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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.... .. ' ,. ... and Suppoq§ . . _* .. . . . ' . . .... . . NUCLEAR PLANT . <;) consumers Power POWERING M.IUllQ,llrS 9509120015 950906 PDR 'ADOCK 05000255 Q PPR ... . " ,' . : : ,,,,-, .... --** ** . *) 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 .,

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 A. Historical Background B. Upgrading Criteria C. References D. General E. Inspection Section 2 Outline of the Third 10 Year Program A. Vessels B. Piping C. Extent of the Program Section 3 Basis Statements A. Exemptions B. --Deferrals C. iAugmentation D. Additional Examinations E. Consumers Company -Palisades Nuclear Power Plant: Position F. Repairs and Replacements G. Containment Penetrations Section 4 Palisades Plant Code Use A. Code Cases for the Third Interva1 Section 5 Relief Requests Section 6 Verification of Section XI Compliance A. Introduction B. Determination of Compliance C. Number of Components D. Interval Compliance E. Midinterval Requirement Changes F. Verification of Compliance -Third Interval by Category * -G. Verification of Compliance -Third Interval by Category and Item Number Section 7 Piping and Instrument Diagrams Section 8 Third Interval, Section XI Category and Item Number -Scheduled Examinations by Period 1 2 2-3 3 4-6 6-7 8 9 9-14 14-17 18 19-21 22 22-27 28 28 28-29 29 30-31 32 33-48 49 50 50 51 51 51 52 53 54 55 TABLE OF CONTENTS (CONTINUED) Section 9 Third Interval, Section XI Category and Item Number -Scheduled Examinations by Period and Outage Year Section 10 ISI Isometrics for Scheduled Examinations during the Third Inservice Inspection Interval PAGE(S) 56 57

• SECTION 1 INTRODUCTION 1
• *-* _) 1. 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 2 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 L i st. 2. The 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Rressure Vessel Code, Section XI "Rules for

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

1. 10CFR50.55a (g) 3 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, 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, 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

.) 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. E. a. EM-09-02, "Inservice Testing of Plant Valves" b. EM-09-04, 11 Inservice Testing of Selected Safety Related Pumps 11 c. EM-09-05, 11 Steam Generator Inservice Inspection 11 d. EM-=p-9--11, "Palisades Eddy Current Test Procedure for Data Management for Replacement Steam Generators 11 e. EM-09-13, "Inservice Inspection Pressure Testing Program 11 f. EM-09-07, 11 Testing of Plant Snubbers" 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 6
• 4th Interval 1st Period 2nd Period 3rd Period From 05/12/2005 7 to 05/11/2015 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 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 of alphabetic codes which designate the main piping system under 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 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 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 -Main Reactor Coolant Loop RE -Recirculaticin*Ltne RHC -Return Header Combined 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 The fourth character set consists of the following:

11 a .. A number-letter 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 11 8 11 Hot 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 1 oop, the number 11 l11 wi 11 be used for consistency of the code. 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 11 4 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 11 Lo 11 facing direction of flow. The Number 2 weld would then be the next.weld clockwise, etc. In an elbow, Number 11 1 11 would be inside and 11 2 11 outside. 6LD1 is longitudinal weld downstream from circumferential weld 11 6 11 when there 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 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 c. 7. For branch 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 (0) Contains the branch line identifier.

8. Some examples of weld numbering are as follows: ESS-6-SIS-lAl-6 PCS-30-RCL-18-4/

ESS-12-SIS-181 PCS-12-PSL-1H-3PL 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. 12" Safety Injection Line lB Branch Connection weld into 30" Cold Leg Loop "B" downstream from circumferential weld number "4". 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." 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. .I 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 h. Valves 2. 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 15

• * -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 16 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
• * ** ) 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 19 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. Component T-82A T-828 T-82C T-820 TABLE Al CLASS 2 OPERATIONAL EXEMPTIONS IWC-122l(e) 1989 Edition Function SI Bottle SI Bottle . SI Bottle SI Bottle

• BASIS:
• 20 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 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 .

• BASIC:
• Component T-53A T-538 T-102 T-103 TABLE A2 c:.ASS 2 PRESSURE TEMPERATURE EXEMPTION IWC-1222(C) 1989 Edition Function Boric Acid Storage Tank Boric Acid Storage Tank NaOH Tank NaOH Makeup Tank Maximum Operating Pressure Atm Atm 16 Psig 16 Psig Maximum Operating 21 Temperature 200°F 200°F 150°F 150°F 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 .

• 22 B. . DEFERRALS 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. Section XI Category B-L-1 B-L-2 *B-M-1 B-M-2 B-N-3 B-0 C. AUGMENTATION COMPONENTS DEFERRABLE TO THE END OF INTERVAL Component* Pump Casing Welds Pump Casing Valve Body Welds Valve Bodies Core Support Structures Welds in CROM Housing Comments Code Case N-481 Code Case N-481 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.

• 23 TABLE 4.3.2 Miscellaneous Surveillance Items 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 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.

24 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 period, such as refueling,. a volumetric 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 refueling outage). 2. Second 3-1/3 years (or nearest* refueling outage). 3. Third 3-1/3 years (or nearest refueling outage). Successive Inspection Intervals Every 10 years thereafter (or nearest refueling outage) 100% Volumetric inspection of all welds. 100% volumetric inspection of all welds. 100% volumetric inspection of all welds. Volumetric inspection of 1/3 25 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 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):

• 26 A. Welds in the main steam lines including the safety valve attachment 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 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 27 4. Peiform ultrasonic wall thinning examinations beginning at the Elbow-to -Pipe weld downstream of the Generators.

5. Perform visual 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.

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

29 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 .

• SECTION 4 PALISADES PLANT . Code Usage 30
• INSERVICE INSPECTION PROGRAM -CODE USE APPLICABLE EDITIONS & ADDENDA OF .A .. BOILER & PRESSURE VESSEL CODE SECTION XI 31 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 .
• * -----32 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 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 Code Case N-457 Code Case N-460 Code Case N-461 Code Case N-463-1 Code Case N-481 Code Case N-489 Code Case N-491 Code Case N-494 Alternate Examination of Outlet Nozzle on Secondary Side of Steam Generators Qualification Specimen Notch Location for Ultrasonic Examination of Bolts and Studs Alternate Examination Coverage for Class 1 and 2 Weld Alternate Rules for Piping Calibration Block Thickness Evaluation Procedures and Acceptance Criteria for Flaws in Class 1 Ferritic Piping that Exceed the Acceptance Standards of IWB-3514.2 Alternate Examination requirements for Cast Austenitic Pump Casings (see attachment 1 to this submittal)

Alternate Rules for Level Ill NOE Qualification Examinations Alternate Rules for Examination of Class 1, 2, 3 and MC Component Supports of Light-Water Cooler Power Plants 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. --1

• 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 ABB Combustion Engineering Nuclear Power jl 1111 ,.,,,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 EXPRFSS OR IMPUm INCLUDING THE WARllANTm FOR A PARTICULAR.

PURPOSE OR MERCHANTABILITY, WITH THE ACCURACY, OF THE INFORMATION CONTAINm IN THIS REPORT, OR mAT THE USE OF ANY INFORMATION, APPARATIJS, ME'IHOD, OR 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 LIST OF TABLES LIST OF FIGURES 1.0

2.0 INTRODUCTION

2.1 Purpose 2.2 Scope and Applicability

3.0 BACKGROUND

3.1 Goals 3.2 RCP Casing Descriptions 3.2.1 General 3.2.2 Casing Design 3.2.3 Fabrication 3.2.4 Inservice Inspections 3.2.5 Codes and Standards 3.3 RCP Casing Material Degradation Mechanisms 3.3.1 Fatigue and Crack Growth Rates 3.3.2 Thermal Embrittlement Section 3 References 4 2 6 7 13 16 16 16 17 17 17 17 18 18 19 19 27 27 29 33 SECTION TITLE 4.0 CURRENT INSPECTION REQUIREMENTS FOR RCP CASINGS 4.1 NRC Requirements 4.2 ASME Requirements 4.3 Applicability of ASME Code Case N-481 Alternative 4.4 ASME Code Case Reference Flaw 4.5 Postulated Initial Crack Depth Section 4 References 5.0 METHODOLOGY FOR EVALUATING PUMP CASINGS 5.1 crack Growth Analysis 5.1.1 5.1.2 5.1.3 _5.1.4 5.1.5 Selection of Locations for Most Analysis Initial Flaw Size Plant Operatinq History Calculation of Crack Growth Rates Plant-specific Results 5.2 Thermal Embrittlement Analysis 5.2.1 Material Heat Data 5.2.2 Estimate of Minimum Charpy Impact Enerqy 5.2.3 Conversion of Charpy Impact Enerqies to J-inteqrals 5.2.4 Conversion of J-inteqrals to Plane Strain Fracture Touqhness 5.2.5 Plant-specific Results 5 35 35 35 36 37 38 38 39 e 39 39 40 41 42 48 56 57 59 60 68 e 69 ..

• 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 5.3.4 Crack Depth for Emergency Condition*

and Faulted Condition Loads 5.3.5 Plant Specific Results 6.0 INSPECTION INTERVAL DETERMINATION 6.1 Safety Margins 6.2 Inspection Alternatives 6.2.1 Volumetric Examination 6.2.2 Visual Examination Inspection Interval 6.4 Plant-specific Inspection Intervals

7.0 CONCLUSION

S TABLE NUMBER 3.2-1 5.1-1 5.3-1 7.0-1 LIST OF TABLES Reactor Coolant Pump Fabrication History Coefficients for Calculating KI Comparison of KI(applied) and KJc Touqhness End-point Crack Sizes as a Percent of Wall Thickness 6 74 74 91 94 96 96 97 98 99 100 20 45 72 101 FIGURE NUMBER 3.2-1 3.2-2 3 .* 2-3 3.2-4 3.2-5 3.2-6 3.3-1 5.1-1 5.1-2 5.1-3 5.1-4 5.1-5 5.2-1 5.2-2 5.2-3 5.2-4 5.2-5 5.3-1 5.3-2 .5. 3-3 5.3-4 5.3-5 5.3-6 5.3-7 5.3-8 5.3-9 LIST OF FIGURES TITLE Typical Pump Vertical Cross-section Typical Casing Vertical Cross-section Typical Casing Horizontal Cross-section Hub Section After Casting, Before Welding Volute Section After Casting, Before Welding Welded Casing Assembly Before Machining Time-Temperature Curve Growth Curves for Palisades RCPs Growth Curves for Fort Calhoun RCPs Growth curves for Calvert Cliffs RCPs Growth Curves for San Onofre RCPs Growth Curves for St. Lucie RCPs Flow Diagram for Fracture Toughness Curves Comparison of J-R Curves, Comparison of Ferrite Content Predictions curves to Define Jic Flow Stress Values for Aged and Unaged Cast SS Palisades curves With Design Condition Limits Fort Calhoun Curves With Design Condition Limits Calvert Cliffs Curves With Design Condition Limits San Onofre Curves With Design Condition Limits St. Lucie Curves With Design Condition Limits Palisades Curves With Emergency Condition Limits Ft. Calhoun Curves With Emergency Condition Limits Calvert Cliffs Curves With Emergency Condition Limits San Onofre Curves With Emergency Condition Limits 7 ' e 21 22 23 24 25 26 34 51 52 53 54 55 63 e 64 65 66 67 76 77 78 79 80 81 82 83 84 FIGURE NUMBER 5.3-10

5. 3-12 5.3-13 LIST OF FIGURES (Cont'd) TITLE PAGE St. Lucie Curves With Emergency Condition Limits 85 Palisades Curves With Faulted Condition Limits 86 Fort Calhoun curves With Faulted Condition Limits 87 Calvert Cliffs Curves .With Faulted Condition Limits 5.3-14 San Onofre curves With Faulted condition Limits 88 89 90 5.3-15 St. Lucie Curves With Faulted Condition Limits APPENDIX TITLE A. APPLICATION OF METHODOLOGY TO PALISAJ;>ES 1.0 Purpose 2.0 Pre-service Inspection Data Evaluation 3.0 Operatinq History 4.0 Thermal Embrittlement 5 6 8 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 5.6 Stresses Under Emerg_ency and Faulted Conditions 5.7 Results 6.0 Inspection Interval Appendix A References PAGE e 32 33 34 35 B. APPLICATION OF METHODOLOGY TO FORT CALHOUN 1.0 Purpose 2.0 Pre-service Inspection Data Evaluation 3.0 *Operating History 4.0 Thermal Embrittlement 4.1 Material Identification and Chemical Properties 4.2 Material Specifications and 5 6 8 10 10 Mechanical Properties 10 4.3 Thermal Aging Behavior 11 4.4 Toughness Properties of Aged Materials 11 4.5 Limiting Values 5.0 crack Growth Analysis 5.1 Scope 5.2 Reference stress Reports 12 30 30 30 5.3 Selection of High Stress Locations 30 5.4* Stresses and Wall Thicknesses at Limiting Locations 5.5 Calculation of. Crack Growth Rates 5.6 Stresses Under Emergency and Faulted Conditions 5.7 Results 6.0 Inspection Interval Appendix B References 9 31 32 37 39 40 41 APPENDIX TITLE C. APPLICATION OF METHODOLOGY TO CALVERT CLIFFS 1&2 1.0 Purpose 2.0 Pre-service Inspection Data Evaluation 3.0 Operating History 4.0 Thermal Embrittlement 4.1 Material Identification and Chemical Properties 4.2 Material Specifications and Mechanical Properties 4.3 Thermal Aging Behavior 4.4 Toughness Properties of Aged Materials 4.5 Limiting Values 6 7 9 12 12 12 13 13 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 Locations 5.5 Calculation of Crack Growth Rates ' 5.6 Stresses Under Emergency and Faulted Conditions 5.7 Results 6.0 Inspection Interval Appendix c References D. APPLICATION OF METHODOLOGY TO SAN ONOFRE 2&3 1.0 Purpose 2.0 Pre-service Inspection Data Evaluation 10 40 42 48 51 52 53 6 7 APPENDIX TITLE 3.0 Operating History 4.0 Thermal Embrittlement 4.1 Material Identification and Chemical Properties 4.2 Material Specifications and Mechanical Properties 4.3 Thermai Aging Behavior 4.4 Toughness Properties of Aged Materials 4.5 Limiting Values 5.0 Crack Growth Analysis 5.1 Scope 5.2 Reference Stress Reports 5.3 Selection of High Stress Locations 5.4 Stresses and Wall Thicknesses at Limiting Locations 5.5 Calculation of Crack Growth Rates 5.6 Stresses Under Emergency and Faulted Conditions 5.7 Results 6.0 Inspection Interval Appendix D References E.* APPLICATION OF METHODOLOGY TO ST. -LUCIE 1&2 1.0 Purpose 2.0 Pre-service Inspection Data Evaluation 3.0 Operating History 4.0 Thermal Embrittlement 4.1 Material Identification and Chemical Properties 4.2 Material Specifications and Mechanical Properties 4.3 Thermal Aging Behavior 11 PAGE 9 12 12 12 13 13 14 43 43 43 43 44 46 47 56 57 58 6 7 9 12 12 12 13 e e 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 5.5 Calculation of Crack Growth Rates 5.6 Stresses Under Emergency and Faulted Conditions 5.7 Results 6.0 Inspection Interval Appendix E References F. COMPUTER CODE LISTINGS Description of Listings Database File Structure Computer dBase Program CASEINFO.PRG computer dBase Pro9ram FRACTOUG.PRG Computer dBase Program TEARMOD.PRG 12 40 42 48 51 52 53 2 4 6 7 9 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: Plant Palisades Fort Calhoun Calvert Cliffs 1&2 San Onofre 2&3 st. Lucie 1&2 Limiting End-point .

