. 990 13.8 04313 18.0 17. 983 14.019 1.28 16.7 20.3 22.960 12.185 13.0 03493 17.0 17.381 13.619 1.28 14.7 19.7 22.240 12.150 11. 0 04455 16.0 17.101 13.953 1.23 13.7 15.7 22.095 12.135 10.0 04286 15.0 17.006 14.208 1.20 12.3 13.6 22.020 12.370 15.3 X43439 14.0 17.878 13. 477 1.33 17 .2 24.3 22.750 11. 895 11. 0
• RCP PUMP CASING 681-N-0440 04146 23.0 19.463 14.129 1.38 24.0 29.4 25.385 12.680 19.0 45164 14.0 17. 577 14.081 1.25 15.7 17.4 23.105 12.400 14.0 44734 13.0 17.560 14.386 1.22 15.7 15.3 23.440 12.640 15.0 03063 15.0 18.555 14.840 1.25 13.7 17 .6 23.495 13.420 10.0 03036 16.0 17.646 14.138 1.25 13.7 17 .4 22.525 12.625 10.0 03003 . 12.0 17.332 14.525 1.19 11.0 13 .3 . 22.250 13.060 14.2 03036A 13.0 17.362 14.474 1.20 11.4 13.8 22.235 12.990 13.8 04313 18.0 17.983 14.019 1.28 16.7 20.3 22.960 12.185 13.0 04455 16.0 17 .101 13.953 1.23 13.7 15.7 22.095 12.135 10.0 04286 15.0 17.006 14.208 1.20 12.3 13.6 22.020 12.370 15.3 X43439 14.0 17.878 13 .477. 1.33 17.2 24.3 22.750 11.895 11 .,0 N

0

                                                              .- - .
                                                                  ~-

x:. Table 4-4

"'C "'C ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR fl)

3 COOLANT PUMP CASINGS AT 70F a.

-'*

x REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL ('") HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (%) @ 70F @ 70F (ft-lbs) (1 b/i n) (ksiVin)

         ** PLANT I.O. CALVERT CLIFFS UNIT 1

• RCP PUMP CASING 681-N~0437 04146 23.0 28.29 3434.0 0.38 21.29 594.5 . 158.5 50.9 41588 14.0 22.32 4290.0 0.38 28.81 756.7 178.8 62.9 03063 15.0 21.27 4496.0 0.39 30.68 757.7 178.9 65.1 03165 14.0 21.81 4389.0 0.38 29.70 775.6 181.0 64.0 41141 8.0 17 .34 5483.0 - 0.39 39.92 947.8 200.1 84.0 03036 16.0 16.41 5782.0 0.39 42. 79 . 1003.0 205.8 82.0 04313 18.0 15.61 6065.0 0.40 45.54 1006.0 206.1 85.1 03003 12.0 15.21 6217.0 0.40 47.02 1035.0 209.l 87.0 03793 16.0 14.86 6356.0 0.40 48.39 1060.0 211.6 88.8 03493 17.0 14.25 6614.0 0.40 50.94 1109. 0 216.4 92.0 X43439 14.0 13.97 6735.0 0.40 52.14 1133 .0 218.7 93.5 04286 15.0. 10.42 8761.0 0.41 72.66 1460.0 . 248.3 117 .4 04195 8.0 10.24 8891.0 0.41 73.99 1484.0 250.4 118. 9
• RCP PUMP CASING 681-N-0438 43836 12.0 23.18 4138.0 0.38 27.44 728.6 175.4 62.8 43393 14.0 22.39 4279.0 0.38 28.70 757.5 178.9 67.9 03063 15.0 '21.27 4496.0 0.39 30.68 . 757. 7 178.9 65.1

          .03165              14.0           21.81       4389.0           0.38       29.70        775.6          181.0   64.0 X43439             14.0          19.81        4804.0           0.39       33.51        815.6          185.6   69.2 03036              15.0           16.0        5910.0           0.40       44.03        976.6         .203.l   83.2 04313              18.0           15.6        6065.0           0.40       45.54       1006.0          206. l  85. I 03003              12.0          15.21        6217.0           0.40       47.02       1035.0          209. l  87.0 N
  ...... 03036A             13.0        . 15.13        6247.0           0.40       47.32       1040.0          209.6   87.4 03793              16.0           14.86       6356.0           0.40       48.39       1060.0        . 211.6   88.8 03493              17 .0         ~34          7031.0           0.40       55.09       1189. 0         224.1   97.

04.286-* 15.0 42 8761.0 941 72.66 1460.0 248.3 llJ

Table 4-4 (Continued)

)>

ANALYSIS OF THERMAL AGING OF CAST STAINlESS STEEL REACTOR -c -c COOLANT PUMP CASINGS AT 70F (I)

I 0..

......                            REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x

n HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jic KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (%) @ 70F @ 70F (ft-lbs) (1 b/i n) (ksiVin)

Table 4-5

)>                                             ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "tJ "tJ CD REACTOR COOLANT PUMP CASINGS AT SSOF

I 0..

-'*

x REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL ("") HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING *CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @ 550F @ 550F (ft-lbs) (lb/in) (ksi/in)

     ** PLANT I .D. CALVERT CLIFFS UNIT 1

• RCP PUMP CASING 681-N-0437 04146 23.0 28.29 2840.0 0.28 21.29 813.4 151.0 70.2 41588 14.0 22.32 3357.0 0.28 28.81 977 .5 165.5 81.6 03063 15.0 21.27 3477.0 0.28* 30.68 1016.0 168.7 83.9 03165 14.0 21.81 3414.0 0.28 29.70 995.3 167.0 82.5 41141 8.0 17.34 4033.0 0.29 39.92 1150 .0 179.5 105.5 03036 16.0 16.41 4197.0 0.29 42.79 1195.0 183.0 99.2 04313 18.0 15.61 4349.0 0.29 45.54 1244.0 186. 7 102.2 03003 12.0 15.21 4430.0 0.29 47.02 1270.0 188.6 103.8 03793 . 16.0 14.86 4504.0 0.29 48.39 1294.0 190.4 105.3 03493 17.0 14.25 . 4639.0 0.30 50.94 1274.0 188.9 108.6 X43439 14.0 . 13. 97 4702.0 0.30 52.14 1294.0 190.4 109.8 04286 15.0 . 10.42 5715.0 0.31 72.66 1543.0 208.0 129.8 04195 8.0 10.24 5777 .o 0.31 73.99 1562.0 209.2 130.9
• RCP PUMP CASING 681-N-0438 43836 12.0 23.18 3267.0 0.28 27.44 951.3 163.3 83.2 43393 14.0 22.39 3350.0 0.28 28.70 981.1 165.8 90.5 03063 15.0 21.27 3477 .o 0.28 30.68 1016.0 168.7 83.9 03165 14.0 ~1.81 3414.0 0.28 29.70 995.3 167.0 82.5 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 169.3 88.0 03036 15.0 16.01 4266.0 0.29. 44.03 1216.0 184.6 100.5 04313 18.0 15.6 4349.0 0.29. 45.54 1244.0 ,-186. 7 102.2 03003 12.0 15.21 4430.0 0.29 47.02 1270.0 188.6 103.8 03036A 13.0 15.13 4446.0 0.29 47.32 1274.0 189.0 104.1 N

w 03793 16.0 . 14.86 4504.0 0.29 48.39 1294.0 190.4 105.3 03493 17 .0 13.34 4855.0 0.30 55.09 1342.0

• 193. 9 112. 7

                         .15.0          10.42        5715.0       fl.3.1 .      72.66       1543.0      208.0  .29.8 0428 * .

Table 4-5 (Continued)

)>

"'C ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "'C Cl> REACTOR COOLANT PUMP CASINGS AT 550F .::3 a.

-'*
)<                                   REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL C"')

HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @ 550F @ 550F (ft-lbs) (1 b/1 n) (ks1/1n)

• RCP PUMP CASING 681-N-0439 44581 13.0 26.90 2941.0 0.28 22.68 839.3 153.4 63.8 44936 14.0 23.61 3224.0 0.28 26.79 924.5 161.0 64.5 03063 15.0 21.27 3477.0 0.28 30.68 1016.0 168.7 83.9 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 169.3 88.0 03036 16.0 16.41 4197.0 0.29 42.79 1195.0 183.0 99.2 04313 18.0 15.61 4349.0 0.29 45.54 1244.0 186.7 102.2 03003 12.0 15.21 4430.0 0.29 47.02 1270.0 188.6 103.8 03036A 13.0 14.90 4495.0 0.29 48.24 1291.0 190.2 105.l 03493 17 .0 14.25 4639.0 0.30 50.94 1274.0 188.9 108.6 04455 16.0 11.43 5383.0 0.30 65.70 1512.0 205.8 122.8 04286 15.0 10.42 5715. 0 0.31 72.66 1543.0 208.0 129.8
• RCP PUMP CASING 681-N-0440 04146 23.0 28.29 2840.0 0.28 21.29 813.4 151.0 70.2 45164 14.0 23.44 3241.0 0.28 27.05 940.5 162.3 78.9 44734 13.0 23.04 3282.0. 0.28 27.66 951.3 163.3 76.0 03063 15.0 21.27 3477.0 0.28 30.68 1016.0 168.7 83.9 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 169.3 88.0 03036 16.0 16.41 4197.0 0.29 42.79 1195.0 183.0 99.2 04313 18.0 15.61 4349.0 0.29 45.54 1244.0 186.7 102.2 03003 12.0 15.21 4430.0 0.29 47.02 1270.0 188.6 103.8 03036A 13.0 14.90 4495.0 0.29 48.24 1291. 0 190.2 105.1

     . 04455             16.0         11.41       5383.0         0.30         65.70       1512.0    ,.205.8   122.8 04286             15.0         10.42       5715 .o        0.31         72.66       1543.0      208.0    129.8
 "'
 ~
                                                    . '

Table 4 Calvert Cliffs I Limiting and Controlling Values of J 10 and KJc of 70°F RCP PUMP CASING HEAT # Jic KJc (lb/in) (ksi/in) 681-N-0437 04146 594.5 158.5 681-N-0438 43836 728.6 175.4 681-N-0439 44581 621. 3 162.0 681-N-0440 04146 594.5 158.5 N U1

Table 4-7 Calvert Cliffs I Limiting and Controlling Values of JIC and KJC at 550°F RCP. PUMP CASING HEAT # Jic KJc (lb/in) (ksi/ in) 681-N-0437 04146 813.4 151.0 681-N-0438 43836 951.3 163. 3' 681-N-0439 44581 839.3 153.4 681-N-0440 04146 813.4 151. 0

)>                                            ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL "C

"C Table 4-8 REACTOR COOLANT PUMP CASINGS 11>

s a.

--'*

x REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS (""') MATERIAL HEAT NO. c Mn Si s p Cr Ni Mo N Cb

         ** PLANT l.D. CALVERT CLIFFS UNIT 2

• RCP PUMP CASING 681-N-0441 HUB/DIFFUSER 48396 0.06 0.60 0.85 0.004 0.040 19.08 9.63 2.52 0.04 0.00 CASING WELD 04460 0.02 1.00 0.51 0.015 0.012 20.35 10.00 2.37 0 .04 . 0 .00 CASING WELD 04459 0.02 0.91 0.51 0.014 0.013 19.82 9.44 2.46 0.04 0.00 CASING WELD 03036A 0.04 1.52 0.47 0.015 0.014 18.69 9.89 2.84 0.04 0.00 CASING WELD 04509 0.02 0.94 0.45 0.018 0.015 19.67 9.75 2.51 0.04 ' 0 .00 CASING WELD 04635 0.02 1.00 0.49 0.015 0.013 19.40 10.00 2. 71 0.04 0.00 CASING WELD T03951 0.04 1.55 0.50 0.014 0.025 19.02 10.03 2.44 0.04 0.00 CASE SCROLL 48734 0.08 0.68 0.87 0.015 .0.038 18.73 9.47 2.10 0.04 0.00 CASING WELD 01953 0.02 1.68 0.44 0.021 0.015 19.11 9.69 2.83 0.04 0.00 CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 19.90 9.10 2.31 0.04 0.00 CASING WELD 57203 0.02 0.66 0.48 0.017 0.039 18.65 10.20 2.41 0.04 0.00
• RCP PUMP CASING 681-N-0442 CASE SCROLL 49122 0.06 0.75 1.35 0.016 0.037 19.10 9.42 2.06 0.04 0.00 HUB/DIFFUSER 48898 0.07 0.64 1.06 0.004 0.037 19.59 9.51 2.22 0.04 0.00 CASING WELD 04460 0.02 1.00 0.51 0.015 0.012 20.35 10.00 2.37 0.04 0.00 CASING WELD 04635 0.02 1.00 0.49 0.015 0.013 19.40 10.00 2. 71 0.04 0.00 CASING WELD 03951 0.04 1.55 0.50 0.014 0.025 19.02 10.03 2.44 0.04 0.00 CASING WELD 03036A 0.04 1.52 0.47 0.015 0.014 18.69 9.89 2.84 0.04 0.00 CASING WELD 04509 . 0.02 0.94 0.45 0.018 0.015 19.67 9.75 2.51 0.04 0.00 CASING WELD T03951 0.04 1.55 0.50 0.014 0.025 19.02 10.03 2.44 0.04 0.00 CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 19.90 9.10 2.31 0.04 0.00 CASING WELD 57203 0.02 0.66 0.48 0.021 0.039 18.65 10.20 2.41 0.04 0.00 N
• RCP PUMP CASING 681-N-0443

   -...J HUB/DIFFUSER 51799      0.05 0.66   1.14 0.012 0.040 19.45       9.69 2.42 0.04 0.00 e                                                    e

x

. ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL "C

"C Table 4-8 (Continued) REACTOR COOLANT PUMP CASINGS ft>

J a.

......

x REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS ('") MATERIAL HEAT NO. c Mn Si s p Cr Ni Mo N Cb CASING WELD 04528 0.02 0.98 . 0.51 0.014 0.013 20.29 10.30 2.62 0.04 0.00 CASING WELD 04818 0.03 1.01 0.49 0.020 0.012 19.56 10.05 2.39 0.04 0.00 CASING WELD 04754 0.02 0.97 0 . .47 0.017 0.014 19.51 9.87 2.46 0.04 0.00 CASING WELD 04509 0.02 0.94 0.45 0.018 0.015 19.67 9.75 2.51 0.04 0.00 CASING WELD 04635 0.02 1.00 0~49 0.015 0.013 19.40 10.00 2.71 0.04 0.00 CASING WELD T03951 0.04 1.55 0.50 0.014 0.025 19.02 10.03 2.44 0.04 0.00 CASE SCROLL 50404 0.03 0.51 1.03 0.015 0.035 18.32 9.52 2.20 0.04 0.00 CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 19.90 9.10 2.31 0.04 '0.00 CASING WELD 57203 " 0.02 0.66 0.48 0.017 0.039 18.65 10.20 2.41 0.04 0.00

• RCP PUMP CASING 681-N-0444 CASE SCROLL 51226 0.06 0.66 0.80 0.015 0.039 19.39 9.57 2.21 0.04 0.00 HUB/DIFFUSER 50658 0.04 0.58 1.31 0.015 0.036 18.37 9.37 2.30 0.04 0.00 CASING WELD 04754 0.02 0.97 0.47 0.017 0.014 19.51 9.87 2.46 0.04 0.00 CASING WELD 04635 0.02 1.00 0.49 0.015 0.013 19.40 10.00 2. 71 0.04 0.00 CASING WELD 57203 0.02 0.66 0.48 0.017 0.039 18.65 10.20 2.41 0.04 0.00 N

co

   )>
  "O Table 4-9                 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL "O

l'D REACTOR COOLANT PUMP CASINGS

s a.

   --'*
   ><

REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES n MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) {%) {%) STRESS STRESS STRESS

                                                                                             @ 70F   @ 70F   @ 550F J.

i

         ** PLANT I.D. CALVERT CLIFFS UNIT 2

• RCP PUMP CASING 681-N-0441 80728 60134 HUB/DIFFUSER A 351 CF8M 48396 45.4 87.2 57.0 76.0 66 CASING WELD A 371 ER-316 04460 o.o 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04459 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03036A 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04509 0.0 0.0 - 0.0 o.o 0 78600 58000 CASING WELD 04635 0.0 0.0 0.0 o.o 0 78600 58000 CASING WELD A 298
• E316-16 T03951 o.o . o.o o.o 0.0 0 78600 58000 CASE SCROLL A 351 CFSM 48734 35.7 77 .1 51.0 65.0 56 65053 44459*

CASING WELD A 371 ER-316 01953 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD. A 298 E316-16 X43439 0.0 o.o o.o 0.0 0 78600 58000 CASING WELD A 298 E316-16 57203 0.0 o.o 0.0 0.0 0 78600 58000

• RCP PUMP CASING 68l~N-0442 56888 CASE SCROLL A 351 CF8M 49122 43.5 85.0 50.0 75.0 64 77482 HUB/DIFFUSER A 351 CF8M 48898 43.0 86.3 58.0 75.0 65 78115 57521 CASING WELD A 371 ER-316 04460 0.0 o.o 0.0 0.0 0 78600 58000 CASING WELD 04635 0.0 0.0 o.o 0.0 0 78600 58000 CASING WELD A 298 E316-16 03951 . 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03036A o.o 0.0 0.0 0.0 0 78600 .: 58000 CASING WELD A 371 ER-316 04509 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 T03951 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 x43439* 0.0 o.o 0.0 o.o 0 78f;OO 58000 CASING WELD A 298 E316-16 57203 0.0 0.0 0.0 0.0 0 78600 58000 N

     \.0

• RCP PUMP CASING 681-N-0443 HUB/DIFFUSER A 351 CF8M 51799 41.4 86.7 58.0 73.0 64 77165 56571 0.0 0.0 0.0 0 78600

                                                                    ~*

CASIN~ELD 04528 5800. CASIN ELD A 5.4 . AS.469 04818 0.0 0.0 o.o 0 78600 58000

.?; Table 4-9 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL "'C n> REACTOR COOLANT PUMP CASINGS

s 0..

--'*
)<

REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES n AGED AGED MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                         @ 70F   @ 70F @ 550F
                                                                     ,.

CASING WELD A 351 CF8M 04754 o.o 0.0 0.0 0.0 0 78600 58000 58000 CASING WELD A 371 ER-316 04509 0.0 0.0 0.0 0.0 0 78600 CASING WELD 04635 o.o 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 T03951 0.0 0.0 o.o 0.0 0 78600 58000 CASE SCROLL A 351 CFAM 50404 38. l 74.9 53.0 75.0 56 65211 44617 CASING WELD A 298 E316-16 X43439 o.o 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 57203 o.o 0.0 0.0 0.0 0 78600 58000

• RCP PUMP CASING 681-N-0444 50001 CASE SCROLL A 351 CF8M 51226 39.l 80.7 50.0 75.0 60 70595 HUB/DIFFUSER A 351 CF8M 50658 35.2 89.7 52.0 54.0 62 74632 54038 CASING WELD A 351 CF8M 04754 o.o o.o o.o 0.0 0 78600 58000 CASING WELD 04635 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 57203 0.0 o.o 0.0 0.0 0 78600 58000 w

C>

)>           Table 4-10                     ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

-c -c C'D REACTOR COOLANT PUMP CASINGS

I a.

-'*

x REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR n HEAT MEASURED CHROMIUM NICKEL Cre/Nie CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS #1 (%) #2 (%) WELDS WELDS {%)

       ** PLANT 1.0. CALVERT CLIFFS UNIT 2

• RCP PUMP CASING 681-N-0441 12.930 10.0 48396 12.0 17 .547 14.669 1.20 13.2 13.5 22.875 04460 20.0 18.473 14.097 1.31 17 .8 22.7 23.485 12.300 15.o 04459 21.0 18.051 13.529 1.33 19.5 25.0 23.045 11. 695 17.0 03036A 13.0 17. 362 14.474 1.20 11.4 13.8 22.235 12.990 13.8 04509 19.0 17 '933 13.842 1.30 17 .2 21.4 22.855 12.020. 14.0 04635 18.0 17 '924 14.097 1.27 16.2 19.3 22.845 12.300 13.0 T03951 ~.o 17. 222 14.568 1.18 .. 11.4 12.6 22.210. 13.085 13.8 48734 7.0 16.699 14.958 1.12 9.3 8.6 22.135 13.350 14.2 01953 18.0 17. 755 13.798 1.29 15.7 20.6 22.600 12.270 12.0 X43439 14.0 17 .878 13.477 1.33 17 .2 24.3 22.750 11. 895 11. 0 57203 14.0 16.806 14.289 1.18 11.8 . 12' 2 21.780 12.360 14.0
• RCP PUMP CASING 681-N-0442 12.0 49122 14.0 17.251 14.474 1.19 14.2 13.2 23.185 12.795 48898 12.0 17.795 14.798 1.20 13.7 14.0 23.400 13.130 0.0 04460 20.0 18.473 14.097 1.31 17 .8 .. 22.7 23.485 12.300 15.0 04635 . o.o 17. 924 14.097 1.27 16.2 19.3 22.845 12.300 13.0 03951 13.0 17' 222 14.568 1.18 10.5 12.6 22.210 13.085 12.3 03036A 13.0 17.362 14.474 1.20 11.4 13.8 22.235 12.990 13.8 04509 19.0 17.933 13.842 1.30 17 .2 21.4 22.855 12.020 14.0 T03951 o.o 17.222 14.568 1.18 10.5 12.6 22.210 13.085 13.8 X43439 14.0 17 .878 13.471 1.33 17 .2 24.3 22.750 11. 895 11:. 0 57203 0.0 16.806 14.289 1.18 11.8 12.2 21. 780 12.360 14.0 w

 .....

• RCP PUMP CASING 681-N-0443 51799 15.0 17. 935 14.490 1.24 16.2 16.6 23.580 12.720 14.0 04528-04818 19.0 *18.715 14.396 0.0 17.697 14.271 1.30 1.24 17 .2 14.2 e 21.8 16.8 23.675 22.685 12.590 12.505 14.0

11. 0

)>   Table 4-10 (Continued)                 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

'U 'U REACTOR COOLANT PUMP CASINGS rD

3 a..

-'*

REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR

)(

("') HEAT MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS #1 (%) #2 (%) WELDS WELDS (%) 04754 18.0 17.722 13.965 1.27 15.7 19.1 22.675 12.155 14.0 04509 19.0 17. 933 13.842 1.30 17.2 21.4 22.855 12.020 14.0 04635 18.0 17.924 14.097 1.27 16.2 19.3 22.845 12.300 13.0 T03951 0.0 17.222 14.568 1.18 10.5 12.6 22.210 13.085 13.8 50404 13.0 16.486 13.815 1.19 14.7 . 13.3 22.065 11. 875 12.0 X43439 14.0 17 .878 13. 477 1.33 17 .2 24.3 22.750 11.895 11.0 57203 0.0 16.806 14.289 1.18 11.8 12.2 21.780 I 12,360 14.0

• RCP PUMP CASING 681-N-0444 51226 12.0 17. 458 14.615 1.19 12.8 13.4 22.800 12.900 11.0 50658 14.0 16.792 13.917 1.21 15.7 14.3 22.635 12.060 13.0 04754 18.0 17.722 13.965 1.27 15.7 19.1 22.675 12.155 14.0 04635 18.0 17. 924 14.097 1.27 16.2 19.3 22.845 12.300 13.0 57203 14.0 16.806 .14. 289 1.18 11.8 12.2 21. 780 12.360 14.0 w

N

e Table 4-11 .

    )>

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

   "'C
   "'C COOLANT PUMP CASINGS AT 70F Cl>

I

    .....

0.. REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x ("") HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (%) @ 70F @ 70F (ft- lbs) (lb/in) (ksiVin)

           ** PLANT 1.0. CALVERT CLIFFS UNIT 2

• RCP PUMP CASING 681-N-0441 48396 12.0 21.54 4441.0 0.38 30.18 783.3 181.9 61. 5 X43439 14.0 19.87 4804.0 o.39 33.51 815.6 . 185.6 69.2 04460 20.0 18.40 5176.0 *o.39 37.00 885.2 193.4 74.1 04459 21.0 17.66 5387.0 0.39 39.00 926.2 197 .8 . 76.8 04509 19.0 15.76 6009.0 0.40 45.00 994.7 205.0 84.4 48734 7.0 16.26 5834.0 0.39 43.30 1040.0 209.5 118.3 03036A 13.0 14.90 6341.0 0.40 48.24 1057.0 211.3 88.6 I

04635 18.0 14.83 6366.0 0.40 48.49 1061.0 211. 7 88.9 T03951 0.0 13.79 6818.0 0.40 52.96 1149 .o 220.3 94.5 I

• 01953 18.0 13.75 6836.0 0.40 53.14 1152.0 220.6 94.8 57203 14.0 9.58 9394.0 0.42 79.18 1506.0 252.2 124.2
• RCP PUMP CASING 681-N-0442 48898 12.0 26.29 3673.0 0.38 23.35 639.5 164.4 54.9 49122 14.0 22.27 4301.0 0.38 28.90 760.0 179.2 64.6 X43439 14.0 19.87 4804.0 0.39 33.51 815.6 185.6 69.2 04460 20.0 18.40 5176.0 0.39 37.00 885.2 193.4 74.1 04509 19.0 15.76 6009.0 0.40 45.00 994.7 205.0 84.4 03036A 13.0 14.90 6341.0 0.40 48.24 1057.0 211.3 88.6

          ' 04635               0.0         14.83       6366.0          0.40         48.49      1061.0         211. 7    88.9 T03951              0.0         13.71       6818.0          0.40         52.96      1149.0         220.3     94.5 03951              13.0         13.7        6818.0 .        0.40         52.96      1149.0        .220.3     94.5 57203               0.0          9.58       9394.0          0.42         79.18      1506.0       : 252.2    124.2 w

w

• RCP PUMP CASING 681-N-0443 3857.0 0.38 24.94 675.3 . 168. 9* 58.8 51799 X434.

15.0 14.0 24.96 19.87 4804.0 ~.39. 33.51 815.6 185.6 6tt

e Table 4-11 (Continued) . )>

 "'C ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR
 "'C C1)

C.OOLANT PUMP CASINGS AT 70F

s a.

  --'*

x REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL ("") HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @70F @ 70F (%) @ 70F @ 70F (ft-lbs) (1 b/i n) (ksiVin) 04528 19.0 18.60 5120.0 0.39 36.47 874.0 192.1 73.4 04509 19.0 15.76 6009.0 0.40 45.00 994.7 205.0 84.4 04818 0.0 14.93 6329.0 0.40 48.12 1055.0 211.1 88.5 04635 18.0 14.83 6366.0 0.40 48.49 1061.0 211. 7 88.9 04754 18.0 14.34 6571.0 0.40 50.51 1101.0 . 215. 7 91. 5 T03951 0.0 13.79 6818.0 0.40 52.96 1149.0 220.3 94.5 50404 13.0 12.50 7461.0 0.41 59.41 1248.0 229.6 144.7 57203 0.0 9.58 9394.0 0.42 79.18 1506.0 252.2 124.2

• RCP PUMP CASING 681-N-0444 51226 12.0 21.25 4500.0 0.39 30. 71 766.8 180.0 80.1 50658 14.0 16.72 5677.0 0.39 41. 78 988.1 204.3 89.0 04635 18.0 14.83 6366.0 0.40 48.49 1061. 0 211. 7 88.9 04754 18.0 14.34 6571.0 0.40 50.51 1101. 0 215.7 91. 5 57203 14.0 9.58 9394.0 0.42 79.18 1506.0 252.2 124.2

e Table 4-12

)>

"O ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "O (I) REACTOR COOLANT PUMP CASINGS AT 550F

I 0..

-'*

x REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL n HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ SSOF (%) @550F @ 550F (ft-lbs) ( 1 b/ in) (ksi/1n)

    ** PLANT I.D. CALVERT CLIFFS UNIT 2

• RCP PUMP CASING 681-N-0441 48396 12.0 21.54 3445.0 0.28 30.18 1002.0 167.6 77.8 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 . 169.3 88.0 04460 20.0 18.40 3864.0 0.29 37.00 1088.0 174.6 92.4 04459 21.0 17.66 3980.0 0.29 39.00 1126.0 177 .6 94.8 04509 19.0 15.76 4319.0 0.29 45.00 1234.0 186.0 101. 6 48734 7.0 16.26 4225.0 0.29 43.30 1250.0 187.2 162.1 03036A 13.0 14.90 4495.0 0.29 48.24 1291. 0 190.2 105.l 04635 18.0 14.83 4509.0 0.29 48.49 1295.0 190.5 105.3 103951 0.0 13.79 4745.0 0.30 52.96 1307.0 191.4 110.6 01953 18.0 13.75 4754.0 0.30 53.14 1311. 0 191. 7 110.8 57203 14.0 9.58 6016.0 0.31 79.18 1638.0 214.3 135.3
• RCP PUMP CASING 681-N-0442 48898 12.0 26.29 2988.0 0.28 . 23.35 859.9 155.2 74.6 49122 14.0 22.27 3363.0 0.28 28.90 981. 5 165.8 84.5 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 169.3 88.0 04460 20.0 18.40 3864.0 0.29 37.00 1088.0 174.6 92.4 04509 19.0 15.76 4319.0 0.29 45.00 1234.0 186.0 101. 6 03036A 13.0 '14.90 4495.0 0.29 48.24 1291. 0 190.2 105.1 04635 0.0 14.83 4509.0 0.29 48.49 1295.0 190.5 105.3 103951 0.0 13.7~ 4745.0 0.30 52.96 1307.0 191.4 110. 6 03951 13.0 13. 7' 4745.0 0.30 52.96 1307.0 .191.4 110.6 57203 0.0 9.58 6016.0 0.31 79.18 1638.0 ' 214.3 135.3 w

01

• RCP PUMP CASING 681-N-0443 51799 15.0 24.96 3099.0 0.28 24.94 897.2 158.6 79.5

                                                             -

X43439 14.0 19.87 3654.0 o.~9 33.51 1023.0 169.3 88.0

e. e

e Table 4-12 (Continued)

)>

"'C ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "'C

'1>

REACTOR.COOLANT PUMP CASINGS AT 550F

J 0..

......                                    REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x

n HEAT MEASURED MATERIAL MATERIAL MATERIAL . MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER .c N ENERGY * @ 550F @ 550F @ 550F (%) @ 550F @ 550F (ft-lbs) (lb/in) (ksi/in) 04528 19.0 18.60 3832.0 0.29 36.47 1079.0 173.9 91.8 04509 19.0 15.76 4319.0 0.29 45.00 1234.0 186.0 101. 6 04818 0.0 14.93 4489.0 0.29 48.12 1289.0 190.1 104.9 04635 18.0 14.83 4509.0 0.29 48.49 1295.0 190.5 105.3 04754 18.0 14.34 4617 .0 0.30 50.51 1268.0 188.5 108 .1 T03951 0.0 13.79 4745.0 0.30 . 52. 96 1307.0 191.4 110.6 50404 13.0 12.50 5074.0 0.30 59.41 1473.0 203.2 187 .8 57203 0.0 9.58 6016.0 0.31 79.18 1638.0 214.3 135.3

• RCP PUMP CASING 681-N-0444 51226 , 12.0 21.25 3479.0 0.28 30.71 1035.0 170.3 110.5 50658 14.0 16.72 4140.0 0.29 41.78 1187 .0 182.4 .111.6 04635 18.0 14.83 4509.0 0.29 48.49 1295.0 190.5 105.3 04754 18.0 14.34 4617 .0 0.30 50.51 1268.0 188.5 108 .1 57203 14.0 9.58 6016.0 0.31 79.18 1638.0 . 214.3 135. 3 w

O'I

Table 4 Calvert Cliffs 2 Limiting and Controlling Values of Jic and KJc at 70°F RCP PUMP CASING HEAT # Jic KJc (lb/in) (ksi/in) 681-N-0441 48396 783.3 181.9 681-N-0442 48898 639.5 164.4 681-N-0443 51799 675.3 168.9 681-N-0444 51226 766.8 180.0 w -.J

Table 4 Calvert Cliffs 2 Limiting and Controlling Values of Jic and KJc at 550°F RCP PUMP CASING HEAT # Jic KJc (lb/ in) (ksi/in) 681-N-0441 48396 1002.0 167.6 681-N-0442 48898 859.9 155.2 681-N-0443 51799 897.2 158.6 681-N-0444 51226 1035.0 170.3

5.0 CRACK GROWTH ANALYSIS In this section, the methodologies discussed in Section 5.1 of the main report are applied to Calvert Cliffs plant-specific conditions. The growth history of a worst-case hypothetical crack is conservatively developed based on information contained in the vendor's stress analysis report. 5.1 Scope The analysis which follows pertains to the 35x35x33 DFSS Reactor Coolant Pump casings, Serial Numbers 681-N-0437 to 681-N-0444, inclusive, at the Calvert Cliffs Units 1&2. 5.2 Reference Stress Reports The stresses used in the hypothetical crack growth analyses are from the stress report prepared by the Byron-Jackson company in August, 1974 (Reference 3-2). 5.3 Selection of High Stress Locations The methodology described in s.1.1 of the main report was applied to identify five regions as potentially limiting: (1) Diffuser Vane 8--Level D (2) Discharge Nozzle--Section c, adjacent to Crotch Region (3) Suction Nozzle--Level.C {4) Junction, Volute with Lower Flange {5) Hanger Bracket #1 Vicinity All other regions in the stress summary were considered and were found to have lower stresses than the above regions. Appendix *c_ 39

5.4 Stresses and Wall Thicknesses at Limiting Locations Membrane and through-wall bending components of the limiting regions were obtained from Reference 3-2 under the Design Condition basis, as follows: (1) For Diffuser Vane 8--Level D: Design Condition= # 103, plus thermal gradient stresses Key Elements = # 3828 & 3832 in Finite-Element Model Membrane stress= 20.95 Ksi (Figure 4-3(b)) Bending stress = 19.87 Ksi (Figures 3-1 & 4-3(b)) Thickness = 4.75" (Figure 4-3(b)) Bending stress includes 3.4 Ksi due to a surface-to-interior temperature difference of l8°F during heatup/cooldown. (2) For Discharge Nozzle--Section c, adjacent to crotch Region: Design Condition = # 107, plus thermal gradient stresses Key Element = # 5125 in Finite-Element Model Membrane stress = 17.4 Ksi (Table 4-6) Bending stress = 21.1 Ksi (Figure 4-8, p. 66 & Figure 3-1) Thickness = 3.3 in. (Table 4-6) Bending stress includes 5.5 Ksi due to a through-wall temperature difference of 29°F during heatup/cooldown. Appendix C 40

(3) For Suction Nozzle--Level C: Design Condition = # 104, plus thermal gradient stresses Key Elements = # 2125 & 2129 in Finite-Element Model Membrane stress = 22.15 Ksi (Figure 4-5(b)) Bending stress = 13.15 Ksi (Figures 3-1 & 4-5(b)) Thickness = 3.0 in. (Figure 4-5(b)) Bending stress includes 5.5 Ksi due to a through-wall temperature difference of 29°F during heatup/cooldown. (4) For Junction, Volute with Lower Flange {vicinity of Vanes 1 & 2): Design condition = #112, plus thermal gradient stresses Key Elements = # 1279, 1283, & 1291 in Finite-Element Model Membrane stress = 17.82 Ksi (page 78) Bending stress_ = 13.2 Ksi {page 78 & Figure 3-3) Thickness = 3.375 in. Bending stress includes 3.0 Ksi due to a through-wall temperature difference of 16°F during heatup/cooldown {Hydraulic Section 3). Bending stress also includes 10.2 Ksi under Design Condition 112, conservatively derived from the Report declaration that surface stresses do not exceed 28.05 Ksi (1.5 Sm). (5) For Hanger Bracket #1 Vicinity: Design Condition = Maximum (pp. 74, 75) Key Elements = # 7461 in Finite-Element Model Membrane stress = 18.9 Ksi (page 75) Appendix c 41

Bendirig stress = 22.1 Ksi (page 75, 97) Thickness = 3.6 in. Bending stress includes 13.7 Ksi due to through-wall temperature difference of 72.7°F for Hydraulic Section s during the heatup/cooldown transients. 5.5 Calculation of Crack Growth Rates The methodology described in Section 5.1.4 of the generic report was applied to the above values, using the annual rate of stress-cycling given in Referenees 3-1 and 3-2. An integration procedure was used to predict dimensionless crack depth, a/t, as a function of time, T. Results are summarized in Tables 5-1 through 5-5. For each region the calculated entries are listed against crack depth, a/t, as follows: (1) Applied Stress Intensity Factor, KI' was calculated using the ASME Section XI procedure, as further described in Section 5.1.4 of the generic report *.. Units for KI are Ksi--squareroot inch. (2) Crack growth rate, da/dT, was calculated using the Bernard & Slama equation (with R=O), multiplied by the design value of the annual rate of stress cycling, as further described in Section 5.1.4. The final equation is: da/dT = 5.435 x 10-lO K 4 .o I Units for da/dT are inches/year. (3) Incremental time, dT, in which the crack will grow~ through the indicated dimensionless crack depth ~ Appendix c 42

Table 5-1 Calvert Cliffs 1 & 2 crack Growth Rates At Vane Number 8 Level D a/t Interval KI da/dT ~Time {fraction} {KSI /Ill} {IN/YEAR} {YEARS} 0.08 -- 0.10 49.98 3.39 x 10-3 28.0 0.10 -- 0.15 60.08 7.08 x 10-3 45.6* 10- 2

                                                                               .;;

0.15 -- 0.20 68.76 1.21 x 24.4* 10- 2

                                                                               *H 0.20  --  0.25                76.76              1.89 x                  12.5 0.25  --  0.30                84.17              2.73 x  10- 2             8.7 0.30  --  0.35                91.69              3.84 x  10- 2             6.2 0.35  --  0.40                99.35              5.29 x  10- 2             4.5 0.40  --  0.45               108.3               7.48 x  10- 2             3.2 0.45  --  0.50               118.4               0.107                     2.2 (um= 20.95, ab* 19.87, t

• 4.75")
• Sum of five time steps through 1% a/t increments using interpolated KI values.

Appendix c 43

Table 5-2 Calvert Cliffs 1 & 2 Crack Growth Rates At Discharge Nozzle Crotch Vicinity - Section c a/t Interval KI da/dT ~Time {fraction} {KSI /IN} {IN/YEAR} {YEARS} 0.08 -- 0.10 38.9 I. 25 x 10- 3 52.8 0 .10 -- 0 .15 46.6 2.56 x 10- 3 64.5 0.15 -- 0.20 53.0 4.30 x 10- 3 38.4 0.20 -- 0.25 58.9 6.55 x 10- 3 25.2 0.25 -- *o.3o 64.3 9.28 x 10- 3 17.7 0.30 -- 0.35 69.7 I. 28 x 10- 2 12.8 0.35 -- 0.40 75.2 I. 74 x 10- 2 9.5 0.40 -- 0.45 81.6 2.41 'x 10- 2 6.8 0.45 -- 0.50 88.9 3.39 x 10- 2 4.8 (O'm = 17.4, O'b = 21.1, t = 3.3") Appendix C 44.

Table 5-3 Calvert Cliffs 1 & 2 Crack Growth Rates At Suction Nozzle - Level C a/t Interval KI da/dT AT (fraction} (KSI /IN) (IN/YEAR) (YEARS} 0.08 -- 0.10 34.92 8.08 x 10- 4 74.2 0.10 -- 0.15 42.37 1. 75 x 10- 3 85.6 0.15 -- 0.20 48.93 3.12 x 10- 3 48. l 0.20 -- 0.25 55.08 5.00 x 10- 3 30.0 0.25 -- 0.30 60.93 7.49 x 10- 3 20.0 0.30 -- 0.35 66.88 1.09 x 10- 2 13.8 0.35 -- 0.40 73.05 1. 55 x 10- 2 9.7 0.40 -- 0.45 80.20 2.25 x 10- 2 6.7

0. 45 -- 0. 50 . 88.25 3 .30 x 10- 2 . 4.5 Appendix c 45

Table 5-4 Calvert Cliffs 1 & 2 Crack Growth Rates At Volute Junction with Lower Flange Near Vanes 1 & 2 From - To KI da/dT 6T {aLt = } {KSI /IN} {INLY EAR} (YEARS} 0.08 -- 0.10 32.27 5.9 x 10- 4 114. 0.10 -- 0.15 38.99 1. 26 x 10- 3 134. 0.15 -- 0.20 44.85 2.2 x 10- 3 77. 0.20 -- 0.25 50.29 3.48 x 10- 3 48.5 0.25 -- 0.30 55.41 5.12 x 10- 3 33. 0.30 -- 0.35 60.61 7.33 x 10- 3 23. 0.35 -- 0.40 65.97 1.03 x 10- 2 16.4 0.40 -- 0.45 72.20 1.48 x 10- 2 11.4 0.45 -- 0.50 79.21 2.14 x 10- 2 7.9 e (um= 17.8, ub = 13.2, t = 3.375) Appendix c 46 ---~----

Table 5-5 Calvert Cliffs 1&2 crack Growth Rates At Hanger Bracket #1 Vicinity From - To Kr da/dT ~T {aLt = } (KSr Im) (rNLYEAR) {YEARS} 0.08 -- 0.10 43.40 1.92 x 10- 3 37.4 0 .10 -- 0 .15 51.90 3.94 x 10- 3 61.3* 0.15 -- 0.20 59 .10 6.65 x 10- 3 33.5* 0.20 -- 0.25 65.80 0.0102 17 .6 0.25 -- 0.30 71.80 0.0145 12.4 0.30 -- 0.35 77.90 0.02 9.0 0.35 -- 0.40 84.10 0.0272 6.6 0.40 -- 0.45 91.40 0.0379 4.7 0.45 -- 0.50 99.50 0.0533 3.4

                              -*

(um = 18.9, ub z 22.1, t

• 3.6")
• The sum of five time steps through 0.01 a/t increments using interpolated Kr values.