• crack Size 0.36t 0.32t 0. 38t 0. 43t 0.38t Minimum Time to Reach . . ** End-point Crack Size 46 years 165 years 130 years 77 years 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 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 . i

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 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 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 \ .. * :O::.;... .. -: 1** -. =-::*-* ",\, .. 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 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 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 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 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. 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 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 1000 UGO ...... GOO 0 w a: eoo :::> I-< a: 100 w -...........: 0... ,--600 ' --->"2 w ' t ------t-' a ---------... ', -----....... 600 ....... 1 2 .c e e 10 "*'-............ .coo 0.01 0.1 1 10 100 TIME (hours) 800 eoo 200 e.. 700 w a: :::> eoo I-< a: w 600 a.. 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 B-L-2 Volumetric examination of all welds 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 10 = ll4t x 6l4t 10 = t x 6t 4./10 4./10 = .oat x .48t Reference crack length 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, 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 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 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 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. 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 By algebraic manipulation, equation A-3300 (1) is made into a function of dimensionless crack depth, a/t, as follows: j!. t (Eqn. 5-2) 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 0.1 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 Fig. 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 _Q_ -m 1.10 0.95 1.05 1.12 0.89 1. 05 1.15 0.94 " 1.185 0.80 " 1. 225 0.76 " 1.27 0.73 " 1.325 0.70 " 1.395 0.685 " 1.48 0.675 1.05 from from from 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. cycle (Eqn. 5-3) 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. K 1 {maximum) is associated with the Applied Stress Intensity Factor from the Design Condition {plus heatup/cooldown thermal gradients), and hence, =KI (maximum). The Bernard & Slama equation then simplifies to: in cycle (Eqn. 5-4) 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 e atmospheric-to-operating pressure cycles, and dN/dT is 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 dT (Eqn. 5-6) 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, are calculated using Time = Distance/Velocity where incremental distances correspond to a change in a/t a percent of thickness interval) and velocity, da/dT. A conservative integration procedure expedites the process considerably. Crack growth rate increases 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 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 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 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 .50 G) .45 .., 0 .40 =s-0 :::c .r: .35 -*c a !e.3 CD 0 CD c .30 ::::! * .., "Tl CD en -* en -CC c 0 .25 0 ""C .., -O CD cu :::c :E 01 c .2 '< CD

• ts .20 "'C .., ..... I! 0 I I lL g. ""C ..... .15 CD P> --c;r .10 Om -.., *en .05 en 0 0 U1 .... e Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1 /6 (1) (2) -, j j _, J /* I I / j / / _/,. > 20 40 . (1) Crotch Region -60 80 Years (2) Junction of Volute.with Suction Nozzle Flange (3) Discharge Nozzle e 100 120 ---

140 160 e 01 I\) 0 G)3 .., Sl> 0 ::r Sl> ::r ""'O ::r: 6-e u;* == -o 0 ""'O "Tl ::::!. 0 -* CD <CC en < c 0 CD CD -.., ::r: q (J1 '< en

• Q ::::!. I -nN ::r 0 -::l. Oo iU Sl> 0 -" ::r en o c :::J e e

& Slama Growth Equation @ 715 cy/40 yr; all = 1 /6 .50 .45 .40 i .35 .30 l! 0 .25 ti I! .20 LL .15 .10 .05 0 *O ' 20 40 60 80 (1) Junction Scroll with Suction Nozzle Flange (2) Scroll (3) Junction Scroll with Bolt Flange (4) Diffuser Vane 7 100 Years {4) r 7 (1) I / / ,. v -----{3) 7?' \ 120 140 160 180 200 CD P> G> ::+ a3 :::J"' CD :r: G> -* P> CJ) Q_Ro'TI CD m -* CJ> -CO 0 CD C 0 .., -t::;' CD :r: c;* 01 g, 0 I :::J"' e!. (,) !. < -* CD &l -o 0= ii1 = 0 CJ) CJ) Ro N .r:. ti CD c 0 'ii c .2 u l! LL Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1 /6 .50 .45 .40 (1) (5) (2) --, I I I r I j I I / .35 .30 ! I / .25 .20 .15 .10 / / / / V. v .... / ::::::------_,,,,, ------ __......-______.. --.05 -0 0 20 40 60 80 100 120 Years 140 160 180 200 (1) Vane 8 (2) Discharge Nozzle, Crotch Vicinity (3) Suction Nozzle (4) Volute @ Junction with Lower Flange (5) Hanger Bracket #1 Vicinity (3) ... _../ (4) 220 240 Ci) CJ> -, 0 0C g. =r CD I3 -* 0 Pl 0 ::-:l. m "Tl CD -* CJ) 9: cg 0 CJ) -, -0 CD I ::J U1 '< I . "'O ...&. Q (/) I -Pl . =r ::J .J:ii. CD e.Q 0 ::J e!.o , -, CD Pl I\) CJ) (A) .r. a CD c 0 ca c .2 u l! LL

• Bernard & Slama Growth Equation @ 700 cy/40 yr; a/L = 1 /6 .50 .45 .40 .35 .30 (3) (2) (4) (1 )(5) -I fj' I I j ) f f .... I I /j / ' / f If ..I .25 .20 / / /j ir / / / v A 7 / ./ .15 v ./ , ;;iiio""'"" . 10 ___.......

---

. .05 0 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 01 01 G) ,, ., -0 0 s: ::::!. ;:i. ::::r I>> :i: ""'O -*o == 0 CD ::! . ., ,, CD -* (/) Ro cc -ce* CD :i: ::::r 01 '< -. "O I ...... Q I -0001 ::::r -CD * =.r fJ 5 0 a;* ., ...... I>> 0 Ro Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1 /6 .c .50 (1) (2) (5) .45 ' I I j I r I J I I /r .40 i .35 c i .30 , I 7 _V 0 .25 .Q .20 .15 .10 .05 0 gj;;fi -/ v

-/' / _/ v v _.. v L-----" -------0 20 40 60 80 100 120 140 160 180 200 Years (1) Vane 8 (4) Volute @ Junction with Lower Flange (2) Discharge Nozzle, Crotch Vicinity (5) Hanger Bracket #1 Vicinity (3) Suction Nozzle (3) _.. (4) 220 240 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 =.Ni + 0.11 (Mn) -0.0086 (Mn)2 + 18.4 (N) + 24.5 eq (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: %F = 100.3 (Creq/Nieq> 2 -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 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 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 e* = [%F x Cr x (C + 0.4N) x (Ni+Si)2]/100 (Eqn. 5-10) 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 1.15 + 1.532 exp(-0.0467 (Eqn. 5-11) This equation gives a value for in joules/cm 2. 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 = {91 (25 4)n (C " d

• Vs at (Eqn. 5-12) where the exponent n is defined as: n = 0.35 + 0.0025(C.._

t]0*67 . -vsa (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[C ]0.49 Vs at (Eqn. 5-15) For these equations, the room temperature impact energy <=vsat is in (J/cm 2) and Jd and 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 Input l>111 UI 11111111111 Lower bound Lower bound J-R curves J-R curves Eqs. 2.1-2.4 Eqs. Section C e known T 8 (*C) e Input* e <280 3.3 280-330 2.9 2.5 Calculate Q, P Eq. 4.1, 4.2 . Input C Vint CVint

• 200 Calculate a,p,OI Eqs. 4.3-4.5 Calculate a,p,OI Eqs. 4.3-4.5 qs. 4.6-4.13 Section B UU Lower-Bound Un aged J-R curve s. 3.16, 3.,17 Lower-Bound Unaged J-R curve Eqs. 3.16, 3.1 FIGURE 5.2-1 Input compositio .Cr,Mo,Si.Mn,Ni,&C N known N c-0.04 Calculate

' .3.4 Calculate CVm Eq.3.5 Service time oughness desired Yes FLOW DIAGRAM FOR ESTIMATING FRACTURE TOUGHNESS J-R CURVES OF CAST STAINLESS STEEL IN LWR SYSTEMS (REF. 5-1) 63 I . ! Crack Extension (In.) 0.0 0.1 0.2 0.3 ---C'I E aoo -----efpy at 290°C ------------ 16 -""') ---------32 -------<48 ""') 600 c .52 -ca 400 E .. 0 -cu c 5000 c :c 4000 -I. 3000 2000 1000 c :::.. .., c 0 ; ca E .... 0 -CD Q 0 -C'll E 800 -""') -""') 600 c 0 ;:: ca 400 E .. 0 -cu 200 c o 2 4 6 8 Crack Extension, da (mm) I Crack Extension d8 (In.) 0.1 0.2 0.3 Heat L CF-SM 290-320°C / I """ ,,, ,, / lJnaQed efpy at 320"C -*---*--** 16 0 2 4 6 8 Crack Extension, l1.a (mm) FIGURE 5.2-2 10 5000 c ]5 4000 i: .5 -3000 .., . c 0 2000 ; ca E .. 1000 0 -CD Q 0 10 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

• * * ) ' --* RELIEF REQUEST NUMBER -RR-10 {cont'd) PROPOSED ALTERNATE EXAMINATION 48 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 .

600 500 300 100 .J--------.,.---------. G.25 GJiO

• POINTS useo FOR ReGRESSfON ANAl.YSIS , l I ( 1.5 mm El$CUIStON UNE I I ' Cl.JS . 1.G 1.50 1.75 CRACK EXTENSION Cmml 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 '
• I I I ** ! ' I I I ! I I I ' ! I ' I ' ! i I ** ! I i
• I ! ' I i i I I I i ! I I I i ! i l.oo

-*-*-*'.-***-. .; .. _ ..... ; ......... ..... JI:.; *... 2 90C AGED t . ! i i i Q: 290C UNAGED : : : : : : : : ; : . . . j . --;

j t-: : : . : .: ; :-: : : :_ *.: =. .= :_* =.. i I-: : : : : : .

• f . .., * ; : :

2s<; 1 I I I i , i = * , : : : : -I .. ttncge'd bfoi: J * * *

• I : ; : : 1 * * * : : * ; "'1 1 : : : : : 1 . . . . : i : : : : :

-; * : : : : : .J * * * * * * * * * : ;

• I ;.. :

-=*-: : : : : : : :

: : =j L. * * * * * * * * * . * * *

• iBe :
  • 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 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 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 1 , most probably the median toughness, is 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 ...... N Table 5.3-1 Applied KI From Design Condition vs Actual Toughness at 550 F Minimum KI KJc Time, yrs Applied Toughness Plant/Limiting Location & Catt) CKsi-yinch) (Ksi-yinch) Calvert Cliffs/Vane 8 133 (0.45) 108. 151. 0 Palisades/ Crotch 49 (0.45) 130. 152.3 Ft Calhoun/Vane 7 173 (0.45) 109. 141. 5 San Onofre/Crotch 77 (0.45) 101. 142.8 st . Lucie/Vane 8 133 (0.45) 108. 142.1 ------------Median KJc Toughness CKsi-vinch) 186.7 223.5 223.8 188.4 184.9 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 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 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 ...... O> (i) a ::J" I !!?. 0 ::::!. CD en 0 -(') "C 0 0 :::J -en i5c t:!'. 3 0 CD "Tl Pl @ -* (') -<g .., ""O .., Pl 0 CD == 01 en c,J (/) I I ::J" ""O _,,, 0 Pl == -::;* c;;* cc OCD CD en en cc* :::J (') 0 :::J a. 0 :::J r: 3 ;::::;: en a G> c 0 a; c .2 LL .50 .45 .40 .35 .30 .25 .20 .15 .10 * .05 0 Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1 /6 (1) (2) -, j j I I I I I / J / 0 20 40 60 80 Years (1) Crotch Region (2) Junction of Volute with Suction Nozzle Flange (3) Discharge Nozzle --"T>>16o Years 100 120 140 T 50 Ksi Flow Stress Limit for Design Conditions 160 G') a ! ::r ::r: c;;* 6 ::J. CD Q en 3 g_ P> ::r: ::r '< P> "O ""'O 0 c -O" ::r -CD -* -o o""'O :!! -ca 0 == c ., CD ., P> ., CD oo " -* 01 en 2?. CA> (/) ::J. I :::T !l I\) I ::r ti1 ca ;:i. Oo m P> -* :::T ca o ::::Jc O::::> 0 ::::J a. ;:+" o* ::::J c 3. ;:+" en Bernard & Slama Growth Equation @ 715 cy/40 yr; all = 1 /6 .50 .45 .40 .c i .35 c .30 0 ii c .25 I ..... 20 .15 -.10 .05 0 0 20 40 60 80 (1) Junction Scroll with Suction Nozzle Flange (2) Scroll (3) Junction Scroll with Bolt Flange (4) Diffuser Vane 7 100 Years (4) -' j l 7 (1 ) I / " v / I/_, V' -120 140 160 180 Y 50 Ksi Flow Stress Limit for Design Conditions (3) 19' ' -, 200 'Y>200 Years G) ..., 0 :::T I c;;* -0 ..., -*m m Pl -3 Io '< ..., "'O CD 9. G) .c :::T Pl a Q) CD en c --*go m :r1 e --CC 0 0 CDC . n; iil & m c 0 -* 01 .2 "0 . u en I w l! (/) 0 I lL :::T Pl w 0-:E < -*CD ::J ;:::i. cc 0 O= CD :::i: en en <O" ...... ::J go ON 0 ::J a. ;::::;: ()" ::J c: 3 ;::::;: en (if Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1 /6 .50 (1) ' .45 J I I / ./ v t::::---- -.40 .35 .30 .25 .20 .15 .10 .05 0 (5) I r J I / / / v . ...... ,_,,,---___.,,,,.. (2) 1 I/ v (3) -_,., /" -(4) Y>240 Years 0 20 40 60 80 100 120 Years 140 160 180 200 220 240 (1) Vane 8 (4) Volute @ Junction with Lower Flange (2) Discharge Nozzle, Crotch Vicinity (5) Hanger Bracket #1 Vicinity (3) Suction Nozzle y 50 Ksi Flow Stress Limit for Design Conditions G> a ! ::J" :::c 6 ::::!. CD en (J) 0 0 -r:: :::c -"< ::J" "C CD 0 .., -:::J ::J" 0 Pl 11 -a.cc 0 -* r:: .., en .., Pl 0 CD 0 :::J 01 " I . en CA> (/) (/) I ::J" Pl 0 :::J :E. 0 :::J :::J cc a o. co m I\) cc* Ro :::J CA) 0 0 :::J a. 0 :::J r: 3 ;:::::;." en. .c a Cl> c 0 'ii c .2 l3 l! u. Bernard & Slama Growth Equation @ 700 cy/40 yr; a/L = 1 /6 .50 .45 .40 (3) (4) (2) (1 )(5) -,, I I ) f I I /) / .35 .30 / I / r ' ' / / /j ir .25 .20 .15 .10 / / 7 A v I./ :iii""'° ,..... l/ & -----------.05 0 0 10 20 30 40 50 60 70 80 90 100 Years (1) Suction Nozzle (4) Vane (2) Junction Lower Flange with Volute (5) Volute in Vane 5 Vicinity (3) Crotch / 110 120 130 140 150 160 Y 50 Ksi Flow Stress Limit for Design Conditions CXI 0 G') ..., 0 ! :J"' I iii" -0 :::!. CD CJ) 0 "Tl --IQ '< -* "U a. 0 Pl --u !:!". :e £ T! -Qo<O (') r c ..., ..., Pl ca* CD 0 :::::r U1 """ -. CJ> I CA> (/) I :J"' CJ) OI 0 r+ :E r -*c :::::J 0 <O -* 0 CD CD __., cg I\) (') 0 :::::J a. a: 0 :::::J c 3 ;::+" en .50 .45 .40 .c: .35 1i G> c iS .30 I!! 0 ca .25 c .2 ts I!! .20 u. .15 .10 .05 0 0 20 Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1 /6 (1) (5) (2) -, I I J I r I J I I /r I / v / / ' / v / ,_,-/ ::::::-------------) -------,si:fi -(4) J ---40 60 80 100 120 Years 140 160 180 200 220 240 'Y>240 Years (1) Vane 8 (4) Volute @ Junction with Lower Flange 'Y 50 Ksi Flow Stress Limit for Design Conditions (2) Discharge Nozzle, Crotch Vicinity (5) Hanger Bracket #1 Vicinity (3) Suction Nozzle Q) .... G) a :J"' ::r: ::::::!. CD U> a ::r: '< "C 0 2.o :J"' ::J CD U> =*c fl1 3 -co 0.., "Tl U> -* .., -co P.> "tJ c U> U1 (/)CD * :J"' .., w 0 I I "tJ CJ) -* P.> ::J -co ur 3 CD CD U> ca CD ::J 0 0 ::J. c. ;::::+" o* ::J r: 3 ;::::+" U> .50 .45 .40 s= .35 a G> c .30 l! .25 c .2 .20 LL .15 .10 .05 0