Appendix C 47

values, a/t, was calculated as described in ~ Section 5.1.4 of the generic report. Units for dT are years. The summation of time increments yields the total Time for a crack to grow to a given a/t value. The predicted growth curves for hypothetical cracks (Figure 5.1-5 of the generic report) show the functional relationship between a/t and total Time, using the initial condition of a/t = 0.08 at Time = o. The first incremental time listed in Tables 5-1 through 5-5 is based on a change in a/t in the amount 0.02 (i.e. 2% of thickness), to indicate the time needed for the crack to grow from a/t = 0.08 to a/t = 0.10. Subsequent incremental times are based on the time needed for the crack to grow through changes in a/t of 0.05 (i.e. 5% of thickness). The first such incremental time is qiven for the ranqe a/t = 0.10 to 0.15. The final incremental time is qiven ~ for the ranqe a/t = 0.45 to 0.50, whereupon the analysis is terminated. 5.6 stresses Under.Emergency and Faulted conditions In order to verify that limitinq sections containinq hypothetical cracks can withstand Emerqency Condition and Faulted Condition Loads, the methodoloqy described in 5.1.4 of the qeneric report was again applied. Applied stress intensity factors were calculated at the limiting locations, based on available data from Reference 3-1. 5.6.1 Emergency Condition Stresses (1) Diffuser Vane 8--Level D: Key elements = # 3828 & 3832 Condition = # 505 Appendix c 48

Membrane stress = 15.5 Ksi (p. 109) Bending stress = 13.2 Ksi (p. 109) These are conservative values, bounded by Vane 9--Level A results, with secondary stresses removed. (2) Discharge Nozzle--Section C, adjacent to Crotch Region: Key element = # 5125 Condition = # 506 Membrane stress = 26.66 Ksi (p. 118) Bending stress = 16.85 Ksi (p. 118) (3") Suction Nozzle--Level C: Key elements = # 2125 & 2129 Condition = # 511 Membrane stress = 25.5 Ksi (p. 112) Bending stress = 11.88 Ksi (p. 114) Bendinq stress is conservatively bounded by Condition 503 results, with secondary stresses removed. (4) Junction, Volute with Lower Flanqe: Key elements = # 1279, 1283 & 1291 No results are published specifically for this reqion. Appendix c 49

(5) Hanger Bracket Vicinity: Key elements = I 7461 Condition = emergency, worst case Membrane stress = 23.2 Ksi (p. 106) Bending stress = 11. 6 Kai (p. 106) Stresses are conservatively set to Emergency Conditions allowables for a worst case analysis. 5.6.2 Faulted Condition Stresses (1) Diffuser Vane 8--Level D Membrane stress = 23.22 Ksi (p. 143) Bending stress = 23.55 Ksi (p. 143) These are conservative values, bounded by Vane 9--Level A results from elements 3904 and 3905 under condition 606. (2) Discharge Nozzle, adjacent to Crotch Region: Membrane stress = 32.0 Ksi (p. 141), upper bound Bending stress = 6.4 Ksi (p. 148) These are worst case results under Faulted Condition 606, with secondary stresses removed. (3) Suction Nozzle--Level C: Membrane stress = 25.96 Ksi (p. 145) Bending stress = 16.7 Ksi (p. 145) Appendix c

  --------- = = = = = = =                                -----~=

50 ---- - - - - -

These are worst case results under Faulted Condition 603, with secondary stresses removed. (4) Junction, Volute with Lower Flange: No Faulted Condition results are published specifically for this region. (5) Hanger Bracket Vicinity: Membrane stress = 29.0 Ksi (pp. 137, 139) Bending stress = 14.5 Ksi (p. 139) These are conservative values based on meeting Faulted Conditions allowables. 5.7 Results Results reported above and shown in Figure 5.3-13 of the generic portion of this report for the Calvert Cliffs 1&2 RCPs indicate that the postulated 8%t initial crack will grow to 25%t in abo?t 110 years under the influence of the conservatively defined stress cycles in t~e design specification. The hypothesized crack will then grow larger until it r~aches an end-point crack size of 38%t, limited by flow stress, in about 130 years. Appendix C 51

6.0 INSPECTION INTERVAL Results reported in this appendix support the position that the 10-year inspection interval for volumetric examination, as required by ASME Section XI, is not necessary to ensure safe operation during the 40-year licensed life of the plant. The

                                  '

conservatively calculated end-point crack size is not reached until 130 years after initial operation. The demonstrated casing integrity also supports a relaxation of the 10-year interval for visual examinations, as currently required by ASME Section XI and Code Case N-481. Such examinations add unnecessarily to personnel exposure with no benefit to plant safety. The ASME Section XI requirement for VT-3 examination of internal surfaces is an appropriate low-marginal-exposure monitoring activity to the extent practicable, but only when the pump is disassembled for maintenance or repair. Based upon the results obtained in this evaluation, relaxation of the casing inspection interval for the Calvert Cliffs RCPs from 10 years to 40 years is considered to be justified.- Appendix c 52

APPENDIX C REFERENCES 3-1 Engineering Specification for Reactor Coolant Pumps for Baltimore Gas and Electric Company, Calvert Cliffs Station, 8067-31-3, Rev. 7, (March 1971). 3-2 Report TCF 1015-STR, Vol. 1, Rev. 1, "Pump Case Structural Analysis", dated August 7, 1974 3-3 Letter, L. D. Smith, BG&E to A. G. Schoenbrunn, ABB C-E Nuclear Power,,dated 8/27/91 4-1 o. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems", NUREG/CR-4513 (ANL-90/42), U.S. Nuclear Regulatory Commission, Washington, D.C., June 1991. Appendix c 53

APPENDIX D APPLICATION OF GENERIC METHODOLOGY FOR RELAXATION OF THE SAN ONOFRE 2&3 REACTOR COOLANT PUMP CASING INSPECTION INTERVAL Appendix D 1 -- ---------

ABSTRACT Appendix D was prepared to demonstrate the amount of inspection interval relaxation appropriate for the reactor coolant pump casings at the San Onofre 2 & 3 plants, based on application of the generic methodology presented in the main body of this report. Appendix D 2

APPENDIX D TABLE OF CONTENTS Section Title Page

1. 0 PURPOSE 6 2.0 PRE-SERVICE INSPECTION DATA EVALUATION 7 3.0 OPERATING HISTORY 9 3.1 Design Specifications 9 3.2 Stress Cycles Used in Evaluation

• 10 3.3 Stress Cycles at San Onofre 2&3 To-date 10 4.0 THERMAL EMBRITTLEMENT 12 4.1 Material Identification and Chemical Properties 12 4.2 Material Specifications and Mechanical Properties 12 4.3 Thermal Aging Behavior 13 4.4 Toughness Properties of Aged Materials 13 4.5 Limiting Values 14 5.0 CRACK GROWTH ANALYSIS 43 5.1 Scope 43 5.2 Reference ?tress Reports 43 5.3 Selection of High Stress Locations 43 5.4 Stresses and Wall Thicknesses at Limiting Locations 44 5.5 Calculation of Crack Growth Rates 46

    .5.6      Stresses Under Emergency and Faulted Conditions                          47 5.7      Results                                     56 6.0       INSPECTION INTERVAL                              57 APPENDIX D REFERENCES                            58 Appendix D                                             3

LIST OF TABLES TABLE 4-1 Material Identification and Chemical Compositions - Unit 2 15 TABLE 4-2 Material Specifications and Tensile Properties - Unit 2 17 TABLE 4-3 Predicted Thermal Aging Behavior - Unit 2 19 TABLE 4-4 Predicted Toughness Properties of Aged Materials (70°F) - Unit 2 21 TABLE 4-5 Predicted Toughness Properties of Aged Materials (550°F) - Unit 2 23 TABLE 4-6 Limiting and Controlling Values of Jic and KJc at 70°F - Unit 2 25 TABLE 4-7 Limiting and Controlling Values of Jic and KJc at 550°F - Unit 2 26 TABLE 4-8 Material Identification and Chemical Compositions - Unit 3 27 TABLE 4-9 Material Specifications and Tensile Properties - Unit 3 29 TABLE 4-10 Predicted Thermal Aging Behavior - Unit 3 31 TABLE 4-11 Predicted Toughness Properties of Aged Materials (70°F) - Unit 3 33 TABtE 4-12 Predicted Toughness Properties of Aged Materials (550°F) - Unit 3 37 Appendix D 4

LIST OF TABLES (Cont'd) Page e TABLE 4-13 Limiting and Controlling Values of Jic and KJc at 70°F - Unit 3 41 TABLE 4-14 Limiting and Controlling Values of Jic and KJc at 550°F - Unit 3 42 TABLE 5-1 Crack Growth Rate at Suction Nozzle 51 TABLE 5-2 crack Growth Rate at Junction of Volute to Lower Flange 52 TABLE 5-3 Crack Growth Rate at Crotch Region 53 TABLE 5-4 Crack Growth Rate at Diffuser Vane 54 TABLE 5-5 Crack Growth Rate at Volute (Vane Region 5) 55 Appendix D 5

1. 0 PURPOSE The purpose of Appendix D is to document the application of methodology presented in the main body of this report to the plant-specific data for the reactor coolant pump casings at the San Onofre 2&3 plants, and to quantify the extent of inspection interval relaxation available.

Appendix D 6

2.0 PRE-SERVICE INSPECTION DATA EVALUATION Pre-service inspection data for the San Onofre 2&3 reactor coolant pumps numbered 701-N-0557 through 701-N-0564 was collected from QA data packages originally prepared by the Byron Jackson Company and stored in archives by ABB Combustion Engineering Nuclear Power. Information in these data packages concerning welding procedures, radiographic inspections, non-destructive testing and dye penetrant testing were examined. The testing and inspection procedures that were followed for the eight casings all were found to be the same in all significant aspects. The most relevant information items obtained from this review of the QA data packages were the reports on radiographic examination of the RCP casing castings, pressure retaining welds, and repair welds. Radiographic examination requirements invoked ASME Section III rules for examination procedures and sensitivity. Th~ required radiograph sensitivity was 2-2T according to applicable ASTM Standard Reference Radiograph Procedure requirements (i.e. ASTM ~71, El86, E280) as determined by the casting thickness. The 2-2T sensitivity is con~istent with a 2% initial flaw size, because the requisite image quality indicator (IQI) for this level of examination is specified as a penetrameter with a minimum hole-size diameter equal to 2% of the casting thickness. The acceptance criteria for interpretation of the radiographs was Severity Level 2 for sand, porosity or shrinkage indications. Linear indications such as cracks, hot tears, and unfused chaplets or chills were unacceptable at any level. Any such discernible indications required rejection of weld repair and a repeated radiographic examination of the affected casting or weldment. Appendix D 7

The results of this review of pre-service RCP casing examinations confirm that cracks of 2% depth were within detection sensitivity as described in Section 4.4 of the generic report but none were left unrepaired. Appendix D 8

3.0 *OPERATING HISTORY 3.1 Design Specifications The San Onofre 2&3 RCPs were delivered to the site in 1979 and 1980 respectively and were first placed in commercial operation in 1983 and 1984. Reactor coolant system design pressure and temperature are 2500 psia and 650°F. Each pump is designed to deliver 99,000 gpm of coolant at a head of 300 feet. These pumps have 30 inch diameter suction and discharge piping. The design specification calls for the pumps to be capable of withstanding the following transient conditions during the 40-year license period: Assumed Occurrences Transient During 40 Year Condition License Period Heat-Up (60.F/hr) 500 Cool-Down _ (l00°F/hr) 500 Hydrostatic Test (3125 psia 100-400.F) 10 Leak Test (2250 psia 100-400*F) 200 Loss of Secondary Pressure 5 Reactor Trip 400 Appendix D 9

3.2 Stress Cycles Used in Evaluation As indicated in Section 5.1.4 of the generic portion of this report, crack growth was evaluated on the basis of the design number of stress cycles between atmospheric and operating pressures during heatup and cooldown over the nominal 40-year life of the plant. The number of such cycles used in the stress analysis performed by the Byron Jackson Company is 700, as given on Page 6-55 of Reference (3-2). This total is greater than the 500 heat-up cooldown cycles specified in the RCP design specifications. On an annual basis, the average number of stress cycles, based on 700 per 40 years, is 17.5 per year, and the hypothetical crack growth calculations and curves were prepared accordingly. 3.3 Stress Cycles at San Onofre 2&3 To-date Details of the actual operating history of the San Onofre 2&3 RCPs from 1983 (Unit 2) and 1984 (Unit 3) to 1991 were furnished in Reference (3-3) and are as follows: Unit 2 Unit 3 Heatup - 20 14 Cooldown - 19 13 Hydrostatic Test - 1 1 Leak Test - 24 15 Loss of Secondary Pressure - 0 0 Reactor Trip - 35 27 Heatup-plus-cooldown, taken together, constitute one stress cycle; the remaining events represent relatively minor stresses. The average number of stress cycles per year over the eight operation years for Unit 2 and seven years for Appendix D 10

Unit 3, ending in 1991, is 2.5 and 2.0 cycles per year -- respectively. The actual rate of stress cycle accrual has been only 14% of the design rate for Unit 2 and 11% for Unit 3, a significant conservatism. Appendix D 11

---

4.0 THERMAL EMBRITTLEMENT Thermal ernbrittlement evaluation of the San Onofre 2&3 casings is discussed and plant specific data are presented in the five following reports. All equations referenced below are found in the main body of this report, which is also referred to as the generic report. 4.1 Material Identification and Chemical Properties The chemical compositions provided in Report #1 (Table 4-1) for each RCP casing at San Onofre 2&3 were obtained from Quality Assurance documents originally supplied by the Byron Jack~on Company and stored at ABB Combustion Engineering Nuclear Power. For each individual pump casing, chemical compositions are given for specific casing welds as well as for individual castings. 4.2 Material Specifications and Mechanical Properties The material specifications and mechanical properties found in Report #2 (Table 4~2) for each RCP casing were obtained from the same data source as in Section 4.1 above. For each individual pump casing the material specification, material type and heat nwnber are given for specific casing welds as well as for individual castings. It is evident from the report that data obtained for mechanical properties (i.e. yield strength, tensile strength, total elongation and reduction in area) for each material was only available for the castings, and was not available for the casing welds. 0 The unaged flow stress at 7.0°F and 550 F and the aged flow stress at 70°F and 550°F were calculated as discussed in Section 5.2.3 of the generic report.

• Appendix D 12

4.3 Thermal Aging Behavior Report #3 (Table 4-3) contains predicted thermal aging behavior data for all of the San Onofre 2&3 RCP casings. The measured ferrite contents listed for specific casing welds and individual castings were supplied by the Byron Jackson Company in the same QA package as referenced in Section 4.1 above. In most cases a value was obtainable for the measured ferrite content. In cases where a value was not given, a zero was recorded. The chromium and nickel equivalents for the castings and weld metal, as well as the chromium/nickel ratio for the castings, were calculated using equations 5-7 and 5-8 respectively, as discussed in Section 5.2.1 of the generic report. Values for ferrite content of the castings were computed using two methods: for ferrite content #1, the values were computed using the

   *method which follows ASTM ASOO/ASOOM; for ferrite content
    #2, the values were computed using equation 5-9 as discussed in Section 5.2.1 of the generic report. The latter method follows work performed by O.K. Chopra (Reference 4-1).

4.4 Toughness Properties of Aged Materials The predicted toughness properties of aged material at 70°F and 550°F a~e given respectively in Report #4 (Table 4-4) and Report #5 (Table 4-5). The measured ferrite contents listed for all heat numbers are the same as the values given in Report #3. The material aging parameter was calculated using equation 5-10 of the generic report. The room-temperature charpy impact energy, cvsat' of the various materials was calculated using equation 5-11. The Jic values were determined in accordance with the methods of ASTM E813 as discussed in Section 5.2.3 of the generic report. The plane strain fracture toughness, KJc' and Appendix D 13

minimum tearing modulus, T, at 70°F and 550°F were calculated using equations 5-17 and 5-18. The values listed for the material constants N and c at 70°F and 550°f were calculated using equations 5-12 through 5-15. These constants were needed in computing the values for Jic' KJc and T. All equations used in Report #4 are found in Section 5.2 of the generic report. 4.5 Limiting Values The limiting and controlling values for Jic and KJc at 70°F and 550°F for each individual pump at San Onofre are given in Tables 4-6 and 4-7 for Unit 2 and 4-13 and 4-14 for Unit 3. Appendix D 14

ANALVIS OF THERMAL AGING OF CAST STAINLESS STEEL

)>

"O Table 4-1 REACTOR COOLANT PUMP CASINGS "O REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS rt)

s a.

......

x 0 MATERIAL HEAT NO. c Mn Si s p Cr. N1 Mo N Cb

            ** PLANT I.D. SOUTHERN CAL. UNIT 2

• RCP PUMP CASING 701-N-0557 CASING WELD 6074 0.06 1.29 0.55 0.014 0.032 20.92 9.84 2.52 0.04 0.05 HUB/DIFFUSER 00766-1 0.04 0.85 1.26 0.006 0.034 19.07 9.61 2.21 0.04 0.00 CASE SCROLL 01003-1 0.05 0.74 1.23 0.008 0.033 18.72 . 9.42 2.12 0.04 0.00 CASING WELD 5733 0.03 1.33 0.41 0.014 0.017 19.33 10.60 2.90 0.04 0.05 CASING WELD 6546 . 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.04 CASING WELD 74726 0.03 1.13 0.44 0.006 0.016 20.18 10.89 2.27 0.04 0.02 CASING WELD 7553A 0.05 1.24 0.42 0.016 0.012 18.61 10.52 2.74 0.04 0.04 CASING WELD 63683 0.03 1.03 0.54 0.020 0.005 18.95 10.76 2.34 0.04 0.02 CASING WELD 7242 0.01 0.91 0.52 0.018 0.008 19.38 10.06 2.58 0.04 0.03 CASING WELD 64621 0.02 1.07 0.48 0.021 0.006 19.39 11.03 2.40 0.04 0.01 CASING WELD 77144 0.03 1.03 0.52 0.020 0.006 18.76 10.46 2.25 0.04 0.00
• RCP PUMP CASING 701-N-0558 o.oo HUB/DIFFUSER 05202-1 0.07 0.70 1.28 0.004 0.035 19.48 9.78 2.20 0.04 CASE SCROLL 06243-1 0.05 0.52 1.24 0.010 0.034 19.38 9.36 2.17 0.04 0.00 CASING WELD *6546 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.04 CASING WELD 74726 0.03 1.13 0.44 0.006 0.016 20.18 10.89 2.27 0.04 0.02 CASING*WELD 7553A 0.05 1.24 0.42 0.016 0.012 18.91 10.52 2.74 0.04 0.04 CASING WELD 63683 0.03 1.03 0.54 0.020 0.005 18.95 10.76 2.34 0.04 0.02 CASING WELD 64621 0.02 1.07 0.48 0.021 0.006 19.39 11.03 2.40 0.04 0.01 CASING WELD 77144 0.03 1.03 0.52 0.020 0.006 18.76 10.46 2.25 0.04 0.00 CASING WELD 7242 0.01 0.91 0~52 0.018 0.008 19.38 10.06 2.58 0.04 0.03
• RCP PUMP CASING 701-N-0559 HUB/DIFFUSER 06381-1 0.06 0.64 1.30 0.008 0.034 19.03 9.58 2.23 0.04 0.00

     ...... CASE SCROLL 06652-1      0.05 0.69   1.25 0.005 0.035 19.22        9.43 2.16 0.04 0.00 U1

e ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL

x:o Table 4-1 (Continued) REACTOR COOLANT PUMP CASINGS

"'C "'C fl)

s

a. REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS

-'*

x 0 MATERIAL HEAT NO. c Mn Si s p Cr Ni Mo N Cb CASING WELD 6546 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.04 CASING WELD 74726 0.03 1.13 0.44 0.006 0.016 20.18 10.89 2.27 0.04 0.02 CASING WELD 7533A 0.05 1. 24 0.42 0.016 0.012 18.91 10.52 2.74 0.04 0.04 CASING WELD 63683 0.03 1.03 0.54 0.020 0.005 18.95 10.76 2.34 0.04 0.02 CASING WELD 64621 0.02 1.07 o*.48 0.021 0.006 19.39 . 11.03 2.40 0.04 0.01 CASING WELD 77144 0.03 1.03 0.52 0.020 0.006 18.76 .10.46 2.25 0.04 0.00

                                                   '

• RCP PUMP CASING 701-N-0560 CASE SCROLL 07082-1 0.07 0.60 1.31 0.008 0.037 19.63 . 9.67 2.27 0.04 0.00 HUB/DIFFUSER 06068-1 0.08 1.03 1.38 0.006 0.036 19.39 10.82 2.22 0.04 0.00 CASING WELD 60525 0.04 1.15 0.70 0.020 0.. 006 20.05 11.53 2.29 0.04 0.01 CASING WELD 6546 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.04 CASING WELD 74726 0.03 1.13 0.44 0.006 0.016 20.18 10.89 2.27 0.04 0.02 CASING WELD 7553A 0.05 1.24 0.42 0.016 0.012 18.91 10.52 2.74 0.04 0.04 CASING WELD 63683 0.03 1.03 0.54 0.020 0.005 18.95 10.76 2.34 0.04 0.02 CASING WELD 64621 0.02 1.07 0.48 0.021 0.006 19.39 11.03 2.40 0.04 0.01 CASING WELD 77144 0.03 1.03 0.52 0.020 0.006 18.76 10.46 2.25 0.04 0.00

Table 4-2 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

;r.:.                                                REACTOR COOLANT PUMP CASINGS "O

"O fl)

I REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES a.

......
)(

MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED . AGED c OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW . FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                               @ 70F   @ 70F @' 550F
           ** PLANT I.D. SOUTHERN CAL. UNIT 2

• RCP PUMP CASING 701-N-0557 CASING WELD A 351 CF8M 6074 0.0 o.o 0.0 o.o 0 78600 58000 HUB/DIFFU~ER A 351 CF8M 00766-1 42.3 85.1 49.0 72.0 64 76611 56017 CASE SCROLL A 351 CF8M 01003-1 1

39.3 81.8 51.0 69.0 61 11624 51030 CASING WELD A 351 CF8M 5733 o.o 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 6546 o.o 0.0 0.0 o.o 0 78600 58000 CASING WELD A 351 CF8M 74726 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 7553A o.o . 0.0 0.0 o.o 0 78600 58000 CASING WELD A 351 CF8M 63683 0.0 o.o o.o o.o 0 78600 58000 CASING WELD A 351 CF8M 7242 o.o 0.0 o.o 0.0 0 78600 58000 CASING WELD A 351 CF8M 64621 0.0 o.o 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 77144 o.o o.o 0.0 0.0 0 78600 58000

• RCP PUMP CASING 701-N-0558 76374 55780 HUB/DIFFUSER A 351 CF8M 05202-1 43.2 83.9 56.0 74.0 64 CASE SCROLL A 351 CF8M 06243-1 44.9 89.6 60.0 73.0 67 82232 61638 CASING WELD A 351 CF8M 6546 o.o o.o 0.0 o.o 0 78600 58000 CASING WELD A 351 E316-16 74726 o.o o.o o.o 0.0 0 78600 58000 CASING WELD A 5.4 E316-16 7553A o.o 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 63683 0.0 o.o 0.0 0.0 0 78600 . 58000 CASING WELD A 351 CF8M 64621 0.0 o.o 0.0 o.o 0 78600 58000 CASING WELD A 351 CF8M 77144 o.o 0.0 o.o 0.0 0 78600 58000 CASING WELD A 351 CF8M 724.2 0.0 0.0 0.0 o.o 0 78900 58000
• RCP PUMP CASING 701-N-0559 HUB/DIFFUSER A 351 CF8M 06381-1 42.8 83.4 59.0 74.0 63 75661 55067

   .......                                                                                                     58788
   ....... CASE SCROLL A 351     CF8M      06652-1         43.8     87.1     55.0     73.0       65   79382 CAS. WELD A 351       CF8M      6546             0.0      o.o      o.o      o.o         0  78600   58000 CAS    WELD A 351

• CF8M 74726 o.o 90 0.0 o.o 0 78600 580-

)>

"'O

-

Table*4-2 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS e "'O 11>

s

.....

0.. REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES

)(

c MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                               @ 70F  @ 70F @ 550F CASING WELD  A 351     CF8M      7533A          o.o        o.o      0.0      o.o        0  78600  58000 CASING WELD  A 351     CF8M      63683          0.0        0.0      0.0      0.0        0  78600  58000 CASING WELD  A 351     CF8M      64621          o.o        o.o      0.0      0.0        0  78600  58000 CASING WELD  A 351     CF8M      77144          o.o        0.0      o.o      0.0        0  76600  58000

• RCP PUMP CASING 701-N-0560 CASE SCROLL A 351 CF8M .07082*1 42.5 85.9 51.0 65.0 64 77403 56809 HUB/DIFFUSER A 351 .CFSM 06068-1 36.4 76.3 68.0 71.0 56 64974 44380 CASING WELD A 351 CF8M 60525 o.o 0.0 o.o 0.0 0 78600 58000 CASING WELD A 351 CF8M 6546 o.o . 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 74726 0.0 o.o 0.0 0.0 0 78600 58000 CASING WELD A 351 CFSM 7553A 0.0 0.0 0.0 o.o 0 78600 58000 CASING WELD A 351 CFSM 63683 o.o o.o 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 64621 o.o 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CFBM 77144 o.o 0.0 0.0 0.0 0 78600 58000

  ......

00

Table 4-3 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

)>                                                          REACTOR COOLANT PUMP CASINGS

-0 -0 f1)

I REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR 0.

-'*
><         HEAT     MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM              NICKEL CALCULATED c         NO.        FERRITE EQUIV. EQUIV.          RATIO FERRRITE       FERRITE   EQUIV.       EQUIV.      FERRITE CONTENT       FOR      FOR       FOR CONTENT GONTENT              FOR           FOR WELD METAL

(%) CASTINGS CASTINGS CASTINGS 11 (%) 12 (%) WELDS WELDS (%)

          ** PLANT I. D. SOUTHERN CAL. UNIT 2

• RCP PUMP CASING 701-N-0557 15.0 6074 17.0 19.243 14.944 1.29 15.7 20.7 24.290 13.485 00766-1 15.0 17 .359 14.183 1.22 15.7 15.5 23 .170 12.435 lit. 0 01003-1 14.0 16.886 14.228 1.19 13.7 12.9 22.685 12.490 11' .o 5733 14.0 18.046 15.021 1.20 11.8 13.9 22.870 1*3. 425 15.0 6546 15.0 18.046 14.721 1.23 12.3 15.7 22.950 13.140 11.0 74726 13.4 18.148 15.195 1.19 11.4 13.4 23.120 13.495 11.0 7553A 10.0 17.137 15.374 1.11 .. 8.1 8.5 22.000 13.840 13.0 63683 10.5 17.051 15.154 1.13 8.9 9.1 22 .110 13.435 14.2 7242 18.0 17. 761 14.026 1.27 16.2 18.9 22.755 12.165 12.0 64621 12.0 17. 534 15.207 1.15 10.1 10.7 22.515 13.455 13.8 77144 10.5 16.742 14.903 1.12 8.9 9.0 21.790 13.195 12.5
• RCP PUMP CASING 701-N-0558 0.0 05202-1 12.0 17.766 15.074 1.18 13.2 12.3 23.600 13.430 06243-1 14.0 17 .611 14.146 1.24 17.2 17.1 23.410 12.320 14.0 6546 15.0 18.046 14. 770 1.22 12.3 .. 15.4 22.950 13.200 11.0 74726 13.4 18.148 15.244 1.19 11.4 13.1 23.120 13.555 11.0 7553A 10.0 17. 431 15.374 1.13 8.9 9.6 22.300 13.840 13.0 63683 10.5 17. 051 15.105 1.13 9.3 9.3 22 .110 13.375 14.2 64621* 12.0 17.534 15.134 1.16 10.5 11. l 22.515 13.365 13.8 77144 10.5 16.742 14.805 1.13 9.3 9.4 21.790 13.075 12.5 7242 18.0 17. 761 13.904 1.28 16.7 19.8 22.755 12.015 12!0
• RCP PUMP CASING 701-N-0559

  ....... 06381-1       13.0 17 .362    14.623       1.19        14.2        12.9  23.210       12.900         10.0 l.O      06652-1       14.0 17 .444    14.233       1.23        15. 7.      15.6  23.255       12.475         13.0
                                                                         -

6546 15.0 18.046 14. 721 1.23 12.3 15.7 22.950 13 .140 11. 0 74726 e 13.4 18.148 15.195 1.19 11.4 13.4 23.120 13.495 11. 0

)>
               -

Table 4-3 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS "'t:J "CJ CD

3

a. REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR

 ......

x 0 HEAT MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTE~~ FOR . FOR WELD METAL {%} CASTINGS CASTINGS CASTINGS #1 (%) #2 (%) WELDS WELDS (%) 7533A 10.0 17 .437 15.374 1.13 8.9 9.6 22.300 13.840 13.0 63683 10.5 17.051 15.154 1.13 8.9 9.1 22 .110 13.435 14.2 64621 12.0 17.534 15.207 1.15 10. l 10.7 22.515 13.455 13.8 77144 10.5 16.742 14.903 1.12 8.9 9.0 21. 790 13.195 12.5

• RCP PUMP CASING 701-N-0560 07082-1 14.0 18.015 14.954 1.20 14.7 14.1 23.865 13. 270 11.0 06068-1 10.0 17. 749 16.390 1.08 7.3 7.0 23.680 14.935 13.8 60525 9.0 18.167 16 .180 1.12 8.9 9.0 23.395 14.565 13.6 6546 15.0 18.046 14.721 1.23 12.3 15.7 22.950 13.140 11.0 74726 13.4 18.148 15.195 1.19 11.4 13.4 23.120 13.495 11.0 7553A 10.0 17.437 15.374 1.13 8.9 9.6 22.300 13.840 13.0 63683 10.5 17 .051 15 .154 1.13 8.9 9.1 22 .110 13.435 14.2 64621 12.0 17. 534 15.207 1.15 10.l 10.7 22.515 13.455 13.8 77144 10.5 16.742 14.903 1.12 8.9 9.0 21. 790 13.195 12.5 N

C>

l Table 4-4 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

):a COOLANT PUMP CASINGS AT 70F

"'C "'C Ill

s REPORT #4 - PREDiCTED TOUGHNESS PROPERTIES OF AGED MATERIAL a..

......

x HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM c NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @70F @ 70F @ 70F (%) @ 70F @ 70F (ft-lbs) (lb/1n) (ksiVin)

        ** PLANT I.D. SOUTHERN CAL. UNIT 2

• RCP PUMP CASING 701-N-0557 6074 17 .0 35.53 2823.0 0.37 16.25 506.0 146.2 42.6 00766-1 15.0 19.58 4872.0 0.39 34.14 830.2 . 187 .3 73.6 01003-1 14.0 18.05 5272.0 0.39 37.91 914.5 i 196.5 90.0 6546 15.0 16.00 5924.0 0.40 44.17 978.9 203.3 83.3 5733 14.0 15.62 6063.0 0.40 45.52 1005.0 206.1 85.1 74726 13.4 15.27 6196.0 0.40 46.82 1029.0. 208.5 86.8 7242 18.0 12.69 7362.0 0.40 58.40 1255.0 230.2 101. 2 7553A 10.0 12.56 7429.0 0.40 59.09 1266.0 231.3 102.1 64621 12.0 10.74 8534.0 0.41 70.33 1415.0 244.5 114.8 63683 10.5 10.58 8644.0 0.41 71.46 1436.0 246.3 116.0 77144 10.5 10.19 8923.0 0.41 74.32 1492.0 251.0 119.3
• RCP PUMP CASING 701-N-0558 05202-1 12.0 25.28 3810.0 0.38 24.53 665.9 167.7 59.3 06243-1 14.0 24.63 3905.0 0.38 25.37 679.9 169.5 52.6 6546 15.0 16.40 5784.0 0.39 42.81 1003.0 205.9 82.0 74726 13.4 15.65 6052.0 0.40 45.41 1004.0 . 205. 9 85.0 7553A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 7242 18.0 . .11.11 8247.0 0.41 67.38 1359.0 239.6 111. 4 63683 10.5 10.36 8798.0 0.41 73.03 1466.0 248.9 117. 9 77144 10.5 9.81 9217 .o 0.42 77 .35 1471.0 249.2 122.2 64621 12.0 l~.2' 8895.0 0.41 74.04 1485.0 250.4 119.0
• RCP PUMP CASING 701~N-0559 06652-1 14.0 . 22.64 4233.0 0.38 28.29 745.3 177.4 60.8 N

 ...... 06381-1             13.0          22.12       4329.0          0.38        29.16       766.4         179.9     68.0 978.9       . 203.3     83.3
                                                                     -

6546 15.0 16.00 5924.0 0.40 44.17 9* e

e Table 4-4 (Continued) e e ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

)::o                                                            COOLANT PUMP CASINGS AT 70F

-0 -0 11)

s REPORT #4 - PREDlCTED TOUGHNESS PROPERTIES OF AGED MATERIAL

.....

0. I x HEAT . MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM c NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jic KJc T MOO. CONTENT PARAMETER c N ENERGY @ 70F @70F. @ 70F {%) @ 70F @70F {ft-lbs) (1 b/1 n) (ks1V1n) 74726 13.4 15.27 6196.0 0.40 46.82 1029.0 208.5 86.8 7533A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 64621 12.0 10.74 8534.0 0.41 70.33 1415.0 244.5 114.8 63683 10.5 10.58 8644.0 0.41 71.46 1436.0 246.3 116 .0 77144 10.5 10.19 8923.0 0.41 74.32 1492.0 ' 251.0 119.3

                                                                                                          '

I

• RCP PUMP CASING 701-N-0560 07082-1 14.0 28.74 3385.0 0.38 20.88 586.0 157.3 51.8 06068-1 10.0 19.30 4941.0 0.39 34.79 860.0 . 190.6 .102.2 6546 15.0 16.00 5924.0 0.40 44.17 978.9 203.3 83.3 60525 9.0 15.62 6061.0 0.40 45.50 1005.0 206.0 85.1 74726 13.4 15.27 6196.0 0.40 46.82 1029.0 208.5 86.8 7553A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91. 4 64621 12.0 10.74 8534.0 0.41 70.33 1415.0 244.5 114.8 63683 10.5 10.58 8644.0 0.41 71.46 1436.0 246.3 116.0 77144 10.5 10.19 8923.0 0.41 74.32 1492.0  : 251.0 119.3 N

N

                                                        -. ; -.

Tal;>le 4-5 ANALYSIS OF THERMAL AGING OF STAINLESS STEEL

)::>

-c REACTOR ~OOLANT PUMP CASINGS AT 550F -c ct>

s 0.' REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL

......

x 0 HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @550F @ 550F (ft-lbs) (lb/in) (ksi/in)

       ** PLANT I.D. SOUTHERN CAL. UNIT 2

• RCP PUMP CASING 701-N-0557 6074 17.0 35.53 2452.0 0.27 16.25 727 .4 142.8 61. 2 00766-1 15.0 19.58 3~92.0 0.29 34.14 1038.0 i 170.6 94.8 01003-1 14.0 18.05 3917.0 0.29 37.91 1124 .0 177. 5 118.5 6546 15.0 16.00 4273.0 0.29 44.17 1219.0 184.9 100.7 5733 14.0 15.62 4348.0 0.29 45.52 1244.0 186.7 102.2 74726 13.4 15.27 4419.0 0.29 46.82 1267.0 188.4 103.6 7242 18.0 12.69 5023.0 0.30 58.40 1396.0 197.8 116.0 7553A 10.0 12.56 5058.0 0.30 59.09 1407.0 198.6 116. 7 64621 12.0 10.74 5605.0 0.31 70.33 1509.0 205.6 127.6 63683 10.5 10.58 5658.0 0.31 71.46 1525.0 206.7 128.7 77144 10.5 10.19 5792.0 0.31 74.32 1567.0 209.6 ] 31.1
• RCP PUMP CASING 701-N-0558 05202-1 12.0 25.28 3071.0 0.28 .24. 53 889.l 157.8 81.0 06243-1 14.0 24.63 3128.0 0.28 25.37 899.2 158.7 68.2 6546 15.0 16.40 4198.0 0.29 42.81 1195. 0 183.0 99.2 74726 13.4 15.65 4342.0 0.29 45.41 1242.0 186.5 102.0 7553A 10.0 14.36 4613.0 0.30 50.44 1266.0 188.4 108.0 7242 18.0 11.17 5464.0 0.30 67.38 1538.0 207.6 124.3 63683 10.5 10.36 5732.0 0.31 73.03 1548.0 208.3 130.0 77144 10.5 9.81 5932.0 0.31 77.35 1610.0 212.4 133.7 64621 12.0 10.2, 5779.0 0.31 74.04 1562.0 ,-209.2 130.9
• RCP PUMP CASING 701-N-0559 N 06652-1 14.0 . 22. 64 3323.0 0.28 28.29 965.4 164.5 78.6 w

06381-1 13.0 22.12 3380.0 0.28 29.16 989.8 *166.5 90.l 15.0 16.00 4273.0 ,,0.29. 44.17 1219.0 184.9 100.7 6546-. e

e Table 4-5 (Continued) . )> ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "O REACTOR COOLANT PUMP CASINGS AT 550F "O CD

i 0..

 ......

REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x c HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 550F @550F @ 550F (%) @ 550F @ 550F (ft-lbs) (lb/in) (ks1/1n) 74726 13.4 15.27 4419.0 0.29 46.82 1267.0 188.4 103.6 7533A 10.0 14.36 4613.0 0.30 50.44 1266.0 188.4 108.0 64621 12.0 10.74 5605.0 0.31 70.33 1509.0 205.6 127.6 63683 10.5 10.58 5658.0 0.31 71.46 1525.0 206.7 128.7 77144 10.5 10.19 5792.0 0.31 74.32 1567.0 209.6 131. l

• RCP PUMP CASING 701-N-0560 07082-1 14.0 28.74 2810.0 0.28 20.88 804.8 150.2 72.3 06068-1 10.0 19.30 3731.0 0.29 34.79 1083.0 174.2 147.0 6546 15.0 16.00 4273.0 0.29 44.17 1219.0 184.9 100.7 60525 9.0 15.62 4347.0 0.29 . 45. 50 1243.0 186.7 102.1 74726 13.4 15.27 4419.0 0.29 46.82 1267.0 188.4 103.6 7553A 10.0 14~36 4613.0 0.30 50.44 1266.0 188.4 108.0 64621 12.0 10.74 5605.0 0.31 70.33 1509.0 205.6 127.6 63683 10.5 10.58 5658.0 0.31 71 . 46 1525.0 206.7 128.7 77144 10.5 10.19 5792.0 0.31 74.32 1567.0 209.6 131.1

Table 4.~ - San Onofre 2 Limiting and Controlling Values of Jic and KJc at 70"F RCP PUMP CASING Heat # Jic KJc (lb/in) {ksi/ in) 701-N-0557 6074 506.0 146.2 701-N-0558 05202-1 665.9 167.7 701-N-0559 06652-1 745.3 177.4 701-N-0560 07082-1 586.0 157.3 l\) Ul

Table 4.7 - San Onofre 2 Limiting and Controlling Values of Jic and KJc at 550°F RCP PUMP CASING Heat # Jic KJc (lb/in) (ksi/in) 701-N-0557 6074 727.4 142.8 701-N-0558 05202-1 889.l 157.8 701-N-0559 06652-1 965.4 164.5 701-N-0560 07082-1 804.8 150.2

Table 4-8 ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL

x:o "C

REACTOR COOLANT PUMP CASINGS "C 11>

3 0..

REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS

  -'*

x I CJ MATERIAL I HEAT NO. c Mn S1 s p Cr N1 Mo N Cb

          ** PLANT l.D. SOUTHERN CAL. UNIT 3

• RCP PUMP CASING 701-N-0561 CASE SCROLL 08426-1 0.04 0.64 1.17 0.011 0.035 19.01 9.31 2.22 0.04 0.00 HU~/DIFFUSER 07903-1 0.05 0.92 1.16 0.008 0.036 18.91 9.38 2.27 0.04 0.00 CASING WELD. 60525 0.04 1.15 0.70 0.020 0.006 20.05 11.53 2.29 0.04 0.01 CASING WELD 6546 0.03 1.42 0.38 0.01'7 0.016 20.01 10.39 2.35 0.04 0.04 CASING WELD 74726 0.03 1.13 0.44 0.006 0.016 20. 18 10.89 2.27 0.04 0*.02 CASING WELD 7553A a.as 1.24 0.42 0.016 0.012 18.91 10.52 2.74 0.04 0.04 0.02 CASING WELD 63683 0.03 1.03 0.54 0.020 0.005 18.95 10.76 2.40 0.04 CASING WELD 64621 0.02 1.01 0.48 0.021 0.006 19.39 11.03 2.40 0.04 0.01 CASING WELD 77144 0.03 1.03 0.52 0.020 0.006 18.76 10.46 2.25 0.04 o.oo CASING WELD 8609 0.02 0.91 0.54 0.023 0.005 19.33 10.97 2.22 0.04 o. 11 CASING WELD 8610 0.02 0.89 0.48 0.021 0.005 19.44 11.13 2.26 0.04 0.07 CASING WELD 776060 0.02 I. 78 0.36 0.012 0.009 19.12 12.87 2.39 0.04 0.22
• RCP PUMP CASING 701-N-0562 HUB/DIFFUSER 08783-1 0.06 0.73 1.18 0.007 0.038 19.29 9.34 2.25 0.04 0.00 CASE SCROLL 11994-1 0.03 0.59 1.28 0.010 0.035 18.92 9.42 2.17 0.04 0.00 CASING WELD 60525 0.04 1.15 0.70 0.020 0.006 20. 05 -11. 53 2.29 0.04 0.01 CASING WELD 6546 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.04 CASING*WELD 74726 0.03 1.13 0.44 0.006 0.016 20.18 10.89 2.27 0.04 0.02 CASING WELD 7553A 0.05 1.24' 0.42 0.016 0.012 18.91 10.52 2.74 0.04 0.04 CASING WELD 8609 0.02 0.91 0.54 0.023 0.005 19.33 10.97 2.22 0.04 0.11 CASING WELD 8610 0.02 0.89 0~48 0.021 0.005 19.44 11.13 2.26 0.04 0.07 CASING WELD 776060 0.02 I. 78 . 0"'36. 0.012 0.009 19.12 12.87 2.39 0.04 0.22
• RCP PUMP CASING 701-N-0563 N HUB/DIFFUSER 10202-1 0.06 0. 72 1.35 0.007 0.036 19.47 9.79 2.22 0.04 0.00

    ""-.!

e e

e Table 4-8 (Continued) ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL

)>

"O REACTOR COOLANT PUMP CASINGS "O Cl>

I

a. REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS

.....
><

0 MATERIAL HEAT NO. c Mn Si s p Cr. . Ni Mo N Cb CASING WELD 7971C 0.05 1.41 0.* 56 0.010 0.012 19.97 '10.44 2.43 0.04 0.05 CASE SCROLL 11294-1 0.04 0.63 1.23 0.010 0.031 19.15 9.40 2.06 0.04 0.00 CASING WELD 8346 0.04 1.29 0.32 0.017 0.010 19.60 10.30 2.56 0.04 0.08 CASING WELD 7970 0.05 1.36 0.43 0.017 0.005 19.56 11.17 2.52 0.04 0.05 CASING WELD 6546 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.04 CASING WELD 74726 0.03 1.13 0.44 0.006 0.016 20.18 10,89 2.27 0.04 0.02 CASING WELD 7553A 0.05 1.24 0.42 0.01'6 0.012 18.91 10.52 2.74 0.04 0.04 CASING WELD 8346A 0.04 1.27 0.29 0.016 0.005 19.16 10.31 2.64 0.04 (LOB CASING WELD 8610 0.02 0.89 0.48 0.021 0.005 19.44 11.13 2.26 0.04 0.07 CASING WELD 8609 0.02 0.91 0.54 0.023 0.005 19.33 10.97 2.22 0.04 0.11 CASING WELD 776060 0.02 1. 78 0.36 0.012 0 ..009 19.12 12.87 2.39 0.04 0.22

• RCP PUMP CASING 701-N-0564 CASE SCROLL 07342-1 0.06 0.91 1.22 0.007 0.039 19. 77 9.70 2.15 0.04 0.00 HUB/DIFFUSER 96755-1 0.06 0.80 1.32 0.006 0.037 19 .15 9.87 2.26 0.04 0.00 CASING WELD 74819 0.06 1.86 0.34 0.002 0.017 21. 78 9. 71 0.11 0.04 0.03 CASING WELD 7971 0.04 1.46 0.66 0.018 0.012 19.59 10.60 2.47 0.04 0.04 CASING WELD 60525 0.04 1.15 0.70 0.020 0.006 20.05 11.53 2.29 0.04 0.01 CASING WELD 6546 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.04 CASING WELD 74726 0.03 1.13 0.44 0.006 0.016 20.18 10.89 2.27 0.04 0.02 CASING WELD 6B5C-16 0.04 1.30 0.28 0.007 0.025 19.32 11.53 2.18 0.04 0.04 CASING WELD 63683 0.03 1.03 0.54 0.020 0.005 18.95 10.76 2.34 0.04 0.02 CASING WELD 64621 0.02 1.07 0.48 0.021 0.006 19.39 11.03 2.40 0.04 0.01 CASING WELD 77144 0.03 1.03 0.52 0.020 0.006 18.76 10.46 2.25 0.04 0.00 CASING WELD 7956 0.03 1.25 0~25 0.014 0.005 19.43 10.15 2.04 0.04 0.06 CASING WELD 8609 0.02 0.91 ' 0~54 0.023 0.005 19.33 10.97 2.22 0.04 0.11 N

co

Table 4-9 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

)>                                                REACTOR COOLANT PUMP CASINGS "C

"C rt>

s

• REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES a.

......

x MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED c OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                            @ 70F   @ 70F   @ SSOF
       ** PLANT I.D. SOUTHERN CAL. UNIT 3

• RCP PUMP CASING 701-N-0561 64 77245 56651 CASE SCROLL A 351 CF8M 08426-1 42.5 85.7 45.0 63.0 HUB/DIFFUSER A 351 CF8M 07903-1 ,39.9 81.9 60.0 75.0 61 72il 78 51584 CASING WELD A 351 CF8M 60525 0.0 . 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 6546 0.0 o.o 0.0 0.0 0 78600 58000 CASING WELD A 351 CFBM 74726 o.o o.o o.o 0.0 0 78600 58000 58000 CASING WELD A 351 CF8M 7553A 0.0 0.0 0.0 0.0 0 78600 CASING WELD A 351 CFBM 63683 o.o o.o o.o o.o 0 78600 58000 CASING WELD A 351 CFBM' 64621 0.0 0.0 o.o 0.0 0 78600 58000

       . CASI NG WELD A 351    CF8M      77144            0.0      o.o       0.0     o.o         0  78600     58000 CASING WELD  A 351    CF8M      8609             0.0      o~o       0.0     o.o         0  78600     58000 CASING WELD  A 351    CF8M      8610             0.0      0.0       0.0     0.0         0  78600     58000 CASING WELD  A 5.9    ER-316    776060           0.0"     o.o       0.0     o.o         0  78600     58000

• RCP PUMP CASING 701-N-0562 62 73286 52692 HUB/DIFFUSER A 351 CF8M 08783-1 40.3 82.9 58.0 75.0 CASE SCROLL A 351 CF8M 11994-1 45.7 89 .3- 60.0 74.0 68 82628 62034 CASING WELD A 351 CF8M 60525 0.0 o.o 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 6546 o.o o.o 0.0 0.0 0 78600 58000

                                                                                                            . 58000 CASING WELD  A 351    CF8M      74726            0.0      0.0       0.0     0.0         0  78600 CASING WELD  A 351    CF8M      7553A            o.o      o.o       o.o     0.0         0  78600     58000 CASING WELD  A 351    CF8M      8609             0.0      0.0       0.0     0.0         0  78600     58000
        *CASING WELD  A 351    CF8M      8610             o.o      0.0       0.0     0.0         0  78600     58000 CASING WELD  A 5.9    ER-316    776060           o.o      o.o       o.o     o.o         0  78600     58000

• RCP PUMP CASING 701-N-0563 N

HUB/DIFFUSER A 351 CF8M 10202-1 0.0 0.0 0.0 o.o 0 78600 58000

    '°   CASI.WELD 1T3454 CASE .. ROLL A 351 S782516 7971C CF8M      11294-1 o.o o.o      w         0.0 0.0 0.0 o.o 0

0 78600 78600 58000 58009

e Table 4-9 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

)>

"C REACTOR COOLANT PUMP CASINGS "C 11)

s 0..

REPORT- #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES

-'*

x c MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                          @ 70F   @ 70F @ 550F CASING WELD  A 351    CF8M      8346             0.0       0.0      0.0    .0.0          0  78600   58000 CASING WELD  1T3454   S782516   7970             o.o       0.0      0.0     0.0          0  78600   58000 CASING WELD  A 351    CF8M      6546             o.o       0.0      0.0     0.0          0  78600   58000 CASING WELD  A 351    CF8M      74726            o.o       0.0      o.o     o.o          0  78600   58000 CASING WELD. A 351    CF8M      7553A.           o.o       0.0      0.0     0.0          0  786PO   58000 CASING WELD  A 351    CF8M      8346A            '0.0      0.0      0.0     0.0          0  78600   58000 CASING WELD  A 351    CF8M      8610             o.o       0.0      0.0     0.0          0  78600   58000 CASING WELD  A 351    CF8M      8609             o.o       o.o      0.0     o.o          0  78600   58000 CASING WELD  A 5.9    ER-316    776060           0.0       0.0      0.0     o.o          0  78600   58000

• RCP PUMP CASING 701-N-0564 CASE SCROLL A 351 CF8M 07342-1 43.8 86.6 55.0 74.0 65 78986 58392 HUB/DIFFUSER A 351 CF8M 96755-1 40.1 85.0 60.0 77.0 63 74790 54196 '

                                                                                                                  '

I CASING WELD A5.9-69 ER-308 74819 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD 113454 S782516 7971 0.0 0.0 o.o o.o 0 78600 58000 CASING WELD A 351 CF8M 60525 o.o* o.o 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 6546 0.0 0.0 o.o o.o 0 78600 58000 CASING WELD A 351 CF8M 74726 o.o 0.0 o.o o.o 0 78600 58000 CASING WELD A 5.4 E316-16 6B5C-16 0.0. o.o. o.o 0.0 . 0 78600 58000 CASING WELD A 351 CF8M 63683 o.o 0.0 0.0 o.o 0 78600 58000 CASING WELD A 351 CF8M 64621 o.o* 0.0 0.0 o.o 0 78600 58000 CASING WELD A 351 CF8M 77144 o.o 0.0 o.o* o.o 0 78600 *58000 CASING WELD 1T3454 S-7825 7956 0.0 . o.o o.o 0.0 0 78600 58000 CASING WELD A 351 CF8M 8609 0.0 0.0 o.o 0.0 0 78600 58000 w 0

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

):>

"Cl Table 4-10 REACTOR COOLANT PUMP CASINGS "Cl l'D

l REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR

.....
~

x c HEAT MEASURED CHROMIUM NICKEL ere/Nie CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO

• FERRRITE FERRITE *EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL

                       . (%) CASTINGS CASTINGS CASTINGS        11 (%)     #2 (%)     WELDS        WELDS          (%)
         ** PLANT I.D. SOUTHERN CAL. UNIT 3

• RCP PUMP CASING 701-N-0561 08426-1 16.0 17. 268 13.863 1.25 17 .2 17 .2 22 *. 985 12.030 15.0 07903-1 14.0 17.224 14.205 1.21 14.7 14.7 22.920 12.540 13.0 60525 9.0 18.167 16.180 1.12 , 8.9 9.0 23.395 14.565 13.6 6546 15.-0 18.046 14.721 1.23 12.3 15.7 22.950 '13 .140 11. 0 74726 13.4 18.148 15.195 1.19 11.4 13.4 23.120 13.495 11.0 7553A 10.0 17.437 15.374 1.13 8.9 9.6 22.300 13.840 13.0 63683 10.5 17.123 15.154 1.13 9.3 9.4 22 .170 13 .435 . 14.2 64621 12.0 17.534 15.207 1.15 *10.1 10.7 22.515 13.455 13.8 77144 10.5 16.742 14.903 1.12 8.9 9.0 21. 790 13.195 12.5 8609 12.0 17 .285 15.059 1.15 10.5 10.4 22.415 13.225 14.2 8610 0.0 17.415 15. 217 1.14 10.1 10.2 22.455 13.375 14.0 776060 0.0 17 .195 16.986 1.01 4.5 4.2 22.160. 15.500 7.4
• RCP PUMP CASING 701-N-0562 08783-1 14.0 17.589 14.392 1.22 15.2 15.4 23.310 12.705 13.0 11994-1 17 .o 17 .170 13.723 1.25 17 .2 17.6 23.010 11.815 15.0 60525. 9.0 18.167 . 16.180 1.12 8.9 9.0 23.395 14.565 13.6 6546 15.0 18.046 14.721 1.23 12.3 15.7 22.950 13.140 11. 0 74726 13.4 18.148 15.195 1.19 11.4* 13.4 23 .120 13.495 11. 0 7553A 10.0 17.437 15.374 1.13 8.9 9.6 22.300 13.840 13.0 8609 12.0 17. 285 15.059 1.15 10.5 10.4 22.415 13.225 14.2 8610 0.0 17.415 15.217 1.14 10.1 10.2 22.455 13.375 14... o 776060 o.o 17.195 16.986 1.01 4.5 4.2 22.160 15.500 i.4 w
• RCP PUMP CASING 701-N-0563

  ...... 10202-1      15.0    17 .814  14.841       1.20         14.7      13.8    23. 715    13.150          12.0 7971C        13.0    18.189   15.309       1.19         11.0      13.0   23.265      13.845          10.0 17.0    17.243   13.952       1.24         16.7      16.4   23.055      12 .115        15.0 11294--

9

e e ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

)>

Table 4-10 (Continued) REACTOR COOLANT PUMP CASINGS "O "O tD

3 REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR a.

-'*

x HEAT MEASURED CHROMIUM NICKEL Cre/N1e CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED c NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS ll (%) 12 (%) WELDS WELDS (%) 8346 13.0 17.861 14.914 1.20 11.4 13.6 22.680 13.345 10.0 7970 9.0 17.826 16.035 l. ll 7.7 8.4 22.750 14.550 12.2 6546 15.0 18.046 14.721 1.23 12.3 15.7 22.950 13.140 11. 0 74726 13.4 18.148 15.195 1.19  : 11.4 13.4 23.120 13.495 11.0 7553A 10.0 17.437 15.374 1.13  : 8.9 9.6 22.300 13.840 13~0 8346A 11.0 17. 504 14.922 1.17 10.1 12.0 22.275 13.345 13 .8 1 8610 0.0 17 .415 15.217 1.14 ., 10 .1 10.2 22.455 13.375 14 .o*

    . 8609           0.0   17 .285  15.059      1.15          10.5     10.4   22.415       13.225       14.2 776060         0.0   17 .195  16.986      1.01        . 4. 5      4.2   22.160       15.500        7.4

• RCP PUMP CASING 701-N-0564 07342-1 14.0 17.967 14.769 1.22 .. 14. 7 15.0 23.750. 13.155 13.0 96755-1 12.0 17.528 14.928 1.17  : 13. 2 12.0 23.390 . 13.270 10.0 74819 11.0 17 .086 14.861 1.15 6.9 10.5 22.415 13.640 12.7 7971 13.0 17.906 15.228 1.18 11.0 12.2 23.070 13.730 14.S 60525 9.0 18.167 16.180 1.12 8.9 9.0 23.395 14.565 13" 6 6546 15.0 18.046 14.721 1.23 12.3 15.7 22.950 13.140 11.0 74726 13.4 18.148 15.195 1.19 11.4 13.4 23.120 13.495 11. 0 6B5C-16 9.0 17.102 16.022 1.07 5.9 6.3 21.940 14.430 10.0 63683 . 10. 5 17.051 15.154 1.13 9.3 9.1. 22 .110 13.435 14.2 64621 12.0 17. 534 15.207 1.15 10.1 10.7 22.515 13.455 13.8 77144 10.5 16.742 14.903 1.12 . 9.3 9.0 21.790 13.195 12.5 7956 12.0 17.028 14.515 1.17 . 10. l 12.0 21.875 12.875 13.8 8609 12.0 17.285 15.059 1.15 10.5 10.4 22.415 13.225 14.2 w

N

                                                                                                                          ---

Table 4-11 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

)>

"'C COOLANT PUMP CASINGS AT 70F "'C ro

I a.. REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL

.....

x 0 HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @70F @ 70F @ 70F (%} @*70F @70F (ft-lbs) (lb/in} (ksiVin)

      ** PLANT I. D. SOUTHERN CAL. UNIT 3

• RCP PUMP CASING 701-N-0561 79.8 07903-1 14.0 20.35 4692.0 0.39 32.48 801.0 183.9 08426-1 16.0 20.10 ~750.0 0.39 33.01 806.4  ; 184.6 70.8 6546 15.0 16.00 5924.0 0.40 44.17 978.9 203.3 83.3 60525 9.0 15.62 6061.0 0.40 45.50 1005.0 206.0 85.1 74726 13.4 15.27 6196.0 0.40 46.82 1029.0 208.5 86.8 7553A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 63683 10.5 10.89 8433.0 0.41 69.29 1396.0 242.9 113 .6 64621 12.0 10.74 8534.0 0.41 70.33 1415.0 244.5 114.8 77144 10.5 10.19 8923.0 0.41 74.32 1492.0 251.0 119.3 8610 0.0 9.63 9356.0 0.42 78.80 1498.0 251.5 123.7 8609 12.0 9.60 9383.0 0.42 79.07 1502.0 251.9 124.0 776060 0.0 4.76 15212.0 0.45 140.00 2326.0 . 313 .4 184.0
• RCP PUMP CASING 701-N-0562 65.0 08783-1 14.0 24.97 3855.0 0.38 24.93 677.5 169.2 11994-1 17.0 17 .57 5412.0 0.39 39.24 925.5 197.7 70.1 6546 15.0 16.00 5924.0 0.40 44.17 978.9 203.3 83.3 60525 9.0 15.62 6061.0 0.40 45.50 1005.0 206.0 85.1 74726 13.4 15.27 6196.0 0.40 46.82 1029.0 208.5 86.8 7553A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 8610 0.0 9.63 9356.0 0.42 78.80 1498.0 251.5 123.7 8609 12.0 9.69 9383.0 0.42 79.07 1502.0  :' 251. 9 124.0 776060 0.0 4.76 15212.0 0.45 140.00 2326.0 . 313.4 184.0 w
• RCP PUMP CASING 701-N-0563 . 167 .3 56.0 w

10202-1 15.0 25.37 3798.0 0.38 24.43 662.5 20.69 4618.0 0.39 31.80 779. 7 181. 5 66.7 79~~. 13.0 e e

e e e Table 4-11 (Continued)

)>

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR "O "O COOLANT PUMP CASINGS AT 70F m

J a.

......                                     REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x

0 HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F {%) @ 70F @70F (ft-lbs) (1b/in) (ks1V1n) 11294-1 17 .0 19.91 4794.0 0.39 33.42 813.9 185.4 69.l 8346 13.0 16.87 5631.0 0.39 41.34 973.3 202.8 80.0 6546 15.0 16.00 5924.0 0.40 44.li7 978.9 203.3 83.3 74726 13.4 15.27 6196.0 0.40 46.82 1029.0 208.5 86.8 7970 9.0 14.57 6475.0 0.40 49.56 1083.0 213.9 90.3 8346A 11.0 14.43 6'533.0 0.40 50.14 1093.0 214.9 91.0 7553A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 8610 0.0 9.63. 9356.0 0.42 78.80 1498.0 251. 5 123.7 8609 0.0 9.60 9383.0 0.42 79.07 1502.0 251. 9 124.0 776060 0.0 4.76 15212.0 0.45 140.00 2326.0 313.4 184.0

• RCP PUMP CASING 701-N-0564 07342-1 14.0 26.83 3605.0 0.38 22.'15 625.3 162.5 52.8 96755-1 12.0 21.95 4360.0 0.38 29.44 774.6 180.9 70.0 74819 11.0 17.59 5405.0 0.39 39.17 929.3 . 198.1 77 .1 7971 13.0 16.90 5619.0 0.39 41.22 971.3 202.5 79.9 6546 15.0 16.00 5924.0 0.40 44.17 978.9 203.3 83.3 60525 9.0 15.62 6061.0 0.40 45.50 1005.0 206.0 85.1 74726 13.4 15.27 6196.0 0.40 46.82 1029.0 208.5 86.8 7956 12.0 11.58 7990.0 0.41 64. 77 1310.0 235.2 108.3 64621 12.0 10.74 8534.0 0.41 70.33 1415.0 244.5 114 .8 63683 10.5 10.58 8644.0 0.41 71.46 1436.0 .246.3 116.0 77144 10.5 10.19 8923.0 0.41 74.32 1492.0 251.0 119.3 8609 12.0 9.6p 9383.0 0.42 79.07 1502.0 251.9 124.0 685C-16 9.0 8.6, 10230.0 0.42 87.86 1667.0 265.3 133.6 w

 ~

Table .4-11 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

)::o "O                                                            COOLANT PUMP CASINGS AT 70F "O

Cl>

I 0.. REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL

-'*
><

a HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (%) @ 70F @ 70F (ft-lbs) (lb/in) (ksfVf n)

        ** PLANT I.D. TEST PROGRAM

• RCP PUMP CASING ANL-HEATS 292 28.0 44.97 2378.0 0.37 12.79 422.1 133.5 36.2 75 27.8 42.57 2467.0 0.37 13.47 438.8 136.1 37.5 286 22.0 40.36 2561.0 0.37 14.19 456.4 138.8 38.9 KRB 34.0 39.79 2588.0 0.37 14.40 461.2 139.6 39.2 65 23.4 33.10 2992.0 0.37 17 .62 538.4 150.8 45.0 64 28.4 32.24 3060.0 0.37 18.17 551.6 152.6 46.0 280 40.0 29.72 3285.0 0.38 20.04 566.5 154.7 48.9 60 21.1 24.44 3934.0 0.38 25.63 688.8 170.6 57.9 74 18.4 23.47 4089.0 0.38 27.00 717 .3 174.1 60.0 68 23.4 22.65 4230.0 0.38 28.26 744.8 177 .4 61. 9 66 19.8 21.29 4491.0 0.39 30.63 756.8 178.8 65.0 70 18.9 22.04 4344.0 0.38 29.29 767.6 180.l 63.4 Pl 24.1 18.04 5277 .0 0.39 37.95 904.1 195.4 75.5 205 15.9 17.90 5315.0 0.39 38.32 912.2 196.3 75.9 61 13.1 17.28 5500.0 0.39 40.08 947.3 200.0 78.3 59 13.5 15.48 6113. 0 0.40 46.00 1014.0 206.9 85.7 278 15.0 16.10 5889.0 0.39 43.83 1023.0 207.9 83.4 758 19.2 15.12 6253.0 0.40 47.38 1041.0 209.7 87.5 P4 10.0 '14.17 6647.0 0.40 51.26 1116.0 217. I 92.4 69 23.6 13.69 6866.0 0.40 53.44 1157.0 221.0 95.2 63 10.4 12.9i 7252.0 0.40 57.31 1232.0 228.1 99.9 56 10.1 12.0 7739.0 0.41 62.22 1262.0 230.9 105.3 I 17 .1 11.92 7789.0 0.41 62.73 1273.0 231.8 105.9 73 7.7 11.58 7992.0 0.41 64.79 1310.0 235.2 108.3 w 53 8.7 11.32 8151.0 0.41 66.41 1341.0 238.0 110.2 U1 P2 15.6 9.39 9549.0 0.42 80.79 1534.0 . 254.5 125.9 51 18.0 9.39 9555.0 0.42 80.86 1535.0 254.6 126.0 e e e

e Table 4-11 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

)>

"O COOLANT PUMP CASINGS AT 70F "O ft)

s

a. REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL

.....

x c HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F *@ 70F @ 70F (%) @ 70F @70F (ft- lbs). (lb/in) (ks1V1n) Cl 2.2 9.12 9777 .0 0.42 83.16 1578.0 258.2 128.5 54 1.8 7.97 10859.0 0.43 94.42 1700.0 268.0 140.0 52 13.5 6.37 12733.0 0.44 114.00 1959.0 287.6 159.4 47 16.3 6.67 12349.0 0.43 110.00 1988.0 289.8 156.1 57 4.0 6.61 12421.0 0.43 110.30 2003.0 290.8 156.9 62 4.5 6.48 12587.0 0.43 112.50 2035.0 293.2 158.6 58 2.9 6.45 12634.0 0.43 113.00 2046.0 294.0 159.1 48 8.7 5.23 14408.0 0.44 131.60 2287.0 310.8 176.7 49 7.2 4.20 16253.0 0.45 150.80 2532.0 327.0 194.2 P3 1.9 4.01 16639.0 0.45 154.80 2610.0 332.0 197.9 50 4.4 3.68 17341.0 0.46 162.10 2610.0 332.0' 203.8 w O'I

Table 4-12 ANALYSIS OF THERMAL AGING OF STAINLESS STEEL

)>

"'C REACTOR COOLANT PUMP CASINGS AT 550F "'C ti>

I a..

REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL

-'*

x MINIMUM MINIMUM MINIMUM HEAT MEASURED. *MATERIAL MATERIAL MATERIAL MINIMUM c FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. NO. @ 550F @ 550F @ 550F CONTENT PARAMETER c N ENERGY (ks1/1n) (%) @ 550F @ 550F (ft-lbs) (lb/1n)

            ** PLANT 1.0. SOUTHERN CAL. UNIT 3

• RCP PUMP CASING 701-N-0561 1068.0 173.0 107.1 07903-1 14.0 20.35 3590.0 0.28 32.48 20.10 3.623.0 0.29 33.01 1015.0 i 168 .. 7 91.3 08426-1 16.0 184.9 100.7 6546 15.0 16.00 4273.0 0.29 44.17 1219.0 9.0 15.62 4347.0 0.29 45.50 1243.0 186.7 102.1 60525 1267.0 188.4 103.6 74726 13.4 15.27 4419.0 0.29 46.82 14.36 4613.0 0.30 50.44 1266.0 188.4 108.0 7553A 10.0 204.5 126.B 63683 10.5 10.89 5556.0 0.31 69.29 1492.0 10.74 5605.0 0.31 70.33 1509.0 205.6 127.6 64621 12.0 209.6 131. l 77144 10.5 10.19 5792.0 0.31 74.32 1567.0 9.63 5998.0 0.31 78.80 1632.0 213.8 134. 9 8610 0.0 214.2 135.2 8609 12.0 9.60 6011.0 0.31 79.07 1637.0 4.76 8551.0 0.33 140.00 2259.0 251.6 180.4 776060 0.0
• RCP PUMP CASING 701-N-0562 24.93 903.1 159.l 90.8 08783-1 14.0 24.97 3098.0 0.28 17 .57 3994.0 0.29 39.24 1121.0 177 .2 83.9 11994-1 17 .0 1219.0 184.9 100.7 6546 15.0 16.00 4273.0 0.29 44.17 15.62 4347.0 0.29 45.50 1243.0 186.7 102.1 60525 9.0 1267.0 188.4 103.6 74726 13.4 "15. 27 4419.0 0.29 46.82 14.36 4613.0 0.30 50.44 1266.0 188.4 108.0 7553A 10.0 213.8 134.9 8610 0.0 9.6, 5998.0 0.31 78.80 1632.0 9.6 6011.0 0.31 79.07 1637.0 ,.214.2 135.2 8609 12.0  : 251.6 180.4 776060 0.0 4.76 8551.0 0.33 140.00 2259.0 w
• RCP PUMP CASING 701-N-0563 883.9 . 157 .4 75.1

   ........

10202-1 15.0 25.37 3064.0 0.28 24.43 13.0 20.69 3547.0 0.28 31.80 1039.0 170.6 85.3 7971 * . e e

Table 4-12 (Continued) ANALYSIS OF THERMAL AGING OF STAINLESS STEEL

  ):o
 "'C                                                REACTOR COOLANT PUMP CASINGS AT 550F
 "'C CD

s I Q.

  ......

REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x CJ HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jic KJc T MOD. CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @ 550F @ 550F. (ft* lbs) (lb/in) (ks1/1n)

         ** PLANT I.D. TEST PROGRAM

• RCP PUMP CASING ANL-HEATS 292 28.0 44.97 2156.0 0.27 12.79 633.3 133.2 54.5 75 27.8 42.57 2216.0 0.27 13.47 652.5 135.2 55.8 286 22.0 40.36 2279.0 0.27 14.19 672. 7 137 .3 57.3 KRB 34.0 39.79 2297.0 0.27 14.40 678.l 137.8 57.7 65 23.4 33.10 2562.0 0.27 17.62 762.8 146.2 63.7 64 28.4 32.24 2605.0 0.27 18.17 777 .4 147.6 64.6 280 40.0 29.72 2748.0 0.28 20.04 784.2 148.2 68.2 60 21. l 24.44 3146.0 0.28 25.63 909.7 159.7 76.8 74 18.4 23.47 3238.0 0.28 27.00 939.6 162.3 78.8 68 23.4 22.65 3321.0 0.28 28.26 965.9 164.5 80.6 66 19.8 21.29 3474.0 0.28 30.63 1015.0 168.7 83.8 70 18.9 22.04 3388.0 0.28 . 29.29 987.7 166.4 82.0 Pl 24.l 18.04 3920.0 0.29 37.95 1106.0 176.l 93.5 205 15.9 17.90 3941.0 0.29 38.32 1114.0 176.6 94.0 61 13.l 17.28 4043.0 0.29 40.08 1145.0 179.1 96 .1 59 13.5 15.48 4374.0 0.29 46.00 1251. 0 187.2 102.7 278 15.0 16.10 4254.0 0.29 43.83 1213.0 184.3 100.3 758 19.2 15.12 4449.0 0.29 47.38 1277.0 189.1 104.1 P4 10.0 14.17 4656.0 0.30 51.26 1280.0 189.4 108.8 69 23.6 13.69 4770.0 0.30 53.44 1315.0 192.0 111.1 63 10.4 12.~0 4968.0 0.30 57.31 1377.0 196.5 114. 9 56 10.1 12.00 5213.0 0.30 62.22 1456.0 202.0 119.6 I 17 .1 11.92 5238.0 0.30 62.73 1465.0 202.6 120.1 73 7.7 11.58 5339.0 0.30 64.79 1498.0 204.9 122.0 w 53 8.7 11.32 5417.0 0.30 66.41 1523.0 206.6 123.4

  \.0 P2                 15.6         9.39       6089.0         0.31        80.79        1661. 0  . 215.8    136.6 51                 18.,0        9.39       6092.0         0.31        80.86        1662.0     215.8    136.6 e                                                 e                                             e

e Table 4-12 (Continued) ANALYSIS OF THERMAL AGING OF STAINLESS STEEL

  )>
 "'C REACTOR COOLANT PUMP CASINGS AT 550F

' "'C tt>

I REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL

  .....

0. x HEAT MEASURED MATERIAL

• MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM c NO. T MOD.

FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc CONTENT PARAMETER c N ENERGY @ 550F @550F @ 550F (%) @ 550F @ 550F (ft-lbs) (lb/in) (ks 1/1 n) Cl 2.2 9.12 6196.0 0.31 83.16 1696.0 218.0 138.5 54 1.8 7.97 6692.0 0.32 94.42 1767. 0 222.5 148.1 52 13.5 6.37 7517. 0. 0.32 114.00 2031.0 238.6 162.4 47 16.3 6.67 7351.0 0.32 110.00 1977.0 235.4 159.6 57 4.0 6.61 7383.0 0.32 110.80 1987.0 236.0 160.l 62 4.5 6.48 7455.0 0.32 112.50 2010.0 237.4 161. 3 58 2.9 6.45 7475.0 0.32 113 .00 2016.0 237.7 161.6 48 8.7 5.23 8222.0 0.33 131.60 2153.0 245.6 175.1 49 7.2 4.20 8969.0 0.34 150.80 2280.0 252.8 188.2 P3 1.9 4.01 9121.0 0.34 154.80 2327.0 255.4 190.6 50 4.4 3.68 9395.0 0.34 162.10 2415.0 260.l 194.9

Table 4-~3 - San Onofre 3 Limiting and Controlling Values of Jic and KJc at 70°F RCP PUMP CASING Heat # JIC KJc (lb/in) (ksi/in) 701-N-0561 07903-1 801.0 183.9 701-N-0562 08783-1 677.5 169.2 701-N-0563 10202-1 662.5 167.3 701-N-0564 07342-1 625.3 162.5

Table 4-~4 - San Onofre 3 Limiting and Controlling Values of Jic and KJc at 550°F RCP PUMP CASING Heat # Jic KJc (lb/in) (ksi/in) 701-N-0561 07903-1 1068.0 173.0 701-N-0562 08783-1 903.l 159.1 701-N-0563 10202-1 883.9 157.4 701-N-0564 07342-1 845.8 154.0 _J

5.0 Crack Growth Analysis In this section, the methodologies discussed in Section 5.1 of the main report are applied to plant-specific conditions at San Onofre 2&3. The growth history of a worst-case hypothetical crack is conservatively developed based on information contained in the vendors stress analysis report. 5.1 scope The analysis which follows pertains to the 36x36x38 DFSS Reactor Coolant Pump casings, Serial Numbers 701-N-0557 to 701-N-0564, inclusive, at the San Onofre Station, Units 2&3. 5.2 Reference Stress Reports The stresses used in the hypothetical crack growth analyses are from the stress reports prepared by the Byron-Jackson Company in July, 1978 (References 3-2). 5.3 Selection of High Stress Locations The methodology described in 5.1.1 of the main report was applied to identify five locations as potentially limiting: (1) suction Nozzle (2) Junction, Volute to Lower Flange (3) Crotch Region (4) Diffuser Vane Number 9 (5) Volute, Vane Number 5 Vicinity All other regions in the stress summary were considered and were found to have lower stresses than the above regions. Appendix D 43

5.4 Stresses and Wall Thicknesses at Limiting Locations -- Membrane and through-wall bending components of the limiting regions were obtained from Reference 3-2 under the design condition basis, as follows: (1) For the Suction Nozzle: Design Condition= # 110, plus thermal gradient stresses Key Element = # 2978 in Finite-Element Model Membrane stress = 22.84 Ksi (p. A-18, Appendix A) Bending stress = 19.87 Ksi (p. A-18, Appendix A) Thickness = 3.625 11 (BJ Drawing SK-0173-4 Rev. A) Bending stress includes 9.45 Ksi due to an assumed through-wall temperature difference of 50°F during heatup/cooldown. (2) For Junction, Volute to Lower Flange: Design Condition= # 104, plus thermal gradient stresses Key Elements = # 1179 & 1183 in Finite-Element Model Membrane stress = 16.0 Ksi (pp. 6-18 & A-76) Bending stress = 23.1 Ksi (pp * . 6-18 & A-76) Thickness = 3. 5" (BJ As-built drawings A-39972, and Vane Regions 1--3, dimension 'B') Bending stress includes 9.45 Ksi due.to an assumed through-wall temperature difference of 50°F during heatup/cooldown. Appendix D 44

(3) For Crotch Region: Design Condition= # 106, plus thermal gradient stresses Key Elements = # 6177 & 6178 in Finite-Element Model Membrane stress = 10.9 Ksi (Figure 6-9 & pp. 6-25 to 6-27) Bending stress = 35.95 Ksi (Figure 6-9 & pp.6-25 to 6-27) Thickness = 4.75 in. (BJ As-built drawings

                            -39973, -39973; Vane Region 9, Dimension 'B' )

Bending stress includes 14.2 Ksi due to an assumed through-wall temperature difference of 75°F during heatup/cooldown. (4) For Diffuser Vane Number 9 Design Condition= #104, plus thermal gradient stresses Key Elements = # 3901 & 3905 in Finite-Element Model Membrane stress = 22.8 Ksi (pp. 6-8, 6-14, 6-15, A-02, A-70) Bending stress = 20.5 Ksi (pp. 6-8, 6-14, 6-15, A-02, A-70) Thickness = 4.0" (p. A-02) Bending stress includes 4.7 Ksi due to an assumed surface-to-interior temperature difference of 25°F during heatup/cooldown. (5) For Volute, Vane 5 vicinity: Design Condition= # 104, plus thermal gradient stresses Key Element = #4501 Appendix D 45

Membrane stress= 27.35 Ksi (pp. A-31 & 6-36). Includes 12.81 Ksi thermal stress derived from Basic Condition 8, Heatup minus Cooldown. Bending stress = 13.9 Ksi (pp. A-31, A-32, & 6-36). Includes 11.29 Ksi thermal stress derived from Basic Condition 8, Heatup minus Cooldown. Thickness = 3.33" (As-built drawing, Scroll Section 5, Dimension 'A') 5.5 Calculation of Crack Growth Rates The methodology described in Section 5.1.4 of the main report was applied to the above values, using the annual rate of stress-cycling given in Reference 3-2. An integration procedure was used to predict dimensionless crack depth, a/t, as a function of time, T. Results are summarized in Tables 5-1, through 5-5. For each region the calculated entries are listed against crack depth, a/t, as follows: (1) Applied Stress Intensity Factor, KI' was calculated using the ASME Section XI procedure, as further described in Section 5.1.4 of the main body of the report. Units for KI are Ksi--squareroot inch. (2) Crack growth rate, da/dT, was calculated using the Bernard & Slama equation (with R=O), multiplied by the design value of the annual rate of stress cycling, as further described in Section 5.1.4. The final equation is:

                                 -10    4.0 da/dT =  7.534 x 10     KI Appendix D                                             46

Units for da/dT are inches/year. (3) Incremental time, dT, in which the crack will grow through the indicated range of dimensionless crack depth values, a/t, was calculated as described in 5.1.4 of the main report. Units for dT are years. The summation of time increments yields the total Time for a crack to grow to a given a/t value. The predicted growth curves for hypothetical cracks show the functional relationship between a/t and total Time, using the initial condition of a/t = 0.08 at Time = o. The first incremental time listed is based on a change in a/tin the amount 0.02 (i.e. 2% of thickness), to indicate the time needed for the crack to grow from a/t

         = 0.08 to a/t = 0.10. Subsequent incremental times are based on the time needed for the crack to grow through changes in a/t of 0.05 (i.e. 5% of thickness). The first such incremental time is given for the range a/t
         = 0.10 to 0.15. The final incremental time is given for the range a/t = 0.45 to 0.50, whereupon the analysis is terminated.