• Bernard & Slama Growth Equation @ 505 cy/40 yr; a/L = 1 /6 (1) (2) -r J j J I I , I I -/ J / / /. 0 20 40 60 80 Years (1) Crotch Region (2) Junction of Volute with Suction Nozzle Flange (3) Discharge Nozzle ____. L----(3) 'f'>160 Years 100 120 140 'f' 50 Ksi Flow Stress Limit for Emergency Conditions 160 G> a ! ::J"' :r: en -0 <D en 0 0 -3 :r: Pl '< ::J"' "'C Pl 9. \J ::J"' c s! -o* Q. CD c -\J ,, oo -* ..., :e co Pl <D c 0 m 1i c en CJ .2 (/) -* U1 N en * ::J"' c..> Q -* I u.. :e n......, :;*I co ,, mo <D 0 ..., Pl co -<D ::J"' ::J 0 0 c '< ::J 0 0 ::J a. ;;:::+ a* ::J c: 3 ;:;.-en Bernard & Slama Growth Equation @ 715 cy/40 yr; all = 1 /6 .50 .45 .40 .35 .30 .25 .20 .15 .10 .05 0 0 20 40 60 80 (1) Junction Scroll with Suction Nozzle Flange (2) Scroll (3) Junction Scroll with Bolt Flange (4) Diffuser Vane 7 100 Years T>0.5 alt (4) r J I (1) I / /' / ,,,,,,---__,,,,,. 120 140 160 180 T 50 Ksi Flow Stress Limit for Emergency Conditions (3) (?' ' 200 T>200 Years G) a ::r I !! 0 :::!. CD en llJ a '1> -I-* '< 3 "C 0 9. <D ffi G) d'. '1> en -Ro om'! -. -CO '1> CD c 0 0 .... "° CD en c;* 01 en I CA> ::r Q 0 I =:* '1> CX> s*< <C CD m ;:i. 30 CD -* .... ::i: <C en CD _.. 5 Ro '< I\) 0 0 :::::J a. ;::::;: ()" :::::J r: 3 ;:;: en .c 1i CD c 0 1i c .2 tS t! u.. .50 .45 .40 .35 .30 .25 .20 .15 .10 .05 0 0 20 40 Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1 /6 'Y>0.5 alt 60 80 100 120 Years (1) (5) 140 160 180 (2) 200 220 240 'Y>24 0 Years (1) Vane 8 (4) Volute @ Junction with Lower Flange 'Y 50 Ksi Flow Stress Limit for Emergency Conditions (2) Discharge Nozzle, Crotch Vicinity (5) Hanger Bracket #1 Vicinity (3) Suction Nozzle G) -, 0 r-+ :::r I u;* ....+ 0 -, ro* CJ> 0 -U> IO '< c -0 ..-+ 0 :::r ..... m :::r -, m :::::> 0 Sll -. Sll :-0 rn -n P> D: ca* 0 CJ> c "0 -, CJ> ::J CD U> I ?1 :::r U> (..> 0 Sl) I :! ::J <O :Jo 0 co ::J mo CD CD <O I\) CD Qo '< 0 0 ::J 0. 0 :::J c 3. ;::::+ CJ> .c li Q) 0 .lC 0 '1S ... 0 cu c .Q u £!! u.. -e Bernard & Slama Growth Equation @ 700 cy/40 yr; a/L = 1 /6 Y>0.5 alt Y>0.5 alt ----------

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.50 .45 .40 (1 )(5) (2) (3) (4) ) If I I }' J I I ... I I II / .35 .30 / I If /' .. .25 .20 / / _fi 'r / / / r .15 _......... ....-'/ L..-----" .10 -----Iii .05 0 0 10 20 30 40 50 60 70 80 90 100 (1) Suction Nozzle (2) Junction Lower Flange with Volute (3) Crotch Years (4) Vane (5) Volute in Vane 5 Vicinity / ,,.... v 11 0 120 130 140 150 160 Y 50 Ksi Flow Stress Limit for Emergency Conditions G) a I c;;-5 CD fl) 9. I "11 '< 0 "C 0 a. :T P> CD -0 =*o fi1 .,, (").., <C" ii) Ro c r.., 0 -*CD "cc 01 fl) =t' * (/) -CJ,) =t' I I 0 (/) -&. :e. =-+ 0 ::s r cc c m 52. 3 CD CD -"' ca Ro CD I\) ::J (") 0 ::J a. ;::::;." c:r ::J c 3 ri .50 .45 .40 s::. .35 a CD c iS .30 I!! 0 cu .25 c .Q 13 l! .20 lL .15 .10 .05 0 0 20 Bernard & Slama Growth Equation @ 505 cy/40 yr; all = 1 /6 T >0.5 alt (1) (5) I ' I j J ' J I I , I / / / , / ./ , v .... v t:::------------__,.,.,....-i----------__........

(2) -I I/ /r v (3) -.,_.... (4) 40 60 80 100 120 140 160 180 200 220 240 Years Y>240 Years (1) Vane 8 (4) Volute @ Junction with Lower Flange (2) Discharge Nozzle, Crotch Vicinity (5) Hanger Bracket #1 Vicinity Y 50 Ksi Flow Stress Limit for Emergency Conditions (3) Suction Nozzle * ------, G) ..., 0 ::r I en ....... 0 ::::!. CD en 0 -I '< 0 -0 0 -8. :::J ::r en CD C O CD 11 Pl ..., -* -en co 0 -c -, -u -, Pl o CD 0 :E 01 . en CD c...:> -, (/) I I ::r ....... 0 -u ....... :: Pl :::J. en <O Pl 11 a. Pl CD c en ....... CD a. 0 0 :::J a. 0 :::J r 3 ;:;: CJ> .50 .45 .40 £ .35 a. Q) 0 iS .30 0 Ci .25 c .2 .20 u.. .15 .10 .05 0

• 0 --Bernard & Slama Growth Equation @ 505 cy/40 yr; a/L = 1 /6 (1) (2) ' , J J j J J I . I .. -* I / J . v / _/' _/_ 20 40 60 80 Years (1) Grotch Region (2) Junction of Volute with Suction Nozzle Flange (3) Discharge Nozzle ----L------(3).

T>160 Years 100 120 140 T 50 Ksi Flow Stress Limit for Faulted Conditions 160 ,------------Bernard & Slama Growth Equation @ 715 cy/40 yr; all = 1 /6 .50 .45 .40 2-.35 c .30 e 0 ! .25 i;; .20 lL .15 .10 .05 0 ' 0 20 40 60 80 (1) Junction Scroll with Suction Nozzle Flange (2) Scroll (3) Junction Scroll with Bolt Flange (4) Diffuser Vane 7 100 Years 'Y>0.5 alt (4) r j I (1) 7 / / II" v V" -__.,,--120 140 160 180 T 50 Ksi Flow Stress Limit for Faulted Conditions (3) /')1 ' 200 °Y>200 Years =* 0 ::J 3 ...... U> e. .c li Q) 0 .:it:. 0 as ... 0 (ij c: .Q u as ... u. Bernard & Slama Growth Equation @ 505 cy/40 yr; a/L = 1 /6 .50 (1) , .45 .40 .35 J I .30 .25 .20 .15 .10 / _/ v t::::----i----.05 0 J -; / , / v J/ -----(5) (2) I 'rf f *-/ (3) -__,..,. L-----'" ........ (4) T>240 Years 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 T 50 Ksi Flow Stress Limit for Faulted Conditions Ci) a ::r I en -0 ::::!. CD en CJ) g_o :cS. '< ::r "C CD 0 .., -::I i°O o*,, n> mcc* .., en .., Pl 0 CD 0 ::I 01 (/) (/) I ::r P> ....&. 0 ::I .i:i.. ==. 0 ::I ::I cc 0 -,, .., Pl CD c: I\) m Ro a. CA> 0 0 ::I a. ;;:::+ a* ::I c: 3 ;;:::+ en .c 1i Cl> c iS l!! 0 a; c* .Q LL .50 .45 .40 .35 .30 .25 .20 .15 .10 Iii .05 0 Bernard & Slama Growth Equation @-700 cy/40 yr; a/L = 1 /6 alt (3) (4) (1 )(5) ) I/ 1 I f I I /J , J I I/ / / // r / / / v-A v i/' _/ / , _......-__,,,,,,. _,,,,,,__ --(2) -I f .. / 'fl'" 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 Y 50 Ksi Flow Stress Limit for Faulted Conditions G) ...... 0 ,.... :::r I CJ) ,.... 0 -, CD CJ) 0 11 --'< -* "O 0. 0 ll> ,.... -u 0 CD 11 ll> ...... -* -co OQo c iiJ c ro () co U1 :::r . CJ) ,_ c...> (f) I I :::r (f) 0 CJl -*c :::J () co -* 11 CD ll> c Qo ,.... CD I\) 0. 0 0 :::J a. 0 :::J r: 3 ,.... CJ) .50 .45 .40 .c li .35 0 -n ca .30 .... (.) ca .25 c .Q u ca .20 .... LL. .15 .10 .05 0 0 20., Bernard & Slama Growth Equation @ 505 cy/40 yr; a/L = 1 /6 (1) (5) (2) I I I j I r I j ! 7 /' 7 l7 / , / / (3) / v v .... v v L.-----i--JI e::::::::- i.---(4) -t:;::::::--: 40 60 80 100 120 Years 140 160 180 200 220 240 'Y'>240 Years (1) Vane 8 (4) Volute@ Junction with Lower Flange T 50 Ksi Flow Stress Limit for Faulted Conditions (2) Discharge Nozzle, Crotch Vicinity (5) Hanger Bracket #1 Vicinity (3) Suction Nozzle 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 July 30, 1991. 93 6.0 INSPECTION INTERVAL DETERMINATION 6.1 Safety Margins The results obtained in assessing s.tabili ty of RCP casings subjected to postulated cracks.in .high stress regions support a relaxation of the inspection intervals. For all plants reviewed, the 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 specifications which establish conservatively the 94

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 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 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 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 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 44 49 47 S4 46 Ernerg. Condition 57 42 42 46 46 Faulted Condition 39 33 38 34 34 San Onofre 2&3 Design Condition 42 49 46 41 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 APPENDIX A APPLICATION OF GENERIC METHODOLOGY FOR RELAXATION OF THE PALISADES REACTOR COOLANT PUMP CASING INSPECTION INTERVAL
• 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 e 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 26 5.1 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 e I APPENDIX A REFERENCES 35 i I Appendix A 3 L -----

TABLE 4-1 TABLE 4-2 TABLE 4-3 TABLE 4-4 TABLE 4-5 TABLE 4-6 TABLE 4-7 TABLE 5-1 TABLE 5-2 TABLE 5-3 Appendix A LIST OF TABLES Material Identification and Chemical Compositions Material Specifications and Tensile Properties Predicted Thermal Aging Behavior Predicted Toughness Properties of Aged Materials (70°F) Predicted Toughness Properties of Aged Materials (550°F) Limiting and Controlling Values of Jic and KJc at 70°F Limiting and Controlling Values of Jic and KJc at 550°F Crack Growth Rate at Suction Nozzle Flange Crack Growth Rate at Crotch Region Crack Growth Rate at Discharge Nozzle 13 16 18 20 22 24 25 29 30 31 4 I I 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 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 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: Transient Condition Heat-Up Cool-Down Hydrostatic Test (3110 psia) Assumed Occurrences During 40 Year License Period 500 500 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 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 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. 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

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rl Table 4-1 ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL )> REACTOR COOLANT PUMP CASINGS "'O "'O REPORT fl* MATERIAL.IDENTIFICATION

& CHEMICAL COMPOSITIONS CD ::I a. -'* 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 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 0.011 CASING WELD A7B24R 0.06 1.42 o.so 0.012 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 0.018 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 REACTOR COOLANT PUMP CASINGS "'C "'C ct> REPORT 11

• MATERIAL IDENTIFICATION

& CHEMICAL COMPOSITIONS

s a. -'* 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 REACTOR COOLANT PUMP CASINGS "O "O tD :::s REPORT 11

• MATERIAL IDENTIFICATION

& CHEMICAL COMPOSITIONS 0.. -'* x MATERIAL )::o 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 STEEL ):> REACTOR COOLANT PUMP CASINGS "'O "'O CD ::I REPORT 12

• MATERIAL SPECIFICATION

& TENSILE PROPERTIES

a. --'* 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. (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 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 o.o o.o o.o o.o 0 78600 58000 CASING WELD E316-15 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 ....... CASING WELD E316-15 7H32A o.o o.o o.o o.o 0 78600 58000 O'I 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 )> REACTOR COOLANT PUMP CASINGS 0 Ill :::s REPORT 12

• MATERIAL SPECIFICATION

& TENSILE PROPERTIES

a. )( 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

\ CASING WELD A 351 CF8M B7Hl5R o.o o.o. o.o 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 78600 58000 CASING WELD 1T3454 E316-16 03067A o.o o.o o.o o.o 0 . 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 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 ....... CASING WELD E316-16 10184 o.o o.o o.o o.o 0 78600 58000 -...i e -e ---Table 4-3 .*** I -' ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL )> REACTOR COOLANT PUMP.CASINGS "O "O 11> REPORT 13

• PREDICTED THERMAL AGING BEHAVIOR ::s 0.. _,, 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 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 1 I 7 16 78143A 5.0 17.495 16.481 i 1.06 5.9 6.0 22.535 14.995 10.0 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 9.7 02475C 0.0 16.955 15.892 1.07 6.2 6.2 22.005 14.430 10.0 I-' 7H32A o.o 16.737 16.237 1.03 4.5 4.8 21.650 14.785 7.6 co 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 REACTOR COOLANT PUMP CASINGS "'C "'C Cb REPORT 13

• PREDICTED THERMAL AGING BEHAVIOR ::s a.. ...... 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 ...... 10184 o.o 16.612 17. 051 0.97 3.0 3.1 21.670 15.490 1.0 \.0 e --

e e Table 4-4 \ )> ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR "'C COOLANT PUMP CASINGS AT 70F "'C tt> :::I REPORT #4 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL 0. .... x )> 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 7E34B 6.0 6.75 12249.0 0.43 109.00 1969.0 288.4 155.0 N 03100 5.5 6.45 12626.0 0.43 112. 90 2045.0 293.9 159.0 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 . 0.45 147.10 322.2 190.7 .... '-::'...*

r------------ N ...... HEAT NO. MEASURED FERRITE CONTENT {%)