5.6 Stresses Under Emergency And Faulted Conditions I In order to _verify that limiting sections containing hypothetical cracks could withstand Emergency Condition and Faulted Condition Loads, the methodology described in 5.1.4 was again applied. Applied stress intensity factors were calculated at the limiting locations, based on data from Reference 3-2. 5.6.1 Emergency Condition Stresses (1) suction Nozzle: Key element = # 2978 Condition = # 401 Appendix D 47

Membrane stress = 19.36 Ksi (Table A-12, pp. A-58, A-59) Bending stress = 3.07 Ksi (Table A-12, pp.A-58, A-59) These values are an upper bound for Element # 2978. (2) Junction, Volute to Lower Flange: Key element = # 1179 Condition = # 402 Membrane stress = 20.08 Ksi (Table A-12, pp. A-58, A-59) Bending Stress = 7.56 Ksi (Table A-12, pp. A-58, A-59) (3) Crotch Region: Key element = # 6178 Condition = # 401 Membrane stress = 23.25 Ksi (pp. 6-71, A-29) Bending stress .. = 2.65 Ksi (Table A-12, p. A-59) These are conservative values based on Element 6108 results. (4) Diffuser Vane 9: Key element = # 3905 Condition = # 401 Membrane stress = 22.29 Ksi (pp. 6-69, 6-70) Bending stress = 16.14 Ksi (pp. 6-69, 6-70) Appendix D 48

        - - - - - - - - - - = = - = - - - - - -::=:-* --===------- --==--=--=------ --- -------------__ -1I

These results are from Element 3805, which is representative, with secondary stresses removed. (5) Volute, Vane 5 vicinity: Key element = # 4501 Condition = # 401 Membrane stress = 22.15 Ksi (Table A-12, pp. A-56, A-57) Bending stress = 0.90 Ksi (Table A-12, pp.A-56, A-57) These are conservative values from the adjacent element 4502 results. 5.6.2 Faulted Condition Stresses (1) Suction Nozzle: Key element = # 2978 Condition = # 507 Membrane stress = 28.67 Ksi (Table A-13, pp. A-62, A-63) Bending stress = 15.21 Ksi (Table A-13, pp.A-62, A-63) (2) Junction, Volute to Lower Flange: Key element = # 1179 Condition = # 508 Membrane stress = 27.78 Ksi (Table A-13, pp. (A-62, A-63) Bending stress = 8.90 Ksi (Table A-13, pp.A-62, A-63) Appendix D 49

(3) Crotch Region: Key element = # 6178 Condition = # 508 Membrane stress = 18.2 Ksi (Table A-13, pp. A-62, A-63) Bending stress = 27.77 Ksi (Table A-13, pp. A-62, A-63) This membrane stress is a conservative upper bound. (4) Diffuser Vane 9: Key element = # 3905 Condition = # 508 Membrane stress = 26.95 Ksi (Table A-13, pp. A-62, A-63) Bending stress = 9.59 Ksi (Table A-13, pp. A-62, A-63) (5) Volute, Vane 5 vicinity: Key element = # 4501 Condition = # 505 Membrane stress = 14.14 Ksi (Table A-13, pp. A-60, A-61) Bending stress = 3.75 Ksi (Table A-13, pp. A60, A-61) These are conservative values from the adjacent element (4502) results. The bending stress is a conservative upper bound. Appendix D so

)ii l1j                                                 Table 5-1 l1j CD

sp, San Onofre 2&3 RCP Casing

.....                                          crack Growth Rate x

0 At Suction Nozzle a/t Interval KI da/dl DT (Fraction) (KSI !TN) (IN/YEAR) (YEARS) 0.08 -- 0.10 44.91 3.06 x 10- 3 23.6 0.10 -- 0.15 54.13 6.47 x 10- 3 28.0 (a) 0.15 -- 0.20 62.12 1.12 x 10- 2 16.1 (b) 0.20 -- 0.25 69.52 1. 76 x 10- 2 10.3 0.25 -- 0.30 76.44 2.57 x 10- 2 7.0 0.30 -- 0.35 83.46 3.66 x 10- 2 4.9 0.35 -- 0.40 90.66 5.09 x 10- 2 3.5 0.40 -- 0.45 99.06 7.26 x 10- 2 2.5 0.45 -- 0.50 108.5 0.104 1. 7 U1 (a) DT = 38.1 1% £Steps

 .....

(b) DT = 20.2 1% f Steps

~ '!'able 5-2 "CS (I)

sp, San Onofre 2 & 3 RCP Casing

.....                                          crack Growth Rate
><

0 At Junction of Volute to Lower Flange a/t Interval KI da/dl DT (Fraction) (KSI !TR) (IN/YEAR) (YEARS) 0.08 -- 0.10 40.46 2.02 x 10- 3 34.7 0.10 -- 0.15 48.23 4.08 x 10- 3 42.9 0.15 -- 0.20 54. 71 6.75 x 10- 3 25.9 0.20 -- 0.25 60.58 1.01 x 10- 2 17 .2 0.25 -- 0.30 65.86 1.42 x 10- 2 12.3 0.30 -- 0.35 71.19 1.93 x 10- 2 9.0 0.35 -- 0.40 76.48 2.58 x 10- 2 6.8 0.40 -- 0.45 82.80 3.54 x 10- 2 4.9 0.45 -- 0.50 89.87 4.91 x 10- 2 3.5 (rm= 16.0, rb = 23.1, t = 3.5") U1 l\J

~ '.fable 5-3 tO (D

s San Onofre 2 & 3 RCP Casing

~
.....                                             Crack Growth Rate
><

c At

                                                   .crotch Region a/t Interval                    KI                da/dT .             OT (Fraction)                 (KSI ITR)           (IN/YEAR)         (YEARS) 0.08  -- 0.10                55.01             6.90 x   10- 3        13.7 0.10  -- 0.15                64.54             1.31 x   10- 2        18.2 (a) 0.15  -- 0.20                72.05             2.03 x   10- 2        11. 7 (b) 0.20  -- 0.25                78.56             2.87 x   10- 2         8.2 0.25  -- 0.30                 83.99            3.75 x   10- 2         6.3 0.30  -- 0.35                 89.41            4.81 x   10- 2         4.9 0.35  -- 0.40                 94.43            5.99 x   10- 2         3.9 0.40  -- 0.45                 100.7            7.76 x   10- 2         3.0 0.45  -- 0.50                 107.7            0.101                  2.3 (rm .. 10.9, rb = 35.95, t = 4.75")

U1 LJ a . (a) 1% I Steps RT 1 = 23.6 (a) (b) 1% I Steps RT 1 a 13.9 (b)

Table 5-4 San Onofre 2 & 3 RCP Casing Crack Growth Rate At Diffuser Vane a/t Interval KI da/dT OT (fraction) (KSI ITIO (IN/YEAR) (YEARS) 0.08 -- 0.10 48. 74 4.25 x 10- 3 18.8 0.10 -- 0.15 58.66 8.92 x 10- 3 22.4 (a) 0.15 -- 0.20 67.21 1.54 x 10- 2 13.0 (b) 0.20 -- 0.25 75.10 2.40 x 10- 2 8.3 0.25 -- 0.30 82.45 3.48 x 10- 2 5.7 0.30 -- 0.35 89.89 4.92 x 10- 2 4.0 0.35 -- 0.40 97.50 6.81 x 10- 2 2.9 0.40 -- 0.45 106.4 9.66 x 10- 2 2.1 0.45 -- 0.50 116.4 0.138 1.4 ( rm = 22 . 8 ' rb = 20 *5' t = 4.0") (a) 1% £Steps RT 1 = 30.4 (b) 1% £Steps RT 1 = 16.2

Table 5-5 San Onofre 2 & 3 RCP Casing Crack Growth Rate At Volute (Vane Region 5) a/t Interval KI da/dT DT (Fraction) {KSI !TN) (IN/YEAR) {YEARS} 43.21 2.63 x 10- 3 25.3 0.10 -- 0.15 52.57 5.75 x 10- 3 28.9 (a) 0.15 -- 0.20 60.88 1.03 x 10- 2 16.1 (b} 0.20 -- 0.25 68.70 1.68 x 10- 2 9.9 0.25 -- 0.30 76.19 2.54 x '10- 2 6.5 0.30 -- 0.35 83.81 3.72 x 10- 2 4.4 0.35 -- 0.40 91. 77 5.34 x 10- 2 3 .1 0.40 -- 0.45 100.9 7.82 x 10- 2 2 .1 0.45 -- 0.50 111.3 0.116 1..4 (rm = 27.35, rb = 13.9, t = 3.33") (a) 1% Ia Steps, RT 1 z 40.0 Ul Ul (b) a 1% {Steps, RT 1 = 20.4

5.7 Results Results of the crack growth analysis for the San Onofre 2&3 RCPs are shown in Figure 5.3-14 of the generic portion of this report. It shows that the postulated 8%t initial crack will grow to 25%t in about 60 years under the influence of the conservatively defined stress cycles in the design specification. The hypothesized crack will then grow larger until it reaches an end-point crack size of 43%t, limited by flow stress, in about 77 years. Appendix D 56

6.0 INSPECTION INTERVAL Results reported in this appendix support the position that the 10-year inspection interval for volumetric examination, as required by ASME Section XI, is not necessary to ensure safe operation during the 40-year licensed life of the plant. The conservatively calculated end-point crack size is not reached until 77 years after initial operation. The demonstrated casing integrity also supports a relaxation of the 10-year interval for visual examinations, as currently required by ASME Section XI and Code Case N-481. Such examinations add unnecessarily to personnel exposure with no significant benefit to safety. The ASME Section XI requirement for VT-3 examination of internal surfaces is an appropriate low-marginal-exposure monitoring activity to the extent practical, but only when the pump is disassembled for maintenance or repair. Based on the results obtained in this evaluation, relaxation of the casing inspection interval for the RCPs at San Onofre 2 and 3 from 10 years to 40 years is considered to be justified . Appendix D . 57

APPENDIX D REFERENCES 3-1 Project Specification for Reactor Coolant Pumps for Southern California Edison, San Onofre Station, Units 2 and 3, 1370-PE-480, Rev. 05 (10/11/77). 3-2 Pump Case Analysis for So. Cal. Ed. San Onofre Sta. 2 & 3, TCF1025-STR, Vol. 2, Rev. 1, dated July 14, 1978. 3-3 Letter: L. A. Wright, So. Cal. Ed. Co., to A. G. Schoenbrunn, ABB C-E Nuclear Power, dated 7/12/91. 4-1 o. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems", NUREG/CR-4513 (ANL-90/42), U.S. Nuclear Regulatory Commission, Washington, D.C., June 1991. Appendix D 58

APPENDIX E APPLICATION OF GENERIC METHODOLOGY FOR RELAXATION OF THE ST. LUCIE 1&2 REACTOR COOLANT PUMP CASING INSPECTION INTERVAL Appendix E 1

ABSTRACT Appendix E was prepared to demonstrate the amount of-inspection interval relaxation appropriate for the reactor ~ coolant pump casings at the St. Lucie 1 & 2 plants, based on application of the generic methodology presented in the main body of this report. Appendix E 2

                                    ---------     --** ---------

APPENDIX E TABLE OF CONTENTS Section Title Page

1. 0 PURPOSE 6

2.0 PRE-SERVICE INSPECTION DATA EVALUATION 7 3.0 OPERATING HISTORY 9 3.1 Design Specifications 9 3.2 Stress cycles Used In Evaluation 10 3.3 Stress Cycles at st. Lucie 1&2 To-date 10 4.0 THERMAL EMBRITTLEMENT 12 4.1 Material Identification and Chemical Properties 12 4.2 Material Specifications and Mechanical Properties 12 4.3 Thermal Aging Behavior 13 4.4 Toughness Properties of Aged Materials 13 4.5 Limiting Values 14 5.0 CRACK GROWTH ANALYSIS 39 5.1 Scope 39 5.2 Reference Stress Reports 39 5.3 Selection of High Stress Locations 39 5.4 Stresses and Wall Thicknesses at Limiting Locations 40 5.5 Calculation of Crack Growth Rates 42 5.6 Stresses Under Emergency and Faulted Conditions 48 5.7 Results 51 6.0 INSPECTION INTERVAL 52 APPENDIX E REFERENCES 53 Appendix E 3

LIST OF TABLES TABLE 4-1 Material Identification and Chemical Compositions - Unit 1 15 TABLE 4-2 Material Specification and Tensile Properties - Unit 1 17 TABLE 4-3 Predicted Thermal Aging Behavior - Unit 1 19 TABLE 4-4 Predicted Toughness Properties of Aged Material (70°F) - Unit 1 21 TABLE 4-5 Predicted Toughness Properties of Aged Material (550°F) - Unit 1 23 TABLE 4-6 Limiting and Controlling Values of Jic and KJc at 70°F - Unit 1 25 TABLE 4-7 Limiting and Controlling Values of Jic and KJc at 550°F - Unit 1 26 TABLE 4-8 Material Identification and Chemical Compositions - Unit 2 27 TABLE 4-9 Material Specification and Tensile Properties - Unit 2 29 TABLE 4-10 Predicted Thermal Aging Behavior - Unit 2 31 TABLE 4-11 Predicted Toughness Properties of Aged Material (70°F) - Unit 2 33 Appendix E 4

LIST OF TABLES (Continued) TABLE 4-12 Predicted Toughness Properties of Aged Materials (550°F) - Unit 2 35 TABLE 4-13 Limiting and Controlling Values of Jic and KJc at 70°F - Unit 2 37 TABLE 4-14 Limiting and Controlling Values of Jic and KJc at 550°F - Unit 2 38 TABLE 5-1 Crack Growth Rates at Vane Number 8 43

           - Level D TABLE 5-2   Crack Growth Rates at Discharge Nozzle crotch Vicinity - Section  c     44 TABLE 5-3   Crack Growth Rates at suction Nozzle-Level c                          45 TABLE 5-4   Crack Growth Rates at Volute Junction With Lower Flange Near Vanes 1 & 2      46 TABLE 5-5   Crack Growth Rates at Hanger Bracket    47
           #1 Vicinity Appendix E                                        5

1.0 PURPOSE

• i The purpose of Appendix E is to document the application of methodology presented in the main body of this report to the plant-specific data for the reactor coolant pump casings at the St. Lucie 1&2 plants, and to quantify the extent of inspection interval relaxation available.

Appendix E 6

2.0 PRE-SERVICE INSPECTION DATA EVALUATION Pre-service inspection data for the St. Lucie 1&2 reactor coolant pumps numbered 681-N-0445 through 681-N-0448 for Unit 1 and 741-N-0001 through 741-N-0004 for Unit 2 were collected from QA data packages originally prepared by the Byron Jackson Company and stored in archives by ABB Combustion Engineering Nuclear Power. Information in these data packages concerning welding procedures, radiographic inspections, non-destructive testing and dye penetrant testing were examined. The testing and inspection procedures that were followed for all reactor coolant pumps at Calvert Cliffs were found to be the same in all significant aspects. The most relevant information obtained from this review of the QA data packages were the reports on radiographic examination of the

• RCP casing castings, pressure retaining welds, and repair welds.

Radiographic examination requirements invoked ASME Section III rules for examination procedures and sensitivity. The required radiograph sensitivity was 2-2T according to applicable ASTM Standard Reference Radiograph Procedure requirements (i.e. ASTM E71, E186, E280) as determined by the casting thickness. The 2-2T sensitivity is consistent with a 2% initial flaw size, because the requisite image quality indicator (IQI) for this level of examination is specified as a penetrameter with a minimum hole-size diameter equal to 2% of the casting thickness. The acceptance criteria for interpretation of the radiographs wa~ Severity Level 2 for sand, porosity or shrinkage indications. Linear indications such as cracks, hot tears, and unfused chaplets or chills were unacceptable at any level. Any such discernible indications required rejection of weld repair and a repeated radiographic examination of the affected casting or weldment. Appendix E 7

The results of this review of pre-service RCP casing examinations - - confirm that cracks of 2% described in Section 4.4 of the generic report were detectable but none were left unrepaired. Appendix E 8

3.0 Operating History 3.1 Design Specifications The st. Lucie 1&2 RCPs were delivered to the site in 1973 and 1977 respectively and were first placed in commercial operation in 1976 and 1983. Reactor coolant system design pressure and temperature are 2500 psia and 650°F respectively. Each pump is designed to deliver 81,200 gpm of coolant at a head of 310 feet. These pumps have 30 inch diameter suction and discharge piping. The design specifications (Reference 3-1 and 3-2) call for the pumps to be capable of withstanding the following transient conditions events during the 40-year license period: Assumed Occurrences Transient During 40 Year Condition License Period Heat-Up (100°F/hr) 500 Cool-Down (100°F/hr) 500 Hydrostatic Test (3125 psia 100-400°F) 10 Leak Test (2250 psia 100-400°F) 200 Loss of Secondary Pressure 5 Reactor Trip or Loss of Load 400 Appendix E 9

3.2 Stress Cycles Used in Evaluation As indicated in Section 5.1.4 of the generic portion of this report, crack growth was evaluated on the basis of an design number of stress cycles between atmospheric and operating pressures during heatup and cooldown over the nominal 40-year life of the plant. The number of such cycles used in the stress analyses performed by the Byron Jackson Company is 500, as given on page 94 of both References 3-3 and 3-4. This total was increased to 505 to include Loss of Secondary Pressure. The total is therefore 5 transient events greater than the 500 heat-up cooldown cycles specified in the RCP design specifications. The average annual number of stress cycles is therefore 12.625, and the hypothetical crack qrowth calculations and curves were prepared accordingly. 3.3 Stress cycles at St. Lucie To-date Details of the actual operating history of the St. Lucie RCPs from 1976 (Unit 1) and 1983 (Unit 2) to 1991 were furnished in Reference 3-5 and are as follows: Unit 1 Unit 2 Heatup/Cooldown - <50 <25 Reactor Trip - 222 90 Heatup-plus-cooldown, taken together, constitute one cycle. The average number of stress cycles per year over the 15 operating years for Unit 1 and 8 years for Unit 2, ending in mid 1991 is <3.3 and <3.1 cycles per year respectively. This time period is equivalent to <3.9 years for Unit 1 and

    <2.0 years for Unit 2 at the design rate of stress cycling Appendix E                                                  10

(12.6 per year). The actual rate of cycle accrual for each plant is seen to be <26% and <25% of the design rate for Units 1 and 2, respectively, a significant conservatism. Appendix E 11

4.0 THERMAL EMBRITTLEMENT Thermal embrittlement evaluation of the st. Lucie 1&2 casings is discussed and plant-specific data are presented in the five following reports. All equations referenced below are found in the main body of this report, which is also referred to as the generic report. 4.1 Material Identification and Chemical Properties The chemical compositions provided in Report #1 (Table 4-1 and 4-8) for each RCP casing at st. Lucie 1&2 were obtained from Quality Assurance documents originally supplied by the Byron Jackson Companyo A copy of these data packages is stored at ABB Combustion Engineering Nuclear Power. For each individual pump casing, chemical compositions are given for specific casing welds as well as for individual castings. 4.2 Material Specifications and Mechanical Properties The material specifications and mechanical properties found in Report #2 {Table 4-2 and 4-9) for each RCP casing were obtained from the same data source as in Section 4.1 above. For each individual pump casing the material specification, material type and heat number are qiven for specific casing welds as well as for individual castinqs. It is evident from the report that data obtained for mechanical properties (i.e. yield strenqth, tensile strenqth, total elongation and reduction in area) for each material was only available for the castinqs, and was not available for the casinq welds. In all cases, however, the unaqed flow stress at 70°F and the aged flow stress at 70°F and 550°F were calculated as discussed in Section 5.2.3 of the qeneric report. Appendix E 12

                             - ---- - ---------- -- -~------              ----
                                                             ---- -------------

4.3 Thermal Aging Behavior Report #3 (Taple 4-3 and 4-10) contains predicted thermal aging behavior data for all of the St. Lucie 1&2 RCP casings. The measured ferrite contents listed for specific casing welds and individual castings were supplied by the Byron Jackson Company in the same QA package as referenced in Section 4.1 above. In most cases a value was obtainable for the measured ferrite content. In cases where a value was not given, a zero was recorded. The chromium and nickel equivalents for the castings and weld metal, as well as the chromium/nickel ratio for the castings, were calculated using equations 5-7 and 5-8 respectively, as discussed in Section 5.2.1 of the generic report. Values for ferrite content of the castings were computed using two methods: for ferrite content #1, the values were computed using the method which follows ASTM ASOO/ASOOM (Reference 5-2 of the generic report); for ferrite content #2, the values were computed using equation 5-9 as discussed in Section 5.2.1 of the generic report. The latter method follows work performed by O.K. Chopra (Reference 4-1). 4.4 Toughness Properties of Aged Materials The predicted toughness properties of aged material at 70°F and 550°F are given respectively in Report #4 (Table 4-4 and 4-11) and Report #5 (Table 4-5 and 4-12). The measured ferrite contents listed for all heat numbers are the same as the values given in Report #3. The material aging parameter was calculated using equation 5-10 of the generic report. The room-temperature Charpy impact energy, CVsat' of the various materials was calculated using equation s-11. The Jic values were determined in accordance with the methods of Appendix E 13

ASTM E813 as discussed in Section 5.2.3 of the generic report. The plane strain fracture toughness, KJc' and minimum tearing modulus, T, at 70°F and 550°F were calculated using equations 5-17 and 5-18. The values listed for the material constants N and c at 70°F and 550°F were calculated using equations 5-12 through 5-15. These constants were needed in computing the values for Jic' KJc and T. All equations used in Reports #4 and #5 are found in Section 5.2 of the generic report. 4.5 Limiting Values The limiting and controlling values for Jic and KJc at 70°F and 550°F for each individual pump at St. Lucie Units 1&2 are given in Tables 4-6 and 4-7 for Unit 1 and 4-13 and 4-14 for Unit 2. Appendix E 14

)>

"'O Table 4-1 "'O C1>

I a.

......

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

)(                                          REACTOR COOLANT PUMP CASINGS l'T1 REPORT l l - MATERIAL IDENTIFICATION & CHEMICAL COMPOSITIONS MATERIAL HEAT NO.      c    Mn      Si      s      p      Cr     Ni    Mo    N   Cb
       ** PLANT I. D. ST. LUCIE UNIT 1

• RCP PUMP CASING 681-N-0445 CASING WELD 04146 0.02 1.84 1.47 0.014 0.014 . 20.48 9.96 2.70 0.04 0.00 HUB/DIFFUSER 40116 0.06 0.67 0.96 0.009 0.036 19.15 9.1J8 2.48 0.04 0.00 CASING WELD 03063 0.04 1.70 0.53 0.016 0.011 19.89 10.29 2.81 0.04 0.00 CASE SCROLL 46737 0.06 0.70 1.22 0.003 0.034 18.66 9.33 2.29 0.04 0.00 CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 19.90 9.10 2.31 0.04 0.00 CASING WELD 04367 0.02 1.64 1.37 0.018 0.013 19.01 9.78 2.98 0.04 0.00 CASING WELD 04459 0.02 0.91 0.51 0.014 0.013 19.82 9.44 2.46 0.04 0.00 CASING WELD 03036 0.03 1.53 0.47 0.016 0.013 19.01 9.70 2.81 0.04 0.00 CASING WELD 04313 0.02 0.91 0.52 0.016 0.015 19.59 9.93 2.59 0.04 0.00 CASING WELD 03036A 0.04 1.52 0.47 0.015 0.014 18.69 9.89 2.84 0.04 0.00 CASING WELD 04455 0.02 0.95 0.51 0.017 0.013 18.87 9.86 2.46 0.04 0.00 CASING WELD 04286 0.02 0.90 0.52 0.015 0.015 18.83 10. ]2 2.41 0.04 0.00
• RCP PUMP CASING 681-N-0446 HUB/DIFFUSER 46993 0.06 0.72 1.16 0.003 0.036 19.45 9.54 2.24 0.04 0.00 CASE SCROLL 48368 0.04 0.78 1.07 0.003 0.037 19.06 9.19 2.28 0.04 0.00 CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 19.90 9.10 2.31 0.04 0.00 CASING WELD 04460 0.02 1.00 0.51 0.015 0.012 20.35 10.CIO 2.37 0.04 0.00

....... CASING WELD 04459 0.02 0.91 0.51 0.014 0.013 19.82 9.44 2.46 0.04 0.00 U1 CASING WELD 04509 0.02 0.94 0.45 0.018 0.015 19.67 9. /'5 2.51 0.04 0.00 CASING WELD 04313 0.02 0.91 0.52 0.016 0.015 19.59 9.93 2.59 0.04 0.00 CASING WELD 03036A 0.04 1.52 0.47 0.015 0.014 18.69 9.89 2.84 0.04 0.00 CASING WELD 04635 0.02 1.00 0.49 0.015 0.013 19.40 10.0IO 2 .71 0.04 0.00 CASING WELD T03951 '0.04 1.55 0.50 0.014 0.025 19.02 10.0l3 2.44 0.04 0.00 CASING WELD 01953 0.02 1.68 0.44 0.021 0.015 19.11 9.69 2.83 0.04 0.00 CASING WELD 04455 0.02 0.95 0.51 0.017 0.013 18.87 9.86 2.46 0.04 0.00

---

-

 ):ll "C

Table 4-1 (Continued)

"C 11>

s ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

 .....

0..

 )(

REACTOR COOLANT PUMP CASINGS l'T1 REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS MATERIAL HEAT NO. c Mn S1 s p Cr Ni Mo N Cb CASING WELD S7203 0.02 0.66 0.48 0.017 0.039 18.6S 10.20 2.41 0.04 0.00

• RCP PUMP CASING 681-N-0447 CASING WELD 04146 0.02 1.84 1.47 0.014 0.014 20.48 9.96 2.70 0.04 0.00 CASE SCROLL 4S920 o.os 0.78 1.13 0.003 0.036 19.88 9.63 2 .19 0.04 0.00 HUB/DIFFUSER 4S871 0.06 0.68 1.08 o.oos 0.036 19.lS 9. 77 2.17 0.04 0.00 CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 19.90 9.10 2.31 0.04 0.00 CASING WELD 04367 0.02 1.64 1.37 0.018 0.013 19.01 9.78 2.98 0.04 0.00 CASING WELD 04S09 0.02 0.94 0.45 0.018 O.OlS 19.67 9.7S 2.51 0.04 0.00 CASING WELD 03036 0.03 I.S3 0.47 0.016 0.013 19.01 9.70 2.81 0.04 0.00 CASING WELD 04313 0.02 0.91 O.S2 0.016 0.015 19.59 9.93 2.S9 0.04 0.00 CASING WELD 03036A 0.04 1.52 0.47 0.015 0.014 18.69 9.89 2.84 0.04 0.00 CASING WELD T03951 0.04 I.SS o.so 0.014 0.02S 19.02 10.03 2.44 0.04 0.00 CASING WELD 04455 0.02 0.9S O.Sl 0.017 0.013 18.87 9.86 2.46 0.04 0.00
• RCP PUMP CASING 681-N-0448 HUB/DIFFUSER 46406 0.08 0.96 0.94 0.003 0.033 19.00 9.17 2.21 0.04 0.00 CASE SCROLL 47380 0.06 0.70 1.22 0.003 0.034 18.66 9.33 2.29 0.04 0.00 CASING WELD X43439 0.03 1.39 0.36 0.017 0.016 19.90 9.10 2.31 0.04 0.00 CASING WELD 04460 0.02 I.OD O.Sl O.OlS 0.012 20.35 10.00 2.37 0.04 0.00 CASING WELD 044S9 0.02 0.91 O.Sl 0.014 0.013 19.82 9.44 2.46 0.04 0.00

  -

Ol CASING WELD CASING WELD CASING WELD 04S09 03036 04313 0.02 0.03 0.02 0.94 I.S3 0.91 0.4S 0.47 O.S2 0.018 0.016 0.016 O.OlS 0.013 O.OlS 19.67 9.7S 2.51 0.04 0.00 19.01 9.70 2.81 0.04 0.00 19.S9 9.93 2.59 0.04 0.00 CASING WELD 03036A 0.04 I.S2 0.47 0.015 0.014 18.69 9.89 2.84 0.04 0.00 CASING WELD T039Sl 0.04 I.SS o.so 0.014 0.02S 19.02 10.03 2.44 0.04 0.00 CASING WELD 019S3 0.02 1.68 0.44 0.021 O.OlS 19.11 9.69 2.83 0.04 0.00 CASING WELD 04455 0.02 0.9S 0.51 0.017 0.013 18.87 9.86 2.46 0.04 0.00 CASING WELD S7203 0.02 0.66 0.48 0.017 0.039 18.6S 10.20 2.41 0.04 0.00 e e

)>

"O TABLE 4-2 "O n>

s a..

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

-'*

x REACTOR COOLANT PUMP CASINGS

 ,.,.,

REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH HONG. AREA FLOW FLOW FLOW PART HEAT NO. (ksi) (ksi) (%) (%) STRESS STRESS STRESS

                                                                                             ~ 70F  @ 70F @ 550F
           ** PLANT I .D. ST. LUCIE UNIT 1

• RCP PUMP CASING 681-N-0445 CASING WELD 04146 0.0 0.0 0.0 0.0 0 78600 58000 HUB/DIFFUSER A 351 CF8M 40116 41.9 85.0 60.0 73.0 63 76215 55621 CASING WELD A 298 [316-16 03063 0.0 0.0 0.0 0.0 0 78600 58000 CASE SCROLL A 351 CF8M 46737 45.5 88.0 55.0 71.0 67 81440 60846 CASING WELD A 298 E316-16 X43439* 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD 04367 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04459 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03036 0.0 o.o 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04313 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 [316-16 03036A 0.0 o.o 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04455 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04286 0.0 0.0 0.0 0.0 0 78600 58000
• RCP PUMP CASING 681-N-0446 HUB/DIFFUSER A 351 CF8M 46993 45.0 84.0 'J4. 0 72.0 64 77878 57284 CASE SCROLL A 351 CF8M 48368 43.3 86.8  !;a.o 70.0 65 78749 58155 CASING WELD A 298 E316-16 X43439 0.0 0.0 0.0 0.0 0 78600 58000

    ......

CASING WELD A 371 ER-316 04460 0.0 0.0 o.o 0.0 0 78600 58000

    ...... CASING WELD   A 371    ER-316   04459           0.0      0.0        0.0     0.0      0  78600  58000 CASING WELD   A 371    ER-316   04509           0.0      0.0        0.0     0.0      0  78600  58000 CASING WELD   A 371    ER-316   04313           0.0      0.0        0.0     0.0      0  78600  58000 CASING WELD   A 298    E316-16  03036A          0.0      0.0        0.0     0.0      0  78600  58000 CASING WELD                     04635           0.0      0.0        0.0     0.0      0  78600  58000 CASING WELD   A 298    E316-16 T03951           0.0      0.0        0.0     0.0      0  78600  58000 CASING WELD   A 371    ER-316 01953             0.0      0.0        0.0     0.0      0  78600  58000 CASING WELD   A 371    ER-316 04455             0.0      0.0        0.0     0.0      0  78600  58000 CASING WELD   A 298    E316-16 57203            0.0      0.0        0.0     0.0      0  78600  58000

l> "'C TABL~ 4-2 (Continued) "'C CD

J 0..

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

......
)(

REACTOR COOLANT PUMP CASINGS

""                          REPORT 12 - MATERIAL SPECIFICATION &TENSILE PROPERTIES MATERIAL     MATERIAL MATERIAL                 YIELD TENSILE     TOTAL RED. IN UNA GED   AGED   AGED OR        SPEC. TYPE                  STRENGTH STRENGTH   HONG. AREA   FLOW    FLOW   FLOW PART                           HEAT NO.      (ks1)     (ks1)     (%)     (%) STRESS  STRESS STRESS

• RCP PUMP CASING 681-N-0447 CASING WELD . 04146 0.0 0.0 0.0 0.0 0 78600 58000 CASE SCROLL A 351 CF8M 45920 47.8 85.l 36.0 37.0 66 80965 60371 HUB/DIFFUSER A 351 CF8M 45871. 42.3 85.6 57.0 72.0 64 77007 56413 CASING WELD A 298 E316-16 X43439 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD 04367 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04509 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03036 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04313 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 03036A 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 298 E316-16 103951 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 371 ER-316 04455 o~o 0.0 0.0 0.0 0 78600 58000
• RCP PUMP CASING 681-N-0448 HUB/DIFFUSER A 298 CF8M 46406 42.4 83.4 56.0 75.0 63 75345 54751 CASE SCROLL A 351 CF8M 47380 43.6 87.7 63.0 73.0 66 79699 59105

       . CASING WELD  A 298    E316-16    X4~439          0.0       0.0     0.0     0.0       0  78600  58000 CASING WELD  A 371    ER-316     04460           0.0       0.0     0.0     0.0       0  78600  58000 CASING WELD  A 371    ER-316     04459           0.0       o.o     0.0     0.0       0  78600  58000 CASING WELD  A 371    ER-316     04509           0.0       0.0     0.0     0.0       0  78600  58000 CASING WELD  A 298    E316-16    03036           0.0       0.0     0.0     0.0       0  78600  58000
......                A 371    ER-316     04313           0.0       0.0     0.0     0.0       0  78600  58000 CD       CASING WELD CASING WELD  A 298    E316-16    03036A          0.0       0.0     0.0     0.0       0  78600  58000 CASING WELD  A 298    [316-16    103951          0.0       0.0     0.0     0.0       0  78600  58000 CASING WELD  A 371    ER-316     01953           0.0       0.0     0.0     0.0       0  78600  58000 CASING WELD  A 371    ER-316     04455           0.0       0.0     0.0     0.0       0  78600  58000 CASING WELD  A 298    [316-16    57203           0.0       0.0     0.0     0.0       0  78600  58000 e

)>

"'C Table 4-3 "'C (1)

s 0..

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

......                                         REACTOR COOLANT PUMP CASINGS x
,..,,

REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR HEAT MEASURED CHROMIUM NICKEL Cre/Nie CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS #1 (%) #2 (%) WELDS WELDS (%)

          ** PLANT 1.0. ST. LUCIE UNIT 1

• RCP PUMP CASING 681-N-0445 04146 23.0 19.463 14.129 1.38 24.0 29.4 25.385 12.680 18.0 40116 12.0 17 .622 14.526 1.21 14.2 14.7 23.070 12.815 11.0 03063 15.0 18.555 14.840 1.25 13.2 17 .6 23.495 13.420 15.0 46737 14.0 17.026 14.379 1.18 13.7 12.7 22.780 12.680 11.0 X43439 14.0 17 .878 13 .477 1.33 17.2 24.3 22.750 11.895 16.0 04367 30.0 18.283 13.933 1.31 20.7 22.9 24.045 12.400 18.0 04459 21.0 18.051 13.529 1.33 19.5 25.0 23.045 11. 695 16.0 03036 15.0 17.646 14.138 1.25 13.7 17 .4 22.525 12.625 11. 0 04313 18.0 17.983 14.019 1.28 17.8 20.3 22.960 12.185 15.0 03036A 13.0 17.362 14.474 1.20 11.0 13.8 22.235 12.990 8.0 04455 16.0 17 .101 13.953 1.23 13.7 15.7 22.095 12.135 11. 0 04286 15.0 17.006 14.208 1.20 12.3 13.6 22.020 12.370 11. 0
• RCP PUMP CASING 681-N-0446 46993 14.0 17.727 14.591 1.21 14.7 14.9 23.430 12.900 16.0 48368 12.0 17 .342 13.757 1.26 17.2 18.4 22.945 11. 980 15.0

  ....... X43439       14.0    17 .878   13 .477     1.33       17 .2      24.3   22.750 11. 895       16.0 04460        20.0    18.473    14.097      1.31       17.8       22.7   23.485 12.300        16.0
  '°       04459        21.0    18.051    13.529      1.33       19.5       25.0   23.045 11.695        16.0 04509        19.0    17.933    13.842      1.30       17 .2      21.4   22.855 12 .020       16.0 04313        18.0    17 .983   14.019      1.28       17.8       20.3   22.960 12.185        15.0 03036A       13.0    17 .362   14.474      1.20       13.8       13.8   22.235 12.990         8.0 04635        18.0    17 .924   14.097      1.27       16.2       19.3   22.845 12.300        15.0 T03951       13.0    17.222    14.568      1.18       10.1       12.6   22.210 13.085         8.0 01953        18.0    17. 755   13.798      1.29       15.2       20.6   22.600 12.270        16.0 04455        16.0    17 .101   13.953      1.23       13.7       15.7   22.095 12.135        11.0 57203        14.0    16.806    14.289      1.18       11.8       12.2   21.780 12.360        11. 0

 )>
"C                                                Tabl~  4-3 (Continued)
"C ID

J

a. ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

 .....
 )(

REACTOR COOLANT PUMP CASINGS rr1 REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR HEAT MEASURED CHROMIUM NICKEL Cre/Nie CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS #1 (%) #2 (%) WELDS WELDS (%)

• RCP PUMP CASING 681-N-0447 04146 23.0 19.463 14.129 1.38 24.0 29.4 25.385 12.680 18.0 45920 15.0 18.082 14.442 ' 1.25 16.7 17. 7 23.765 12. 720 15.0 45871 12.0 17.304 14.817 1.17 12.3 11.6 22.940 13 .110 11.0 X43439 14.0 17.878 13.477 1.33 17 .2 24.3 22.750 11. 895 16.0 04367 30.0 18.283 13.933 1.31 20.7 22.9 24.045 12.400 18.0 04509 19.0 17.933 13.842 1.30 17 .2 21.4 22.855 12.020 16.0 03036 15.0 17 .646 14.138 1.25 13.7 17 .4 22.525 12.625 11.0 04313 18.0 17 .983 14.019 1.28 17 .8 20.3 22.960 12.185 15.0 03036A 13.0 17 .362 14.474 1.20 11.0 13.8 22.235 12.990 8.0 T03951 13.0 17.222 14.568 1.18 11.0 12.6 22.210 13.085 8.0 04455 16.0 17 .101 13.953 1.23 13.7 15.7 22.095 12.135 11.0
• RCP PUMP CASING 681-N-0448 46406 10.0 17 .135 14.734 1.16 11.0 11.3 22.620 13.250 11.0 47380 12.0 17 .026 14.379 1.18 13.7 12.7 22.780 12.680 11.0 X43439 14.0 17.878 13.477 1.33 17 .2 24.3 22.750 11.895 16.0 04460 20.0 18.473 14.097 1.31 17 .8 22.7 23.485 12.300 16.0 04459 21.0 18.051 13.529 1.33 19.5 25.0 23.045 11. 695 16.0 04509 19.0 17.933 13.842 1.30 17 .2 21.4 22.855 12.020 16.0 1~ 03036 15.0 17 .646 14.138 1.25 13.7 17.4 22.525 12.625 11.0 04313 18.0 17 .983 14.019 1.28 17 .8 20.3 22.960 12 .185 15.0 03036A 13.0 17.362 14.474 1.20 11.0 13.8 22.235 12.990 8.0 T03951 13.0 17.222 14.568 1.18 10. l 12.6 22.210 13 .085 8.0 01953 18.0 17. 755 13. 798 .1. 29 15.2 20.6 22.600 12.270 16.0 04455 16.0 17 .101 13.953 1.23 13.7 15.7 22.095 12 .135 11.0 57203 14.0 16.806 14.289 1.18 11.8 12.2 21. 780 12.360 11.0 e e e

):II

"'C Table 4-4 "'C CD

s ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR Cl.

-'*
)(

COOLANT PUMP CASINGS AT 70F rr1 REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (%) @ 70F @ 70F (ft-lbs) (lb/in) (ksi\/in)

        ** PLANT I. D. ST. LUCIE UNIT 1

• RCP PUMP CASING 681-N-0445 04146 23.0 28.29 3434.0 0.38 21.29 594.5 158.5 50.9 40116 12.0 23.36 4108.0 0.38 27 .17 723.3 174.8 63.9 03063 15.0 21.27 4496.0 0.39 30.68 757.7 178.9 65.l 46737 14.0 20.05 4760.0 0.39 33.11 803.6 184.2 64.1 X43439 14.0 19.87 4804.0 0.39 33.51 815.6 185.6 69.2 04367 30.0 19.49 4894.0 0.39 34.35 832.4 187. 5 70.4 04459 21.0 17.66 5387.0 0.39 39.00 926.2 197.8 76.8 03036 15.0 16.41 5782.0 0.39 42.79 1003.0 205.8 82.0 04313 18.0 15.61 6065.0 0.40 45.54 1006.0 206.1 85.1 03036A 13.0 14.90 6341.0 0.40 48.24 1057.0 211.3 88.6 04455 16.0 11.43 8082.0 0.41 65.70 1327.0 236.8 109.4 04286 15.0 10.42 8761.0 0.41 72.66 1460.0 248.3 117 .4
• RCP PUMP CASING 681-N-0446 46993 14.0 25.15 3829.0 0.38 24. 71 669.2 168.1 57.4 48368 12.0 20.68 4620.0 0.39 31.82 780.0 181.5 66.5 N

X43439 14.0 19.87 4804.0 0.39 33.51 815.6 185.6 69.2

 ...... 04460               20.0        18.40       5176.0         0.39         37.00    885.2      193.4    74 .1 04459               21.0        17.66       5387.0         0.39         39.00    926.2      197.8    76.8 04509               19.0        15.76       6009.0         0.40         45.00    994.7      205.0    84.4 04313               18.0        15.61       6065.0         0.40         45.54   1006.0      206.1    85.1 03036A              13.0        14.90       6341.0         0.40         48.24   1057.0      211.3    88.6 04635               18.0        14.83       6366.0         0.40         48.49   1061.0      211. 7   88.9 T03951              13.0        13.79       6818.0         0.40         52.96   1149.0      220.3    94.5 01953               18.0        13.75       6836.0         0.40         53 .14  1152 .0     220.6    94.8 04455               16.0        11.43       8082.0         0.41         65.70   1327.0      236.8   109.4 57203               14.0         9.58       9394.0         0.42         79.18   1506.0      252.2   124.2

~-----------------------------

  )>                                              Tabl~ 4-4 (Continued)
 "C
  ,,,
 "C

,

Cl. ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

  ~-
  )(

COOLANT PUMP CASINGS AT 70F

  ,.,,                     REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL HEAT           MEASURED      MATERiAL       MATERIAL     MATERIAL     MINIMUM   MINIMUM    MINIMUM MINIMUM NO.             FERRITE         AGING       CONSTANT     CONSTANT      IMPACT        Jlc        KJc T MOD.

CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (I) @ 70F @ 70F (ft-lbs) (lb/1n) (ksi\/1n)

• RCP PUMP CASING 681-N-0447 04146 23.0 28.29 3434.0 0.38 21.29 594.5 158.5 50.9 45920 15.0 26. 90, 3596.0 0.38 22.68 622.5 162.2 50.2 45871 12.0 19.95 4785.0 0.39 33.34 813.9 185.4 71. 7 X43439 14.0 19.87 4804.0 0.39 33.51 815.6 185.6 69.2 04367 30.0 19.49 4894.0 0.39 34.35 832.4 187.5 70.4 04509 19.0 15.76 6009.0 0.40 45.00 994.7 205.0 84.4 03036 15.0 16.41 5782.0 0.39 - 42. 79 1003.0 205.8 82.0 04313 18.0 15.61 6065.0 0.40 45.54 1006.0 206.1 85.1 03036A 13.0 14.90 6341.0 0.40 48.24 1057.0 211.3 88.6 T03951 13.0 13.79 6818.0 0.40 52.96 1149.0 220.3 94.5 04455 16.0 11.43 8082.0 0.41 - 65. 70 1327.0 236.8 109.4
• RCP PUMP CASING 681-N-0448 46406 10.0 21.13 4524.0 0.39 30.93 766.6 179.9 71.0 47380 12.0 20.05 4760.0 0.39 33.11 805.1 184.4 66.8 X43439 14.0 19:81 4804.0 0.39 33.51 815.6 185.6 69.2 04460 20.0 18.40 5176.0 0.39 37.00 885.2 193.4 74.1 04459 21.0 17.66 5387.0 0.39 39.00 926.2 197.8 76.8 04509 19.0 15.76 6009.0 0.40 45.00 994.7 205.0 84.4 N

N 03036 15.0 16.41 5782.0 0.39 42.79 1003.0 205.8 82.0 04313 18.0 15.61 6065.0 0.40 45.54 1006.0 206.l 85.1 03036A- 13.0 14.90 6341.0 0.40 48.24 1057.0 211.3 88.6 T03951 13.0 13.79 6818.0 0.40 52.96 1149.0 220.3 94.5 01953 18.0 13.75 6836.0 0.40 53 .14 1152 .o 220.6 94.8 04455 16.0 11.43 8082.0 0.41 ' 65.70 1327.0 236.8 109.4 57203 14.0 9.58 9_394.0 0.42 79.18 1506.0 252.2 124.2 e e e

):lo

"'C Table 4-5 "'C 11)

J 0..

ANALYSIS OF THERMAL AGING OF STAINLESS STEEL

......                                     REACTOR COOLANT PUMP CASINGS AT 550F x

IT1 REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N EN!ERGY @ 550F @ 550F @ 550F (%) @ 550F @ 550F (ft-lbs) (lb/in) (ksi/in)

       ** PLANT I.D. ST. LUCIE UNIT 1

• RCP PUMP CASING 681-N-0445 04146 23.0 28.29 2840.0 0.28 21.29 813.4 151.0 70.2 40116 12.0 23.36 3249.0 0.28 27.17 947.2 162.9 85.5 03063 15.0 21.27 . 3477 .o 0.28 J,0.68 1016.0 168.7 83.9 46737 14.0 20.05 3629.0 0.29 33.11 1010.0 168.2 80.0 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 169.3 88.0 04367 30.0 19.49 3705.0 0.29 34.35 1038.0 170.6 89. l 04459 21.0 17.66 3980.0 0.29 39.00 1126.0 177 .6 94.8 03036 15.0 16.41 4197.0 0.29 42.79 1195.0 183.0 99.2 04313 18.0 15.61 4349.0 0.29 4.5. 54 1244.0 186.7 102.2 03036A 13.0 14.90 4495.0 0.29 48.24 1291.0 190.2 105.l 04455 16.0 11.43 5383.0 0.30 65.70 1512.0 205.8 122.8 04286 15.0 10.42 5715.0 0.31 72.66 1543.0 208.0 129.8
• RCP PUMP CASING 681-N-0446 46993 14.0 25.15 3083.0 0.28 24. 71 890.5 158.0 77.3 48368 12.0 20.68 3549.0 0.28 31.82 1039.0 170.7 84.9 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 169.3 88.0 N

w 04460 20.0 18.40 3864.0 0.29 37.00 1088.0 174.6 92.4 04459 21.0 17.66 3980.0 0.29 39.00 1126.0 177 .6 94.8 04509 19.0 15.76 4319.0 0.29 45.00 1234.0 186.0 101.6 04313 18.0 15.61 4349.0 0.29 45.54 1244.0 186.7 102.2 03036A 13.0 14.90 4495.0 0.29 48.24 1291.0 190.2 105.1 04635 18.0 14.83 4509.0 0.29 48.49 1295.0 190.5 105.3 T03951 13.0 13.79 4745.0 0.30 52.96 1307.0 191.4 110.6 01953 18.0 13.75 4754.0 0.30 53.14 1311.0 191. 7 110.8 04455 16.0 11.43 5383.0 0.30 65.70 1512.0 205.8 122.8 57203 14.0 9.58 6016.0 0.31 79.18 1638.0 214.3 135.3

                                                                            ."

)>

"C Tabl~ 4-5 (Continued) "C tD

I c..

ANALYSIS OF THERMAL AGING OF STAINLESS STEEL

.....                                        REACTOR COOLANT PUMP CASINGS AT 550F x

l'T1 REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @ 550F @ 550F (ft-lbs) (lb/in) (ksi/in)

• RCP PUMP CASING 681-N-0447 04146 23.0 28.29 2840.0 0.28 21.29 813.4 151.0 70.2 45920 15 .. 0 26.90 2941.0 0.28 22.68 841.3 153.5 67.2 45871 12.0 19.95 3643.0 0.29 33.34 1022.0 169.3 92.5 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 169.3 88.0 04367 30.0 19.49 3705.0 0.29 34.35 1038.0 170.6 89.1 04509 19.0 15.76 4319.0 0.29 45.00 1234.0 186.0 101.6 03036 15.0 16.41 4197.0 0.29 42.79 1195.0 183.0 99.2 04313 18.0 15.61 4349.0 0.29 45.54 1244.0 186.7 102.2 03036A 13.0 14.90 4495.0 0.29 48.24 1291.0 190.2 105.1 T03951 13.0 13.79 4745.0 0.30 52.96 1307.0 191.4 110.6 04455 16.0 11.43 5383.0 0.30 65.70 1512.0 205.8 122.8
• RCP PUMP CASING 681-N-0448 46406 10.0 21.13 3493.0 0.28 30.93 1028.0 169.7 93.7 47380 12.0 20.05 3629.0 0.29 33.11 1012.0 168.4 84.5 X43439 14.0 19.87 3654.0 0.29 33.51 1023.0 169.3 88.0 04460 20.0 18.40 3864.0 0.29 37 .00 1088.0 174.6 92.4 04459 21.0 17.66 3980.0 0.29 39.00 1126. 0 177 .6 94.8 04509 19.0 15.76 4319.0 0.29 45.00 1234.0 186.0 101.6 N

~      03036              15.0           16.41        4197.0          0.29       42.79       1195. 0    183.0       99.2 04313              18.0           15.61        4349.0          0.29       45.54       1244.0     186.7      102.2 03036A             13.0           14.90        4495.0          0.29       48.24       1291. 0    190.2      105.l T03951             13.0           13.79        4745.0          0.30       52.96       1307 .o    191.4      110.6 01953              18.0           13.75        4754.0          0.30       53 .14      1311. 0    191. 7     110.8 04455              16.0           11.43        5383.0          0.30       65.70       1512.0     205.8      122.8 57203              14.0            9.58        6016.0          0.31       79.18       1638.0     214.3      135.3 e                                                       e                                             9'

)>

"'C Table 4 St. Lucie Unit 1 "'C n>

I

0. Limiting and Controlling Values of

......
><                                  J 10 and KJc at 70°F l'T1 RCP PUMP CASING HEAT  #                     Jic           KJc (lb/in)       (ksi/1n) 681-N-0445      04146                       594.5         158.5 681-N-0446      46993                       669.2         168.1 681-N-0447      04146                       594.5         158.5 681-N-0448      46406                       766.6         179.9 N

U1

>                             Table 4-7  ~ St. Lucie Unit 1 "C

"C Cl)

s

.....

0.

)(

Limiting and Controlling Values of fT1 J 10 and KJc at 550°F RCP PUMP CASING HEAT # Jic KJc (lb/in) (ksi/Iil) 681-N-0445 04146 813.4 151.0 681-N-0446 46993 890.5 158.0 681-N-0447 04146 813.4 151.0 681-N-0448 46406 1028.0 169.7 N m

)>

"'C Table 4-8 "'C ct>

I

a. ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

.....                                      REACTOR COOLANT PUMP CASINGS
><

rr1 REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS MATERIAL HEAT NO. c Mn Si s p Cr Ni Mo N Cb

        **PLANT I.D. ST. LUCIE UNIT 2

• RCP PUMP CASING 741-N-0001 HUB/DIFFUSER 91097-1 0.06 0.72 1.18 0.010 0.038 19.76 9.14 2.62 0.04 0.00 CASING WELD 6074 0.06 1.29 0.55 0.014 0.032 20.92 9.84 2.52 0.04 0.00 CASE SCROLL 91402-1 0.05 0.62 1.28 0.010 0.038 19.38 9.50 2 .18 0.04 0.00 CASING WELD 7174 0.03 1.26 0.62 0.018 0.014 19.65 10.20 2.65 0.04 0.00 CASING WELD 5952C 0.06 1.20 0.58 0.017 0.026 19.08 10.80 2.87 0.04 0.00 CASING WELD 05929 0.02 0.91 0.72 0.015 0.013 19.70 9.62 2.54 0.04 0.00 CASING WELD 5733 0.03 1.33 0.41 0.014 0.017 19.33 10.60 2.90 0.04 0.00 CASING WELD 9317-051 0.02 1.10 0.60 0.026 0.006 19.70 9.98 2.28 0.04 0.00 CASING WELD 5280 0.03 1.26 0.47 0.013 0.013 19.60 10.36 2.30 0.04 0.00 CASING WELD 05936 0.01 0.87 0.76 0.017 0.013 19.60 9.66 2.30 0.04 0.00 CASING WELD 5386 0.03 1.19 0.53 0.011 0.023 18.76 10.03 2.60 0.04 0.00 CASING WELD 7242 0.01 0.91 0.52 0.018 0.008 19.38 10.06 2.58 0.04 0.00
• RCP PUMP CASING 741-N-0002 CASING WELD 6074 0.06 1.29 0.55 0.014 0.032 20.92 9.84 2.52 0.04 0.00 CASE SCROLL 97947-1 0.06 0.50 1.23 0.010 0.033 19.50 9.59 2.25 0.04 0.00 HUB/DIFFUSER 95211-1 0.06 0.58 1.09 0.010 0.034 19.00 9.43 2.10 0.04 0.00 N

CASING WELD 7174 0.03 1.26 0.62 0.018 0.014 19.65 10.20 2.65 0.04 0.00

 ...... CASING WELD 05929      0.02  0.91    0.72  0.015  0.013  19.70  9.62  2.54   0.04 0.00 CASING WELD 5733       0.03  1.33    0.41  0.014  0.017  19.33 10.60  2.90   0.04 0.00 CASING WELD 6546       0.03  1.42    0.38  0.017  0.016  20.01 10.39  2.39   0.04 0.00 CASING WELD 9317-051 0.02    1.10    0.60  0.026  0.006  19.70  9.98  2.28   0.04 0.00 CASING WELD 7553A      0.05  1.24    0.42  0.016  0.012  18.91 10.52  2.74   0.04 0.00 CASING WELD 7242       0.01  0.91    0.52  0.018  0.008  19.38 10.06  2.58   0.04 0.00

    )>
   -0                                             Tabl~ 4-8 (Continued)
   -0 CD

I

a. ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

   ......                                     REACTOR COOLANT PUMP CASINGS
    ><

I l'T'I REPORT #1 - MATERIAL IDENTIFICATION &CHEMICAL COMPOSITIONS MATERIAL HEAT NO. C* Mn S1 s p Cr N1 Mo N Cb

• RCP PUMP CASING 741-N-0003 CASING WELD 6074 0.06 1.29 0.55 0.014 0.032 20.92 9.84 2.52 0.04 0.00 HUB/DIFFUSER 99346-1 0.06 0.67 1.27 0.006 0.036 19.14 9.57 2.26 0.04 0.00 CASING WELD 7174 0.03 1.26 P.62 0.018 0.014 19.65 10.20 2.65 0.04 0.00 CASING WELD 6546 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.00 CASING WELD 5733 0.03 1.33 0.41 0.014 0.017 19.33 10.60 2.90 0.04 0.00 CASE SCROLL 99918-1 0.04 0.49 1.21 0.013 0.030 18.76 9.85 2.11 0.04 0.00 CASING WELD 7553A 0.05 1.24 0.42 0.016 0.012 18.91 10.52 2.74 0.04 0.00 CASING WELD 7242 0.01 0.91 0.52 0.018 0.008 19.38 10.06 2.58 0.04 0.00
• RCP PUMP CASING 741-N-0004 CASING WELD 6074 0.06 1.29 0.55 0.014 0.032 20.92 9.84 2.52 0.04 0.00 HUB/DIFFUSER 99161-1 0.06 0.66 1.27 0.010 0.037 19.21 9.50 2.13 0.04 0.00 CASE SCROLL 00233-1 0.07 0.58 1.21 0.006 0.034 18.85 9.42 2.11 0.04 0.00 CASING WELD 6546 0.03 1.42 0.38 0.017 0.016 20.01 10.39 2.35 0.04 0.00 CASING WELD 5733 0.03 1.33 0.41 0.014 0.017 19.33 10.60 2.90 0.04 0.00 CASING WELD 5280 0.03 1.26 0.47 0.013 0.013 19.60 10.36 2.30 0.04 0.00 CASING WELD 7553A 0.05 1.24 0.42 0.016 0.012 18.91 10.52 2.74 0.04 0.00 CASING WELD 5386 0.03 1.19 0.53 0.011 0.023 18.76 10.03 2.60 0.04 0.00 CASING WELD 7242 0.01 0.91 0.52 0.018 0.008 19.38 10.06 2.58 0.04 0.00 iN I co I

 !

)>

-0 Table 4-9 -0 11>

s ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL a.

......

x REACTOR COOLANT PUMP CASINGS l'T1 REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH HONG. AREA FLOW FLOW FLOW PART HEAT NO. (ksi) (ksi) (%) (%) STRESS STRESS STRESS

                                                                                          @ 70F   @ 70F @ 550F
       ** PLANT I. D. ST. LUCIE UNIT 2

• RCP PUMP CASING 741-N-0001 HUB/DIFFUSER A 351 CF8M 91097-1 48.0 92.1 50.0 71.0 70 86665 66071 CASING WELD A 351 CF8M 6074 0.0 0.0 0.0 0.0 0 78600 58000 CASE SCROLL A 351 CF8M 91402-1 43.2 89.0 55.0 74.0 66 80411 59817 CASING WELD A 351 CF8M 7174 0.0 0.0 0.0 0.0 0 . 78600 58000 CASING WELD A 351 CF8M 5952C 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 05929 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 5733 0.0 0.0 0.0 . 0.0 0 78600 58000 CASING WELD A 351 CF8M 9317-051 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 Cf 8M 5280 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 05936 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 5386 0.0 0.0 0.0 0.0 0 78600 5~000 CASING WELD A 351 CF8M 7242 0.0 0.0 0.0 0.0 0 78600 58000
• RCP PUMP CASING 741-N-0002 CASING WELD A 351 CfSM 6074 0.0 0.0 0.0 0.0 0 78600 58000 CASE SCROLL A 351 CF8M 97947-1 44.7 88.2 50.0 71.0 66 80965 60371 HUB/DIFFUSER A 351 CFSM 95211-1 40.0 84.6 58.0 72.0 62 74395 53801 CASING WELD A 351 CF8M 7174 0.0 0.0 0.0 0.0 0 78600 58000 N

CASING WELD A 351 CF8M 05929 0.0 0.0 0.0 0.0 0 78600 58000

 '°     CASING WELD   A 351    CFSM     5733            0.0      0.0       0.0       0.0      0   78600  58000 CASING WELD   A 351    CF8M     6546            0.0      0.0       0.0       0.0      0   78600  58000 CASING WELD   A 351    CF8M     9317-051        0.0      0.0       0.0       0.0      0   78600  58000 CASING WELD   A 351    CF8M     7553A          0.0       0.0       0.0       0.0      0   78600  58000 CASING WELD   A 351    CF8M     7242            0.0      0.0       0.0       0.0      0   78600  58000

l

)>

"'C Tabl~ 4-9 (Continued) "O (I)

s ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

.....

0.. REACTOR COOLANT PUMP CASINGS x l'T1 REPORT #2 - MATERIAL SPECIFICATION &TENSILE PROPERTIES MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED OR SPEC. TYPE STRENGTH STRENGTH HONG. AREA FLOW FLOW FLOW PART HEAT NO. (ks1) (ks1) (%) (%) STRESS STRESS STRESS

                                                                                         @ 70F  @ 70F @ 550F

• RCP PUMP CASING 741-N-0003 CASING WELD A 351 CF8H 6074 0.0 0.0 0.0 0.0 0 78600 58000 HUB/DIFFUSER A 351 CFBM 99346,-1 40.6 84.0 58.0 70.0. 62 74395 53801 CASING WELD A 351 CF8H 7174 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 6546 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351. CF8M 5733 0.0 0.0 0.0 0.0 0 78600 58000 CASE SCROLL A 351 CF8H 99918-1 40.0 84.1 55.0 75.0 62 73999 53405 CASING WELD A 351 CF8M 7553A 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CFSM 7242 0.0 0.0 0.0 0.0 0 78600 58000
• RCP PUMP CASING 741-N-0004 CASING WELD A 351 CfSM 6074 0.0 0.0 0.0 0.0 0 78600 58000 HUB/DIFFUSER A 351 CF8M 99161-1 44.8 83.6 51.0 7.0.0 64 77403 56809 CASE SCROLL A 351 CFBM 00233-1 40.0 80.5 61.0 67.0 60 71149 50555 CASING WELD A 351 Cf BM 6546 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 5733 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CFSH 5280 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CF8M 7553A 0.0 0.0 0.0 0.0 0 78600 58000 CASING WELD A 351 CFSM 5386 0.0 0.0 0.0 0.0 0 78600 58000 w

CASING WELD A 351 CF8M 7242 0.0 0.0 0.0 0.0 0 78600 58000 0

)> 'C Table 4-10 'C CD ~ 0. ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL -'* >< REACTOR COOLANT PUMP CASINGS l'T1 REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR HEAT MEASURED CHROMIUM NICKEL Cre/Nie CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR FOR FOR CONTENT CONTENT FOR FOR WELD METAL (%) CASTINGS CASTINGS CASTINGS #1 (%) #2 (%) WELDS WELDS (%)

        ** PLANT I.D. ST. LUCIE UNIT 2

• RCP PUMP CASING 741-N-0001 91097-1 18.0 18.507 14.191 1.30 20. l 22.2 24 .150 12.500 17.0 6074 17.0 19.243 14.944 1.29 15.2 20.7 24.265 13.485 13.0 91402-1 16.6 17.642 14.296 1.23 13.2 16.3 23.480 12.510 15.0 7174 17.0 18.164 14.615 1.24 14.2 17.0 23.230 12.990 14.0 5952C 10.0 17 .841 15.896 1.12 8.5 9.0 22.820 14.400 12.0 05929 21.0 18.129 13.611 1.33 20. l 24.8 23.320 11. 755 17.0 5733 14.0 18.046 15.021 1.20 11.8 13.9 22.845 13.425 9.0 9317-051 19.0 17. 757 14.160 1.25 15.2 17 .9 22.880 12.420 14.0 5280 14.0 17.619 14.824 1.19 11.4 13.0 22.605 13.210 9.0 05936 20.0 17.758 13.623 1.30 18.4 22.1 23.040 11. 745 16.0 5386 15.0 17 .170 14.292 1.20 11.8 13.9 22.155 12.605 10.0 7242 18.0 17. 761 14.026 1.27 16.7 18.9 22.740 12.165 15.0
• RCP PUMP CASING 741-N-0002 6074 17 .0 19.243 14.944 1.29 15.2 20.7 24.265 13.485 13.0 97947-1 15.0 17.823 14.619 1.22 15.7 15.2 23.595 12.840 13.0 w

95211-1 14.0 17.074 14.467 1.18 12.8 12.4 22.735 12. 720 11.0

  ..... 7174         17 .o   18.164    14.615      1.24       14.2       17 .0  23.230  12.990         14.0 05929        21.0    18.129    13 .611     1.33       20. l      24.8   23.320  11. 755        17.0 5733         14.0    18.046    15.021      1.20       11.8       13.9   22.845  13. 425         9.0 6546         15.0    18.094    14.721      1.23       12.3       15.9   22.970  13 .140        10.0 9317-051     19.0    17.757    14.160      1.25       15.2       17.9   22.880  12.420         14.0 7553A        10.0    17.437    15.374      1.13        8.5        9.6   22.280  13.840          5.0 7242         18.0    17. 761   14.026      1.27       16.7       18.9   22.740  12.165         15.0

j J>

 "'Cl Table .4-10 (Continued)
 "O (1)

s a..

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL

  .......                                       REACTOR COOLANT PUMP CASINGS
  ><

IT1 REPORT #3 - PREDICTED THERMAL AGING BEHAVIOR HEAT MEASURED CHROMIUM NICKEL Cre/Nie CALCULATED CALCULATED CHROMIUM NICKEL CALCULATED NO. FERRITE EQUIV. EQUIV. RATIO FERRITE FERRITE EQUIV. EQUIV. FERRITE CONTENT FOR . FOR FOR CONTENT CONTENT FOR FOR WHO METAL (%) CASTINGS CASTINGS CASTINGS #1 (%) #2 (%) WE LOS WE LOS (%)

• RCP PUMP CASING 741-N-0003 6074 17 .0 19.243 14.944 1.29 15.2 20.7 24.265 13.485 13.0 99346-1 15.0 17.494 14.616 , 1.20 14.2 13.6 23.305 12.905 12.0 7174 17.0 18.164 14.615 1.24 14.2 17 .0 23.230 12.990 14.0 6546 15.0 18.046 14.721 1.23 12.3 15.7 22.930 13.140 10.0 5733 14.0 18.046 15.021 1.20 11.8 13.9 22.845 13.425 9.0 99918-1 15.0 16.904 14.388 1.17 13.7 12.1 22.685 12.495 9.0 7553A 10.0 17 .437 15.374 1.13 8.5 9.6 22.280 13.840 5.0 7242 18.0 17. 761 14.026 1.27 16.7 18.9 22.740 12.165 15.0
• RCP PUMP CASING 741-N-0004 6074 17.0 19.243 14.944 1.29 15.2 20.7 24.265 13.485 13.0 99161-1 15.0 17 .407 14.545 1.20 14.2 13.6 23.245 12.830 18.0 00233-1 13.0 16.994 14.702 1.16 11.8 10.9 22. 775 13.010 13.0 6546 15.0 18.046 14.721 1.23 12.3 15.7 22.930 13 .140 10.0 5733 14.0 18.046 15.021 1.20 11.8 13.9 22.845 13.425 9.0 5280 14.0 17.619 14.824 1.19 11.4 13.0 22.605 13.210 9.0 7553A 10.0 17 .437 15.374 1.13 8.5 9.6 22.280 13.840 5.0 5386 15.0 17 .170 14.292 1.20 11.8 13.9 22.155 12.605 10.0 7242 18.0 17.761 14.026 1.27 16.7 18.9 22.740 12.165 15.0 w

N

)>

"'C Table 4-11 "'C CD

l a.. ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

.....
)(                                        COOLANT PUMP CASINGS AT 70F f'T'1 REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL HEAT            MEASURED     MATERIAL     MATERIAL     MATERIAL      MINIMUM    MINIMUM    MINIMUM MINIMUM NO.              FERRITE        AGING     CONSTANT     CONSTANT       IMPACT         Jlc        KJc T HOD.

CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (%) @ 70F @ 70F (ft-1 bs) . (lb/in) {ksi\/in)

      ** PLANT I. D. ST. LUCIE UNIT 2

• RCP PUMP CASING 741-N-0001 91097-1 18.0 35.45 2828.0 0.37 16.29 504.0 145.9 35.3 6074 17.0 35.53 2823.0 0.37 16.25 506.0 146.2 42.6 91402-1 16.6 24.21 3969.0 0.38 25.94 693.6 171. 2 55.8 7174 17 .0 18.74 5084.0 0.39 36.13 867.9 l,91. 5 72.9 5952C 10.0 16.83 5644.0 0.39 41.46 975.6 203.0 80.2 05929 21.0 16.69 5686.0 0.39 41.87 . 982.8 203.7 80.7 5733 14.0 15.62 6063.0 0.40 45.52 1005.0 206 .1 85.1 9317-051 19.0 15.36 6159.0 0.40 46.46 1023.0 207.9 86.3 5280 14.0 14.94 6324.0 0.40 48.07 1054.0 211.0 88.4 05936 20.0 14.59 6468.0 0.40 49.50 1082.0 213.8 90.2 5386 15.0 12.20 7627.0 0.41 61.08 1242.0 229.0 103.9 7242 18.0 12.69 7362.0 0.40 58.40 1255.0 230.2 101.2
• RCP PUMP CASING 741-N-0002 6074 17 .0 35.53 2823.0 0.37 16.25 506.0 146.2 42.6 97947-1 15.0 26.31 3670.0 0.38 23.32 636.6 164.0 51. 2 95211-1 14.0 19.88 4799.0 0.39 33.47 819.2 186.0 76.8 w

w 7174 17 .o 18.74 5084.0 0.39 36.13 867.9 191. 5 72.9 05929 21.0 16.69 5686.0 0.39 41.87 982.8 203.7 80.7 5733 14.0 15.62 6063.0 0.40 45.52 1005.0 206 .1 85.1 6546 15.0 16.25 5835.0 0.39 43.31 1012.0 206.8 82.6 9317-051 19.0 15.36 6159.0 0.40 46.46 1023.0 207.9 86.3 7553A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 7242 18.0 12.69 7362.0 0.40 58.40 1255.0 230.2 101. 2

)>

"'C Table .4-11 (Continued) "'C 11>

3 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR

....><

0.. COOLANT PUMP CASINGS AT 70F l'T'I REPORT #4 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 70F @ 70F @ 70F (%) @ 70F @ 70F (ft-lbs) (lb/in) (ksi\/in)

• RCP PUMP CASING 741-N-0003 6074 17 .0 35.53 2823.0 0.37 16.25 506.0 146.2 42.6 99346-1 15.0 23.20 4134.0 0.38 27.40 730.5 175.7 67.3 7174 17 .0 18.74 5084.0 0.39 36.13 867.9 191.5 72.9 6546 15.0 16.00 5924.0 0.40 44 .17 978.9 203.3 83.3 5733 14.0 15.62 6063.0 0.40 45.52 1005.0 206.l 85.l 99918-1 15.0 15.54 6091.0 0.40 45.80 1019.0 207.5 95.8 7553A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 7242 18.0 12.69 7362.0 0.40 58.40 1255.0 230.2 101. 2
• RCP PUMP CASING 741-N-0004 6074 17 .o 35.53 2823.0 0.37 16.25 506.0 146.2 42.6 99161-1 15.0 22.97 4174.0 0.38 27.76 734.9 176.2 63.0 00233-1 13.0 19.96 4782.0 0.39 33.31 819.3 186.0 83.4 6546 15.0 16.00 5924.0 . 0.40 44.17 978.9 203.3 83.3 5733 14.0 15.62 6063.0 0.40 45.52 1005.0 206.1 85.l 5280 14.0 14.94 6324.0 0.40 48.07 1054.0 211.0 88.4 7553A 10.0 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 5386 15.0 12.20 7627.0 0.41 61.08 1242.0 229.0 103.9 7242 18.0 12.69 7362.0 0.40 58.40 1255.0 230.2 101. 2 w

 ~

)>

"C Table 4-12 "C rt>

3 ANALYSIS OF THERMAL AGING OF STAINLESS STEEL

....x Q.

REACTOR COOLANT PUMP CASINGS AT 550F fT1 REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTE.NT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @ 550F @ 550F (ft-lbs) (lb/in) (ksi/in)

      ** PLANT  I. D. ST. LUCIE UNIT 2

• RCP PUMP CASING 741-N-0001 91097-1 18.0 35.45 2455.0 0.27 16.29 721.1 142.1 47.8 6074 17.0 35.53 2452.0 0.27 16.25 727 .4 142.8 61.2 91402-1 16.6 24.21 3167.0 0.28 25.94 913.6 160.0 72.9 7174 17.0 18.74 3812.0 0.29 36.13 1072. 0 173.3 91.3 5952C 10.0 16.83 4122.0 0.29 41.46 1170 .0 181. l 97.6 05929 21.0 16.69 4145.0 0.29 41.87 1178. 0 181. 7 98. l 5733 14.0 15.62 4348.0 0.29 45.52 1244.0 186.7 102.2 9317-051 19.0 15.36 4399.0 0.29 46.46 1260.0 187.9 103.2 5280 14.0 14.94 4486.0 0.29 48.07 1289.0 190.0 104.9 05936 20.0 14.59 4563.0 0.30 49.50 1250.0 187 .2 107.l 5386 15.0 12.20 5157.0 0.30 61.08 1439.0 200.8 118. 5 7242 18.0 12.69 5023.0 0.30 58.40 1396.0 197.8 116.0
• RCP PUMP CASING 741-N-0002 6074 17.0 35.53 2452.0 0.27 16.25 727.4 142.8 61. 2 97947-1 15.0 26.31 2986.0 0.28 23.32 856.3 154.9 68. l 95211-1 14.0 19.88 3651.0 0.29 33.47 1030.0 169.9 101.2 w 36.13 173 .3 91.3 Ul 7174 17.0 18.74 3812.0 0.29 1072 .0 05929 21.0 16.69 4145.0 0.29 41.87 1178.0 181. 7 98. l 5733 14.0 15.62 4348.0 0.29 45.52 1244.0 186.7 102.2 6546 15.0 16.25 4225.0 0.29 iiJ.31 1203.0 183.6 99.7 9317-051 19.0 15.36 4399.0 0.29 46.46 1260.0 187 .9 103.2 7553A 10.0 14.36 4613.0 0.30 50.44 1266.0 188.4 108.0 7242 18.0 12.69 5023.0 0.30 58.40 1396.0 197.8 116.0

)>                                                 Table ~-12 (Continued)

"'O "'O CD

s ANALYSIS OF THERMAL AGING OF STAINLESS STEEL

....><

c.. REACTOR COOLANT PUMP CASINGS AT SSOF fT'I REPORT #5 - PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL HEAT MEASURED MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM NO. FERRITE AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. CONTENT PARAMETER c N ENERGY @ 550F @ 550F @ 550F (%) @ 550F @ 550F (ft-1 bs) (lb/1n) {ks;/in)

• RCP PUMP CASING 741-N-0003 6074 17 .0 35.53 2452.0 0.27 16.25 727 .4 142.8 61.2 99346-1 15.0 23.20 3265.0 0.28 27.40 955.1 163.6 91.4 7174 17 .o 18.74 3812.0 0.29 36.13 1072.0 173.3 91.3 6546 15.0 16.00 4273.0 0.29 44.17 1219.0 184.9 100.7 5733 14.0 15.62 4348.0 0.29 45.52 1244.0 186.7 102.2 99918-1 15.0 15.54 4363.0 0.29 45.80 1263.0 188.1 119.2 7553A 10.0 14.36 4613.0 0.30 50.44 1266.0 188.4 108.0 7242 18.0 12.69 5023.0 0.30 58.40 1396.0 197.8 116.0
• RCP PUMP CASING 741-N-0004 6074 17 .0 35.53 2452.0 0.27 16.25 727 .4 142.8 61.2 99161-1 15.0 22.97 3289.0 0.28 27.76 957.7 163.8 83.l 00233-1 13.0 19.96 3641.0 0.29 33.31 1036.0 170.4 113.3

)>

"t:I Table 4 st. Lucie Unit 2 "t:I n>

3 c.

.....
)(                          Limiting and Controlling Values of
,.,.,

Jic and KJc at 70°F RCP PUMP CASING HEAT # Jic KJc (lb/in) (ksi/in) 741-N-0001 91097-1 504.0 145.9 741-N-0002 6074 506.0 146.2 741-N-0003 6074 506.0 146.2 741-N-0004 6074 506.0 146.2 w

......

)>                          Table 4-14.- St. Lucie Unit 2 "C

"C rD

I 0..

-'*
)(

Limiting and Controlling Values of Jlc and KJc at 550°F RCP PUMP CASING HEAT # Jic KJc (lb/in) (ksi/in) 701-N-0001 91097-1 721.1 142.1 701-N-0002 6074 727.4 142.8 701-N-0003 6074 727.4 142.8 701-N-0004 6074 727.4 142.8 w CD

5.0 CRACK GROWTH ANALYSIS In this section, the methodologies discussed in Section 5.1 of the main report are applied to st. Lucie 1&2 plant-specific conditions. The growth history of a worst-case hypothetical crack is conservatively developed based on information contained in the vendor's stress analysis report. 5.1 Scope The analysis which follows pertains to the 35x35x43 DFSS Reactor Coolant Pump casings, Serial Numbers 681-N-0445 to 681-N-0448 and 741-N-0001 to 741-N-0004, inclusive~ at the St. Lucie Units 1 & 2, respectively. 5.2 Reference Stress Reports The stresses used in the hypothetical crack qrowth analyses are from the stress reports prepared by the Byron-Jackson Company in August, 1974 (Reference 3-3) for Unit 1, and in March, 1977 (Reference 3-4) for Unit 2. 5.3 Selection of High Stress Locations The methodology described in 5.1.1 of the main report was applied to identify five regions as potentially limiting: (1) Diffuser Vane 8--Level D (2) Discharge Nozzle--Section c, adjacent to Crotch Region (3) Suction Nozzle--Level c (4) Junction, Volute with Lower Flange (5) Hanger Bracket #1 Vicinity. All other regions in the stress summary were considered and were found to have lower stresses than the above regions. Appendix E 39

Stress results for Unit 2 are virtually identical to those ~ published for Unit 1. 5.4 Stresses and Wall Thicknesses at Limiting Locations Membrane and through-wall bending components of the limiting regions were obtained from Reference 3-3 and 3-4 under the Design Condition basis, as follows: (1) For Diffuser Vane 8--Level D: Design Condition = # 103, plus thermal gradient stresses Key Elements = # 3828 & 3832 in Finite-Element Model Membrane stress = 20.95 Ksi (Figure 4-3.(b)) Bending stress = 19.87 Ksi (Figures 3-1 & 4-3 (b)) Thickness = 4.75 11 (Figure 4-3(b)) Bending stress includes 3.4 Ksi due to a surface-to~interior temperature difference of 18°F during heatup/cooldown. (2) For Discharge Nozzle--Section C, adjacent to Crotch Region: Design Condition = # 107, plus thermal gradient stresses Key Element = # 5125 in Finite-Element Model Membrane stress = 17.4 Ksi (Table 4-6) Bending stress = 21.1 Ksi (Figure 4-8, p. 66 & Figure 3-1) Thickness = 3.3 in. (Table 4-6) Bending stress includes 5.5 Ksi due to a through-wall temperature difference of 29°F during heatup/cooldown. Appendix E 40

                                ------ ----------- - - * -

(3) For Suction Nozzle--Level C: Design Condition = # 104, plus thermal gradient stresses Key Elements = # 2125 & 2129 in Finite-Element Model Membrane stress = 22.15 Ksi (Figure 4-5(b)) Bending stress = 13.15 Ksi (Figures 3-1 & 4-5(b)) Thickness = 3.0 in. (Figure 4-5(b)) Bending stress includes 5.5 Ksi due to a through-wall temperature difference of 29°F during heatup/cooldown. (4) For Junction, Volute with Lower Flange (vicinity of Vanes 1 & 2): Design Condition = #112, plus thermal gradient stresses Key Elements = # 1279, 1283, & 1291 in Finite-Element Model Membrane stress = 17.82 Ksi (page 78) Bending stress_. = 13.2 Ksi (page 78 & Figure 3-3) Thickness = 3.375 in. Bending stress includes 3.0 Ksi due to a through-wall temperature difference of 16°F during heatup/cooldown (Hydraulic Section 3). Bending stress also includes 10.2 Ksi under Design Condition 112, conservatively derived from the Report declaration that surface stresses do not exceed 28.05 Ksi (1.5 Sm). (5) For Hanger Bracket #1 Vicinity: Design condition = Maximum (pp. 74, 75) Key Elements = # 7461 in Finite-Element Model Membrane stress = 18.9 Ksi (page 75) Appendix E 41

Bending stress = 22.1 Ksi (page 75, 97) Thickness = 3.6 in. Bending stress includes 13.7 Ksi due to through-wall temperature difference of 72.7°F for Hydraulic Section 8 during the heatup/cooldown transients. 5.5 Calculation of Crack Growth Rates The methodology described in Section 5.1.4 of the generic report was applied to the above values, using the annual rate of stress-cycling given in References 3-3 and 3-4. An integration procedure was used to predict dimensionless crack depth, a/t, as a function of time, T. Results are swnmarized in Tables 5-1 through 5-5. For each region the calculated entries are listed against crack depth, a/t, as follows: (1) Applied Stress Intensity Factor, KI' was calculated using the ASME Section XI procedure, as further described in Section 5.1.4 of the generic report. Units for KI are Ksi--squareroot inch. (2) Crack growth rate, da/dT, was calculated using the Bernard & Slama equation (with R=O), multiplied by the design value of the annual rate of stress cycling, as further described in Section 5.1.4. The final equation is: da/dT = 5.435 x 10-lO K 4 .0 I Units for da/dT are inches/year. (3) Incremental time, dT, in which the crack will grow through the indicated dimensionless crack depth Appendix E 42

Table 5-1 St. Lucie 1 & 2 Crack Growth Rates At Vane Number 8 Level D a/t Interval KI da/dT t>.Time (fraction} (KSI mo (IN/YEAR} (YEARS} 0.08 -- 0.10 49.98 3.39 x 10- 3 28.0 0 .10 -- 0.15 60.08 7.08 x 10- 3 45.6* 0.15 -- 0.20 68.76 1.21 x 10- 2 24.4* 0.20 -- 0.25 76.76 1.89 x 10- 2 12.5 0.25 -- 0.30 84.17 2.73 x 10- 2 8.7 0.30 -- 0.35 91.69 3.84 x 10- 2 6.2 0.35 -- 0.40 99.35 5.29 x 10- 2 4.5 0.40 -- 0.45 108.3 7.48 x 10- 2 3.2 O.&i5 -- 0.50 118.4 0.107 2.2 {am

• 20.95, ab* 19.87, t
• 4.75")
• Sum of five time steps through 1% a/t increments using interpolated KI values.

Appendix E 43

Table 5-2 St. Lucie 1 & 2 crack Growth Rates At Discharge Nozzle Crotch Vicinity - Section C a/t Interval KI da/dT ~Time (fraction} (KSI /IN) (IN/YEAR} (YEARS} 0.08 -- 0.10 38.9 1.25 x 10-3 52.8 0.10 -- 0.15 46.6 2.56 x 10-3 64.5 0.15 -- 0.20 53.0 4.30 x 10-3 38.4 0.20 -- 0.25 58.9 6.55 x 10- 3 25.2 0.25 -- 0.30 64.3 9.28 x 10-3 17.7 0.30 -- 0.35 69.7 1.28 x 10- 2 12.8 0.35 -- 0.40 75.2 1. 74 x 10- 2 9.5 0.40 -- 0.45 81.6 2.41 x 10- 2 6.8 e 0.45 -- 0.50 88.9 3.39 x 10- 2 4.8

                           ..