• RCP PUMP CASING 661-N-0766 B7J1R 10.0 B7Hl5R 8.0 B7H13R 11.0 4850 0.0 B7H14R 13.0 4854 o.o 0337BA o.o 033788 0.0 7F43A 6.0 03252 6.0 03033A 7.5 03067A 5.0 0214C 5.0 03100 5.5
• RCP PUMP CASING 661-N-0767 B7J1R 10.0 B7H15R 8.0 B7H13R 11.0 4848 0.0 4852 o.o B7H14R 13.0 7F43A 0.0 02475C 0.0 03252 6.0 02362A 0.0 03100 5.5 10184 0.0 .e 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 MATERIAL AGING PARAMETER 23.41 18.91 17.81 15.91 16.56 12.70 12.70 8.47 7.84 7.20 7.14 6.70 6.45 24.34 23.41 18.91 *17,79 17.34 15."9.l 7.86 7.84 8.26 6.45 4.10 MATERIAL CONSTANT c 9 70F 3949.0 4099.0 5040.0 5343.0 5956.0 5732.0 7358.0 7358.0 10371.0 10994.0 11700.0 11777 .o 12313.0 12626.0 3949.0 4099.0 5040.0 . 5347 .o 5482.0 5956.0 10371.0 10979.0 10994.0 10576.0 12626.0 16459.0 MATERIAL CONSTANT N @ 70F 0.38 0.38 0.39 0.39 0.40 0.39 0.40 0.40 0.42 0.43 0.43 0.43 0.43 0.43 0.38 0.38 0.39 0.39 0.39 0.40 0.42 0.43 0.43 0.42 0.43 0.45 .e MINIMUM IMPACT ENERGY (ft-lbs) 25.76 27.09 35.72 38.58 44.48 42.31 58.37 58.37 89.33 95*.83 103.20 104.00 109.60 112.90 25.76 27.09 35.72 38.62 39.91 44.48 89.33 95.67 95.83 91.46 112. 90 153.00 MINIMUM Jlc @ 70F (1 b/1n) 691.4 720.4 858.8 934.8 984.2 1011.0 1254 .o . 1254.0 1695.0 1727 .o 1861.0 1876.0 1982.0 2045.0 691.4 720.4 858.8 941.8 959.1 984.2 1695.0 1722.0 1727. 0 1734.0 2045.0 2571.0 MINIMUM MINIMUM KJc T MOD. @ 70F @ 70F (ksfVf n) 170.9 58.1 174.4 60.1 190.5 72.3 198.7 102.9 203.9 83.7 206.7 106.4 230.2 101.2 230.2 101. 2 267.6 135.2 270.l 141. 5 280.3 149.1 281.5 150.0 289.4 155.7 "293.9 159.0 170.9 58 .1 174.4 60.1 190.5 72.3 199:5 111. 5 201.3 100.1 203.9 83.7 267.6 135.2 '269. 7 141.3 270. l 141. 5 270.6 137.5 293.9 159.0 "329. 5 196.2 e e Table 4-5 )::o ANALYSIS OF THERMAL AGING OF STEEL "'C REACTOR COOLANT PUMP CASINGS AT SSOF "'C Cl) ::l REPORT 15 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL 0. -'* 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 )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 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 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 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 N 7H37A 0.0 4.39 8826.0 0.33 147.10 2349.0 256.5 184.8 I J:> Cl) ::I a. -'* x )::> N w HEAT NO. MEASURED FERRITE CONTENT (%)
• RCP PUMP CASING 661-N-0766 B7J1R 10.0 B7H15R 8.0 B7Hl3R 11.0 4850 0.0 B7Hl4R 13.0 4854 o.o 03378A o.o 033788 o.o 7F43A 6.0 03252 6.0 03033A 7.5 03067A 5.0 0214C 5.0 03100 . 5.5
• RCP PUMP CASING 661-N-0767 B7JlR 10.0 B7Hl5R 8.0 B7Hl3R 11.0 4848 0.0 4852 o.o B7H14R 13.0 7F43A o.o 02475C o.o 03252 6.0 02362A 0.0 03100 5.5 10184 0.0 e. Table 4-5 (Continued*)

ANALYSIS OF THERMAL AGING OF STAINLESS STEEL REACTOR PUMP CASINGS AT 550F REPORT 15 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL MATERIAL AGING PARAMETER 24.34 23.41 18.91 17 .81 15.91 16.56 12.70 12.70 8.47 7.84 7.20 7.14 6.70 6.45 24.34 23.41 18.91 . 17. 79 17 .34 15.*i1 .8. 7 7.86 .7.84 8.26 6.45 4.10 . MATERIAL CONSTANT c @ 550F 3155.0 3244.0 3787.0 3956.0 4290.0 4170.0 5022.0 5022.0 6470.0 6753.0 7068.0* 7102.0 7336.0 7472.0 3155.0 3244.0 3787.0 3959.0 4033.0 4290.0 6470.0 6746.0 6753.0 6564.0 7472.0 9050.0 MATERIAL CONSTANT N @ 550F 0.28 0.28 0.29 0.29 0.29* 0.29 0.30 0.30 0.31 0.32 0.32 0.32 0.32 0.32 0.28 0.28 0.29 0.29 0.29 0.29 0.31 0.32 0.32 0.31 0.32 0.34 e MINIMUM IMPACT ENERGY (ft-lbs) 25.76 27.09 35.72 38.58 44.48 42.31 58.37 58.37 89.33 95.83 103.20 104.00 109.60 112.90 25.76 27.09 35.72 38.62 39.91 44.48 89.33 95.67 95.83 . 91.46 112. 90 153.00 MINIMUM Jlc @ 550F (1 b/1 n) 912.3 941.4 1064.0 1151.0 1224.0 1218.0 1395.0 . 1395.0 1783.0 1787.0 1886.0 1897.0 1973.0 2015.0 912.3 941.4 1064.0 1162. 0 1170.0 1224.0 . 1783.0 1784.0 1787.0 1814.0 2015.0 2305.0 I MINIMUM* KJc @ 550F (ks1/1n) 159.9 162.4 172. 7 179.6 185.2 184.7 197.7 197.7 223.5 223.8 229.9 230.5 235.1 237.6 159. 9 . 162.4

• 172. 7 180.5 181.0 185.2 .223. 5 '223.6 223.8 225.5 .237.6 254.l . MINIMUM T MOD. @ 550F 77 .0 7B.9 90.B 141.0 101.0 141. 6 116.0 116.0 143.3 149.2 154.7 155.3 159.3 161. 6 77 .0 7B.9 90.8 157.9 133.7 101. 0 143.3 149 .1 149.2 145 .1 161. 6 189.5 e RCP PUMP CASING 661-N-0764 661-N-0765 661-N-0766 661-N-0767 Table 4-6 Limiting and Controlling Values of Jic and KJc at 70°F Jlc HEAT ! (lb/in) 4847 675.2 A7B24R 607.3 B'?JlR 691. 4 B7J1R 691. 4 KJc (ksi/ in) 168.9 160.2 170.9 170.9 RCP PUMP CASING 661-N-0764 661-N-0765 661-N-0766 661-N-0767 Table 4-7 Limiting and Controlling Values of J 1 and KJ at 550°F c c Jic HEAT i (lb/in) 4847 905.6 A7B24R 827.9 B7J1R 912.3 B7J1R 912.3 KJc (ksi/in) 159.3 152.3 159.9 159.9

--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 Membrane stress Bending stress Thickness Appendix.A = #83, 84, 85, and 86 in finite element model = 26.34 Ksi (p. xi, Table 5.0, Rev. 1) = 8.64 Ksi (p. xi, Table 5. 0-, Rev. 1) = 3.5 in. (estimate) 27 e 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 {fraction} 0.08 --0.10 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 a (a) for At = 0.01 a (b) for At = 0.01 Appendix A KI (KSI /IN} 56.97 68.27 77.88 86.68 94.76 102.9 111.2 120.9 131.8 Ll°. = 27 .3 1 +/-T. = 14.8 1 da/dT (IN/YEAR) 5.69 x 10-3 1.17 x 10-2 1. 99 x 10-2 3.05 x 10-2 4.36 x 10-2 6.06 x 10-2 8.26 x 10-2 0.116 0.163

---

---

A Time {YEARS} 16.7 20.2 (a) 11. 9 (b) 7.8 5.4 3.9 2.8 2.0 1.4 29 a/t Interval {fraction 0.08 --0.10 a .10 --a .15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 a (a) for t:,.t = 0.01 a (b) for t:,.t = 0.01 Appendix A Table 5-2 Palisades RCP Casing Crack Growth Rates At Crotch Region KI {KSI /IN) 64.05 76.26 86.45 95.65 103.9 112.2 120.5 130.3 141.4 I:T. = 16. 4 1 I:T; = 9.1 da/dT {IN/YEAR) 9.09 x I0-3 1.83 x io-2 3.02 x Ia-2 4.52 x io-2 6.30 x io-2 8.57 x io-2 0.114 0.156 0.216 t:,.Time {YEARS} 9.9 12.3 (a) 7.4 (b) 5.0 3.5 2.6 2.0 1.4 1.0 30 a/t Interval {fraction} 0.08 --0.10 0 .10 --0 .15 0 .15 --0. 20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Appendix A Table 5-3 Palisades RCP Casing Crack Growth Rates At Discharge Nozzle KI da/dT {KSI Im} {IN/YEAR} 38.05 1.13 x 10-3 46.61 2.55 x 10-3 54.34 4.71 x 10-3 61.60 7.82 x 10-2 68.83 1.21 x 10-2 76.11 1.81 x 10-2 83.8 2.67 x 10-2 92.61 3.97 x 10-2 102.5 5.98 x 10-2 t:..Time {YEARS} 61.8 68.6 37.1 22.3 14.4 9.6 6.5 4.4 e 2.9 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 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. 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 L========--=-:..:.--=--=--=-==-=-=====-=--=-====-=-=-=-=-=-===-------------------------- - 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 5-3 5-4 Byron Byron Byron Jackson, Stress Jackson, Stress Jackson, Stress 1, November 1969 Appendix A Report Report Report Summary, Revision 1, May 1970 Volume IIA, Pump Case, May 1969 Volume IIA, Pump Case, Revision 35 Appendix B APPENDIX B APPLICATION OF GENERIC METHODOLOGY FOR RELAXATION OF THE FORT CALHOUN REACTOR COOLANT PUMP CASING INSPECTION INTERVAL 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 1. 0 2.0 3.0 4.0 3.1 3.2 3.3 PURPOSE PRE-SERVICE INSPECTION DATA EVALUATION OPERATING HISTORY Design Specifications Stress Cycles Used In Evaluation Stress Cycles at Fort Calhoun To-date THERMAL EMBRITTLEMENT 4.1 Material Identification and Chemical Properties 4.2 4.3 Material Specifications and Mechanical Properties Thermal Aging Behavior 4.4 Toughness Properties of Aged Materials

4. 5 Limiting 5.0 CRACK GROWTH ANALYSIS 5.1 sc_ope 5.2 Reference Stress Reports 5.3 Selection of High Stress Locations 5.4 Stresses and Wall Thicknesses at Limiting Locations 5.5 Calculation of crack Growth Rates 5.6 Stresses Under Emergency and Faulted Conditions 5.7 Results 6.0 INSPECTION INTERVAL APPENDIX B REFERENCES Appendix B 3 Page 5 6 8 8 9 9 10 10 10 11 11 12 30 30 30 30 31 32 37 39 40 41 TABLE 4-1 TABLE 4-2 TABLE 4-3 TABLE 4-4 TABLE 4-5 TABLE 4-6 TABLE 4-7 TABLE 5-1 TABLE 5-2 TABLE 5-3 TABLE 5-4 Appendix B APPENDIX B LIST OF TABLES Material Identification and Chemical Compositions Material Specifications and Tensile Pl;'operties Predicted Thermal Aging Behavior Predicted Toughness Properties of Aged Materials (70°F) Predicted Toughness Properties of Aged Materials (550°F) Limiting and Controlling Values of Jic and KJc at 70°F Limiting and Controlling Values of Jic and KJc at 550°F Crack Growth Rate at Junction of Scroll and suction Nozzle Flange Crack Growth Rate at Scroll Crack Growth Rate at Junction of Scroll and Bolt Circle Flange Crack Growth Rate at Diffuser Vane Number 7 13 16 19 22 25 28 29 33 34 35 36 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: Transient Condition Assumed Occurrences During 40 Year License Period Heat-Up (lOO"F/hr) Cool-Down (lOO"F/hr) Hydrostatic Test (3125 psia 100-400.F) Leak Test -In Conjunction with Heatup (2100 psia 100-400.F) Loss of Secondary Pressure Reactor Trip or Loss of Load Appendix B 500 500 10 200 5 400 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 e. Details of the actual operating history of the Fort Calhoun 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 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 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 and 55o*F are and Report #5 listed for all in Report #3. toughness properties of aged material at 70°F given respectively in Report #4 (Table 4-4) (Table 4-5). The measured ferrite contents heat numbers are the same as the values given 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 ----------===

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e e ANALVIS OF THERMAL AGING OF CAST STAINLESS STEEL :;i::. Table 4-1 REACTOR COOLANT PUMP CASINGS "'O "'O CD REPORT 11 G MATERIAL IDENTIFICATION & CHEMICAL COMPOSITIONS

I a. ..4, x MATERIAL OJ 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 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 f i '/ e i ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL* I Table 4-1 (Continued)

REACTOR COOLANT PUMP CASINGS ):::> i "'O "'O REPORT #1 -MATERIAL IDENTIFICATION & CHEMICAL COMPOSITIONS m ::::I a. MATERIAL -'* i x. 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 ! I CASE ASSEM. 517116. 0.05 1.06 0.35 0.020 0.038 19.19 9.73 2.23 0.04 0,.00 I! 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 I' 1/ CASE ASSEM. 714649 0.06 1.21 0.59 0.018 0.027 18.88 10.82 2.24 0.04 0.00 I 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 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 CASE ASSEM. 517571
• 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::.

):> "C "C CD :l a. -'* x OJ ...... U1 Table 4-1 (Continued) . . MATERIAL HEAT NO. c CASING WELD 01765 0.03 CASE ASSEM. 714334 0.04 CASE ASSEM. 517312 0.07 CASE ASSEM. 517246 0.06 CASING WELD L933 0.04 CASING WELD 01746 0.02 CASING WELD 03100 0.03 CASING WELD 03681 0.03 CASING WELD 02964E 0.03-CASING WELD 01124K 0.03 ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS REPORT #1.

• MATERIAL IDENTIFICATION

& CHEMICAL COMPOSITIONS Mn Si s p Cr Ni Mo N Cb 2.32 0.36 0.028 0.010 23.44 12.84 o.oo 0.04 o.oo 0.66 0.38 0.020 0.032 19.29 9.73 2.29 0.04 0.00 1.30 0.75 0.023 0.043 18.98 10.*92 2.21 0.04 0.00 1.28, 0.80 0.023 0.028 18.54 10.63 2.40 0.04 0.00 0.77 0.95 0.012 0.011 19.80 8.30 0.00 0.04 0.00 1.69 0.42 0.012 0.011 20.99 9.47 o.oo 0.04 0.00 1.66 0.40 0.018 0.010 19.06 12.20 2.25 0.04 0.00 1. 73 0.52 0.016 0.010 19.95 9.90 o.oo 0.04 0',55 1. 73 0.52 0.016 0.012 19.22 9.40 0.00 0.04 0.66 1.90 0.47 0.026. 0.009 19.56 10.71 0.00 0.04 0.51 e e ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS )::o Table 4-2

• REPORT 12
• MATERIAL SPECIFICATION

& TENSILE PROPERTIES "t:J -c m MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN UNAGED AGED AGED ::I a. OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA FLOW FLOW FLOW ..... x 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 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 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 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL "t:J "t:J REACTOR COOLANT PUMP CASINGS CD ::I 0.. REPORT 12

• MATERIAL SPECIFICATION

& TENSILE PROPERTIES -'* x 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 . --.