(am* 17.4, ab* 21.1, t

• 3.3")

Appendix E 44

Table 5-3 St. Lucie 1 & 2 Crack Growth Rates At Suction Nozzle - Level C a/t Interval KI da/dT t.T {fraction} {KSI Im} {IN/YEAR} {YEARS} 0.08 -- 0.10 34.92 8.08 x 10-4 74.2 0 .10 -- 0 .15 42.37 1. 75 x 10- 3 85.6 0.15 -- 0.20 48.93 3.12 x 10- 3 48. l 0.20 -- 0.25 55.08 5.00 x 10- 3 30.0 0.25 -- 0.30 60.93 7.49 x 10-3 20.0 0.30 -- 0.35 66.88 1.09 x 10- 2 13.8 0.35 -- 0.40 73.05 1.55 x 10- 2 9.7 0.40 -- 0.45 80.20 2.25 x 10- 2 6.7 0.45 -- 0.50 88.25 3.30 x 10- 2 4.5 Appendix E 45

Table 5-4 st. Lucie 1 & 2 Crack Growth Rates At Volute Junction with Lower Flange Near Vanes 1 & 2 From - To KI da/dT ~T {aLt

• l {KSI mo {INLYEAR} {YEARS}

0.08 -- 0.10 32.27 5.9 x 10- 4 114. 0 .10 -- 0 .15 38.99 1.26 x 10-3 134. 0.15 -- 0.20 44.85 2.2 x 10- 3 77. 0.20 -- 0.25 50.29 3.48 x 10- 3 48.5 0.25 -- 0.30 55.41 5.12 x 10-3 33. 0.30 -- 0.35 60.61 7.33 x 10-3 23. 1.03 x 10- 2 0.35 0.40

     --
     --

0.40 0.45 65.97 72.20 1.48 x 10- 2 16.4 11.4 e 0.45 -- 0.50 79.21 2.14 x 10- 2 7.9 (um* 17.8, ub

• 13.2, t
• 3.375)

Appendix E 46

Table 5-5 St. Lucie 1 & 2 Crack Growth Rates At Hanger Bracket #1 Vicinity From - To KI da/dT l1T (a Lt ,.. l (KSI mn (INLYEAR) (YEARS}" 0.08 -- 0.10 43.40 1.92 x 10- 3 37.4 0.10 -- 0.15 51.90 3.94 x 10- 3 61.3* 0.15 -- 0.20 59.10 6.65 x 10- 3 33.5* 0.20 -- 0.25 65.80 0.0102 17.6 0.25 -- 0.30 71.80 0.0145 12.4 0.30 -- 0.35 77.90 0.02 9.0 0.35 -- 0.40 84.10 0.0272 6.6 0.40 -- 0.45 91.40 0.0379 4.7 0.45 -- 0.50 99.50 0.0533 3.4 (um

• 18.9, ub m 22.1, t
• 3.6")
• The sum of five time steps through 0.01 a/t increments using interpolated KI values.

Appendix E 47

values, a/t, was calculated as described in Section ~ 5.1.4 of the generic report. Units for dT are years. The summation of time increments yields the total Time for a crack to grow to a given a/t value. The predicted growth curves for hypothetical cracks (Figure 5.1-5 of the generic report) show the functional relationship between a/t and total Time, using the initial condition of a/t = 0.08 at Time = o. The first incremental time listed in Tables 5-1 through 5-5 is based on a change in a/t in the amount 0.02 (i.e. 2% of thickness), to indicate the time needed for the crack to grow from a/t = 0.08 to a/t = 0.10. Subsequent incremental times are based on the time needed for the crack to grow through changes in a/t of 0.05 (i.e. 5% of thickness). The first such incremental time is given for the range a/t = 0.10 to 0.15. The final incremental time is given for the range a/t = 0.45 to a.so, wh~reupon the analysis is terminated. 5.6 Stresses Under Emergency and Faulted Conditions In order to verify that limiting sections containing hypothetical cracks can withstand Emergency Condition and Faulted Condition Loads, the methodology described in Section 5.1.4 of the generic report was again applied. Applied stress intensity factors were calculated at the limiting locations, based on available data from References 3-3 and 3-4. 5.6.1 Emergency Condition Stresses (1) Diffuser Vane 8--Level D: Key elements = # 3828 & 3832 Condition = # 505 Appendix E 48

Membrane stress = 15.5 Ksi (p. 109) Bending stress = 13.2 Ksi (p. 109) These are conservative values, bounded by Vane 9--Level A results, with secondary stresses removed. (2) Discharge Nozzle--Section c, adjacent to Crotch Region: Key element = # 5125 Condition = # 506 Membrane stress = 26.66 Ksi (p. 118) Bending stress = 7.3 Ksi (p. 124) Secondary stresses are not included. (3) Suction Nozzle--Level C: Key elements = # 2125 & 2129 Condition = # 511 Membrane $tress = 25.5 Ksi (p. 112) Bending stress = 11.88 Ksi CP~ 114) Bending stress is conservatively bounded by Condition 503 results, with secondary stresses removed. (4) Junction, Volute with Lower Flange: Key elements = # 1279, 1283 & 1291 No emergency condition results are published specifically for this region.

• Appendix E 49

(5) Hanger Bracket Vicinity: Key elements = I 7461 Condition = emergency, worst case Membrane stress = 23.2 Ksi (p. 106) Bending stress = 11. 6 Ksi (p. 106) Stresses are conservatively set to Emergency Conditions allowables for a worst case analysis. 5.6.2 Faulted Condition Stresses (1) Diffuser Vane 8--Level D Membrane stress = 23.22 Ksi (p. 143) Bending stress = 23.55 Ksi {p. 143) These are conservative values, bounded by Vane 9--Level A results from elements 3904 and 3905 under condition 606. (2) Discharge.Nozzle, adjacent to Crotch Region: Membrane stress = 32.0 Ksi (p. 141), upper bound Bending stress = 6.4 Ksi (p. 148) These are worst case results under Faulted Condition 606, with secondary stresses removed. (3) Suction Nozzle--Level c: Membrane stress = 25.96 Ksi (p. 145) Bending stress = 16.7 Ksi (p. 145) __ App_endix E""-=== _________________ -----*- _________________so____ _

These are worst case results under Faulted Condition 603, with secondary stresses removed. (4) Junction, Volute with Lower Flange: No Faulted Condition results are published specifically for this region. (5) Hanger Bracket Vicinity: Membrane stress = 29.0 Ksi (pp. 137, 139) Bending stress = 14.5 Ksi (p. 139) These are conservative values based on meeting Faulted Conditions allowables. 5.7 Results Results reported above and shown in Figure 5.3-15 of the generic portion of this report for the St. Lucie 1&2 RCPs indicate that the postulated S%t initial crack will grow to 25%t in about 110 y~ars under the influence of the conservatively defined stress cycles in t~e design specification. The hypothesized crack will then grow larger until it r~aches an end-point crack size of 38%t, limited by flow stress, in about 130 years. Appendix E 51

6.0 INSPECTION INTERVAL Results reported in this appendix support the position that the 10-year inspection interval for volumetric examination, as required by ASME Section XI, is not necessary to ensure safe operation during the 40-year licensed life of the plant. The conservatively calculated end-point crack size is not reached until 130 years after initial operation. The demonstrated casing integrity also supports a relaxation of the 10-year interval for visual examinations, as currently required by ASME Section XI and Code case N-481. Such examinations add unnecessarily to personnel exposure with no benefit to plant safety. The ASME Section XI requirement for VT-3 examination of internal surfaces is an appropriate low-marginal-exposure monitoring activity to the extent practicable, but only when the pump is disassembled for maintenance or repair. Based upon the results obtained in this evaluation, relaxation of the casing inspection interval for the St. Lucie RCPs from 10 years to 40 years is considered to be justified. Appendix E 52

APPENDIX F COMPUTER CODE LISTINGS Appendix F. 1

APPENDIX E REFERENCES 3-1 Engineering Specification for Reactor Coolant Pumps for Florida Power and Light Co., Hutchinson Island Plant Unit 1, 19367~31-3 Rev. 4, 5/20/71. 3-2 Project Engineering Specification for Reactor Coolant Pumps for St. Lucie Plant, Unit 2, 13172-PE-480, Rev. 05 (10/12/83). 3-3 Pump case Structural Analysis for Florida Power and Light Co. (CE Contract 19367) (i.e., St. Lucie 1) TCF-1017-STR, Vol. 1, Rev. l dated August 7, 1974. 3-4 Pump case Structural Analysis for Florida Power and Light, St. Lucie Plant, Unit 2, TCF-1024-STR, Vol. 1, Rev. 1, dated March 22, 1977. 3-5 Letter, G. B. Crowley (FP&L) to P. W. Richardson, ABB C-E Nuclear Power, dated 4/2/92. 4-1 o. K. Chopra, "Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems", NUREG/CR-4513 (ANL-90/42), U.S. Nuclear Regulatory Commission 1 Washington, D.C., June 1991. Note: The original name of the st. Lucie Plant was "Hutchinson Island". Some of the above documentation was completed prior to the name change.

                                                                  --

Appendix E 53

Listing F-4 is a dBase program file called TEARMOD.PRG. This program calculates an average tearing modules from the power law equation of the J-R Curve. A linear equation is fit for the power law within the range of acceptable data as defined by ASTM E 813. The slope of this equation represents the average tearing modulus of the material within this range. Appendix F 3

DESCRIPTION OF LISTINGS The following listings provide the details of the dBase III Plus database on the RCP casing materials and the dBase programs used to calculate the various material properties needed for the thermal embrittlement evaluation. Listing F-1 presents the file structure of the database with a description of the contents of each field within the file. Some of the data was entered directly from information obtained from the Certified Material Test Reports (CMTR's), the remainder of the material properties were calculated from the available information using several dBase programs. Listing F-2 is the dBase program file called CASEINFO.PRG. This program calculates the ferrite content by Hull's factors. The program continues to calculate material parameter and minimum Charpy impact energy according to the equations developed by tit Chopra. These values are then used to calculate the coefficients and exponents for the power law J-R Curve equations at room* temperature and 550°F. All of the calculated values are then written into the database for use in further calculations and for presentation in report form outputs. Listing F-3 is ~ dBase program file called FRACTOUG.PRG. This program calculates the Jic and KJc values for the materials in accordance with the standard method of ASTM E 813. Values of flow stress used in the Jic determination are based on the room temperature tensile properties and a derived relationship for flow stress of aged material. When no tensile properties are available, a flow stress value is assumed based on the average of all the calculated flow stresses plus one standard deviation. The Jic' KJc and flow stress values at room temperature and 550°F are then stored in the database. Appendix F 2

DATABASE FILE STRUCTURE Field Name Content in Database 29 MIN T25 Minimum Room Temperature Average Tearing Modules 30 MIN JIC290 Minimum 550°F JIC 31 MIN KJC290 Minimum 550°F JJC 32 MIN T290 Minimum 550°F Average Tearing Modules 33 MAT PAR Material Parameter from Chopra 34 c 25 J-R Curve coefficient .at Room Temperature 35 c 290 J-R Curve Coefficient at 550°F 36 N 25 J-R Curve Exponent at Room Temperature 37 N 290 J-R Curve Exponent at 550°F 38 FERR CALC2 Calculated Ferrite Content from Hull's Factors 39 CRM_EQUIV Chromium Equivalent from Hull's Facto~s 40 NIC_EQUIV Nickel Equivalent from Hull's Factors 41 CRNI_EQUIV Rate of Chromium + Nickel Equivalents from

                                                             "'*

Hull's Factors 42 AFS 25 Aged Flow Stress at Room Temperature 43 AFS 290 Aged F_low Stress at 550°F 44 UAFS_25 Unaged Flow Stress at Room Temperature LISTING F-1 (Continued) Appendix F 5

DATABASE FILE STRUCTURE Field Name Content in Database 1 MATERIAL Casing Material Case/Scroll, Hub or Weld 2 COMP ID Pump Casing Serial Number 3 PLANT RCP Plant or Utility Location 4 MAT SPEC Material Specification 5 MAT TYPE Material Type 6 HEAT NO Heat Number of Material 7 c Carbon Content 8 MN Manganese.content 9 SI Silicon Content* 10 s Sulfur Content 11 p Phosphorus Content 12 CR Chromium Content 13 NI Nickel Content 14 MO Molybdenum Content 15 CB Columbium/Niobium Content 16 N Nitrogen Content 17 FERR_MEA$ Measured Ferrite Content 18 FERR_CALCl Calculated Ferrite Content from Schaefer Equation 19 FERR SPAC Ferrite Spacing 20 FERR_WELD Calculated Ferrite Content of Weld Metal from Schoef ler Diagram 21 YIELDSTR Yield Strength 22 TENSILSTR Tensile Strength 23 ELONGATION Elongation 24 REDINAREA Radiation in Area 25 HARDHRB Hardness Rockwell B 26 MIN_CVNE Minimum Charpy Impact Energy 27 MIN_JIC25 Minimum Room Temperature JIC 28 MIN_KJC25 Minimum Room Temperature JJC LISTING F-1 Appendix F 4

• - CLEAR SET STATUS OFF COMPUTER dBASE PROGRAM CASEINFO.PRG SET TALK OFF USE USE CAST SS GOTO TOP-DO WHILE .NOT. EOF()

RECNUH

• RECNO()

912,20 SAY *CALCULATING FOR RECORD NO.:

• 912,50 SAY RECNUM CRMEQUIV
• CR+(l.2l*M0)+(0.48*SI)-4.99 NICEQUIV
• Nl+(.ll*MN)-(0.0086*MN-2)+(18.4*N)+(24.S*C)+2.77 CRMNICEQUIV
• CRMEQUIV/NICEQUIV FERRTCALC2
• 1 .

100.3*(CRMEQUIV/NICEQUIV)-2-170.72*(CRMEQUIV/NICEQUIY)+74.22 FERR

• FERR MEAS IF FERR
• o:

FERR

• FERRTCALC2 ENDIF HATLPAR * (FERR*CR*(C+0.4*N)*((NI+Sl)-2))/l00 MINCVNEG * (10-(l.15+1.374*EXP(-0.0467*MATLPAR)))

I 10,20 SAY MATLPAR 1*12,20 SAY MINCVNEG NCON25

• 0.35+0.0025*(MINCVNEG)-.67 CCON25
• 91.36*(25.4-NCON25)*(MINCVNEG-.67)

NCON290

• 0.24+0.0063*(MINCVNEG)-.49 CCON290
• 200*(25.~-NCON290)*(MINCVNEG-.49)

UNAGFS25 * (YIELDSTR*+ TENSILSTR)/2 REPLACE CRM EQUIV WITH CRMEQUIV REPLACE NIC-EQUIV WITH ~tCEQUIV REPLACE CRNI EQUIV WITH CRMNICEQUIV REPLACE FERR-CALC2 WITH FERRTCALC2 REPLACE MAT PAR WITH MATLPAR MIN. CVNE

• MINCVNEG*.588 REPLACE HIN CVNE WITH HINCVNEG REPLACE C 25 WITH CCON25 REPLACE N-25 WITH NCON25 REPLACE C-290 WITH CCON290 REPLACE N-290 WITH NCON290 REPLACE UAFS 25 WITH UNAGFS25 SKIP 1 -

ENDDO USE CLOSE DATABASES CLOSE ALL SET STATUS ON LISTING F-2 Appendix F 6

COMPUTER dBASE PROGRAM FRACTOUG.PRG

 *************~*******************

• DBASE III PLUS *
• PROGRAM TO CALCULATE JIC AND KJC *.
• OF A GIVEN MATERIAL, USING THE *
• BEST FIT TREND FOR FLOW STRESS OF *
• AGED CAST STAINLESS STEEL AT 25C *
• AND AT 290C *

 *************************************

CLEAR USE . SET STATUS OFF SET TALK OFF USE CAST SS GOTO TOP-00 WHILE .NOT. EOF() RtCNUM

• RECNO ()

 @ 8,20 SAY *CALCULATING FOR RECORD NO.: *
 @ 8,40 SAY RECNUM                                .

FS25 * (l.58~3*((YIELpSTR+TENSILSTR)/2)-24.245)*1000 IF FS25 *< 0 FS25

• 78600 ENDIF DIFF*lOO A25*0.008 DO WHILE ABS(DIFF) > 10 *

    @10,20 SAY *vALUE OF A25 * *
    @10,30 SAY A25 A25*A2S+O .*oooos DIFF*(C_25*(A25-N_25))-(4~FS25*(A25-0.008))
   @11,30 SAY DIFF ENDDO JIC25

• C_25*(A25-N_25)

 @ 12,20 SAY *JIC25 AT 25C * *
 @ 12;40 SAY JIC25-KJC25 * {{{JIC25*28300000)/.67)-.5)/1000
 @ 13,20 SAY *KJC25 AT 2SC * *
 @ 13,45 SAY KJC25 REPLACE MIN JIC25 WITH JIC25 REPLACE MIN-KJC25 WITH KJC25 REPLACE AFS-25 WITH FS25 FS290 * (l.S833*((YIELDSTR+TENSILSTR)/2)-44.839)*1000 IF FS290 < O FS290

• 58000 END IF DIFF*lOO A290*0.008 LISTING F-3 Appendix F 7

DO WHILE ABS(D.IFF) > 10

 @ 14,20 SAY *VALUE OF A290 * *
 @ 14,30 SAY A290 A290*A290+0.0000S OIFF*(C_290*(A290-N_290))-(4*FS290*(A290-0.008))
  @ 16,30 SAY DIFF              .

ENDOO JIC290

• C 290*(A290*N_290)

@ 18,20 SAY *JIC290 AT 290C * * @ 18,40 SAY JIC290 . KJC290 * (((JIC290*25500000)/.91)-.S)/1000 @ 20,20 SAY *KJC290 AT 290C * * @ 20,40 SAY KJC290 REPLACE MIN JIC290 WITH JIC290 REPLACE MIN-KJC290 WITH KJC290 REPLACE AFS-290 WITH FS290 SKIP 1 - CLEAR ENDDO DO TEARMOD USE CLOSE CLOSE ALL END LISTING F-3 (Continued) 8 Appendix F

COMPUTER dBASE PROGRAM TEARMOD.PRG

 *************************************************

• DBASE III PLUS *
• PROGRAM TO CALCULATE AVERAGE TEARING MODULUS *
• FROM POWER LAW EQUATION FOR J-R CURVE *
• USING HEA PROCEDURE FROM NUREG/CR-3506 APP. H *

 *************************************************

CLEAR USE SET STATUS OFF SET TALK OFF USE CAST SS

• GOTO TOP-00 WHILE .NOT. EOF()

RECNUM

• RECNO()

  @ 8,20 SAY *CALCULATING FOR RECORp NO.: *
  @ 8,40 SAY RECNUH
    .E25

• 28300000 FS25 * ((l.5833~((YIELOSTR+TENSILSTR)/2))-24.245)*1000 IF FS25 < 0 FS25
• 78600 ENO IF DIFF*IOO .

A25*0.006 DO WHILE ABS(OIFF) > 10 I 10,20 SAY *vALUE OF A25 * *

        @ 10,30 SAY A25 A25*A25+0.00005
      *oIFF*(C_25*(A2S*N_25))-(4*FS2S*(A25-0.006))
        @ 11,30 SAY DIFF      .

ENO DO DIFF*lOO 825*.06 DO WHILE ABS(DIFF) > 10

         @ 12,20 SAY *vALUE OF 825 * *
         @ 12,30 SAY 825 825*825+0.0000S DIFF*(C_25*(825*N_25))-(4*FS25*(825-.06))
         @ 13,30 SAY DIFF ENDDO LISTING F-4 Appendix F                                                  9
                                                             ---------- - - - -

1125 -= B25-A25 1225 * {B25-2-A25-2}/2 1325 * (C_25/(N_25+1))*((825-(N_25+1))-(A25:(N_25+1))) 1425 * (C 25/(N 25+2))*((825-(N 25+2))-(A25 (N 25+2))) 1525 * (B25-3-A25-3)/3 - .- S25 * ((I425*Il25}-(l225*1325))/((1525*1125)-(1225-2)) TAVG25 * (E25/FS25-2)*S25

 @ 14,20 SAY *AVERAGE TEARING MODULUS AT 25C - *
 @ 14,40.SAY TAVG25 REPLACE MIN T25 WITH TAVG25 REPLACE AFS-25 WITH FS25 E290

• 25600000 FS290 * ((l.5833*((YIELDSTR+TENSILSTR)/2))-44.839)*1000 IF FS290 < 0
• FS290
• 58000 END IF DIFF*lOO A290*0.006 DO WHILE ABS(OIFF) > 10

   @ 16i20 SAY *vALUE OF A290 * *
   @ 16,30 SAY A290 A290*A290+o.oooos OIFF*(C_290*(A290*N.:._290))-(4*FS290*{A290-0.006))
   @ 17,30 SAY DIFF ENODO OIFf*100 8290*.06 DO WHILE ABS(DIFF) > 10
   @ 18,20 SAY *vALUE OF 8290 * *
   @ 18,30 SAY 8290        .

8290=8290+0.00005 I .DIFF*(C_290*(8290-N_290))-(4*FS290*(8290-0.06))

   @ 19,30 SAY DIFF ENO DO Il290

• B290-A290 -

I2290 * (8290-2-A290-2)/2 I3290 * (C_290/(N_290+1))*((8290-(N_290+1))-(A290-(N_290+1))) 14290 * (C_290/(N_290+2))*((8290-(N_290+2))-(A290-(N 290+2))) 15290 * (B290-3-A290-3)/3 - 5290 * ((I4290*11290)-(I2290*I3290))/((I5290*Il290)-(I2290-2)) TAVG290 * (E290/FS290-2)*S290 .

@ 20,20 SAY *AVERAGE TEARING MODULUS AT 290C -= *
@ 20,45 SAY TAVG290 REPLACE MIN T290 WITH TAVG290 REPLACE AFS-290 WITH FS290 SKIP 1        -

END DO USE CLOSE CLOSE All ENO LISTING F-4 (Continued) Appendix F 10

33

• SECTION 5 RELIEF REQUESTS

*

• I

  )

• CONSUMERS POWER COMPANY PALISADES NUCLEAR POWER PLANT THIRD; INTERVAL INSERVICE INSPECTION PROGRAM RELi EF REQUESTS 34 RELIEF SECTION REQUEST XI ITEM COMPONENT RELIEF ALTERN.

NUMBER CATEGORY . NUMBER DESCRIPT. REQUESTED EXAM RR-1 B-J B9.ll PCS INACCESS. 100% UT B9.12 PIPING FOR O.D. FROM I.D. SURFACE .& EXAM VOL. EXAM RR-2 B-A Bl.40 RV HEAD-TO INACCESS. ACCESS FLANGE FOR 100% VOLUNE &* WELD VOL. EXAM SURFACE EXAM RR-3 C-A Cl.10 SG SHELL- INACCESS. ACCESS TO-CONE FOR 100% VOLUME

• WELDS VOLUME EXAMINE

                    *-         -                        EXAM

• RR-4 RR-5 B-D s.:.D 83 .130 B3 .150

                                          *sG NOZZLE-TO-SHELL WELDS REGEN. HX WELDS INACCESS.

FOR 100% VOLUME EXAM INACCESS. FOR 100% ACCESS. VOLUME EXAMINE . ACCESS. VOLUME VOLUME EXAMINE EXAM RR-6 8-8 82.11 PZ. HEAD, INACCESS. ACCESS. 8-D 82.21 SHELL*& FOR 100% VOLUME 82.22 NOZZLE VOLUME EXAMINE 83 .110 WELDS EXAM WITH ALL MERID. WELDS

  . RR-7         C-A          Cl.10         SOC HX     . INACCESS. ACCESS.

C-8 Cl.30 SHELL, FOR 100% VOLUME AND C2.21 FLG. & TS VOLUME ALL WELDS EXAM: SURFACE EXAMINE RR-8 B-A Bl. 21 RV CLOSURE INACCESS. ACCESS HEAD WELD FOR 100% VOLUME VOLUME EXAMINE EXAM

• RELIEF REQUEST NUMBER SECTION XI CATEGORY ITEM NUMBER

                         !COMPONENT DESCRIPT.

RELIEF REQUESTED ALTERN. EXAM 35 RR-9 8-D 83.90 RV NOZZLE- CODE CASE NO PRIOR 83.100 TO-VESSEL N-451 REPAIRS,

                          & INS.RAD. EXAM       REPLACE SECTION    DEFERRED   OR FLAWS WELDS      TO END OF  IN WELDS INTERVAL RR-10   8-J      89.11   PIPING NPS CODE CASE  8I-DIREC.

C-F-1 C5 .12 4 & LGER. N-524, UT IN C-F-2 C5.52 LONGITUDE EXAMINE LONG./ WELDS ONLY CIRC. WELD INTERSECT AREA CIRC. WELD AREA

          - --

• PALISADES NUCLEAR POWER PLANT THIRD* INTERVAL INSERVICE INSPECTION PROGRAM RELIEF REQUESTS 36 RELIEF REQUEST NUMBER - RR-1 COMPONENT IDENTIFICATION Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-J Item Number 89 .11, 89 .12 Component Description . . Pressure Retainirig Piping PCS-42-RCL-lH-l, PCS-42-RCL-2H-l PCS-30-RCL-1A-16LU-l, 16LU-2, 16 PCS-30-RCL-1B-14LU-l, 14LU-2, 14
• CODE REQUIREMENT PCS-30-RCL-2A-15LU-l, 15LU-2, 15 PCS-30-RCL-2B-15LU-l, 15LU-2, 15 Table IWB-2500-1 requires a surface and volumetric examination of the circumferehtial and longitudinal welds.

BASIS FOR RELIEF The piping welds adjacent to the reactor vessel are buried in cement and are not accessible .for OD examination by the surface or volumetric methods. PROPOSED ALTERNATE EXAMINATION I. The welds will be 100% volumetrically examined (in lieu of the lower l/3t required by Figure IWB-2500-8) from the ID with a m~chanically remote device. Surface examinations will not be performed. As documented in Consumers Power response to the NRC Request for Additional Information TAC No. 72622, _dated March 23, 1990 -for NRC Concern 2.I relating to this relief request for the Second Interval ISI program submittal; the applicable Section XI 1989 edition references for the UT examinations will be

• implemented for thts .relief request.

37 In addition to performing t~e mechanized ID volumetric examination of the entire weld volume and heat affected zone (instead of the lower one-third of the weld volume as required by the Code), Consumers Power commits to demonstrate that the equipment and examination procedures will be capable of detecting OD defects in a laboratory test block with the defects being crack-type defects (not machined notches). Therefore, based on an acceptable demonstration, the proposed alternative, along with the system pressure tests, will provide reasonable assurance of the continued inservice structural integrity .

*I RELIEF REQUEST NUMBER - RR~2 COMPONENT IDENTIFICATION Code Class                    1 Code Reference                 IWB-2500 Table IWB-2500-1 Examination Category          B-A Item Number                    Bl.40 Component Description          Reactor Vessel Head to Flange Weld, 6-118A CODE REQUIREMENT Table IWB-2500-1 requires a surface and volumetric examination of the essentially 100% of the weld length.

BASIS FOR RELIEF Due to the component design -configuration (Ref. CE Dwg. E-232-118) relating to

• the weld and flange proximity, the ultrasonic examination which is performed is limited in that the code required volume for examination cannot be achieved in all scanning paths.

PROPOSED ALTERNATE EXAMINATION The accessible volume of the weld will be examined in lieu of the 100% volumetric examination requirements. The required surface examination will be performed on the entire weld length as required by Section XI .

• )
• RELIEF REQUEST NUMBER - RR-3 COMPONENT IDENTIFICATION 39 Code Class 1 Code Reference IWB-2500 -

Table IWC-2500-1 Examination Category C-A

 .Item Number                    Cl.10 Component Description          St~am Generator Upper Shell to Shell Cone Welds, 1-101-221, 2-101-221 CODE REQUIREMENT Table IWB-2500-1 requires a volumetric examination of welds at.gross struct~ral discontinuities ~hich also.includes essentially 100% of the weld length.

BASIS FOR RELIEF Due to the component design configuration (Ref. CE Dwg(s). E-70277-271-021 and* E-70277-271-001) relating to the snubber attachment, welded pads, and feedwater nozzle interference, the ultrasonic examination which is pe~formed is limited in achieving the 100% code required volume requirement due to scanning limitations. Radiography is not possible due to internal interferences. PROPOSED ALTERNATE EXAMINATION The accessible volume of the welds wi11 be examined in lieu of the 100% volumetric examination requirements.

                                                          -*

• RELIEF REQUEST NUMBER - RR-4 COMPONENT IDENTIFICATION 40 Code Class 1 Code Reference IW8-2500 Table IWB-2500-1 Examination Category B-D Item Number B3.130 Component Description Steam Generator Nozzle to Shell Welds, 1-104-251, l-102-251A, 1-102-2518 2-104-351, 2-102-351A, 2-102-3518 CODE REOU IR EM ENT Table IW8-2500-l requires all nozzle to vessel welds to be volumetrically examined during each inspection interval .
• /

BASIS FOR RELIEF Due to the component design configuration {Ref. CE Dwg{s). E-70277-251-001 and E-70277-251-003) relating to the vessel support skirt and welded lug attachments, the ultrasonic examination which is performed is limited in ach.ie.vjng the code required volume due to scanning limitations. Radiography is not possible due to internal interferences. PROPOSED ALTERNATE EXAMINATION The accessible volume of the welds will be examined in lieu of the 100% volumetric examination requirements.

                                                    .:*:

41 RELIEF REQ\IEST NUMBER - RR-5 COMPONENT IDENTIFICATION Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-D Item Number 83 .150 Component Description Regenerative Heat Exchangers E-56-A and E-56-B Nozzle to Shell Welds E-56-A, Welds 05 and 07 E-56-B, Welds 05 and 07 CODE REQUIREMENT Table IWB-2500-1 requires all nozzle to vessel welds to be volumetrically

                                 -

examined during' each inspection interval.

• ~I BASIS FOR RELIEF Due to the component design configuration (ATLAS Dwg. D-1733 and D-1759}

relating to the vessel support pads, the ultrasonic examination which is performed is limited in achieving the code required volume due to scanning limitations. Radiography is not possible due to internal interferences. PROPOSED ALTERNATE EXAMINATION The accessible volume of the welds will be examined in lieu of the 100% volumetric examination requirements .

• )
• RELiEf REQUEST NUMBER - RR-6 COMPONENT IDENTIFICATION 42 Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-B B-D Item Number B2 .11, B2. 21, B2. 22 B3 .110 Component Description Pressurizer T-72 Upper Shell to Upper Head Weld 5-988 Lower Shell to Lower Head Weld 3-982 Lower Head Circumferential Weld 2-984 Meridional Head Welds:

Upper Head l-983A thru D Lower Head l-984A thru D Nozzle to Shell Welds 1-986, 3-985, 8-986, 8-986A, 8-986B, 8-986C CODE REQUIREMENT Table IWB-2500-1 requires shell and head circumferential welds, one meridional head weld and all nozzle to vessel welds to be volumetrically examined during each inspection interval. BASIS FOR RELIEF Due to the component design configuration (Ref. CE Dwg(s). CE-E-231-982, CE-E-231-983, CE-E-231-984, CE-E-231-985, CE-E-231-986 and CE-E-231-988) relating to the lifting lugs, adjacent nozzles and manway interferences, the ultrasonic examination which is performed is limited in achieving the code required volume due to scanning limitations. Radiography is not possible due to internal interferences .

.. J PROPOSED ALTERNATE EXAMINATION The accessible volume of the welds, including ~ll meridional welds, will be examined in lieu of the 100% volumetric examination requirements of the welds.

• RELIEF REQUEST NUMBER - RR~7 COMPONENT IDENTIFICATION 43 Code Class I Code Reference IWC-2500 Table IWC-2500-1 Examination Category C-A C-B Item Number Cl.10, Cl.30 C2.21 Component Description Shutdown Cooling Heat Exchangers E-60-A and E-60-B Shell to Flange E-60-A and E-60-B Weld 01 Tubesheet to Shell E-60-A and E-60-B Weld Ol Nozzle to Shell E-60-A and E-60-B Weld 03 &

04 CODE REQUIREMENT Table IWC-2500-1 requires C-A welds to be examined volumetrically and C-B welds to be examined with the surface and volumetric methods. BASIS FOR RELIEF Due to the component design configuration (Ref. EFC0-15080) relating to the weld joint to member proximity, the ultrasonic examination which is performed is limited in achieving the code required volume due to scanning limitations. Radiography is not possible due to inaccessability of the ID. PROPOSED ALTERNATE EXAMINATION The accessible volume of the welds will be examined in lieu of the 100% volumetric examination requirements of the welds. The nozzle to shell welds will be examined by the required surface examination technique .

• )
• RELi EF REQUESI NUMBER - . RR-8 COMPONENT IDENTIFICATION 44 Code Class 1 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-A Item Number Bl. 21 Component Description Reactor Vessel Closure Head Circumferential Weld 6-1188 CODE REQUIREMENT Table IWB-2500-1 requires the accessible length of all welds to be volumetrically examined each inspection interval.

BASIS FOR RELIEF .) Due to the component design configuration {Ref. CE-E-232-118 and CE-E-232-139) relating to CRDM assemblies, the ultrasonic examination which is performed is limited in achieving the 100% code required volume requirement due to scanning limitations. Radiography is not possible due to internal interferences. PROPOSED ALTERNATE EXAMINATION The accessible volume of the weld will be examined in lieu of the 100% volumetric examination requirements.

• RELIEF REQUEST NUMBER - RR-9 COMPONENT IDENTIFICATION Code Class 1 45 Code Reference IWB-2500 Table IWB-2500-1 Examination Category B-0 Item Number 83.90, 83.100 Component Description Reactor Vessel Nozzle to Vessel Welds 5-114A, 5-1148, 5-114C, 5-1140, 5-114E, 5-114F and Inside Radius Section Heads 5-114A-IRS, 5-1148-IRS, 5-114C-IRS 5-1140-IRS, 5-114E-IRS, 5-114F-IRS CODE REQUIREMENT Table IWB-2500 requires that nozzle to vessel and inside radius section welds shall be volumetrically examined with a schedule to consist of at least 25%
• but not more than 50% (credited) of the nozzles be examined by the end of the first period, and the remainder by the end of the *inspection interval.

Relating to deferral of inspections, the nozzle to vessel welds are deferrable to the end of the interval provided the examinations are conducted from inside the component and the nozzle weld is examined by straight beam ultrasonic method from the nozzle bore, the remaining examinations required to be conducted from the shell inside diameter may be performed at or near the end of each inspection interval. The inside radius section welds are not deferrable to the end of the interval. BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3) and Footnote 6, the use of the following code case is requested as a relief request. Code case N-521 allows the examination schedule for the aforementioned welds to be deferred to the end of the inspection interval provided the welds have not been repaired or replaced, the welds do not contain identified flaws or relevant indications that currently require successive inspections in

• RELIEF REQUEST NUMBER - RR-9 (cont'd) accordance with IWB-2420(b) and the unit is not in the first inspection 46 interval.

PROPOSED ALTERNATE EXAMINATION Examination records for the nozzle to vessel and inside radius section welds have been reviewed to verify compliance to the conditions in the code case. No repairs, replacements or flaws or relevant indications have been identified with the plant currently nearing the end of the second inspection interval. In addition, these welds will be examined during the last outage of the current second inspection interval. Therefore, the twelve welds will be scheduled to be volumetrically examined during the 2005 outage of the third period of the third interval .

• 47 RELIEF. REQUESI~1NUMBER - RR-10 COMPONENT IDENTIFICATION Code Class 1 and 2 Code Reference IWB-2500, IWC-2500 Table IWB-2500-1, IWC-2500-1 Examination Category 8-J, C-F-1, C-F-2 I tern N.umber 89.12, C-5.12, C5.52 Component Description Pressure Retaining Piping NPS 4 and larger*longitudinal welds CODE REQUIREMENT Table IWB-2500-1 requires longitudinal welds to be examined by the surface .and volumetric methods at least a pipe-diameter length but not more than 12" of each weld.intersecting the circumferential welds required to be examined by examination catiegori es 8-F and 8-J .
• Table IWC-2500-1 requires longitudinal w.elds to be examined by the surface and volumetric methods 2.5t at the associated intersecting circumferential weld.

BASIS FOR RELIEF Pursuant tti 10 CFR 50.55a(a){3) and Footnote 6, the use of the following code case is requested as relief of the code requirements stated. For Code Case N-524, when both surface and volumetric examinations are required, examination and longitudinal piping welds is not required beyond those portions of the welds within the examination boundaries of intersecting circumferential welds.*

49 SECTION 6 VERIFICATION OF SECtION XI COMPLIANCE

  ***                  6. VERIFICATION OF SECTION XI COMPLIANCE 50 A. INTRODUCTION The following tables document compliance, for the third interval, with the examination distribution requirements of Section XI of the ASME B&PV Code. The tables identify the total number of components by category which are subject to distribution and the total number of components for Inspection Program B.

B. DETERMINATION OF COMPLIANCE

l. First Period The minimum and maximum number of components requi~ed to be examin~d during the first period has been determined by applying the minimum and maximum percentages cited in the code, Table{s) IWB-2412-1, IWC-2412-1 and'IWD-241201.

The minimum-number of components to be examined is 16% of the total components. The maximum number of components which can be examined is 34% of* the total number of components.

2. Second; Period

  **        lhe minimum number ~f components to be examined is 50% to the total components. The maximum number of components which can be examined is 67% of the total number of components.

The Section XI creditable percentage for this period is computed by dividing the number of components examined by the total number of components.

3. Third Period The minimum and maximum number of components to be examined is obtained based solely on the number of examinations required.to complete the cumulative total.

C. NUMBER OF COMPONENTS The number of components subject to examination per period will vary throughout the life of the plant due to code changes, relief requests and line walking. Compliance per period is based solely on the number of components subject to the examination for that period .

• ' /

j

• D. INTERVAL COMPLIANCE ,,,.,_,,_.,,., .

Third period cumulative percentage totals which equal or exceed 100% verifies 51 compliance with the distribution requirements of Section XI. A separate table verifies compliance for those examinations which are deferred to the end of the interval. In the event that the number of examinations subject to distribution in a category decreases, and it is determined that it is impossible to achieve 100% without re-examining components, the component will not be reexamined. For example:

1. During the first period, 5 of 20 components in a category are examined (ie, 25%).

2. During the second period, it is determined that only 15 components actually exist in that category and that 5 additional components are examined (ie, 33-1/3% for this interval third and 58.3% cumulative).

3. The 5 remaining components are examined during the last period (ie, a cumulative total of 91.7%).

Under this; situation, which could result from a number of reasons, no further

• examining is required provided adequate documentation substantiates the anomaly .

E. MIDINTERVAL REQUIREMENT CHANGES 10 CFR 50.55a(g) requires periodic updates of ISI programs to the currently approved version of Section XI. Implementation of these changes in this Plan in midinterval may require examinations of areas not previously subject to examination. No attempt is made to "catch up" those examinations.

52

• VERIFICATION OF COMPLIANCE THIRD INTERVAL_,..

BY CATEGORY

• ATTACHMENT: - Database Population Summary

           - Category and Item Number Designation, Legend
           - Category Summary, Page(s) 1-2

PALISADES NUCLEAR PLANT

• DATABASE POPULATION

SUMMARY

(includes all records, regardless of schedule) Class 1 components Class 2* Components Class 3 components B-A components = 28 C-A components = 18 D-A components = 0 B-B components = 36 C-B components = 22 D-B components = 56 B-D components = 42 c-c components =* 63 D-C components = 0 B-E components = 0 C-F-1 components = 414 B-F components = 47 C-F-2 components = 174 B-G-1 components = 175 B-G-2 components = 152 B-H components = 6 component Supports Non-sec. XI Components B-J components = 746 B-K-1 components = 20 F-A components = 892 N/A components = 929 B-L-1 components = 4 B-L-2 components = 4 B-M-1 componen11s- = 2 Inservice Inspection - Database Totals

• B-M-2 components = 14 B-N-1 components = 1 Total Class 1 Components =1513 I

B-N-2 components = 2 Total Class 2 Components = 691 B-N-3 components = 1 Total Class 3 Components = 56 B-0 components = 233 Total Component Supports = 892 B-Q components = 0 Non-Section XI Components = 929 Database Total: 4081

• ) Report: file _pop File: dcprogma Project: file_pop 02/28/95
• Palisades Nuclear Power Plant Third Inservice. Inspection Interval ASME Section XI Category and Item Number Designation Legend

1. An "R" after the Section XI item number for a particular examination

     . category identifies that there is a associated Request for Relief.

2. If exempt due to component thickness per Category C-F-1, Item ~o. CS.IO or Category C-F-2, Item No. CS.SO then; Examination Category is followed by an astrix (*). (e.g., C-F-1*, C-F-2*} Also, the Code Item No. is identified as C-F~l*-NA or C-f-2*-NA as applicable.

3. If excluded by Table IWC-2SOO~l per note (2)(b) for Examination Category C-F-1 or C-F-2 pipe to pipe and associated longitudinal weld, the Examination Category is followed by a pound (#). (e.g., C-F-1#, C-F-2#)

Also, the Cod~ Item No. i~ identified as C-F-1#-NA or C7F-2#-NA as

• applicable .

4. For pipe restraints on lines that are exempt per IWC-1220 in accordance with Code Case N-491, the Examination Category is the applicable Category (i.e., F-A). Also, the applicable Code Item No. is used followed by a astrix (*). (e.g., Fl.20A*)

5. For Examinati~n Category C-C, Integral Attachments where the line is exempt by Table IWC-2500-1 per note (l)(c) and note (4) the Examination Category is followed by a astrix (*). (e.g., C-C*) Also, the Code Item No. is identified as CC*-NA.

For Examination C~tegory C-C, Integral Attachments where the line is not exempt based on line thickness per note (4), but is e~empt based on material design thickness per note (l)(c) then the Examination Category is followed by a pound (#). (e.g., C-C#) Also, the Code Item No. is identified as CC#-NA. \

6. Class 1, 2 and 3 Component suppo~t letter designation after the Section XI Item No. as required by Code Case N-491. *

         "A" Designates one directional "B" Designates multi-directional

.)