)> "O "O rt> :::I 0.. -lo >< OJ ..... 00 e Table 4-2 MATERIAL OR PART CASING WELD CASING WELD CASING WELD CASING WELD CASING WELD CASING WELD . I .e (Continued) ANALYSIS OF THERMAL AGING OF CAST STEEL REACTOR COOLANT PUMP CASINGS REPORT 12 -MATERIAL SPECIFICATION & TENSILE PROPERTIES MATERIAL MATERIAL YIELD TENSILE . TOTAL RED. IN SPEC. TYPE STRENGTH STRENGTH ELONG. AREA . HEAT NO. (ks1) (ks1) (%) (%) A 362 CF8M L933 0.0 0.0 o.o o.o E347 01746 0.0 0.0 0.0 o.o E316*16 03100 o.o o.o o.o 0.0 E347-16 03681 o.o 0.0 o.o o.o E347*16 02964E o.o o.o o.o o.o E347-16 01124K o.o 0.0 o.o 0.0 UNAGED AGED AGED FLOW FLOW FLOW STRESS STRESS STRESS @ 70F @ 70F @ SSOF 0 78600 58000 0 78600 58000 *O 78600 58000 0 78600 58000 0 78600 58000 0 . 78600 58000 ' . e e e )::> ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL "'C REACTOR COOLANT PUMP CASINGS "'C Table 4-3 Cl> ::::s 0.. REPORT 13 -PREDICTED THERMAL AGING BEHAVIOR I x' 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 7 o 0 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

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e Table 4-3 (Continued)

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL )> "'C REACTOR COOLANT PUMP CASINGS -0 fl) ::I REPORT 13 -PREDICTED THERMAL AGING BEHAVIOR 0. -'* 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 714334 13.0 17.253 14.407 1.20 11.8 13.6 22.150 12.610 14.5 0 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 22.140 14.210 10.7

)> ""C ""C CD :::s 0.. ...... >< OJ N ....... Table 4-3 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS . REPORT 13

• 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 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 I '

---

e Table.4-4

)::>> ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR -0 . -0 COOLANT PUMP CASINGS AT 70F fl) :::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) ( 1 b/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 022918 5.0 4.73 15265.0 0.45 140.50 2334.0 314.0 184.6 N 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 (Continued)

t> "'C "'C m :::I a. -'* >< o:J N w HEAT NO. MEASURED FERRITE CONTENT (%)

• RCP PUMP CASING 671-N-0030 714334 13.0 517116 12.0 03314A 8.0 L933 o.o 03314 6.5 Oi765 9.0 714649 8.0 01746 0.0 714737 8.0 03252 6.0 03107 5.5 714637 7.0 03681 0.0 02964E 0.0 01124K 0.0
• RCP PUMP CASING 671-N-0031 L933 0.0 517312 7.0 517246 8.0 714737 8.0 752627 0.0
• RCP PUMP CASING 671-N-0032 517571 10.0 714334 13.0 L933 0.0 01765 9.0 e. ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR . COOLANT PUMP CASINGS AT 70F REPORT 14 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL MATERIAL AGING PARAMETER 16.38 15.41 13.45 12.21 12.78 12.14 11.14 9.78 10.40 7.84 7.15 7.72 5.92 5.02 3.71 12.21 . 12.00 11.73 10.40 5.7;4 16.76 16.38 12.21 12.14 MATERIAL CONSTANT c @ 70F 5793.0 6140.0 6977.0 7624.0 7315.0 7664.0 8268.0 9235.0 8773.0 10994.0 11758.0 11127 .0 13358.0 14769.0 17268.0 7624.0 7742.0 7903.0 8773.0 13618.0 5666.0 5793.0 7624.0 7664.0 MATERIAL CONSTANT N @ 70F 0.39 0.40 0.40 0.41 0.40 0.41 0.41 0.42 0.41 0.43 0.43 0.43 0.44 0.45 0.46 0.41 0.41 0.41 0.41 0.44 0.39 0.39 0.41 0.41 e MINIMUM IMPACl ENERGY (ft-lbs) 42.90 46.27 54.54 61.06 57.93 61.46 67.60 77.54 72.78 95.83 103.80 97.22 120.60 135.30 161.30 61.0G 62.25 63.88 72. 78* 123.30 41.67 42.90 61.06 61.46 MINIMUM Jlc @ 70F ( l b/1n) 1026.0 1041.0 1179.0 1241.0 1245.0 1248.0 1424.0 1476.0 1524.0 1727 .0 1873.0 1890.0 2080.0, 2238.0 2596.0 1241.0 1315.0 1357.0 1535.0 2130.0 1003.0 1037.0 1241.0 1248.0 --MINIMUM MINIMUM KJc T MOD. @ 70F @ 70F (ks1V1n) 208.1 109.9 *209. 7 114.1 223.2 96.6 229.0 103.9 229.3 100.7 229.6 104.4 245.2 172.7 249.7 122.3 253.7 175.0 270.1 141. 5 281.3 149.8 282.5 252.9 *296.4 166.0 307.5 179.6 331.1 203.2 229.0 103.9 235.7 161. 4 239.4 175.1 254.7 187.1 299.9 168.6 205.8 112 .2 209.3 128.3 *229.0 103.9 229.6 104 .*

e Table 4-4 (Continued) I )> ANALYSIS OF THERMAL AGING OF CAST ,*STAINLESS STEEL REACTOR "O .COOLANT PUMP CASINGS AT 70F "O 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" ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "'C REACTOR COOLANT PUMP CASINGS AT 550F fl) :::s 0... -'* REPORT 15 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL x 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 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 022918 5.0 4.73 8573.0 0.33 140.50 2267.0 252.0 180.8 U'1 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> ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "O "O REACTOR COOLANT PUMP CASINGS AT 550F n> :;, c.. REPORT #5 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL ....... x 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 714334 13.0 . 16.38 4203.0 0.29 42.90 1255.0 187.6 182.6 O'I L933 0.0 12.21 5156.0 0.30 61.06 1439.0 200.8 118. 5 0176 *. 9.0 12.14 5176.0 0.30 61.46 1444.0 201.2 118.8 e -

e -e Table 4-5 (Continued) )> ANALYSIS OF THERMAL AGING OF STAINLESS STEEL -0 REACTOR COOLANT PUMP CASINGS AT 550F -0 11> ::I a. REPORT #5

• PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL -'* x o:J HEAT MEASURED MATERIAL MATERIAL MATERIAL 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) 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 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 "CJ I I 0, I I I RCP PUMP CASING 671-N-0029 671-N-0030 671-N-0031 671-N-0032 Table -Fort Calhoun Limiting and Controlling Values of Jic and KJc at 70°F Jic HEAT Jt (lb/in) 516942 495.1 714334 1026.0 L933 1241. 0 517571 1003.0 KJc (ksi/in) 144.6 208.1 229.0 205.8 I RCP PUMP CASING 671-N-0029 671-N-0030 671-N-0031 671-N.;.,0032 Table 4-7 Limiting and Controlling Values of Jic and KJc at 550"F Jic HEAT .i (lb/in) 516942 715.0 714334 1233.0 L933 1439.o* 517471 1214.0 KJc (ksi/ in) 141.5 185.8 200.8 184.5 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, .. (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 Bending stress Thickness = 27.7 Ksi (p. 208) 2.2 Ksi (p. 208) = 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 Bending stress Thickness = 28.33 Ksi (p. 220) = 20.5 Ksi (p. 220) = 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

• a/t Interval (fraction}

0.08 --0.10 0.10 --0.15 0 . .15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Table 5-1. Fort Calhoun RCP Casing crack Growth Rates At of Scroll and Suction Nozzle Flange (Region 3) KI da/dT (KSI /TN) (IN/YEAR) 31.44 7.52 x 10"" 4 38.18 1.63 x 10-3 44.13 2.92 x 10-3 49.72 4.70 x 10-3 55.04 7.06 x 10-3 60.46 1. 03 x 10-2 66.10 1.47 x 10-2 72.62 2.14 x 10-2 79.95 3.14 x 10-2 (am= 24.47, ub = 13.82, t = 2.06") Appendix B fl.Time (YEARS} 54.8 63.0 35.3 21.9 14.6 10.0 7.0 4.8 3.3 /lit = 2. 544 /IN 33 a/t Interval (fraction} 0.08 --0.10 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Table 5-2 Fort Calhoun RCP Casing crack Growth Rate At Scroll (Region 4) KI da/dT (KSI /IN) (IN/YEAR) 25.56 3.29 x 10-4 31. 71 -4 7. 78 x 10 . 37.42 1. 51 x 10-3 42.93 2.61 x 10-3 48.41 4.23 x 10-3 54.03 6.56 x 10-2 60.05 1.00 x 10-2 66.86 1.54 x 10-2 74.58 2.38 x 10-2 (um= 27.7, ub = 2.2, t = 2.06") Appendix B (YEARS} 125. 132. 68.2 39.4 24.3 15.7 10.3 6.7 4.3 e 34 a/t Interval {fraction} 0.08 --0.10 0 .10 --0 .15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Table 5-3 Fort Calhoun RCP Casing Crack Growth Rate At Junction of Scroll and Bolt tircle Flange (Region 5) KI da/dT {KSI JIN) {IN/YEAR) 30.02 6.25 x 10-4 36.72 1. 40 x 10-3 42.75 2.57 x i0-3 48.46 4.24 x 10-3 54.0 6.54 x 10-3 59.65 9.74 x 10-3 65.59 1. 42 x 10-2 72.41 2.12 x 10-2 80.10 3.17 x 10-2 (am= 21.9, ab = 8.0, t = 3.0") Appendix B t.T {YEARS} 96. 107. 58.4 35.3 22.9 15.4 10.5 7.1 e 4.7 35 a/t Interval {fraction} 0.08 --0.10 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Table 5-4 Fort Calhoun RCP Casing crack Growth Rate At Diffuser Vane Number 7 KI da/dT (KSI Im} {IN/YEAR} 41.95 2.38 x 10-3 51.99 5.62 x 10-3 61.28 1. 08 x 10-2 70.24 1.87 x 10-2 79.13 3.02 x 10-2 88.24 4.67 x 10-2 98.0 7 .1. x 1.0-2 109.05 0.109 121.5 0.168 {am= 28.33, ab= 2.77, t = 5.15") -1% a/t steps: = 65.4 (a) = 31.1 (b) Appendix B {YEARS} 43.2 45.8 (a) 23.7 (b) 13.7 8.5 5.5 3.6 2.3 1.5 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. Appendix B 5.6.1 Emergency Condition Stresses (1) Junction, Scroll to suction Nozzle Flange ("Region 3 11) 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 Section 1. 0 2.0 3.0 PURPOSE APPENDIX C TABLE OF CONTENTS Title PRE-SERVICE INSPECTION DATA EVALUATION OPERATING HISTORY Page 6 7 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 4.2 Material Specifications and Mechanical Properties 4.3 4.4 Thermal Aging Behavior Toughness Properties of Aged Materials 4.5 Limiting Values 5.0 CRACK GROWTH ANALYSIS 5.1 5.2 5.3 Scope Reference Stress Reports Selection of High Stress Locations 5.4 Stresses and Wall Thicknesses at Limiting Locations 5.5 Calculation of Crack Growth Rates 5.6 Stresses Under Emergency and Faulted Conditions 5.7 Results 6.0 INSPECTION INTERVAL APPENDIX C REFERENCES Appendix C 3 12 12 13 13 14 39 39 39 39 40 42 48 51 52 53 TABLE 4-1 TABLE 4-2 TABLE 4-3 TABLE 4-4 TABLE 4-5 TABLE 4-6 TABLE 4-7 TABLE 4-8 TABLE 4-9 TABLE 4-10 TABLE 4-11 Appendix C LIST OF TABLES Material Identification and Chemical Compositions

-Unit 1 Material Specifications and Tensile Properties -Unit 1 Predicted Thermal Aging Behavior -Unit 1 Predicted Toughness Properties of Aged Materials (70°F) -Unit 1 Predicted Toughness Properties of Aged Materials (550°F) -Unit 1 Limiting and Controlling Values of Jic and KJc at 70°F -Unit 1 Limiting and Controlling Values of Jic and KJc at 550°F -Unit 1 Material Identification and Chemical Compositions -Unit 2 Material Specifications and Tensile Properties -Unit 2 Predicted Thermal Aging Behavior -Unit 2 Predicted Toughness Properties of Aged Materials (70°F) -Unit 2 4 15 17 19 21 23 25 26 27 29 31 33 TABLE 4-12 TABLE 4-13 TABLE 4-14 TABLE 5-1 TABLE 5-2 TABLE 5-3 TABLE 5-4 TABLE 5-5 Appendix c LIST OF TABLES (Continued) Predicted Toughness Properties of Aged Materials (550°F) -Unit 2 Limiting and Controlling Values of Jic and KJc at 70°F -Unit 2 Limiting and Controlling Values of Jic and KJc at 550°F -Unit 2 Crack Growth Rates at Vane Number 8 Crack Growth Rates at Discharge Nozzle-Crotch Vicinity Crack Growth Rates at Suction Nozzle-Level c Crack Growth Rates at Volute Junction With Lower Flange Crack Growth Rates at Hanger Bracket #1 Vicinity 35 37 38 43 44 45 46 47 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: Appendix c Transient Condition Heat-Up (100.F/hr) Cool-Down {l00°F/hr) Hydrostatic Test (3125 psia 100-400°F) Leak Test-In Conjunction With Heatup (2250 psia 100-400°F) Assumed Occurrences During 40 Year License Period 500 500 10 320 Loss of Secondary Pressure 5 Reactor Trip or Loss of Load 400 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: Heatup/Cooldown Reactor Trip -* Unit 1. Unit 2 84 112 57 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 and 550°F are and Report #5 listed for all in Report #J. using equation toughness properties of aged material at 70°F given respectively in Report #4 (Table 4-4) (Table 4-5). The measured ferrite contents heat numbers are the same as the values given The material aging parameter was calculated 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 )> Table 4-1 "'C "'C . ANALVIS OF THERMAL AGING OF CAST STAINLESS STEEL Cl> ::s REACTOR COOLANT PUMP CASINGS 0.. ..... x 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-' CASING WELD 04286 0.02 o .. 90 0.52 0.015 0.015 18.83 10.12 2.41 0.04 0.00 I Ul 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> REACTOR COOLANT PUMP CASINGS -a -a fl) ::s REPORT #1 -MATERIAL IDENTIFICATION & CHEMICAL COMPOSITIONS 0.. ...... >< ("") 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 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 0.017 0.013 18.87 9.86 2.46 0.04 0.00 CASING WELD 04286 0.02 0.90 .o. 52 0.015 0.015 18.83 10.12 2.41 0.04 0.00 CASING WELD X43439 0.03 1.39 1 0.36 0.017 0.016 19.90 9.10 2.31 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 REPORT #2 -MATERIAL SPECIFICATION & TENSILE PROPERTIES a.. -'* 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 CASING WELD 04146 0.0 0.0 0.0 0.0 0 78600 58000 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 CASE SCROLL A 351 CF8M 43393 42.1 83 .7 63.0 67.0 63 75345 54751 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 Table 4-3 REACTOR COOLANT PUMP CASINGS "'O 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 22.750 11. 895 "° 24.3 11. 0 03493 17,0 17 .381 13.521 1.29 15.2 20.5 22.240 12.030 11. 0 03793 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 REACTOR COOLANT PUMP CASINGS "'C CD :::I REPORT 13 -PREDICTED THERMAL AGING BEHAVIOR a.. ...... 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 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR "'C fl) COOLANT PUMP CASINGS AT 70F ::3 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 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 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)

)> -c -c (I) ::I 0.. ...... x n HEAT NO. MEASURED FERRITE CONTENT (%) ANALYSIS OF THERMAL AGING OF CAST STAINlESS STEEL REACTOR COOLANT PUMP CASINGS AT 70F REPORT #4 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL MATERIAL AGING PARAMETER MATERIAL CONSTANT c @ 70F MATERIAL CONSTANT N @ 70F MINIMUM IMPACT ENERGY (ft-lbs) MINIMUM Jic @ 70F (1 b/i n) MINIMUM MINIMUM KJc T MOD. @ 70F @ 70F (ksiVin) Table 4-5 )> ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "tJ REACTOR COOLANT PUMP CASINGS AT SSOF "tJ CD ::I 0.. REPORT #5 -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) (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 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 03793 16.0 . 14.86 4504.0 0.29 48.39 1294.0 190.4 105.3 w 03493 17 .0 13.34 4855.0 0.30 55.09 1342.0
• 193. 9 112. 7 0428 *. .15.0 10.42 5715.0 fl.3.1 . 72.66 1543.0 208.0 .29.8 Table 4-5 (Continued)