         "t" Designates thermal movement

EXAM

            **
            *._

TOTAL INTERVAL 3 PERCENT *PERIOD 1 PALISADES.GRAM PLAN THIRD INTERVAL PERIOD 1 PERIOD 1 ATEGORY

SUMMARY

PERIOD 2 PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY ACTIVE REC SCHEDULED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE AUG 44 44 100.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0%

B-A 28 27 96.4% 0 0 0% 0 0 0% 27 0 100% 0 0.0% B-B 36. 22 61.1% 7 0 32% 6 0 59% 9 0 100% 0 0.0% B-D 42 38 90.5% 8 0 21% 12 0 53% 18 0 100% 0 0.0% B-F 47 47 100.0% 13 0 28% 9 0 47'/. 25 0 100% 0 0.0% B-G-1 175 171 97. T'!. 56 0 33% 56 0 65% 59 0 100% 0 0.0% B-G-2 152 152 100.0% 41 0 27% 40 0 53% 71 0 100% 0 0.0% B-H 6 2 33.3% 0 50% 0 0 50% 0 100% 0 0.0% B-J 746 217 29.1% 38 0 18% 82 0 55% 97 0 100% 0 0.0% B-K-1 20 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B-L-1 4 25.0% 0 0 0% 0 100% 0 0 100% 0 0.0% B-L-2 4 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B-M-1 2 50.0% 0 0 0% 0 0 0% 0 100% 0 0.0% B-M-2 14 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B-N-1 100.0% 0 100% 0 200% 0 300% 0 0.0% B-N-2 2 2 100.0% 0 0 0% 0 0 0% 2 0 100% 0 0.0% B-N-3 100.0% 0 0 0% 0 0 0% 0 100% 0 0.0% Report: i3catpop Note: All totals .based on Active records. File: sch_cat Period totals must also be Code Credit "Y". Index: key/Exam Category Page nunber 1

EXAM

            **-
            \    .,

TOTAL .INTERVAL 3 PERCENT PERIOD 1 PALISADES.OGRAM PLAN THIRD INTERVA[*-* CATEGORY

SUMMARY

PERIOD 1 PERIOD 1 PERIOD 2 PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY ACTIVE REC SCHEDULED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE B-0 233 6 2.6% 0 0 0% 0 0 0% 6 0 100% 0 O.D%

C-A 18 9 50.0% 2 0 22% 4 0 6 7"-' 3 0 100% 0 0.0% C-B 22 16 72. 7"-' 4 0 25% 6 0 63% 6 0 100% 0 0.0% c-c 63 57 90.5% 11 0 19% 25 0 63% 21 0 100% 0 0.0% c-c# 11 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C-C* 56 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C-F-1 414 215 51.9% 68 0 32% 69 0 64% 77 0 100% 0 0.0% C-F-1# 4 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C-F-1* 1304 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C-F-2 174 86 49.4% 16 0 19% 27 0 50% 43 0 100% 0 0.0% C-F-2# 24 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C-F-2* 138 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% D-B 56 56 100.0% 13 0 23% 21 0 61% 22 0 100% 0 0.0% F-A 892 229 25. 7"-' 67 0 29% 74 0 62% 88 0 100% 0 0.0% N/A 578 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% TOTALS: 5311 1400 346 0 433 0 578 0 0 Report: i3catpop Note: All totals based on Active records. Fi le: sch_ cat Period totals must also be Code Credit "Y". Index: key/Exam Category Page m.ntier 2

53 VERIFICATION OF COMPLIANCE THIRD INTERVAL BY CATEGORY AND ITEM NUMBER

• ATTACHMENT: - Category and Item Number Designation, Legend

           - Category/Item Number Summary, Page(s) 1-12

• Palisades Nuclear Power Plant Third Inservice Inspection Interval ASME Section XI Category and Item Number Oesianation Legend

1. An 11 R11 after the Section XI item number for a particular examination category identifies that there is a associated Request for Relief.

2. If exempt due to component thickness per Category C-F-1, Item No. CS.10 or Category C-F-2, Item No. CS.50 then; Examination Category is followed by an astrix (*).(e.g., C-F-1*, C-F-2*) Also, the Code Item No. is identified as C-F-1*-NA or C-F-2*-NA as applicable.

3. If excluded by Table IWC-2500-1 per note (2)(b) for Examination Category C-F-1 or C-F-2 pipe to pipe and associated longitudinal weld, the Examination Category is followed by a pound (#). (e.g., C-F-1#, C-F-2#)

Also, the Cod~ Item No. i~ identified as C-F-1#-NA or C-F-2#-NA as applicable. *

• 4. For pipe restraints on lines that are exempt per IWC-1220 in accordance with Code Case N-491, the Examination Category is the applicable Category (i.e., F-A). Also, the applicable Code Item No. is used followed by a astrix (*).(e.g., Fl.20A*)

5. For Examination Category C-C, Integral Attachments where the line is exempt by Table IWC-2500-1 per note (l)(c) and note (4) the Examination Category is followed by a astrix (*). (e.g., C-C*) Also, the Code Item No. is identified as CC*-NA.

For Examination Category C-C, Integral Attachments where the line is not exempt based on line thickness per note (4), but is e~empt based on material design thickness per note (l)(c) then the Examination Category is followed by a pound(#). (e.g., C-C#) Also, the Code Item No. is identified as CC#-NA.

6. Class 1, 2 and 3 Component support letter designation after the Section XI Item No. as required by Code Case N-491.

11 A11 Designates one directional "B" Designates multi-directional 11 C11 Designates thermal movement .)

EXAM

            *
            "--"-

ITEM TOTAL INT 3 PERCENT PERIOD 1 PAL!SADE~OGRAM PERIOD 1 PERIOD 1 PERIOD 2 PLAN THIRD INTERVAL --,TEM NUMBER

SUMMARY

PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE AUG MP-4.0 10 10 100.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0%

OVERLAY 2 2 100.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% RG-1.14 4 4 100.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% TS-4.12 28 28 100.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0%

                                                      -----                   -----         ---*------**--*-

CATEGORY TOTAL: 44 44 0 0 0 0 0 0 0 B-A B1 .11 3 3 100.0% 0 0 0% 0 0 0% 3 0 100% 0 0.0% B1 .12 9 9 100.0% 0 0 0% 0 0 0% 9 0 100% 0 0.0% B1.21 100.0% 0 0 0% 0 0 0% 0 100% 0 0.0% B1.21R 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B1.22 12 12 100.0% 0 0 0% 0 0 0% 12 0 100% 0 0.0% B1 .30 100.0% 0 0 0% 0 0 0% 0 100% 0 0.0% B1.40 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B1 .40R 100.0% 0 0 0% 0 0 0% 0 100"1, 0 0.0% CATEGORY TOTAL: 28 27 0 0 0 0 27 0 0 Report: i3it""°p Note: All totals based on Active records. File: sch_item Period totals RlJSt also be Code Credit 11 Y11

• Index: key/Exam Category Page nllJlber 1

                                                                                    .*.:

          *
                                                                                                                   -----------             - ---

EXAM CATEGORY ITEM NUMBER TOTAL ACTIVE INT 3 SCH ED PERCENT TOTAL POP PERIOD 1 SCHEDULED PALISADE~OGRAM PLAN TH I RD INTERVAL ----dEM NUMBER

SUMMARY

PERIOD 1 PERIOD 1 COMPLETE PERCENT PERIOD 2 SCHEDULED PERIOD 2 PERIOD 1+2 COMPLETE PERCENT PERIOD 3 SCHEDULED PERIOD 3 PERIOD 1-3 COMPLETE PERCENT

• 02/28/95 TOTAL PERCENT COMPLETE COMPLETE B-B B2.11 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0%

B2.11R 2 2 100.0% 0 50% 0 100% 0 0 100% 0 0.0% 82.12 4 2 50.0% 0 0 0% 0 50% 0 0 0.0% 82.21 100.0% 0 0 0% 0 0 0% 0 100% 0 0.0% 82.21R 100.0% 0 100% 0 0 100% 0 0 100% 0 0.0% 82.22 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B2.22R 8 8 100.0% 4 0 50% 4 0 100% 0 0 100% 0 0.0% B2.31 4 2 50.0% 0 0 0% 0 0 0% 2 0 100% 0 0.0% 82.32 10 10.0% 0 100% 0 0 100% 0 0 100% 0 0.0% B2.40 2 50.0% 0 0 0% 0 0 0% 0 100% 0 0.0% B2.51 2 2 100.0% 0 0 0% 0 0 0% 2 0 100% 0 0.0% 82.80 2 2 100.0% 0 0 0% 0 0 0% 2 0 100% 0 0.0% CATEGORY TOTAL: 36 22 7 0 6 0 9 0 0 8-D 83.100 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% 83.100R 6 6 100.0% 0 0 0% 0 0 0% 6 0 100% 0 0.0% 83.110 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% Report: i3it~p Note: All totals based on Active records. File: sch_item Period totals must also be Code Credit ,"Y". Index: key/Exam Category Page number 2

EXAM

             *ITEM   TOTAL  INT 3  PERCENT  PERICO 1 PALISADE~*JGRAM PERIOD 1 PERIOD 1         PERICO 2 PLAN THIRD INTERVAL ----,:TEM NUMBER 

SUMMARY

PERIOD 2 PERIOD 1+2 PERIOO 3 PERICO 3 PERIOD 1-3

• 02/28/9S TOTAL PERCENT CATEGORY NUMBER ACTIVE SCHED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE B-D B3.110R 6 6 100.0% 0 1?"" 5 0 100% 0 0 100% 0 0.0%

B3.120 6 6 100.0% 0 5 0 100% 0 0 100% 0 0.0% 63.130 0 0 0.0% 0 0 0% o- 0 0% 0 0 0% 0 0.0% B3.130R 6 6 100.0% 3 0 50% 0 0 50% 3 0 100% 0 0.0% B3.140 6 6 100.0% 3 0 50% 0 0 50% 3 0 100% 0 0.0% B3. 1SO 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B3.150R 4 2 S0.0% 0 0 0% 2 0 100% 0 0 100% 0 0.0%

63. 160 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0%

B3.160R 2 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B3.90 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B3.90R 6 6 100.0% 0 0 0% 0 0 0% 6 0 100% 0 0.0% CATEGORY TOTAL: 42 38 8 0 12 0 18 0 0 B-F BS .130 23 23 100.0% s 0 22% 4 0 39% 14 0 100% 0 0.0% B5. 130R 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% BS .140 13 13 100.0% 7 0 S4% 0 0 S4% 6 0 100% 0 0.0%

85. 150 5 s 100.0% 0 0 0% 0 0 0% 5 0 100% 0 0.0%

Report: i3itrf'4'0p Note: ALL totals based on Active records. Fi le: sch_ item Period totals must also be Code Credit "Y". Index: key/Exam Category Page number 3

EXAM CATEGORY

• ITEM NUMBER TOTAL ACTIVE INT 3 SCH ED PERCENT TOTAL POP PERIOD 1 SCHEDULED PALISADE~~GRAM PLAN THIRD INTERVAL * - EM NUMBER

SUMMARY

PERIOD 1 PERIOD 1 COMPLETE PERCENT PERIOD 2 PERIOD 2 PERIOD 1+2 SCHEDULED COMPLETE PERCENT PERIOD 3 SCHEDULED PERIOD 3 PERIOD 1-3 COMPLETE PERCENT

• D2/20/95 TOTAL PERCENT COMPLETE COMPLETE B-F ES.40 6 6 100.0% 1 0 17"/, 5 0 100% 0 0 100% 0 0.0%

CATEGORY TOTAL: 47 47 13 0 9 0 25 0 0 B-G-1 86.10 54 54 100.0% 18 0 33% 18 0 67% 18 0 100% 0 0.0% 86.180 4 4 100.0% 0 25% 0 50% 2 0 100% 0 0.0%

             &6.190      4      0      0.0%         0       0          0%            0        0        0%        0          0         0%          0        0.0%

86.200 4 4 100.0% 0 25% 0 50% 2 0 100% 0 0.0% 86.30 54 54 100.0% 18 0 33% 18 0 67% 18 0 100% 0 0.0% 86.40 100.0% 0 0 0% 0 0 0% 0 100% 0 0.0% 86.50 54 54 100.0% 18 0 33% 18 0 67% 18 0 100% 0 *0.0% CATEGORY TOTAL: 175 171 56 0 56. 0 59 0 0 B-G-2 . 87.10 16 16 100.0% 0 0 0% 16 0 100% 0 0 100% 0 0.0% 87.20 2 2 100.0% 2 0 100% 0 0 100% 0 0 100% 0 0.0% 87.30 4 4 1oo".0% 0 0 0% 0 0 0% 4 0 100% 0 0.0% 87.50 5 .5 100.0% 0 0 0% 0 0 0% 5 0 100% 0 0.0% Report: i 3 i tf1l>Op . Note: All totals based on Active records. Fi Le: sch_ item Period totals must also be Code Credit "Y". Index: key/Exam Category Page m.1nber 4

                                                                                       *~:*

EXAM

           *
           ,....,....,

ITEM TOTAL INT 3 PERCENT PERIOD 1 PAL I SADE.:OGRAM PLAN THIRD INTERVAL ~--, EM NUMBER

SUMMARY

PERIOD 1 PERIOD 1 PERIOD 2 PERIOO 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE 8-G-2 87.60 8 8 100.0% 0 0 0% 8 0 100% 0 0 100% 0 0.0%

87.70 27 27 100.0% 7 0 26% 16 0 85% 4 0 10!)~ 0 0.0% 87.80 90 90 100.0% 32 0 36% 0 0 36% 58 0 100% 0 0.0% CATEGORY TOTAL: 152 152 41 0 40 0 71 0 0 8-H 88.10 3 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% 88.20 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% 88.30 2 2 100.0% 0 50% 0 0 50% 0 100% 0 0.0% CATEGORY TOTAL: 6 2 0 0 0 0 0 8-J 89. 11 191 50 26.2% 11 0 22% 15 0 52% 24 0 100% 0 0.0%

89. 11R 10 10 100.0% 0 0 0% 0 0 0% 10 0 10,0% 0 0.0%

                                                                                                                                                                     \.

89.12 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0%

89. 12R 118 25 21.2% 2 0 8% 4* 0 24% 19 0 Hl0% 0 0.0%

89.21 65 18 27. 7"1. 5 0 28% 6 0 61% 7 0 100% 0 0.0% 89.31 6 6 100.0% 2 0 33% 3 0 83% 0 100% 0 0.0% Report: i3it~p Note: All totals based on Active records. File: sch_item Period totals must also be Code Credit "Y". index: key/Exam category Page number 5

      - ---------

EXAM CATEGORY

           ~

ITEM NUMBER TOTAL ACTIVE INT 3 SCH ED PERCENT TOTAL POP PERIOD 1 SCHEDULED PALISADES~OGRAM PLAN THIRD INTERVAL PERIOD 1 PERIOD 1 COMPLETE PERCENT

                                                                         ~--.-rEM  NUMBER 

SUMMARY

PERIOD 2 SCHEDULED PERIOD 2 PERIOD 1+2 COMPLETE PERCENT PERIOD 3 SCHEDULED PERIOD 3 PERIOD 1-3 COMPLETE PERCENT

• 02/28/95 TOTAL PERCENT COMPLETE COMPLETE B-J 89.32 5 5 100.0% 0 0 0% 5 0 100% 0 0 100% 0 0.0%

B9.40 351 103 29.3% 18 0 17"-' 49 0 65% 36 0 100% 0 0.0% CATEGORY TOTAL: 746 217 38 0 82 0 97 0 0 B-K-1 B10.10 4 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% B10.20 16 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% CATEGORY TOTAL: 20 0 0 0 0 0 0 0 0 B-L-1 B12.10 4 25.0% 0 0 0% 0 100% 0 0 100% 0 0.0% 8-L-2 812.20 4 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% 8-M-1 B12.40 2 50.0% 0 0 0% 0 0 0% 0 100% 0 0.0% 8-M-2 812.50 14 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% Report: i3itrrpop Note: ALL totals based on Active records. File: sch_item Period totals rrust also be Code Credit "Y". Index: key/Exam Category Page m.rnber 6

EXAM

           *
           \;__,,-'

ITEM TOTAL INT 3 PERCENT PERIOD 1

                                                                 .PALISADE~OGRAM PERIOD 1 PERIOD 1 PLAN THIRD INTERVAL *--'dEM NUMBER 

SUMMARY

PERIOD 2 PERIOD 2 PERIOD 1+2 I PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE. PERCENT SCHEDULED COMPLETE PERCENT I SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE B*N-1 B13.10 100.0% ' 1 0 100% 0 200% 0 300% 0 0.0%

B-N-2 813.50 100.0% 0 0 0% 0 0 0% 0 100% 0 0.0%

           !?13.60                 100.0%        0       0          0%              0       0         0%                    0         100%         0         0.0%
  *CATEGORY TOTAL:    2      2                   0       0                          0       0                      2        0                      0 B-N-3        813.70                 100.0%        0       0          0%              0       0         0%                    0         100%         0         0.0%

8-0 814.10 233 6 2.6% 0 0 0% 0 0 0% 6 0 100% 0 0.0% C*A C1. 10 4 2 50.0% 0 0 0% 0 50% 0 100% 0 0.0% C1.10R 4 2 50.0% 0 0 0% 0 50% 0 100% 0 0.0% C1.20 6 3 50.0% 2 0 67% 0 100% 0 0 100% 0 0.0% C1.30 2 50.0% 0 0 0% 0 100% 0 0 100% 0 0.0% C1.30R 2 50.0% 0 0 0% 0 0 0% 0 100% 0 0.0% Report: i3it~p Note: All totals based on Active records. File: sch_item *Period totals must also be Code Credit "Y" *. Index: key/Exam Category Page nll!lber 7

        - - - -                                                                                                                         \

EXAM

           *'--'

ITEM TOTAL INT 3 PERCENT PERIOD 1 PALISADE~~OGRAM PERIOD 1 PERIOD 1 PERIOD 2 PLAN THIRD INTERVAL --dEM NUMBER

SUMMARY

PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT CCJ4PLETE COMPLETE C-A CATEGORY TOTAL: 18 9 2 0 4 0 3 0 0 C-B C2.21 12 8 66. 7"!. 0 13% 4 0 63% 3 0 100% 0 0.0%

C2.21R 4 2 50.0% 2 0 100% 0 0 100% 0 0 100% 0 0.0% C2.22 6 6 100.0% 0 17% 2 0 50% 3 0 100% 0 0.0% CATEGORY TOTAL: 22 16 4 0 6 0 6 0 0 c-c C3.10 12 6 50.0% 0 0 0% 0 0 0% 6 0 100% 0 0.0% C3.20 51 51 100.0% 11 0 22% 25 0 71% 15 0 100% 0 0.0% CATEGORY TOTAL: 63 57 11 0 25 0 21 0 0 c-c# CC#-NA 11 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C-C* CC*-NA 56 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% Report: i3itTr4J0p Note: ALL totals' based on Active records. File: sch_item Period totals must also be Code Credit "Y". Index: key/Exam Category Page mmber 8

EXAM

         * '-....-*'

ITEM TOTAL INT 3 PERCENT PERIOD 1 PALISADE.WGRAM PLAN THIRD INTERVAL -~.-TEM NUMBER

SUMMARY

PERIOD 1 PERIOD 1 PERIOD 2 PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE. PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE C-F-1 CS.11 187 116 62.0% 38 0 33% 30 0 59% 48 0 100% 0 0.0%

CS.12 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% CS .12R 31 6 19.4% 0 0 0% 6 0 100% 0 0 100% 0 0.0% CS.21 94 30 31.9% 12 0 40% 9 0 70% 9 0 100% 0 0.0% CS.30 64 30 46.9% 10 0 33% 12 0 73% 8 *O 100% 0 0.0% CS.41 38 33 86.8% 8 0 24% 12 0 61% 12 0 97°io 0 0.0%

                                                                                                                                      ';*:

CATEGORY TOTAL: 414 215 68 0 69 0 77 0 0 C-F-1# CF1#-NA 4 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C-F-1* CF1*-NA 1304 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C-F-2 CS.51 69 32 ' 46.4% 8 0 25% 7 0 47°/, 17 0 100% 0 0.0% CS.52 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% C5.52R 72 26 36.1% 3 0 12% 12 0 58% 11 0 100% 0 0.0% cs .81 33 28 84.8% 5 0 18% 8 0 46% 15 0 100% 0 0.0% Report: i3itmpop Note: All totals based on Active records. File: sch_item Period totals must also be Code Credit "Y". Index: key/Exam Category Page minber 9

EXAM CATEGORY

• ITEM NUMBER TOTAL ACTIVE INT 3 SCH ED PERCENT TOTAL POP PERIOD 1 SCHEDULED PALISADE~OGRAM PERIOD 1 PERIOD 1 COMPLETE PERCENT PERIOD 2 PLAN TH I RD INTERVAL **--.rEM NUMBER

SUMMARY

SCHEDULED PERIOO 2 PERIOD 1+2 COMPLETE PERCENT PERIOD 3 SCHEDULED PERIOO 3 PERIOD 1-3 COMPLETE PERCENT

• 02/28/95 TOTAL PERCENT COMPLETE COMPLETE C-F-2 CATEGORY TOTAL: 174 86 16 0 27 0 43 0 0 C-F-2# CF2#-NA 24 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0%

C-F-2* CF2*-NA 138 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% D-B 02.20 53 53 100.0% 12 0 23% 21 0 62% 20 0 100% 0 0.0% 02.40 3 3 100.0% 0 33% 0 0 33% 2 0 100% 0 0.0% CATEGORY TOTAL: 56 56 13 0 21 0 22 0 0 F-A F1. 10A 102 28 27.5% 11 0 39% 8 0 68% 9 0 100% 0 0.0% F1. 1OB 29 8 27.6% 3 0 38% 3 0 75% 2 0 100% 0 0.0% F1. 1OC 26 9 34.6% 5 0 56% 0 6T'I. 3 0 100% 0 0.0% F1.20A 142 28 19. T'I. 11 0 39% 9 0 71% 8 0 100% 0 0.0% Report: i3itrrpop Note: All totals based on Active records. Fi le: sch_ item Period totals must also be Code Credit "Y". Index: key/Exam Category Page number 10

EXAM

           *... ~....

ITEM TOTAL INT 3 PERCENT PERIOD 1 PALISADE~OGRAM PERIOD 1 PERIOD 1 PERIOD 2 PLAN TH I RD I MTERVAL - *-1 TEM NUMBER

SUMMARY

PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE F-A f1.20A* 11 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0%

F1.20B 76 17 22.4% 6 0 35% 4 0 59% 7 0 100% 0 0.0% F1.20B* 21 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% F1.20C 132 26 19.7% 6 0 23% 11 0 65% 9 0 100% 0 0.0% F1.20C* 14 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% F1.30A 145 60 41.4% 13 0 22% 26 0 65% 21 0 100% 0 0.0% F1.30B 107 15 14.0% 4 0 27% 7 0 73% 4 0 100% 0 0.0% F1 .30C 37 12 32.4% 5 0 42% 2 0 58% 5 0 100% 0 0.0% F1.40A 25 19 76.0% 2 0 11% 3 0 26% 14 0 100% 0 0.0% F1.40A* 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% F1.40B 24 7 29.2% 0 14% 0 0 14% 6 0 100% 0 0.0% CATEGORY TOTAL: 892 229 67 0 74 0 88 0 0 N/A D-PAL 0 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% N/A 578 0 0.0% 0 0 0% 0 0 0% 0 0 0% 0 0.0% CATEGORY TOTAL: 578 0 0 0 0 0 0 0 0 Report: i3itmpop Note: All totals based on Active records. File: sch_item Period totals must also be Code Credit "Y". Index: key/Exam Category Page number 11

EXAM

           *----

ITEM TOTAL INT 3 PERCENT PERIOD 1 PALISADE~OGRAM PERIOD 1 PERIOD 1 PERIOD 2 PLAN THIRD INTERVAL ---, fEM NUMBER

SUMMARY

PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3

• 02/28/95 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE

                    --

GRAND TOTAL: 5311 1400 346 0 433 0 578 0 0 Report: i3itmpop Note: All totals based on Active records. File: sch_item Period totals must also be Code Credit "Y". Index: key/Exam Category Page number 12

• 54 SECTION 7 COLOR CODED PIPING AND INSTRUMENT DIAGRAMS
• KEY:

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                                                                                                                                                                                                               ,__,__~------------~~__,A
                                                                                                                                                                                                                   @                       CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT COVERT, MICHIGAN TO PIPING & INSTRUMENT DIAGRAM CHEMICAL & VOLUME CONTROL SYSTEM
 .

R 7 4

I 4 l H H G G MV-CVC-591 I _ __J~ CHARGING PUMPS F <REFER TO M-209 FOR COOL ING) F OBA & NORMAL SHUTDOWN SEOUENCER (START P- 5581 i

                                                                                              ~
                                                                                              ~

VCT M-202 SH. IA E CC-7l E CC-7-2" D D 2* -N7M3-2195 L. o. c ~~~~~n~ WATER FOR O ISCHARGE c MA INF OLD FLUSH AND HYDROS TAT IC TEST CONNECTION DEDT TRIP PUMP H.P. SAFETY ON LOW' SUCTION - INJECTION PRESS WHEN M-204 SH. lA P55B IS POWERED IC-Bl tftf pmflR ps5 RcE DEDT

                                                      @   CHARO I NG PL !MP SEAL ~ATION p S5A, p- 558, P- 55C
                                                     ~~~~~BF~R cpA~~A !DENT I CAL B                                                     EXCEPT TAG. NO!!- ARE IN PARENTHESIS>                                                                                                                                                         B PRIMARY MAKE-UP \!(ATER M-202 SH. 1A (f-4)

DEDT 9509120015 1g i;f ~R~~%~~0~1c&0~t~ ~~029

                                                                                                                                                                                                                       -05 JGD
                                                                                                                         ~~~~~e~   WATER FOR DISCHARGE MA INFOLD FLUSH                                        WATER WELL                                 AND HYDROSTATIC TEST CONNECT fON                            HC-33-%."

OEOT CHARGING PUMP A LEAKAGE DEOT TEST M-2l0 SH 1

                   <B-7l               * -V2276 HC-1-!I.,'          V2277)

V2278 THIS DWG. IS !SI COLOR CODED. 8 7 5 4 3 7 1

8 7 6 5 4 3 2 1

* ...            NJTROGEN M-222 SH. I GC-7-1'
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• M-211 SH. 2 (0-8l

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                                    ~7 HB GC F                                                                                                                                                                                                             F E                                                                                                                                                                                                             E SIRW TANK M-204 SH 1 CG-Bl CLOSES D                             ~                                                                                                                                                                               D c                                                                                                                                                                                                             c REACTOR COOLANT                                                                                                                                      SAFETY INJECT ION LOOP    lA                                          CC 12°                                                                      cc-4-sn              HEADER
        < M-201. SH. l"E                                                                                                                                      M-203. SH. 2
              <F-3)                                                                                                                                                  <D-Bl REACTOR COOLANT                                                                                                                                      SAFETY INJECT ION LOOP 18                                                                                                                                                HEADER M-201. SH. 1 CC 12"                                                                        M-203. SH. 2 CE-41                                                                                                                                                  tC 81 REACTOR COOLANT                                                                                                                                      SAFETY INJECT !ON B             LOOP 2A M 201, SH. 1 12 1 -ES 3131 CC 4 12P                                           cc   4 6 11 HEADER M-203. SH. 2
              <E-4l                                                                                                                                                  CB-Bl REACTOR COOLANT                          12° -ES 3146                                                                                                SAFETY INJECTION LOOP 28                                                                                                                                                HEADER CC-4-6° M 201. SH. 1                                                                                                                                           M-203, SH. 2 CF-5l                                                                                                                                                  CA-Bl 9509120015 37 5

qr

                                                                                                                                                                             -Db CHANGED CV-306q STATUS TO CLOSED PER OCR Cli-404
                                                                                                                                      ,__~~~~~~~~~~~~~~~...._,A fP'
                                                                                                                                           ".::::+/-:)

CONS~,t1,~fJ. ~p~~~i?MPANY COVERT, MICHIGAN 1' p*oui-.~ PIPING & INSTRUMENT OIAGRc.M*- l=i§ill!i!ODlli<qSAFETY INJECTION. CONTAINMENT ,;_PRAY THIS DWG. IS ISI COLOR CODED. & SHUDOWN COOLING SYSTEM M203 37 8 I_ 7 6 5 4 3 2 1

8 7 1 ....... H OBA SEQUENCER

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A/S I 2 1 -CA 739 I -CA 738 LOW PRESSURE INJECTION GC- 1-12" M-204- SH 1

                "'..;.                                                                                                            <A-B>

I b~b: I A/S D ti C8-3J ,.__ _ __. D 1 1 -ES 3227 z '4* RECIRCULATION TO SIRW TANK

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w FA! :i: -~ _d33 T-829 l'-ES3217 M-203, SH. 1

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0303 GC 1" LC GC 1 1 RECIRCULATION TO SIRW TANK M-21214 SH 1 c @ 0301 ZS GC-1-8 1 SHUTDOWN COOLING HEAT EXCHANGER E-608 M-204 SH 1 CG-Bl c C0-8l Y,*

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                                                                   ---1                     CA-4>

SHUTDOWN COOLING T-B2C M-203. SH. 1 I I~ GC-1-8 1 HEAT EXCHANGER E- S0A M-204 SH 1 CB-2l L) D e3 CE-Bl I B B tI @

                                                                                                                                                                                                                     -
                                                                                                                                                                                                                         ~     B 33

(!-~-~~~---~; J o-~~~~~---~~ EVENT GC IP THIS DWG. IS ISI COLOR CODED.

                                                                                                                                                                                                      +- --- - -

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                                                                                                                                                                                                             ~33 CONTAINMENT SPRAY PUMP TEST INTERLOCK t--~
                                                                                                                                                                                                   ~

IOPEN & CLOSE) HMW A T-820 GC- 1-6" CIC APP A M-21213. SH. l CONTAINMENT SPRAY CONSUMERS POWER COMPANY CB-21 PALJSAOES NUCLEAR PLANT IOUAL ITY GROUP 21 COVE~T. MICHIGAN 01 PIPING & INSTRUMENT.- D!AG::tAM pWUll!il!;:;io;:::jSAFETY INJECTION. CONTAINMENj:' SPRAY 9509120015 & SHUTDOWN COOL~G SYSTEM M203 .""' 2 16 8 7 6 5 4 3 2 1

      -     ij SPRAY REC IRC TO S IRW' HC-23-6 1 M-204 SH. 18                                         -* f' CH-Bl GC-4-4 1 H                           1 1 -Nlc:IM3-3236                                                                                                                                                                                                                                                                       (4' H

RECIRC TO SIRW M-204 SH. 18 CG-Bl GC 10' REFER NOTE 5

                  <D- 1l                                                                                                                                                       C33 GC P                                                                                                                                                                                                                OC 2" G B TRAIN     M-203 SH. 2
                                                                                                                                                                          ~                                                                                                                                                                                                                                                         G CONTAINMENT SPRAY                                                                                                                                                                                                                                                          I   OR        OBA A TRAIN RECIRCULATION M-21213 SH. 2 GC 1'                                                                     GC 10 1
                                                                                                                                                                          ~                                                                                                         I~ SEQUENCER
                  <c- 1>                                                                                                                                                                                             u 0                                                                          DC- 1-3" 1 1 -N 130M3-3378 61 - N26M3DR- 3225
                                                                                                                                                                                    @SHUTDOWN                                                   HP SAFETY COOLING HEAT EXCHANGERS INJECT ION PUMP
                                                                                                                                                                                                                                                  .E=..Qfill_             0                                                                                       SIRW C33 11 -N1312JM3-3380                                                                                                                                                                                                         M-204 SH. LB F                                                                                                                                 WEST SAFEGUARDS ROOM SUMP CE-7l                                                              F M-211 SH. l                                                                             u                                        GC- 10-3a 0

LO SIS

                                                                                                                                                                                                                                                                                                                                 ~

u' CONT A rNMENT SPRAY

                                                                                                                                                                                                                                                                                                                                 "'                     P668 SUBCOOL ING A        GC- 1-81                                                                                                                                                                                                                                                                                                  M-204 SH. 18 M-203 SH. 2 (8- ll                                                                                                                                                                                                                                                                                                                                        <0-7)                                                                E
                                       ~!

E LC

                                       ~

T 8° -N138M3DR-3214 ~ I

                                                                                                                                                                                                                                         @ LP SAFETY INJECT ION PUMP
                                                                                                                                                                                                                                                 .E.::...6.Til.                                                                         l L--H----IJ---i!>M9-~G~C_-~1-~a~*-~G~c-_,_1-~1~2~*-'-~----------1                                3ZAZ4 c CONTAINMENT
                                                                                                                            ~                                                                 SPRAY VALVES
                                                                                                                        =~~

L.C. TEST INTERLOCK CONNECTION 32H24 C TO SPENT 8 t FUEL POOL <D t5 C33 u

                                                                                                                                                                                                                                                                                                                                      "'

ELO SIRW ~ - ------TR-0351 D CONTAINMENT (M-204 SH. IB( ~ ~<C-8> D

                                                                                                                                                                                           ~

SPRAY GC- 1-8 1 GC 10' C0~ LHT 0 CC-7> ( M-2"3 SH. 2 (

                 <C-1l                                                                                                                                                                                                                                                                                                                              42-2 c

F~c 8 1 -N26M3DR-3205

                                                                                                                                ~

0 u A ELC c TEMPORARY

                                                                                                                            ~

CONNECTION CONTAINMENT GC-8-8 FOR SPENT SPRAY PUMPS LC FUEL POOL COOLING M-221

                                                                                                                                                                                                                                                                                                                                                                        <A-5>

C33 u CHP c COMPONENT COOLING WATER

                                                                                                                                                                                     --'

w C03 c

                                                                                                                                ---------------h M-2~G:.S~H.                                                                                                                                                 OBA SEQUENCER                        L   -

00 2-23 9

                                                                                                                                                                                                                                                                              ,--;---1~-"""-'-"~1J 2 (                                                                                                                                                                                cs E-608       ~      )
                                                                                                                                ----------------i---*'
                                                                                                                                                                                                                                     ~

WEST SAFEGUARDS ROOM SUMP

                                                                                                                                                                                                                                     '"u' M-211 SH. 1 GC-1-12a
                                                                                                                                                                                                                                     "'                                                                                                                                             Also Avsneola on SIRW (M-204 SH. IB(                 Aperture Card CA-7l B                                                                                                                                                                                                                                                                                                                                                                                                                                    B 52-111
                                                                                                                                                                                                                     ~

o..:ao ::i 9 5 0 9 12 0 0 1 5 '"CJ{) 0. I CA 233A 9g:; lil:;;;:

                                                                                                                                                                                                     -               t:l                                                                                            ~: ~ g                                   THIS DWG. IS ISI COLOR COOED.

CA 233 52-206

                                                                                                                                                                                                                                                                                                                           "

. It. G~UPtoLOYM.-7----{L>.<OJ-.~---L...C*.~

                                                                                                   --1ijf/

JD-2-Y.:: 1 5 s 5 NOTES1 WEST ROOM EAST ROOM 1-- REV DAT CHANGED GATE VALVE 3235 TO GLOBE VALVE PER OCR 94-613 DESCRIPTION HMW 8'1' APP L REFER TO DWG M- 209 FOR SAFEGUARDS PUMPS COOL ING C33 LP SAFETY INJECTION ~Cfv5Cf~tf 2. REFER TO DWG M- 211 FOR SAFEGUARDS ROOMS SUMPS

3. REFER TO D\llG N555-696- 1 CCE DWG 2'=166-0-3204) FOR L. P. CONSUMERS POWER COMPANY M-21213 SH. 2 GC 12 1 GC- L- 12 1 relb~9relb SAFETY INJECTION PUMPS VENT AND DRAIN PIPING. PALISADES NUCLEAR PLANT m-4l o

c N

k ~

                                                                                                                                                                                                                                           '        4, REFER TO DWG N555-6CJ5- 1 <CE mm 2966-0-512182> FOR H.P.

SAFETY INJECTION PUMPS VENT ANO DRAIN PIF'INO. COVERT. MICHIGAN M21214-1.0GN A SHUTDOWN COOLJNG PREVlOUS RECORD ISSUE

5. REFER TO P&I DIAGRAM M-225 FOR CONTROL AIR SYSTEM DETAILS.

CROSSOVER TO SHUTDOWN COOL ING

                                                                                                                                                                                                                      <r- co-
                                                                                                                                                                                                                      ,....V> <.DV>
                                                                                                                                                                                                                                        ~<r ov 6.HCC-101-2 1 & HCC-102-2 1 SHALL BE TIED INTO EXISTING PIPING & INSTRUMENT DIAGRAM LINE                                                                                                                                                                                                        ~~ ~§:            ff:~             ~~p~~~ ~~E~Li-~~;* r~* R~g5~~~D~OOT VALVES ADJACENT TO TIE                       ~-'~'"~'"~"~'*'~*"*~-jSAFETY INJECTION CONTAINMENT SPRAY THIS OWG                                                                                                                                                                                                                                                                                                                                                       AND SHUTDOWN COOLING SYSTEM (E- D                                                                                                                                                                                                                                  7. s~~E~~'r~~ ~~~~6~ES ARE     q ES"  AND TAGGED AS BUILT UNLESS A I M204 8                                                         7                                                  _6_                                                          5                                                        4                                                 3                                                 2                                                         1

I I 1

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                                                                                                        -     - EASTRoOM      -i - -Q~                           C03 I2~2ig I MO 3189
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                                                                                                                                       "'~                             GC   HC ELC F                                                                                                                                                                                         ~                                                                                                                      F
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CONTAINMENT SUMP

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C33 I

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3286 E ENG. SFG. SUMP HC-1-1' ~* E M-210 SH. 1 CA-8) DC-1-4'

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N DIRTY WASTE D ORA IN HEADER D

                                                                                                                                                                                                                                                                                   -~
                                                                                                                                                                                                                                                                                                       <F-8l
                                                                                                                                                                                                            ~

J ~ 1 CONTAINMENT 14* -N2qM2DR-322s SPRAY PUMP

                                                                                                                                                                                                            ~

P-54A (Q-4) LO c L_OR OBA

                                                                                           ~-s_rs_~1r SEQUENCER 1

2 -NL10M3-3188 C33 c CC-6-3' CHARGING PUMPS 6 1 -N234M3R-31B3

                          -----j;oo*---'
                                         ~r
 )M-202 SH. 18)-----'~~~-                                                                                                                                                                                       NOTESo CB-4l                                           POS 312136                                                                                                                                                     1. REFER TO OWG M- 20'::1 FOR SAFEGUARDS PUMPS COOL I NG.

2. REFER TO OWG M- 211 FOR SAFEGUARDS ROOMS SUMPS.

3. REFER TO DWG NSSS-6CJS-1 {CE DWG 2gss-D-3204l FOR L. P.

                                           ~

SAFETY INJECTION PUMPS VENT AND DRAIN PIPING.

4. REFER TO DWG NSSS-6G5- l CCE OWG 2C366-D-50821 FOR H.P.

SAFETY INJECTION PUMPS VENT ANO DRAIN PIPING.

5. REFER TO P&I DIAGRAM M-225 FOR CONTROL AIR SYSTEM DETAILS.

C33

6. HCC-101-2 1 & HCC-102-2n SHALL BE TIED INTO EXISTING PIPING AT EL. 63C!' -121 1 LOCATE ROOT VALVES ADJACENT TO TIE IN FOR 1 WET TAP 1 IF REQUIRED.

B 7. ALL VALVE PREFIXES ARE 1 ES 1 AND TAGGED AS BUILT UNLESS OTHER WI SE NOTED.

                                                                                                 @HP SAFETY INJECT ION PUMP                                                                          WEST ROOM GC-4-4"            .E=..fil?B 9so9i2oo1 s -                                          OCf THIS DWG. IS ISI COLOR COOED.                                   REV, 8'-N1"M30R-3187 TO 3' PER OCR qJ-638
                                                                                                                                                                                                                                   !----'~~.-~~~~~~~~.,-,-::-~~....,A A
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• CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT COVERT, MICHIOAN

5 l

                                                                                                                                                      ~-----]VENT ES 33718
                                                                                                                                                                                                                                                              ..

H H SPRAY RECIRC.

     ) M- 204 SH. 1)-..,

IH-2J REC I RC M-204 SH. 1 G CH-2> I

                                  ~

SPENT FUEL POOL HC-4-6 1 M-221 SH. 2

                                 <C-Bl F

• SEE NOTE *q F GC-4-sn P668 Also AvaU£ibfte on E M-21214 SH. 1 Aperture Card E OEDT P668 SUBCOOLING GC-4-4 1 1" -N17B'(R-ES 502 M-204 SH. 1 RADWASTE

                                <E-2)                                                                                                                                                               M-210 SH 18
                                                                                                                                                                                                        <C-3>

D l FUEL POOL D

                                                               @l?'