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

N U1 RCP PUMP CASING 681-N-0437 681-N-0438 681-N-0439 681-N-0440 Table 4-6 -Calvert Cliffs I Limiting and Controlling Values of J 10 and KJc of 70°F HEAT # 04146 43836 44581 04146 Jic (lb/in) 594.5 728.6 621. 3 594.5 KJc (ksi/in) 158.5 175.4 162.0 158.5 RCP. PUMP CASING 681-N-0437 681-N-0438 681-N-0439 681-N-0440 Table 4-7 Calvert Cliffs I Limiting and Controlling Values of JIC and KJC at 550°F HEAT # 04146 43836 44581 04146 Jic (lb/in) 813.4 951.3 839.3 813.4 KJc (ksi/ in) 151.0 163. 3' 153.4 151. 0 )> ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL "C Table 4-8 REACTOR COOLANT PUMP CASINGS "C 11> ::::s a. REPORT #1 -MATERIAL IDENTIFICATION & CHEMICAL COMPOSITIONS --'* x (""') 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

. "C "C ft> ::J a. ...... x ('") N co Table 4-8 (Continued)

MATERIAL HEAT NO. c CASING WELD 04528 0.02 CASING WELD 04818 0.03 CASING WELD 04754 0.02 CASING WELD 04509 0.02 CASING WELD 04635 0.02 CASING WELD T03951 0.04 CASE SCROLL 50404 0.03 CASING WELD X43439 0.03 CASING WELD 57203 " 0.02

• RCP PUMP CASING 681-N-0444 CASE SCROLL 51226 0.06 HUB/DIFFUSER 50658 0.04 CASING WELD 04754 0.02 CASING WELD 04635 0.02 CASING WELD 57203 0.02 ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS REPORT #1 -MATERIAL IDENTIFICATION

& CHEMICAL COMPOSITIONS Mn Si s p Cr Ni Mo N Cb 0.98 . 0.51 0.014 0.013 20.29 10.30 2.62 0.04 0.00 1.01 0.49 0.020 0.012 19.56 10.05 2.39 0.04 0.00 0.97 0 . .47 0.017 0.014 19.51 9.87 2.46 0.04 0.00 0.94 0.45 0.018 0.015 19.67 9.75 2.51 0.04 0.00 1.00 0.015 0.013 19.40 10.00 2.71 0.04 0.00 1.55 0.50 0.014 0.025 19.02 10.03 2.44 0.04 0.00 0.51 1.03 0.015 0.035 18.32 9.52 2.20 0.04 0.00 1.39 0.36 0.017 0.016 19.90 9.10 2.31 0.04 '0.00 0.66 0.48 0.017 0.039 18.65 10.20 2.41 0.04 0.00 0.66 0.80 0.015 0.039 19.39 9.57 2.21 0.04 0.00 0.58 1.31 0.015 0.036 18.37 9.37 2.30 0.04 0.00 0.97 0.47 0.017 0.014 19.51 9.87 2.46 0.04 0.00 1.00 0.49 0.015 0.013 19.40 10.00 2. 71 0.04 0.00 0.66 0.48 0.017 0.039 18.65 10.20 2.41 0.04 0.00 )> Table 4-9 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL "O "O REACTOR COOLANT PUMP CASINGS l'D ::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 HUB/DIFFUSER A 351 CF8M 48396 45.4 87.2 57.0 76.0 66 80728 60134 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 CASE SCROLL A 351 CF8M 49122 43.5 85.0 50.0 75.0 64 77482 56888 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 04528 0.0 0.0 0.0 0 78600 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 REACTOR COOLANT PUMP CASINGS "'C n> ::s 0.. --'* )< n w C> REPORT #2 -MATERIAL SPECIFICATION & TENSILE PROPERTIES MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA PART HEAT NO. (ks1) (ks1) (%) (%) ,. CASING WELD A 351 CF8M 04754 o.o 0.0 0.0 0.0 CASING WELD A 371 ER-316 04509 0.0 0.0 0.0 0.0 CASING WELD 04635 o.o 0.0 0.0 0.0 CASING WELD A 298 E316-16 T03951 0.0 0.0 o.o 0.0 CASE SCROLL A 351 CFAM 50404 38. l 74.9 53.0 75.0 CASING WELD A 298 E316-16 X43439 o.o 0.0 0.0 0.0 CASING WELD A 298 E316-16 57203 o.o 0.0 0.0 0.0

• RCP PUMP CASING 681-N-0444 CASE SCROLL A 351 CF8M 51226 39.l 80.7 50.0 75.0 HUB/DIFFUSER A 351 CF8M 50658 35.2 89.7 52.0 54.0 CASING WELD A 351 CF8M 04754 o.o o.o o.o 0.0 CASING WELD 04635 0.0 0.0 0.0 0.0 CASING WELD A 298 E316-16 57203 0.0 o.o 0.0 0.0 UNAGED AGED AGED FLOW FLOW FLOW STRESS STRESS STRESS @ 70F @ 70F @ 550F 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 56 65211 44617 0 78600 58000 0 78600 58000 60 70595 50001 62 74632 54038 0 78600 58000 0 78600 58000 0 78600 58000

)> Table 4-10 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL -c -c REACTOR COOLANT PUMP CASINGS C'D ::I a. -'* REPORT #3 -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 (%) #2 (%) WELDS WELDS {%) ** PLANT 1.0. CALVERT CLIFFS UNIT 2

• RCP PUMP CASING 681-N-0441 48396 12.0 17 .547 14.669 1.20 13.2 13.5 22.875 12.930 10.0 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 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 49122 14.0 17.251 14.474 1.19 14.2 13.2 23.185 12.795 12.0 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 04528-19.0 *18.715 14.396 1.30 17 .2 e 21.8 23.675 12.590 14.0 04818 0.0 17.697 14.271 1.24 14.2 16.8 22.685 12.505 11. 0

)> 'U 'U rD ::3 a.. -'* )( ("') w N Table 4-10 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS 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 e Table 4-11 . )> ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR "'C COOLANT PUMP CASINGS AT 70F "'C 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 I 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 04635 18.0 14.83 6366.0 0.40 48.49 1061.0 211. 7 88.9 I
• T03951 0.0 13.79 6818.0 0.40 52.96 1149 .o 220.3 94.5 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
• RCP PUMP CASING 681-N-0443 w 51799 15.0 24.96 3857.0 0.38 24.94 675.3 . 168. 9* 58.8 X434. 14.0 19.87 4804.0 33.51 815.6 185.6 6tt e Table 4-11 (Continued) . )> ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR "'C C.OOLANT PUMP CASINGS AT 70F "'C C1) ::s a. 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) (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

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e Table 4-12 )> ANALYSIS OF THERMAL AGING OF STAINLESS STEEL "O REACTOR COOLANT PUMP CASINGS AT 550F "O (I) ::I 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 @ 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 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
• RCP PUMP CASING 681-N-0443 01 51799 15.0 24.96 3099.0 0.28 24.94 897.2 158.6 79.5 X43439 14.0 19.87 3654.0 33.51 1023.0 169.3 88.0 e. -e e Table 4-12 (Continued)

)> "'C "'C '1> :::J 0.. ...... x n HEAT MEASURED NO. FERRITE CONTENT (%) 04528 19.0 04509 19.0 04818 0.0 04635 18.0 04754 18.0 T03951 0.0 50404 13.0 57203 0.0

• RCP PUMP CASING 681-N-0444 w O'I 51226 , 50658 04635 04754 57203 12.0 14.0 18.0 18.0 14.0 ANALYSIS OF THERMAL AGING OF STAINLESS STEEL REACTOR.COOLANT PUMP CASINGS AT 550F REPORT #5 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL MATERIAL MATERIAL MATERIAL . MINIMUM MINIMUM MINIMUM MINIMUM AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. PARAMETER .c N ENERGY * @ 550F @ 550F @ 550F @ 550F @ 550F (ft-lbs) (lb/in) (ksi/in) 18.60 3832.0 0.29 36.47 1079.0 173.9 91.8 15.76 4319.0 0.29 45.00 1234.0 186.0 101. 6 14.93 4489.0 0.29 48.12 1289.0 190.1 104.9 14.83 4509.0 0.29 48.49 1295.0 190.5 105.3 14.34 4617 .0 0.30 50.51 1268.0 188.5 108 .1 13.79 4745.0 0.30 . 52. 96 1307.0 191.4 110.6 12.50 5074.0 0.30 59.41 1473.0 203.2 187 .8 9.58 6016.0 0.31 79.18 1638.0 214.3 135.3 21.25 3479.0 0.28 30.71 1035.0 170.3 110.5 16.72 4140.0 0.29 41.78 1187 .0 182.4 .111.6 14.83 4509.0 0.29 48.49 1295.0 190.5 105.3 14.34 4617 .0 0.30 50.51 1268.0 188.5 108 .1 9.58 6016.0 0.31 79.18 1638.0 . 214.3 135. 3 w -.J RCP PUMP CASING 681-N-0441 681-N-0442 681-N-0443 681-N-0444 Table 4-13 -Calvert Cliffs 2 Limiting and Controlling Values of Jic and KJc at 70°F HEAT # 48396 48898 51799 51226 Jic (lb/in) 783.3 639.5 675.3 766.8 KJc (ksi/in) 181.9 164.4 168.9 180.0 RCP PUMP CASING 681-N-0441 681-N-0442 681-N-0443 681-N-0444 Table 4-14 -Calvert Cliffs 2 Limiting and Controlling Values of HEAT # 48396 48898 51799 51226 Jic and KJc at 550°F Jic (lb/ in) 1002.0 859.9 897.2 1035.0 KJc (ksi/in) 167.6 155.2 158.6 170.3 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=

1. 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: Appendix C Design Condition Key Element Membrane stress Bending stress Thickness = # 107, plus thermal gradient stresses = # 5125 in Finite-Element Model = 17.4 Ksi (Table 4-6) = 21.1 Ksi (Figure 4-8, p. 66 & Figure 3-1) = 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. 40 (3) For Suction Nozzle--Level C: Design Condition Key Elements = Membrane stress Bending stress = Thickness = = # 104, plus thermal gradient stresses # 2125 & 2129 in Finite-Element Model = 22.15 Ksi (Figure 4-5(b)) 13.15 Ksi (Figures 3-1 & 4-5(b)) 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 Element Model Membrane stress = Bending stress_ = 17.82 Ksi (page 78) 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 Thickness = 22.1 Ksi (page 75, 97) = 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 Appendix c 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: (3) da/dT = 5.435 x 10-lO K 4.o I Units for da/dT are inches/year. Incremental time, dT, in which the crack will through the indicated dimensionless crack depth 42 a/t Interval {fraction} 0.08 --0.10 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Table 5-1 Calvert Cliffs 1 & 2 crack Growth Rates At KI Vane Number 8 Level D {KSI /Ill} 49.98 60.08 68.76 76.76 84.17 91.69 99.35 108.3 118.4 da/dT {IN/YEAR} 3.39 x 10-3 7.08 x 10-3 1.21 x 10-2 1.89 x 10-2 2.73 x 10-2 3.84 x 10-2 5.29 x 10-2 7.48 x 10-2 0.107 (um= 20.95, ab* 19.87, t

• 4.75")

{YEARS} 28.0 45.6* 24.4* 12.5 8.7 6.2 4.5 3.2 2.2

• Sum of five time steps through 1% a/t increments using interpolated KI values. Appendix c 43 .;; *H Table 5-2 Calvert Cliffs 1 & 2 Crack Growth Rates At Discharge Nozzle Crotch Vicinity -Section c a/t Interval KI da/dT {fraction}

{KSI /IN} {IN/YEAR} 0.08 --0.10 38.9 I. 25 x 10-3 0 .10 --0 .15 46.6 2.56 x 10-3 0.15 --0.20 53.0 4.30 x 10-3 0.20 --0.25 58.9 6.55 x 10-3 0.25 --*o.3o 64.3 9.28 x 10-3 0.30 --0.35 69.7 I. 28 x 10-2 0.35 --0.40 75.2 I. 74 x 10-2 0.40 --0.45 81.6 2.41 'x 10-2 0.45 --0.50 88.9 3.39 x 10-2 (O'm = 17.4, O'b = 21.1, t = 3.3") Appendix C {YEARS} 52.8 64.5 38.4 25.2 17.7 12.8 9.5 6.8 4.8 44. a/t Interval (fraction} 0.08 --0.10 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0. 45 --0. 50 . Appendix c Table 5-3 Calvert Cliffs 1 & 2 Crack Growth Rates At Suction Nozzle -Level C KI da/dT (KSI /IN) (IN/YEAR) 34.92 8.08 x 10-4 42.37 1. 75 x 10-3 48.93 3.12 x 10-3 55.08 5.00 x 10-3 60.93 7.49 x 10-3 66.88 1.09 x 10-2 73.05 1. 55 x 10-2 80.20 2.25 x 10-2 88.25 3 .30 x 10-2 . AT (YEARS} 74.2 85.6 48. l 30.0 20.0 13.8 9.7 6.7 4.5 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 {aLt = } {KSI /IN} {INLY EAR} 0.08 --0.10 32.27 5.9 x 10-4 0.10 --0.15 38.99 1. 26 x 10-3 0.15 --0.20 44.85 2.2 x 10-3 0.20 --0.25 50.29 3.48 x 10-3 0.25 --0.30 55.41 5.12 x 10-3 0.30 --0.35 60.61 7.33 x 10-3 0.35 --0.40 65.97 1.03 x 10-2 0.40 --0.45 72.20 1.48 x 10-2 0.45 --0.50 79.21 2.14 x 10-2 (um= 17.8, ub = 13.2, t = 3.375) Appendix c 46 6T (YEARS} 114. 134. 77. 48.5 33. 23. 16.4 11.4 e 7.9 From -To {aLt = } 0.08 --0.10 0 .10 --0 .15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Table 5-5 Calvert Cliffs 1&2 crack Growth Rates At Hanger Bracket #1 Vicinity Kr da/dT (KSr Im) (rNLYEAR) 43.40 1.92 x 10-3 51.90 3.94 x 10-3 59 .10 6.65 x 10-3 65.80 0.0102 71.80 0.0145 77.90 0.02 84.10 0.0272 91.40 0.0379 99.50 0.0533 -* (um = 18.9, ub z 22.1, t

• 3.6") {YEARS} 37.4 61.3* 33.5* 17 .6 12.4 9.0 6.6 4.7 3.4
• 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. Appendix c 5.6.1 Emergency Condition Stresses (1) Diffuser Vane 8--Level D: Key elements Condition = # 3828 & 3832 = # 505 48 Appendix c 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: (3") Key element = # 5125 Condition = # 506 Membrane stress = 26.66 Ksi (p. 118) Bending stress = 16.85 Ksi (p. 118) 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. 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 Appendix c (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) 50 ---------=======

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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 design specification. The hypothesized crack will then grow larger until it 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 Section 1. 0 2.0 3.0 PURPOSE APPENDIX D TABLE OF CONTENTS Title PRE-SERVICE INSPECTION DATA EVALUATION OPERATING HISTORY Page 6 7 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 4.2 Material Specifications and Mechanical Properties 4.3 4.4 Thermal Aging Behavior Toughness Properties of Aged Materials 4.5 Limiting Values 5.0 CRACK GROWTH ANALYSIS 5.1 5.2 5.3 Scope Reference