COOLING PUMPS SUCTION HC-4-6 1 M-221 SH. 2 HCC-102-2 1 <C-4> J' )I H H eves icr-:E-~h HC-20-3' M-202 SH lA

                                                                                 ~-D!<J---i LO                                                                        CD-6>
                                                                 ~r=l"'=ti~~-~-~-{B-7(

HCC- 102-2 1 HCC-102-2 1 L

                                                                                                          -~-                ~

c P54C M-204 SH.

                               <0-2)

I

                                                                                                           --==========c='-=
                                                                                                                  ---------1 NOTES1

1. REFER TO DWG M-20CJ FOR SAFEGUARDS PUMPS COOLING.

c EVENT RECORDER 2. REFER TO DWG M-211 FOR SAFEGUARDS ROOMS SUMPS.

                                                                          ~-
                                                          <CLOSED>                                                                                                                   3. REFER TO DWG N555-6CJ6- 1 <CE DWG 2966-0-3204) FOR L. P.

SAFETY INJECTION PUMPS VENT AND DRAIN PIPING.

4. REFER TD DWG N555-6'=i5- l <CE DWO 2CJ66-D-5082l FOR H.P.

SAFETY INJECT!ON PUMPS VENT AND DRAIN PIPING.

5. REFER TO P&I D M-225 FOR CONTROL AIR SYSTEM DETAILS.

C33

7. ALL VALVE PREFIXES ARE Q ES 1 AND TAGGED AS BUILT UNLESS OTHER WISE NOTED.

8. REFER TO DWG E-81211210 SH. 1. 2 & 3 FOR HEAT TRACING OF B L T-'1:1332A & 8. AND D\'/G E-358 SH 5 FOR LT-0331.

                                                                                                                                                                                     'Ci, DUE TO POOR RELIABILITY THE DISCS WERE REMOVED FROM CK-ES-3400, 3401. 3404 S. 3405 PER sc-g2-078, REF. D-PAL-C!2-07g.

P54B HC-3-14n

                                                                                                                                                                                                                                     - 10 M-204 SH. 1 IB-2l 9509120015 CHANGED N.C.VALYE ES-3372 TO N.O.PER THIS DWG. IS ISI COLOR CODED.

DCft q'!l-038. HMW A CE6CRIPTJON 81' Ck N>P A CONSUMERS POWER COMPANY

! PALlSADES NUCLEAR PLANT COVERT, MlCHIGAN

                                                                                        ~~M

a. :; 8 PIPING & INSTRUMENT u1!AGRAM

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t p~==1sAFETY INJECTION CONTAINME":;r- SP~ll*

M204 1B 25 8 7 3 2 1

H H I ~~~Gu~F P~~~~Ss

                                                                                                                                                                                                              - <f1-207 SH. 1c<

((3-4) OPENS Cl 80 PSIG K-7A & B HP SIDE

                                                                                                                                                                                                                                       -205 SH. 2A EB- 5-6 1 MAIN STEAM                                                                               MSR E-CJD
                                                                                                                                                                                                                                                                                         -205 SH. 18 G                                                                                                                                           EB 12 11                                                                                 EB 12 1                        EB-s-au CH-31 I                                                                                                                                                                                                                         MSR E-<::is I                                                                                                                                                            m
                                                                                                                                                                                 '" .

WI EB-s-ea 3>'.M-205 SH. 11§> CH-51 GB 16' E-5A ili t ~ (/) w"' '

                                                                                                                                                                                        -::.                                                                                             M-2!216 SH. 1       ~..,

rnrn 11- ~ ~ '=' 0-7

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K ~ STOP VAC\7'ES E-58 M-206 SH. l

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VALVES M-218 SH. 2 MSR E-gs

                                                          -~

F F L--1 s~r CG-5l -205 SH. 1A CG-2l e: ~~ 2; I I I MSR E-go 0 ~w M-205 SH. 18 I ~~ ~ CG-51 I EE]

                                     !-

LPA & B GLAND

               <H-5l                                                                                                                                                                                                                                                                      SEAL EXHAUST HS                                                                                                                                                                                                                                                          -205 SH. lA ft"         1
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(F-7)

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RD* 11 - E I ~~ M-2185.2 CG-5l

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CG- 11 1= --- D

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STACK MONITOR c c GLAND SEAL STEAM SUPPLY M-205 SH. lA B l/ (0-6) 9 5 0 9 12 0 0 15 - THIS DWG. IS ISi COLOR CODED. EB-6-Bn E-CJC M-205 SH. !B (A-3> EB-3-8 1

J MSR E-CJA

                                                                       ~

3>M-2~ 5_:iH* 8

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PALISADES f\l.ICLEAF? PLANT COVERT, MICHICAN M20'3-1.ClGN

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PIPING & INSTRUMENT DIAGRAM MAIN StiJ~~fNf1Al~Y~TE'1U~IL!ARY

                                                                     "

M205 58 8 7 5 5 4 3 2 1

8 5 4 3 2 1 _. ~I OPENS ON LOSS _ _j _ _J CONDENSER OF DC POWER M-206 SH 18 H <C-4l H 1q BLEED VAL YE MS 500A

                                                                                                                                                                                                                                      ~

PS-0741A B. DD, 2/3 LOW rn -,,_.. _ " "!"oo - -~ _j ~ SUCT lON PRESSURE

 ~~ ~~7  SH. 2  CG- 7l 1

I PBB RUNNING G ~ I M-207 SH. 2

            ~

153 MAIN STEAM EB 4P H.P. SIDE MAIN STEAM E-50A M-205 SH. 2A M-21217 SH 1 CG-Bl CG-Bl 2 1 -130-FW 177 F F 4 1 FW 151 1- r 1 I

                                                                                              ~

E I E I

                                                                                                   '

F. P. TURBINE TRIP AUX. F. P. THRUST SPEED CONTROLLER FT-0527 P- 88 EB-10-t~*

                       >----                                                                                                                                                                                                         HlC-0526                                                                -<f'!- 205  SH. 2A<
      >M-207 SH 2
            <H-6l
                                                                                                                                                                                                                           -    7   M-207 SH.1     >                                                                   lF-6l (8 3l D

AUX. FEEDWATER PUMP FT 0524~ FW 617 4 d

      <M-207 SH                                                            TURBINE DRIVER K-8 CG-Sl                                                                                                                                                                                                                                                                                                       MSR E-GB LP STEAM GB 4- 12 1 M-205 SH. LA CG-7) c                                                                                                                                                                                                                                                                                                                                                               c CC,D-4l M-207 SH.1<

CA-2l

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I HB-40-1 1 WATER

                                                                                                                                                                                                                                             -21217 SH lA CA-5l FOR NOTES SEE M- 205 SH. 1
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11 F"-' 559 1'-19-FW 561 10-25 REV'D. YS-0520 & ID'D. YS-0513 JGD

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CG-6l CH-3l NOTES*

1. SHOWN FOR CV 07811 CV 078121.

F 2. CV 077g & CV 121782 ARE IDENTICAL. SX-IZl77g THAU SX-0782 ARE HYDROSTATIC F TEST TAPS TO S. G. TEST COOLER HB- 17-2~

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                                                                                                                                  ~                                                                   'eY)                                                                                                                               CONSUMERS POWER COMPANY TURBlN   TRIP I

CAUTO-STOP OIL PRESS> PIPING & INSTRUMENT DIAGRAM A FEEDWATER & CONDENSATE SYSTEM

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M-21217 SH. IC

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F 11 -SPtci-1588 A/S M-212 SH. 2 E F-46

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ELD L D I I D I I NOTES1

         +--                                                                                                                                                                                                                                   -+                                                                         l. REFER TO P& IO M- 225 FOR CONTROL AIR SYSTEM DETAILS.

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                                                                                                                                                                                                                                                                                         -207 SH. lC                                       CONSUMERS POWER COMPANY
                                                                                                                                                                                                                                                                                            !H-7, Sl A                                                                                                                                                                                                                                                                                                                                PIPING & INSTRUMENT DIAGRAM                 A FEEDWATER & CONDENSATE SYSTEM CONDENSATE GB- c:i- 3 1 M-207 SH. lC SEAL WATER                                                                                                                                CG~ ll TO DRAIN                                                                                                                                                                                                             M-207 SH. lA

~ TANK T-2SA M-21215 SH. 2 HB 1 '

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                                                                                                                                                                                           >

NOTES1

1. REFER TO P&IO M-225 FOR CONTROL *PIPING & INSTRUMENT DIAGRAM A

                                                                                                                                                                                                         ~

A lR SYSTEM OETA ILS INGERSOLL-RAND CONN. NO.' S CF. P. 5q35-FEEDWATER & CONDENSATE SYSTEM

                                                                                                                                                                                                         \::5' M4-4. I-R DWG:  N4 210RET501Xll IZJ"l51ZJ        M-207 SH. lB         22

8 I 7 5 4 3 I' 2 1 "' - PUMP P-8A RUNNING

                                                                                             ,----                                                                                                                                                                                          NOTES                                             ._.,
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                                                                                                    ~I I                                                                                                                                      SEE P & lD M-20710> SKEET J, H             ~~E~F,;,\'~E {00--.--

LOW FLOW LOGIC TO START P-BB CSEE NOTETl

                                                                                                                                                                                  !FOGG A
                                                                                                                                                                                  !FOGG B I

I MEANS STEAM GENERATOR E-500 IS ISOLATED. MEANS STEAM GENERATOR E-50A IS [SOLATED. H OR P-BC I I -filTI 2. REMOVE STRAINER lNTERNALS DURING NORMAL OPERATION I I ~ ~

                                                                                      <SEE NOTE ll T

I I TO CFM _<NDTE4l

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                                                                                                                                                                                 ~NCTIONS MONJTOR fCFMl, EXCEPT H.O. vAEJ 0pci~r~TJCAL

4. ALL AUXILIARY FEEO'lilATER SYSTEM SIGNAL t(@q '

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POWER COMPANY PIPING & INSTRUMENT DIAGRAM SERVICE WATER SYSTEM M208 is'- 20 7 _J_ 1

8 7 L _ _ _ _ Q _ _ __ 6 _L__----2______~ 5 _ __1_4 _ __L---=-~====r==~~==r===:r::~3 2 1

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SpT-01328

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F !i ' c-,~ ~ I ,._" ' ,,. _L -, F

                                                                                                                                        -----i~~;<

WO

                -

[_______ )p----- -*-- 12' -cc 102 ) ffi=J.< .. !i I

                            ------- --- -----

CPERMIT ALARM WHEN v-0q13. cv-0e7q DR E CV-0880 ARE OPENED E 0 0 Alsso A"l~Uabie on Aperture C6rd c c B B 9509120015 THIS DWG. IS ISI COLOR CODED. 18 ~~* ~4~~2REF. TO P-52 PERC-PAL-'l4-0HlB &

                                                                                                                                                                                                                                                          *ONE                                  DR~"11  J. L. STERRETT A                                                                                                                                                                                                                                                     PIPING COMPONENT  & INSc!J'Jl~IENNGT DIAGRAM                        A SYSTEM M-20g SH.                2      18

.,* 8 7 6 5 4 3 2 l _ _ _ __

                                                                                                                                   .'

H

 < M-20G    SH IG-5>

G G F F

                     ~

m-!Iill

                      ~

I RADWASTE EVAPORATOR M-655 E CF- 1l E D D RADWASTE c EVAPORATORS M-655 c CF- ll MV-CC-565 B B

l. l M'l-CC-154 THROTILEO THIS DWG. IS IS! COLOR CODED.

1~§1 REV'D, CC-162 &164 TO L JGD PER C-PAL-94-04l7A & DgR-94-990 l""~@::=:-:::!"--r~~~rcro;N:;;;-;~"~~;"'~'n~~~:-:-.,.-~~~=-~*~':I~

                                                                         ':-+>_                     SUMERS POWER COMPANY A

PALJSADES NUCLEAR PLANT COVEfl:T, MICHIGAN PIPING & INSTRUMENT DIAGRAM WASTE GAS COMPONENT COOLING SYSTEM COMPRESSOR < _l_ 4 <B-!l M20CJ 3 27 __8 7 6 5 3 2 1

8 7 5 4 3 2 1

 .. _ .._...                                                                                                                                                                                                                                                                                                *_1' H                        3v -CAW 407 H

DEGASIFIER PUMP DISCHARGE HC-2-3 11 HC-2-3 M-210 SH 2 CH-2>

                                                                                              "'

N "' N r-=3*~-~C~R~W~l~l=g~~

                                                                                              ,',

c~

                                                                                ~-----~"'                                                                               3 1 -CRW 121 3 11 -CRW  123                   3 1 -cRw 125 G                                                                                                                                                                                                                                                                                                                             G
                                                                                                                                                                                                                                                ~40                    ~

A/S t-~ I c HS- 1-2" F F HB-43-2" ----~~H~B_-~4~3-~2~'--.-1--~---,--=,,,,----<( M- 222 SH 1 <

                                                                                                                                                                                                                                              ~"7~
                                    ~

HB t-HC 2" -CRW 508 Hc-3q-2* HC-38-2U y 2' t-HC t-HC 2"-CRW 510

                                                                                                                                                                                                                                               ~'¢0 HBC 1" S. I. TANK-VENT M-211 SH 2
                                                                                                                                                                                                                                                                      <G-!l CE-8l E                                                                                                                                                                                                                                                                                                                             E C40~
                                                                                                                                                                              ~

CLEAN WASTE RECEIVER TANKS 0 0 HC-2-6 1 c c GI::f40

                                                   ~
                                                                                                  ~

y~ lQiEJ I~ I T t-~ J:Q;J A/S r-~ I I c1"38 A1s I 1_ B B 4' I.BIP 3' 9509120015-2.{,. C40 RECEIVER TANK CIRC. PUMP P-70 4

                                                                                            -

THIS DWG. IS ISI COLOR CODED 11 -CRW 514 3 1 -CRW 765 L'-J DEDT M210-1A.OON PALISADES PLANT CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM A ADIOACTIVE WASTE TREATMENT SYSTEM A tillI.E;. CLEAN

1. ALL VALVES ON THIS PRINT ARE CAW PREF IX UNLESS OTHER'<<' ISE NOTE0.-

2. RUD- 1018 RMOVED TD ALLOW 0950 M-210 SH. lA 12 CONTAINMENT VENT PATH

I 3 z l

                                                                                                                                                                                                                                                                                                                          ......

H H G G A/S CLEAN 'f/ASTE F-57 A/8 HC-2-3 1 M-210 SH 1 CE- ll I HC"-f-HC HC-2-3° __________________________________ HC-2-3 1 _____________________ 1

                               ,--------------------------,-/:-                 ~---------%-'--C-R_W_8_0_5                                                                                                    ~

Hc-2-30 '1-~7 ~ HCC-45-3' ~~§'f~~E~u~~NK

                                                                                \ ..._/~                                         M-2 ~~ ~H 1 1 5 3' -CRW 751 F  CLEAN WASTE CLEAN WASTE TRANSFER PUMP                                                                                                                                                                                                                  F F-57 A/8 M-21121 SH 1 co- ll HCC 3 '

HC 3' M-650 SH. 18 (A-8) AN STEC JB 1~ 11 3 1 -CR'J 13 APERTlJRE HB HC-32-3 1 tYP l' [ CA,RD 3" 31 -CAW 13 3" 3 1 -CRW 411 E=..li9ll I \ Also At:aUable en E C-40@\ Apertur~ C:trd E 42- I RFCF~TANK 4 I -~_JR_!.P, E::::..0fl C-40 ~ '*/

                                                                                                                                                                                                                                                                    \&('

3° -CRW 128 I I 3" 2' 3" 3 1 -CR'W 174 3 1 -CRW 3 1 -CAW D D c~ v~ C40~! CLEAN RESIN TRANSFER TANK

                                                 ~                                                                                                                                                                                                                                                                         L:fil c                                                                     IOl 6

ID>

                                                                                                                                                  "'~

IDl c Ll

                                                                                                                                                  ~                                                                      "'a."
                                                                                                                                                                                                                         ~
                                                                         "'                                                                                                                                              "'~

sa.."' :c:. r;-i

ir:::i Cf)

                                                                                                                                                                                                                                     ~

(/)~ ~ 9 Ll ' ii!"' 210 PURIFICATION & DEBORATING u

                                                                                                                                                                                                                              "'                                                                                  ION EXCHANGERS M-202 SH. 1 r

HC-32-3' B PURIFICATION DEM I NERAL I ZERS DRAINS M-21212 SH. 1

                                          <B-1)

DEDT 9509120015 - THIS DWGw IS ISi COLOR CODED M-210 SH. 1 CB-81 1' r.i

                                                                                                                                                                                  -~;~~~~~~~~E=--{1~~9}---'---1'""1;-----J T-Sg C-40@- Pl 11ZJ64 3 1 -CR'N 147                                                                          3 1 -CRW 15                       M-657 tB, C-Bl NONE PALISADES PLANT STE~flETT CONSUMERS POWER COMPANY A                                                                                                                                                      HC- 13-3P T-Sg                              PIPING & INSTRUMENT DIAGRAM M-657                       RADIOACTIVE WASTE TREATMENT SYSTEM CO-B>

CLEAN tillI.E.;. ALL VALVES ON THIS PRINT ARE t;.Bil PREFIX UNLESS DTHERW I SE NOTED.

                                                                                                                                                                                                                                                                                  @J             0"350         M-210 t-----~-----'-----------c,<iL:-_

SH. lR 11 M2 10- lB. OGN 8 4 3 2 1 P&ID

I H H G 3 1 -cRw 151 G RADllASTE FILTER F-57A OR T- 5q HCC-58-3' HC-2-3* 3 1 -CRW 175 M-210 SH 18 LCI CB-31 HJO...W;!CC 153 31 -CAW 748

                                                                                                                                                                               ~
                                                                                                                                                                               ~~~=<i HC-2-3' DI sC*h'1'.t'l.T~ASPUMPS ~~~lb
                                                                                                                                                           -650 SH. lB         u~    2:
                                                                                                                                                               <G-2>           ....

HC-2-3 1 8 HC-2-2 1 LC 31 -CRV 173 HC-2-2 1 F z* -CAW 170 F C-101 CLEAN WASTE DISTILLATE SYSTEM J. _ - H-650 SH. 1 I IB-2> E 31 -CAW 167 168 E HC-2-3 1 HC-2-3 1

                                                                                                              ~C40 Hc-2e-1*                                                                                     TREATED WASTE MONITOR TANK TREATED WASTE MON !TOR TANK Also AwaUab!e on D                                                                                                                                                                                                                                                Aperturta Card                           o, c                                                                                  1- - - - -

c tiQIE_;_ CILL VALVES ON THJS PRJNTNAO~~Dc:fil!'. PREF IX UNLESS OTHERlllSE

• c-<@}c 42' DILUTION LINE

                                                                                                                                                 '------<- -      zM-6~ ~;,H* 3 8         >

B HC0-23-2' B 3' L.!.J 9509120015 LOCAL THIS DWG. IS ISI COLOR CODED. L - TRIP CLOSED - - + ON HIGH RADIATION 10-18 REMOVED SEISMIC FLAGS PER OCR q4-q0z g q4 HMW LAUNDRY DRAIN PUMP REV DATE DEStRJPTION BY Ck APP cJ & FIL TEREO WASTE

                                            ~L----~(MO]NEI~TIOLR     P~UtMtP~

M-211 1SH. 1 CONSUMERS POWER COMPANY A PALISADES NUCLEAR PLANT CB- 1> COVERT* MICHIGAN MZ10-1C.DGN PIPING & INSTRUMENT DIAGRAM RAD IOACTIVE WASTE TREATMEN;,; SYSTEM CLEAN r- ,

8 7 6 5 4 3 2 1 LAUNDRY DRAIN TANK T-70 H C40

                                                                                                                                                                                                                                                                                          @7----

1 G ~ G

                                                    ~ ~

_r- ~ r-3*-DRW 130

                                                                                                  - - - - - - - - - - EMERGENCY SHOWER
                                                                                                  -""=--------                               ~~~~s~n~JlNTROL HC-12-4' F

HCD-127-2* HB HBO HC-1-2' l'-ORW

   >M-218 SH,2>- -
                                                                                                  -~------~CASk                                   WASHDOWN AREA DRAINS y

ODWH CH-4> I 2 0~S~HQ!3';:COTROLLED CHEM. L.!:H!f;Cc:;-l!:-fl.__<!)M!::-[!_Jl LAB, DRAIN TANK DRAINS

                                                                                                                                      <F-6>

PUMP LEAKOFFS BELOW EL 5q0*-0* LAUNDRY E HC-18-4" VCT RELIEF

                                                                                                                                                                                                                 -202-SH 1 EQUIP.

E CF-SJ AUX. BLDG. VACUUM FLOOR DRAINS HC-1-3' DEGAS FIER RELIEF BELOW EL. 5q0*-0* M-210 SH 2 CF-5l 2 1 -RW'-142 TURBINE BLDG. 582) HC-1-2* LC He-1q-2* su~~~MP DlSCH. HC-l-.HB <C-3> WASHING MACHINE SUMP PUMPS RW 114 DIRTY WASITmTER F-53 1' -DRW 510  ? l-__,~--------1*1--------, D I -DRW 124 VGCH Al.. AYsn~ble on D RW 603 M-211 SH. 3 C4~H

                                                                                                                                                                                                                                                                                               ~

pertMre Card 53 L PUMP LEAKOFFS

                                                                                                                                                                                                                                                                                               ~40
 <BELOW EL. 5q0'-0'>
                                                                                                                                                                                                                                                                                              ~

AUX. BLDG. FLOOR DRAINS

 <BELOW EL.5q0'-0'>

3"-DRW 788 T-78A <EAST ROOM> c ENGINEERED SAFEGUARD ROOMS SUMP PUMPS VENT HCC-80-3' c SPENT RESIN STRG. TANK HCD XCD z

               <T-100>

HCD-17-3" VGCH M-655 FILTERED !i CLEAN Wl>STE RECEIVER TANKS CF-71 ..h'.ASIE MONITOR PUMP RADWASTE HC-1-2" DISCHARGE HEADER HC- 1-2' -210 SH lC MISCELLANEOUS I Rl>OWASTE CA-6> ( 2 1 -DRW 137) SYSTEM 2"-0RW 79

                                                                                                                                                                                                 <M-851 SH. J E                                                                                       MISCELLANEOUS WASTE B                                                                                                                                                                                                         <D-BJ                        HCC-5-3" HCC- 10-31 DISTILLATE SYSTEM M-651 SH, IB B

AUXIUARY BUILDING IG-1> SUMP PUMPS CONSUMERS POWER COMPANY

                                                                                                                                                                                                                                                                                                                                      -

PALISADES PLANT PIPING & INSTRUMENT DIAGRAM A 9509120'015-l-b DIRTY WASTE & GASEOUS WASTE A THIS owe;. IS lSI COLOR CODED 0950 M-211 Sl'i.l 8 7 6 5 4 3 2

7 -1 6 5 4 3 2 1 H H RE-1113

                                                                                                                                                                                                                                                                                                                                                 -1 AND STACK M-211 SH 3
                                                                     ~                                          WASTE GAS SYSTEM                                                                                                   <G-9)

L!!-fU PURFICATION & DEBORATING ION EXCHANGERS VENT HB-1-1 112' VCT VENT M-202 SH 1 M-202 SH lA <D-2l G <G-8> G RADWASTE EVAPORATOR VENT M-650 THIS OWG DIRTY WASTE CD-Bl DRAIN TANK RELIEF HB-1-1 1/2' HB-1-4' M-210 SH 1 HB-1-1' 22 HB I HBC CE-4l T-68'S & T-101'5 HBC-35-2' RELIEF VALVES EQUIPMENT DRAIN TANK RELIEF

                                                                                                                                                                                                                       ~~~~~~~~~~~~~~~~~~~~~~~~---<( M-211SH                                  3<

HB-1-1 112' THIS OWG CF-81 M-210 SH I (E-Gl

                                                                                                                     <E-Bl                                                            M-211 SH 3 VAC~ DEGASIFIER                                                                               HB-20-2' VACUUM PUMP DISCHARGE F                                                                                                                  M-210 SH 2                                                                                                                                                                                                                      F:

I 2"-WG 120 QUENCH TANK VENT HB-1-1' M-201 SH 2 RV-1111 CH-SJ THIS DWG 1' FROM COMPRESSOR IF-4) CONTROLLER WASTE GAS I E CLEAN AW TANK SURGE TANK 112' WG 513 I E VENT HEADER HCC-1-3/4' HBC-23-1'

    -210 SH lA                              LOCATE                                                                                                                                                                                                                             I
        !E-2>                                PILOT                                  DT-1113         HBC 22 1"                                                              I 1'-WG 724              IN AIR ROOM                              THIS OWG IF-3l F-58 DRAIN M-218 SH 2
                                                                                   <C-Bl SAFETY INJECTION TANK VENT                                                  1'-WG 723 M-203 SH 1
       !H-8>

D D I WG 516 112" WG 514 I I I I WASTE GAS DECAY TANKS __ ----~- _ _BY VENDOR _ _ _ _ _ _ _ _ _I M-211 SH 3 (C-8l c c HCC-123-1' DEDT T-80 HCC-125-11/2' M-211 SH 1 (Q-8) 2"-WG 706 GASEOUS WASTE SURGE TANK HBC-21-l" M-211 SH 3

                                                                                                                                                                                                                                                                                                                        <F-4)

WASTE GAS B COMPRESSORS 517 OT 1164 B HCC-125-1" DISCHARGE M-211 SH 3 CB-7l HMW - CONSUMERS POWER COMPANY PALISADES PLANT

                                                                                                                                                                                                                                                           -21 MZll-2,0GN PIPING & INSTRUMENT DIAGRAM                          A 9509120015                                                      RADIO-ACTIVE WASTE TREATMENT SYSTEM GASEOUS WASTE Cl>A~IC:l.   ,"'{

THIS DWG. IS IS! COLOR CODED M-211 SH.2 0'150 7 5 5 4 3  ?

NOTES1

1. ALL VALVE PREFIXES ARE n CA 1 UNLESS OTHERWISE NOTED.

2. ALL PLUG VAL YES ARE 3; 4u H PS- 1203 UNLESS OTHERWISE NOTED. H 18-7) >THIS DWG >-- -

52- 1107/CS C-35

                                         <C- 7) <  TH IS OWG  $--- -

COMPRESSOR PS 1213 PS- 1202 CONTROLLER IC-7l >THIS DWG >---- - CSEE E-167) - - - - - - < STBY START THIS DWG < 52-1207/CS <C-4l l0-8) >THIS DWG >-- G 1,============ I I i------ - --< PS-1214 C THIS DWG

                                                                                                  <C-Bl
                                                                                                           <

G 1 I f '----~~~~ I I Pl 1200A*.__ _ __ L I I I I FILTER I I s ILENCER F I I II F C-13 I I l I -£E- s

                            -

~

I1 I I1 I I1 L ___ _ I1 I1 I1 E I1 E I1 INSTRUMENT A IR

ol'!f:;-~*~------C~J----------~T~U~R~B~I~N~E~B~LDlGC.>

                                                                                                                                                                      '--------'-----'---------_JI I L. -   -    _ - _ - _ - _ -_ _ _ _:JJ      ~I                                                                                          M-212 SH. 2 IH-Sl
                                                                                                                                                                                                                              ~                                                             INSTRUMENT AIR N                                                            AUX IL I ARY BLOG.

D < C-35 THIS DWG D

                                                                                                                                                                                                                              -,                                                                  M-212 SH. 3
                                                                                                                                                                                                                                                                                                       <F-31                                                                      D IG-7>

1 I -263 INST. AIR TO

                                                                                                                                                                                                                                                                                                                                      'N'EST SAFEGUARDS M225 SH 1
                                                                                                                                                                                                          -MV-CA80~*
                                                                                                                                                                                                                                                                                                                                               <H-8>
                                                                                                                          -5028~,.                                                                 (314'                }             t 3 / 4 ' -MV-CA803)
                                    ~5-                                                                                                                                                                                 x
                                                                                                                  ' - CA C-35 C-35                                                                          --'=~==""--'        ~~

8;: ~ g

                                                                                                                                                  )> THIS OWG >                                                                , c""o"'o"'L'-'I'-'-N,_,G, __T!..!O"'-W!!..=E!..!R_,_P_,,UccM.!!.P__,H_cO"'-U'=!.>!.S=.E
                    >THIS DWG      > -                                                                                          ...                   CG-SJ TURB. BLDG

• c CG-7l I Cl PI 1202A*.__ _ __

DELUGE SYSTEM PS-1341 & PS-1344 FILTER M216 SH 2 SILENCER B

                                                                                                                                         *__[j:ji]
                                                                                                                                        .-~
                          < THI~-~=G <-

CG-7) _j 9509120015-~~ THIS OWG. IS ISI COLOR CODED REMOVED CA-705 PER OCR g3-1185 ITT A <Dl CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT COVERT,MICHJVAN

 ---*                                                                                   DRAIN PIPING & INSTRUMENT DIAGRAM SERVICE & INSTRUMENT AlR SYSTEM CONTAINMENT                                                                                                                                                4.

BLDG M-212 R 7 6 5 4 3 I_ 2 1

I l CV CV cv~v CV CV CV 200a2117 2001 2004 2115 2003 2113

                                                                                                                                                                                                                                                                                                                         .*.
                                                                                                                                                                                        -<B                  13 H                                                                                                                                                                                                                                                                                                                                          H
                                                                                                                                                                                                                            *'

CV CV 1058 3044 ~

                                                                                                                                                                                --.....----~~-~

CV 1028 i----~3'fs

                                                                                                                                                                                    <CV-3046>

CV CV 1030 3065 CV CV 112133 ll'Zl24 NEAf!. Y..! 1~-4 G CV CV G 1031 1023 CV CV RE 1063 1056 1917 CV CV CV 101q 1013 3040 CV CV 1018 1012 CV 3043 F F E/P CV 3005 CV 0867 CV 3084 CNEAR VHX-2> IVl f - - - - 0~~4 TO CA-2'qq E cv*s 1100A&8 EAST ESS M-225 SH. I N2 M-222 SH. 2 E SJ TANK CH-1> VENT CV' S CV CE-61 CTO STACI( 0148 OR WGSTl CA-221 CV CV M-212 SH. 3 0157 3003 <G-8> CV 1002 D CV 1007 CV D CV 112L3 1036 I CV CV 1114 CV CV CV 1038 CV CV CV HlC I 121D 1121C 11218 lll'lC lllCJD lllCJA CV 1123 1044 CV 1045 L_

                                                                                                                                                                                    ---

CV 602 IS' PIPEWAY 1001 c CV 200q CV CV cv MIDDLE CORRIDOR CV CV CV CV I 120A I 1208 I 120C I 120D c 1037 1004 1102 602' P'lPEWm:::iY CV 1101 M-212 SH. 5 EVAP DIST CG-6> _TANK .fil:L EVAPS CV B 5018 CV B 5020 f----c-v.,.---.---.---5~~4

                                                                                            ~148 CV 5133 CV      CV
                                                                                                                                                                                                                                                                                           -L..C\

CV CV CV CV 5025 5135 CV 5143 5140 5138 5141~~~==- CCV CV 5132 5007 5136 CV 5122 9509120015 Tl-US DRAWING JS ISl COLOR CODED. 11T ID> CONSUMERS POWER COMPANY A AUX BLQG AODJTION HEADER Plll.JSADES l'tlJCLEM PLANT CD'IERT, HICMIGAN HZlZ-4.DGN CV 5128 PIPING & INSTRUMENT DIAGRAM INSTRUMENT Al.fl WALK DOWN M-212 4. 8 7 ___I_ 6 5 3 2 1

         ',                                                                                                                                                                                                                                                                                      ... f H                                                                                         TO FLOOR DRAIN BIO BOX CORROSION TEST ASSEMBLY
                                                                                                                                                  *-s11 !101
                                                                                                                                                             ~!3610A)
                                                                                                                                                                *-sw s10 IEAOER B                                                                                                                                        HB-23-24"                                      HB-23-12" A/S G

NON-CRITICAL SERVICE 0 llATER M-208 SH.I

       <H-5>

3"-54-SW 10'5 STAND-BY PUMP

                                                         - - - - -,                                                                                                                                                                                          ----*-->   START I

CIRC. llATER LINE HYPOCHLORITE I INJECTION DIFFUSER M-663 SH. ( -4> Q) F 3*

                                                                                                                                                                                                                                                                                  --~-->F
                                                                                                                                                                                                                                                           ---g *"/'~

I';! INTAl<E STRUCT. !l!l'iJ UPSTREAM TRAVELING SCREENS I SW 332/ I

                                                                                      ~

PACKING / I P _- _ 7_ c LEAl<-OFF L. - -, /

                                                                                                                                                                                                                                                                                /
                                                                                                                                                                                                                                                                                           /
                                                                                                                                                                                                                                                                                                   /

TO FLOOR I / CHLORINATOR H11-Sll 342 CONTROL PANEL C48 -, / HY-SW 662 I E L.C. E 820 HY-Sii 821

                                                                                                                                                                                                '     '~""'

SERVICE WATER PUMPS

                                                                                                                                                                                                                   /'

I ( APER1~U ~~~EH:]334[)* LOCAL LI I @C-13 I

                                                                                                                                                                                                                  ~                                     CAfU)aiP~~

301 D I._ ) 0 SCREEN WASH SYSTEM 58 Also AvaULPib!c on Aperture Card A/S

                                                                    *-CHM 103 c                                                                                                                                                                                                                                                                                                                                            c FILL <OUTSIOEI 11--1111---------,

I *- HH 4 SCREEN WASH I "-CHM PUMP P-4 B B THIS DWG. IS ISi COLOR CODED. RTD-1-2" INTAkE STRUCTURE

                                                                                                                                                                                                                                  ~                      CONSUMERS POWER COMPANY 2" -CHM 752                                               .:re-21-4*                                                                                                                           ~                                 P.tL.16AE:JES N.J:U:AR FUNT tmm::;iii*t---;:p:-:-1:::P::1N:G::--:s.:--~::;:;::::~::*;;"':W::~::;=N:::T::-:D::r:--:A::G::::=(:i.::;-=:=:'i A                                 T-lSA S. B HYPOCHLORITE STORAGE TANKS 9509120015             -30                        l::e1i111!!1m~SERV!CE w~~BRcREo~rrsiATSO UC RE, M213 A

7

                                                                                                                                                                      ~~
        *:                                                                                                                                                            y             ~
                                                                                                                                                        <THlS1 Y,~o. ,...._......,_.....,..,rrHr~sgVG>

JB-23-20" I H + H LO

                                                                                                                                                                               ..

I T-3'1 H-655

                                          <E- ll                                                                       <H-8>

T-26 ./1!, M-215 T-24 <E-91 16 SH 1 TO INJECTORS G <G-8> G F~l~~Elo, K-!'I VENT T-40 H216 SH l

                                                                                                                                                   * .. 0130-0E                                                                                                                                                                ttttttt!t

.  :.

I0-81

                                                                                                                                                                      ~@                                                                                                                                                                           ~NOTE te-5-2'                                                                   y             y                                                                                                                                                                        2 HS 1415 & 1452 lTHIS DllG. I LS 1417, 1419, 1462 & 145'3 CTHJS OWG.>

HS 5353A & 1414, LS 53!'13 & 130S lM-2161 HS 1413. LS 1506 & 1507 IM-2151 HS 5600A. LS 5600A & 56008 lM-6551 t' - -FO 111

                                                                                                                                                         *THI'S1?twa.) ...._.,_,__,._,._

I

                                                                                                                                                                                + lTHIS  P18A
                                                                                                                                                                                             <THrd  8Cwo>

RETURN* +++++++1+ TO INJECTORS D\IGI l I . A DETAIL 'A' LOCAL ,. __ j ALARM

     <lll'll
                                              !- - -                                                                                HB-5-2'

r~

                                                                                                                                                                                                                                                               ~

LO EL. 615' FUEL OIL SYSTEM F F I'. -1

                                                                                                                                                                                                                                         ~H                    LOCAL
                                                                                                                                                                                                                                                               <3014) 0 CP-'165 YLL H-C!l07 IG- ll                                                                                             T-24 I                                                                                               H-216 SH 1 HS
                                                                              ..

1419 & 1453 <THIS DllOI

                                                                                                         <F-91 T-40 HS  5353B & 1417 lM-2161            H-216 SH 1 HS   1416 lM-2151                      C0-7)

I -0130-0E F0-402 LO K-SB HS 5600B <M-6551 te-5-2' OIL RESERVE I E I DIESEL OIL STORAGE TANK 2' -01'1-DE 102 lSEE TYPICALI E Em EMERGENCY PIESEL GENERATORS I PAY TANKS I

    @-----@                                       START-STOP COMPRESSOR t          t                                                                   JB-24-1 D                                                                                                                                                                                                                                                                                                                                                                      D FILL LOCAL 13007~

c SYSTEM i BECHTEL ml.ES.; c ENG AIR HANIFOLO PRESSU>E IEAMP-21 ---+. I. ~Dp!:',J1:;{ Rtc!lk"c.'rilfJ~E PLATE WITH "6" HOLE 2* llt.'f;;E~~Wi'o:E~fe:l, i~xWi~ E' g~~E~"J.'l~fiWis's1~fR 1 1 RUNS ATOP DIESEL HOUSING. JACKET WATER 3. TWO TU>BO CHARGER INLET LltES ANO OUTLET COOLER LINE TO DIESEL ARE FLEXIBLE CONNECTIONS.

                                                                                                                                                                  ~300~

B I I 9509120015-~\ B THIS DWG. IS ISI COi.OR CODED. JACKET WATER SYSTEM REVISEO OFF SHEET REF. !CONDENSATE MAKE-UP PRE TO T*BI. M220 SH. I. 0*81 PER OR F-CG-'12-021 & LUBE 46 '44 OCR '14-102" HMW-TIMER T~ olE~LL 8

                                                                                                                                                                                                *  '++'++'++'+'+'+                                                                                           DESCRIPTION                             BY'  CIC   WP
                                                                                                                                                                                                   *********

TO MAIN BEARINGS CONSUMERS POWER COMPANY

                                                                                                                                                                                              , TO PISTONS     12 PER BEARING!,.

A PALl~~~T~~~:;~c:.:-ANT H214-J.OGN A

                                                                                                                                                                                                               +

INTERNAL SUMP PIPING & INSTRUMENT DIAGRAM QETAIL '8' LUBE OIL, FUEL OIL & DIESEL GENERATOR SYSTEMS*-*

• • ~ -"***

114881 INSIDE DIESEL M214 46 8 7 4 3 2 1

8 7 3

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          ,,;                                                              CD-7l                                                                                                                                                                                                              2. DAMPER P0-1546 HAS ITS POSIT!ONERS HTG. VENT. & AIR COND.

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o.o. I ( NOM. l JOJ ' JDJ-1-2

• I 3. UNIT STARTS S. LOADS ONE CYLINDER Q TEMPERATURE i-- __________ J AS DETERMINED BY PLANT (TB). LOADS THIRD S.

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8 7 5 5 4 3 2 1*. H H P - N 208 BULK N2 SUPPLY G G 1'- N2 184 ) CLEAN WASTE RECEIVER HB 111 I HB-43-2 1 TANKS M-210 SH 1A

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CIRCULATING WATER MONITOR THIS DWG. IS ISI COLOR CODED.

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                                                                                                     ~                                                   I                     ~                      ~161                   I                                          SAMPLE             ZERO            SPAN 1

2 55 60 300 300 55 300

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l. HEAVY LINES INDICATE PRIMARY SAMPLE PATH.

F CllA l 2. TUBING TO BE SERIES 3121121 STAINLESS STEEL F

3. HYDROGEN DETECTOR 1 1

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8. SV-2424A & B OPEN i CALIBRATION GAS e-- -

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L ________~-~~-=-v_+--==-=~~~~ REAGENT OR CAL GAS LOW PRESSURE REAGENT GAS 100% 0 2 i I PALISADES PLANT CONSUMERS POWER COMPANY M224-2. OGN A

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100% 1212 FSO-CONTAINMENT

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               <E-3>

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                                                                                                                                 <A-SJ HB0-24-% 1                                                                                                                                  THIS DWG. IS ISI COLOR CODED.

STM. GEN. REC I RC. RETURN M-226 SH. 18 Bl' c~o APP CC-11 STEAM GENERATOR CONSUMERS POWER COMPANY NOTES BLOWDOWN PUMPS PALJSACIES NUCLEAR PLANT COVERT, MICHlVAN l TI M226-1.DGN krRR§~~~E~OD~T~rC~ M-212 SH. l THAU 4 FOR CONTROL 2626 PIPING & INSTRUMENT DIAGRAM STEAM GENERATOR BLOWDOWN 9!5"09120015 MODIFICA T!ON I:' M226 42 8 7 5 5 4 2

8 7 5 5 4 2 1 CVl CLOSE VALVE ON COOLING TOWER PUMPS P-3c:':IB TRIP

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