?tress Reports Selection of High Stress Locations 5.4 Stresses and Wall Thicknesses at Limiting Locations 5.5 Calculation of Crack Growth Rates .5.6 Stresses Under Emergency and Faulted Conditions 5.7 Results 6.0 INSPECTION INTERVAL APPENDIX D REFERENCES Appendix D 3 12 12 13 13 14 43 43 43 43 44 46 47 56 57 58 TABLE 4-1 TABLE 4-2 TABLE 4-3 TABLE 4-4 TABLE 4-5 TABLE 4-6 TABLE 4-7 TABLE 4-8 TABLE 4-9 TABLE 4-10 TABLE 4-11 TABtE 4-12 Appendix D LIST OF TABLES Material Identification and Chemical Compositions -Unit 2 Material Specifications and Tensile Properties -Unit 2 Predicted Thermal Aging Behavior -Unit 2 Predicted Toughness Properties of Aged Materials (70°F) -Unit 2 Predicted Toughness Properties of Aged Materials (550°F) -Unit 2 Limiting and Controlling Values of Jic and KJc at 70°F -Unit 2 Limiting and Controlling Values of Jic and KJc at 550°F -Unit 2 Material Identification and Chemical Compositions -Unit 3 Material Specifications and Tensile Properties -Unit 3 Predicted Thermal Aging Behavior -Unit 3 Predicted Toughness Properties of Aged Materials (70°F) -Unit 3 Predicted Toughness Properties of Aged Materials (550°F) -Unit 3 4 15 17 19 21 23 25 26 27 29 31 33 37 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. required radiograph sensitivity was 2-2T according to applicable ASTM Standard Reference Radiograph Procedure requirements (i.e. ASTM El86, E280) as determined by the casting thickness. The 2-2T sensitivity is 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: Appendix D Transient Condition Heat-Up (60.F/hr) Cool-Down _ (l00°F/hr) Hydrostatic Test (3125 psia 100-400.F) Leak Test (2250 psia 100-400*F) Assumed Occurrences During 40 Year License Period 500 500 10 200 Loss of Secondary Pressure 5 Reactor Trip 400 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 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 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 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 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 )> Table 4-1 REACTOR COOLANT PUMP CASINGS "O "O rt) REPORT #1 -MATERIAL IDENTIFICATION & CHEMICAL COMPOSITIONS

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 HUB/DIFFUSER 05202-1 0.07 0.70 1.28 0.004 0.035 19.48 9.78 2.20 0.04 o.oo 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 0.018 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 Table 4-1 (Continued)

ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL :x:o REACTOR COOLANT PUMP CASINGS "'C "'C fl) REPORT #1 -MATERIAL IDENTIFICATION & CHEMICAL COMPOSITIONS

s a. -'* x MATERIAL 0 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) REPORT #2 -MATERIAL SPECIFICATION

& TENSILE PROPERTIES

I 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 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 HUB/DIFFUSER A 351 CF8M 05202-1 43.2 83.9 56.0 74.0 64 76374 55780 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 ....... CASE SCROLL A 351 CF8M 06652-1 43.8 87.1 55.0 73.0 65 79382 58788 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 "'O 11> ::s 0.. ..... )( c ...... 00 -: Table*4-2 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS REPORT #2 -MATERIAL SPECIFICATION & TENSILE PROPERTIES MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA PART HEAT NO. (ks1) (ks1) (%) (%) CASING WELD A 351 CF8M 7533A o.o o.o 0.0 o.o CASING WELD A 351 CF8M 63683 0.0 0.0 0.0 0.0 CASING WELD A 351 CF8M 64621 o.o o.o 0.0 0.0 CASING WELD A 351 CF8M 77144 o.o 0.0 o.o 0.0

• RCP PUMP CASING 701-N-0560 CASE SCROLL A 351 CF8M .07082*1 42.5 85.9 51.0 65.0 HUB/DIFFUSER A 351 .CFSM 06068-1 36.4 76.3 68.0 71.0 CASING WELD A 351 CF8M 60525 o.o 0.0 o.o 0.0 CASING WELD A 351 CF8M 6546 o.o . 0.0 0.0 0.0 CASING WELD A 351 CF8M 74726 0.0 o.o 0.0 0.0 CASING WELD A 351 CFSM 7553A 0.0 0.0 0.0 o.o CASING WELD A 351 CFSM 63683 o.o o.o 0.0 0.0 CASING WELD A 351 CF8M 64621 o.o 0.0 0.0 0.0 CASING WELD A 351 CFBM 77144 o.o 0.0 0.0 0.0 e UNAGED AGED AGED FLOW FLOW FLOW STRESS STRESS STRESS @ 70F @ 70F @ 550F 0 78600 58000 0 78600 58000 0 78600 58000 0 76600 58000 64 77403 56809 56 64974 44380 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 Table 4-3 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL )> REACTOR COOLANT PUMP CASINGS 0 f1) REPORT #3 -PREDICTED THERMAL AGING BEHAVIOR ::I 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 6074 17.0 19.243 14.944 1.29 15.7 20.7 24.290 13.485 15.0 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 05202-1 12.0 17.766 15.074 1.18 13.2 12.3 23.600 13.430 0.0 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 )> "'t:J "CJ CD ::3 a. ...... x 0 N C> -Table 4-3 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS 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 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 l Table 4-4 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR ):a COOLANT PUMP CASINGS AT 70F "'C "'C REPORT #4 -PREDiCTED TOUGHNESS PROPERTIES OF AGED MATERIAL Ill :::s a.. ...... HEAT MEASURED MATERIAL x 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 8895.0 0.41 74.04 1485.0 250.4 119.0
• RCP PUMP CASING 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 ...... 6546 15.0 16.00 5924.0 0.40 44.17 978.9 . 203.3 83.3 9* -e

)::o 0 11) :::s 0. ..... x c N N e Table 4-4 (Continued) I HEAT . MEASURED NO. FERRITE CONTENT {%) 74726 13.4 7533A 10.0 64621 12.0 63683 10.5 77144 10.5

• RCP PUMP CASING 701-N-0560 07082-1 14.0 06068-1 10.0 6546 15.0 60525 9.0 74726 13.4 7553A 10.0 64621 12.0 63683 10.5 77144 10.5 e e ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS AT 70F REPORT #4 -PREDlCTED TOUGHNESS PROPERTIES OF AGED MATERIAL MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM AGING CONSTANT CONSTANT IMPACT Jic KJc T MOO. PARAMETER c N ENERGY @ 70F @ 70F. @ 70F @ 70F @ 70F {ft-lbs) (1 b/1 n) (ks1V1n) 15.27 6196.0 0.40 46.82 1029.0 208.5 86.8 14.36 6564.0 0.40 50.44 1100.0 215.6 91.4 10.74 8534.0 0.41 70.33 1415.0 244.5 114.8 10.58 8644.0 0.41 71.46 1436.0 246.3 116 .0 10.19 8923.0 0.41 74.32 1492.0 ' 251.0 119.3 ' I 28.74 3385.0 0.38 20.88 586.0 157.3 51.8 19.30 4941.0 0.39 34.79 860.0 . 190.6 .102.2 16.00 5924.0 0.40 44.17 978.9 203.3 83.3 15.62 6061.0 0.40 45.50 1005.0 206.0 85.1 15.27 6196.0 0.40 46.82 1029.0 208.5 86.8 14.36 6564.0 0.40 50.44 1100.0 215.6 91. 4 10.74 8534.0 0.41 70.33 1415.0 244.5 114.8 10.58 8644.0 0.41 71.46 1436.0 246.3 116.0 10.19 8923.0 0.41 74.32 1492.0 : 251.0 119.3 -. ; -.

Tal;>le 4-5 ANALYSIS OF THERMAL AGING OF STAINLESS STEEL )::> REACTOR PUMP CASINGS AT 550F -c -c ct> ::::s REPORT #5 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL 0.' ...... 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 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 6546-. 15.0 16.00 4273.0 ,,0.29. 44.17 1219.0 184.9 100.7 e e Table 4-5 (Continued)

ANALYSIS OF THERMAL AGING OF STAINLESS STEEL . )> REACTOR COOLANT PUMP CASINGS AT 550F "O "O CD ::i REPORT #5 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL 0.. ...... 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 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 l\) Ul RCP PUMP CASING 701-N-0557 701-N-0558 701-N-0559 701-N-0560 Table -San Onofre 2 Limiting and Controlling Values of Jic and KJc at 70"F Heat # 6074 05202-1 06652-1 07082-1 Jic (lb/in) 506.0 665.9 745.3 586.0 KJc {ksi/ in) 146.2 167.7 177.4 157.3 RCP PUMP CASING 701-N-0557 701-N-0558 701-N-0559 701-N-0560 Table 4.7 -San Onofre 2 Limiting and Controlling Values of Jic and KJc at 550°F Heat # 6074 05202-1 06652-1 07082-1 Jic (lb/in) 727.4 889.l 965.4 804.8 KJc (ksi/in) 142.8 157.8 164.5 150.2 Table 4-8 ANALYIS OF THERMAL AGING OF CAST STAINLESS STEEL :x:o REACTOR COOLANT PUMP CASINGS "C "C 11> ::3 REPORT #1 -MATERIAL IDENTIFICATION

& CHEMICAL COMPOSITIONS 0.. -'* x I MATERIAL I CJ 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 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 CASING WELD 63683 0.03 1.03 0.54 0.020 0.005 18.95 10.76 2.40 0.04 0.02 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 0.021 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 )> REACTOR COOLANT PUMP CASINGS "O "O Cl> ::I REPORT #1 -MATERIAL IDENTIFICATION & CHEMICAL COMPOSITIONS

a. ..... >< 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 0.014 0.005 19.43 10.15 2.04 0.04 0.06 CASING WELD 8609 0.02 0.91 '

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> REPORT #2 -MATERIAL SPECIFICATION & TENSILE PROPERTIES

s

• 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 CASE SCROLL A 351 CF8M 08426-1 42.5 85.7 45.0 63.0 64 77245 56651 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 CASING WELD A 351 CF8M 7553A 0.0 0.0 0.0 0.0 0 78600 58000 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 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 HUB/DIFFUSER A 351 CF8M 08783-1 40.3 82.9 58.0 75.0 62 73286 52692 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 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 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 N
• RCP PUMP CASING 701-N-0563

'° HUB/DIFFUSER A 351 CF8M 10202-1 0.0 0.0 0.0 o.o 0 78600 58000 CASI.WELD 1T3454 S782516 7971C o.o w 0.0 0.0 0 78600 58000 CASE .. ROLL A 351 CF8M 11294-1 o.o 0.0 o.o 0 78600 58009 )> "C "C 11) ::s 0.. -'* x c w 0 e Table 4-9 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS REPORT-#2 -MATERIAL SPECIFICATION & TENSILE PROPERTIES MATERIAL MATERIAL MATERIAL YIELD TENSILE TOTAL RED. IN OR SPEC. TYPE STRENGTH STRENGTH ELONG. AREA PART HEAT NO. (ks1) (ks1) (%) (%) CASING WELD A 351 CF8M 8346 0.0 0.0 0.0 .0.0 CASING WELD 1T3454 S782516 7970 o.o 0.0 0.0 0.0 CASING WELD A 351 CF8M 6546 o.o 0.0 0.0 0.0 CASING WELD A 351 CF8M 74726 o.o 0.0 o.o o.o CASING WELD. A 351 CF8M 7553A. o.o 0.0 0.0 0.0 CASING WELD A 351 CF8M 8346A '0.0 0.0 0.0 0.0 CASING WELD A 351 CF8M 8610 o.o 0.0 0.0 0.0 CASING WELD A 351 CF8M 8609 o.o o.o 0.0 o.o CASING WELD A 5.9 ER-316 776060 0.0 0.0 0.0 o.o

• RCP PUMP CASING 701-N-0564 CASE SCROLL A 351 CF8M 07342-1 43.8 86.6 55.0 74.0 HUB/DIFFUSER A 351 CF8M 96755-1 40.1 85.0 60.0 77.0 CASING WELD A5.9-69 ER-308 74819 0.0 0.0 0.0 0.0 CASING WELD 113454 S782516 7971 0.0 0.0 o.o o.o CASING WELD A 351 CF8M 60525 o.o* o.o 0.0 0.0 CASING WELD A 351 CF8M 6546 0.0 0.0 o.o o.o CASING WELD A 351 CF8M 74726 o.o 0.0 o.o o.o CASING WELD A 5.4 E316-16 6B5C-16 0.0. o.o. o.o 0.0 . CASING WELD A 351 CF8M 63683 o.o 0.0 0.0 o.o CASING WELD A 351 CF8M 64621 o.o* 0.0 0.0 o.o CASING WELD A 351 CF8M 77144 o.o 0.0 o.o* o.o CASING WELD 1T3454 S-7825 7956 0.0 . o.o o.o 0.0 CASING WELD A 351 CF8M 8609 0.0 0.0 o.o 0.0 UNAGED AGED AGED FLOW FLOW FLOW STRESS STRESS STRESS @ 70F @ 70F @ 550F 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 786PO 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 65 78986 58392 63 74790 54196 ' 0 78600 58000 ' I I 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 58000 0 78600 *58000 0 78600 58000 0 78600 58000 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL ):> Table 4-10 REACTOR COOLANT PUMP CASINGS "Cl "Cl l'D REPORT #3 -PREDICTED THERMAL AGING BEHAVIOR :l ..... x HEAT MEASURED CHROMIUM NICKEL ere/Nie 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 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 11294--17.0 17.243 13.952 1.24 16.7 9 16.4 23.055 12 .115 15.0 )> "O "O tD ::3 a. -'* x c w N e e Table 4-10 (Continued) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS 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 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 8346A 11.0 17. 504 14.922 1.17 10.1 12.0 22.275 13.345 13 1.8 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

---

, Table 4-11 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR )> COOLANT PUMP CASINGS AT 70F "'C "'C ro ::I REPORT #4 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL a.. ..... 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 07903-1 14.0 20.35 4692.0 0.39 32.48 801.0 183.9 79.8 08426-1 16.0 20.10 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 08783-1 14.0 24.97 3855.0 0.38 24.93 677.5 169.2 65.0 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 w 10202-1 15.0 25.37 3798.0 0.38 24.43 662.5 . 167 .3 56.0 13.0 20.69 4618.0 0.39 31.80 779. 7 181. 5 66.7 e e e e e Table 4-11 (Continued)

ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR )> COOLANT PUMP CASINGS AT 70F "O "O m :::J REPORT #4 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL a. ...... 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) ( 1 b/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 COOLANT PUMP CASINGS AT 70F "O "O Cl> :::I REPORT #4 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL 0.. -'* >< 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

)> "O "O ft) :::s a. ..... x c w O'I e Table 4-11 HEAT NO. Cl 54 52 47 57 62 58 48 49 P3 50 (Continued) MEASURED FERRITE CONTENT (%) 2.2 1.8 13.5 16.3 4.0 4.5 2.9 8.7 7.2 1.9 4.4 ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR COOLANT PUMP CASINGS AT 70F REPORT #4 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL MATERIAL MATERIAL MATERIAL MINIMUM MINIMUM MINIMUM MINIMUM AGING CONSTANT CONSTANT IMPACT Jlc KJc T MOD. PARAMETER c N ENERGY @ 70F *@ 70F @ 70F @ 70F @ 70F (ft-lbs). (lb/in) (ks1V1n) 9.12 9777 .0 0.42 83.16 1578.0 258.2 128.5 7.97 10859.0 0.43 94.42 1700.0 268.0 140.0 6.37 12733.0 0.44 114.00 1959.0 287.6 159.4 6.67 12349.0 0.43 110.00 1988.0 289.8 156.1 6.61 12421.0 0.43 110.30 2003.0 290.8 156.9 6.48 12587.0 0.43 112.50 2035.0 293.2 158.6 6.45 12634.0 0.43 113.00 2046.0 294.0 159.1 5.23 14408.0 0.44 131.60 2287.0 310.8 176.7 4.20 16253.0 0.45 150.80 2532.0 327.0 194.2 4.01 16639.0 0.45 154.80 2610.0 332.0 197.9 3.68 17341.0 0.46 162.10 2610.0 332.0' 203.8 Table 4-12 ANALYSIS OF THERMAL AGING OF STAINLESS STEEL )> REACTOR COOLANT PUMP CASINGS AT 550F "'C "'C ti> ::I REPORT #5 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL a.. -'* 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/1n) (ks1/1n) ** PLANT 1.0. SOUTHERN CAL. UNIT 3
• RCP PUMP CASING 701-N-0561 07903-1 14.0 20.35 3590.0 0.28 32.48 1068.0 173.0 107.1 08426-1 16.0 20.10 3.623.0 0.29 33.01 1015.0 i 168 .. 7 91.3 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 63683 10.5 10.89 5556.0 0.31 69.29 1492.0 204.5 126.B 64621 12.0 10.74 5605.0 0.31 70.33 1509.0 205.6 127.6 77144 10.5 10.19 5792.0 0.31 74.32 1567.0 209.6 131. l 8610 0.0 9.63 5998.0 0.31 78.80 1632.0 213.8 134. 9 8609 12.0 9.60 6011.0 0.31 79.07 1637.0 214.2 135.2 776060 0.0 4.76 8551.0 0.33 140.00 2259.0 251.6 180.4
• RCP PUMP CASING 701-N-0562 08783-1 14.0 24.97 3098.0 0.28 24.93 903.1 159.l 90.8 11994-1 17 .0 17 .57 3994.0 0.29 39.24 1121.0 177 .2 83.9 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 8610 0.0 9.6, 5998.0 0.31 78.80 1632.0 213.8 134.9 8609 12.0 9.6 6011.0 0.31 79.07 1637.0 ,.214.2 135.2 776060 0.0 4.76 8551.0 0.33 140.00 2259.0 : 251.6 180.4 w
• RCP PUMP CASING 701-N-0563

........ 10202-1 15.0 25.37 3064.0 0.28 24.43 883.9 . 157 .4 75.1 7971 *. 13.0 20.69 3547.0 0.28 31.80 1039.0 170.6 85.3 e e

Table 4-12 (Continued) ANALYSIS OF THERMAL AGING OF STAINLESS STEEL ):o REACTOR COOLANT PUMP CASINGS AT 550F "'C "'C CD ::s REPORT #5 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL I Q. ...... 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 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 )> REACTOR COOLANT PUMP CASINGS AT 550F "'C ' "'C tt> REPORT #5 -PREDICTED TOUGHNESS PROPERTIES OF AGED MATERIAL ::I 0. ..... 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 @ 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 RCP PUMP CASING 701-N-0561 701-N-0562 701-N-0563 701-N-0564 Table

-San Onofre 3 Limiting and Controlling Values of Jic and KJc at 70°F Heat # 07903-1 08783-1 10202-1 07342-1 JIC (lb/in) 801.0 677.5 662.5 625.3 KJc (ksi/in) 183.9 169.2 167.3 162.5 RCP PUMP CASING 701-N-0561 701-N-0562 701-N-0563 701-N-0564 Table -San Onofre 3 Limiting and Controlling Values of Jic and KJc at 550°F Heat # 07903-1 08783-1 10202-1 07342-1 Jic (lb/in) 1068.0 903.l 883.9 845.8 KJc (ksi/in) 173.0 159.1 157.4 154.0 _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=

1. 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=

1. 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=

1. 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=

1. 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=

1. 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). Thickness Includes 11.29 Ksi thermal stress derived from Basic Condition 8, Heatup minus Cooldown. = 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 Condition Appendix D = # 2978 = # 401 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 I ----------==-=----- -::=:-* --===---------==--=--=------

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___ -1 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. Bending stress A-56, A-57) = 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 Appendix D (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. so )ii l1j l1j CD ::s p, ..... x 0 U1 ..... a/t Interval (Fraction) 0.08 --0.10 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 (a) DT = 38.1 1% £Steps (b) DT = 20.2 1% f Steps KI (KSI !TN) 44.91 54.13 62.12 69.52 76.44 83.46 90.66 99.06 108.5 Table 5-1 San Onofre 2&3 RCP Casing crack Growth Rate At Suction Nozzle da/dl DT (IN/YEAR) (YEARS) 3.06 x 10-3 23.6 6.47 x 10-3 28.0 (a) 1.12 x 10-2 16.1 (b) 1. 76 x 10-2 10.3 2.57 x 10-2 7.0 3.66 x 10-2 4.9 5.09 x 10-2 3.5 7.26 x 10-2 2.5 0.104 1. 7 "CS (I) ::s p, ..... >< 0 U1 l\J '!'able 5-2 San Onofre 2 & 3 RCP Casing crack Growth Rate 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") tO (D ::s ..... >< c U1 LJ a/t Interval KI (Fraction) (KSI ITR) 0.08 --0.10 55.01 0.10 --0.15 64.54 0.15 --0.20 72.05 0.20 --0.25 78.56 0.25 --0.30 83.99 0.30 --0.35 89.41 0.35 --0.40 94.43 0.40 --0.45 100.7 0.45 --0.50 107.7 (rm .. 10.9, rb = 35.95, t = 4.75") a . (a) 1% I Steps RT 1 = 23.6 (a) (b) 1% I Steps RT 1 a 13.9 (b) '.fable 5-3 San Onofre 2 & 3 RCP Casing Crack Growth Rate At .crotch Region da/dT . OT (IN/YEAR) (YEARS) 6.90 x 10-3 13.7 1.31 x 10-2 18.2 (a) 2.03 x 10-2 11. 7 (b) 2.87 x 10-2 8.2 3.75 x 10-2 6.3 4.81 x 10-2 4.9 5.99 x 10-2 3.9 7.76 x 10-2 3.0 0.101 2.3 a/t Interval (fraction) 0.08 --0.10 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 KI (KSI ITIO 48. 74 58.66 67.21 75.10 82.45 89.89 97.50 106.4 116.4 ( 22 8 20 5 t = 4.0") rm = . ' rb = * ' (a) 1% £ Steps (b) 1% £ Steps RT 1 = 30.4 RT 1 = 16.2 Table 5-4 San Onofre 2 & 3 RCP Casing Crack Growth Rate At Diffuser Vane da/dT (IN/YEAR) 4.25 x 10-3 8.92 x 10-3 1.54 x 10-2 2.40 x 10-2 3.48 x 10-2 4.92 x 10-2 6.81 x 10-2 9.66 x 10-2 0.138 OT (YEARS) 18.8 22.4 (a) 13.0 (b) 8.3 5.7 4.0 2.9 2.1 1.4 Ul Ul KI a/t Interval (Fraction) {KSI !TN) 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 43.21 52.57 60.88 68.70 76.19 83.81 91. 77 100.9 111.3 (rm = 27.35, rb = 13.9, t = 3.33") a (a) 1% I Steps, RT 1 z 40.0 a (b) 1% {Steps, RT 1 = 20.4 Table 5-5 San Onofre 2 & 3 RCP Casing Crack Growth Rate At Volute (Vane Region 5) da/dT (IN/YEAR) 2.63 x 10-3 5.75 x 10-3 1.03 x 10-2 1.68 x 10-2 2.54 x '10-2 3.72 x 10-2 5.34 x 10-2 7.82 x 10-2 0.116 DT {YEARS} 25.3 28.9 (a) 16.1 (b} 9.9 6.5 4.4 3 .1 2 .1 1..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 1. 0 2.0 3.0 4.0 3.1 3.2 3.3 PURPOSE PRE-SERVICE INSPECTION DATA EVALUATION OPERATING HISTORY Design Specifications Stress cycles Used In Evaluation Stress Cycles at st. Lucie 1&2 To-date THERMAL EMBRITTLEMENT 4.1 Material Identification and Chemical Properties 4.2 Material Specifications and Mechanical Properties 4.3 Thermal Aging Behavior 4.4 Toughness Properties of Aged Materials 4.5 Limiting Values 5.0 CRACK GROWTH ANALYSIS 5.1 Scope 5.2 Reference Stress Reports 5.3 Selection of High Stress Locations 5.4 Stresses and Wall Thicknesses at Limiting Locations 5.5 Calculation of Crack Growth Rates 5.6 Stresses Under Emergency and Faulted Conditions 5.7 Results 6.0 INSPECTION INTERVAL APPENDIX E REFERENCES Appendix E 3 Page 6 7 9 9 10 10 12 12 12 13 13 14 39 39 39 39 40 42 48 51 52 53 TABLE 4-1 TABLE 4-2 TABLE 4-3 TABLE 4-4 TABLE 4-5 TABLE 4-6 TABLE 4-7 TABLE 4-8 TABLE 4-9 TABLE 4-10 TABLE 4-11 Appendix E LIST OF TABLES Material Identification and Chemical Compositions -Unit 1 Material Specification and Tensile Properties -Unit 1 Predicted Thermal Aging Behavior -Unit 1 Predicted Toughness Properties of Aged Material (70°F) -Unit 1 Predicted Toughness Properties of Aged Material (550°F) -Unit 1 Limiting and Controlling Values of Jic and KJc at 70°F -Unit 1 Limiting and Controlling Values of Jic and KJc at 550°F -Unit 1 Material Identification and Chemical Compositions -Unit 2 Material Specification and Tensile Properties -Unit 2 Predicted Thermal Aging Behavior -Unit 2 Predicted Toughness Properties of Aged Material (70°F) -Unit 2 15 17 19 21 23 25 26 27 29 31 33 4 TABLE 4-12 TABLE 4-13 TABLE 4-14 TABLE 5-1 TABLE 5-2 TABLE 5-3 TABLE 5-4 TABLE 5-5 Appendix E LIST OF TABLES (Continued) Predicted Toughness Properties of Aged Materials (550°F) -Unit 2 Limiting and Controlling Values of Jic and KJc at 70°F -Unit 2 Limiting and Controlling Values of Jic and KJc at 550°F -Unit 2 Crack Growth Rates at Vane Number 8 -Level D Crack Growth Rates at Discharge Nozzle crotch Vicinity -Section c Crack Growth Rates at suction Nozzle-Level c Crack Growth Rates at Volute Junction With Lower Flange Near Vanes 1 & 2 Crack Growth Rates at Hanger Bracket #1 Vicinity 35 37 38 43 44 45 46 47 5 1.0 PURPOSE 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

• i 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 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: Appendix E Transient Condition Heat-Up (100°F/hr) Cool-Down (100°F/hr) Hydrostatic Test (3125 psia 100-400°F) Leak Test (2250 psia 100-400°F) Assumed Occurrences During 40 Year License Period 500 500 10 200 Loss of Secondary Pressure 5 Reactor Trip or Loss of Load 400 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: Heatup/Cooldown -Reactor Trip -Unit 1 <50 222 Unit 2 <25 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 -----------------

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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 )> Table 4-1 "'O "'O C1> :::I ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL a. ...... REACTOR COOLANT PUMP CASINGS )( l'T1 REPORT ll -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 Table 4-1 (Continued) "C "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 CASING WELD 04S09 0.02 0.94 0.4S 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 Ol CASING WELD 04313 0.02 0.91 O.S2 0.016 O.OlS 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

)> TABLE 4-2 "O "O n> ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL ::::s a.. REACTOR COOLANT PUMP CASINGS -'* x ,.,., 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>

4-2 (Continued) "'C "'C CD ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL ::J 0.. 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 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 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 ...... CASING WELD A 371 ER-316 04313 0.0 0.0 0.0 0.0 0 78600 58000 CD 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

)> Table 4-3 "'C "'C (1) :::::s ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL 0.. ...... 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

)> 4-3 (Continued) "C "C ID :J ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL a. ..... 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 MET AL (%) 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 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 Table 4-4 "'C "'C CD ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL :::s 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 X43439 14.0 19.87 4804.0 0.39 33.51 815.6 185.6 69.2 N 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

)> 4-4 (Continued) "C "C ,,, ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR :;:, Cl. 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 03036 15.0 16.41 5782.0 0.39 42.79 1003.0 205.8 82.0 N 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 Table 4-5 "'C "'C 11) ANALYSIS OF THERMAL AGING OF STAINLESS STEEL ::J 0.. ...... 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 04460 20.0 18.40 3864.0 0.29 37.00 1088.0 174.6 92.4 w 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 . "

)> 4-5 (Continued) "C "C tD ANALYSIS OF THERMAL AGING OF STAINLESS STEEL ::I c.. 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 "'C n> ::I 0. ...... >< l'T1 N U1 RCP PUMP CASING 681-N-0445 681-N-0446 681-N-0447 681-N-0448 Table 4-6 -St. Lucie Unit 1 Limiting and Controlling Values of J 10 and KJc at 70°F HEAT # 04146 46993 04146 46406 Jic (lb/in) 594.5 669.2 594.5 766.6 KJc (ksi/1n) 158.5 168.1 158.5 179.9 > "C "C Cl) ::::s 0. ..... )( fT1 N m RCP PUMP CASING 681-N-0445 681-N-0446 681-N-0447 681-N-0448 Table 4-7 St. Lucie Unit 1 Limiting and Controlling Values of J 10 and KJc at 550°F HEAT # 04146 46993 04146 46406 Jic (lb/in) 813.4 890.5 813.4 1028.0 KJc (ksi/Iil) 151.0 158.0 151.0 169.7 )> Table 4-8 "'C "'C ct> ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL ::I a. ' ..... 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 CASING WELD 7174 0.03 1.26 0.62 0.018 0.014 19.65 10.20 2.65 0.04 0.00 N 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

)> 4-8 (Continued) 0 CD :::I ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL a. ...... 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 !

)> Table 4-9 0 11> ::s ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL a. ...... 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. (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 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 Cf SM 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 )>

4-9 (Continued) "'C "O (I) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL :::s 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 Cf SM 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 CASING WELD A 351 CF8M 7242 0.0 w 0.0 0.0 0.0 0 78600 58000 0

)> Table 4-10 "'C "'C CD ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL 0. 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 95211-1 14.0 17.074 14.467 1.18 12.8 12.4 22.735 12. 720 11.0 w 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) ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL ::::s a.. 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 MET AL (%) 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

)> Table 4-11 "'C "'C CD :l ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR a.. ..... 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 7174 17 .o 18.74 5084.0 0.39 36.13 867.9 191. 5 72.9 w 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

)> Table .4-11 (Continued) "'C "'C 11> ANALYSIS OF THERMAL AGING OF CAST STAINLESS STEEL REACTOR ::3 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 ___ J

)> "C Table 4-12 "C rt> ANALYSIS OF THERMAL AGING OF STAINLESS STEEL ::3 Q. .... REACTOR COOLANT PUMP CASINGS AT 550F x 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 7174 17.0 18.74 3812.0 0.29 36.13 1072 .0 173 .3 91.3 Ul 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 (Continued) "'O "'O CD ANALYSIS OF THERMAL AGING OF STAINLESS STEEL :::s 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 "t:I n> ::3 c. ..... )( ,.,., w ...... RCP PUMP CASING 741-N-0001 741-N-0002 741-N-0003 741-N-0004 Table 4-13 -st. Lucie Unit 2 Limiting and Controlling Values of Jic and KJc at 70°F HEAT # Jic KJc (lb/in) (ksi/in) 91097-1 504.0 145.9 6074 506.0 146.2 6074 506.0 146.2 6074 506.0 146.2 )> "C "C rD ::I 0.. -'* )( w CD RCP PUMP CASING 701-N-0001 701-N-0002 701-N-0003 701-N-0004 Table 4-14.-St. Lucie Unit 2 Limiting and Controlling Values of Jlc and KJc at 550°F HEAT # 91097-1 6074 6074 6074 Jic (lb/in) 721.1 727.4 727.4 727.4 KJc (ksi/in) 142.1 142.8 142.8 142.8 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, 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 temperature difference of 18°F during heatup/cooldown. (2) For Discharge Nozzle--Section C, adjacent to Crotch Region: Design Condition Key Element = Membrane stress Bending stress = = # 107, plus thermal gradient stresses # 5125 in Finite-Element Model = 17.4 Ksi (Table 4-6) 21.1 Ksi (Figure 4-8, p. 66 & Thickness = Figure 3-1) 3.3 in. (Table 4-6) Appendix E Bending stress includes 5.5 Ksi due to a through-wall temperature difference of 29°F during heatup/cooldown. 40 ----------------- --*- (3) For Suction Nozzle--Level C: Design Condition = # 104, plus thermal gradient stresses Key Elements = # 2125 & 2129 in Finite-Element Model Membrane stress Bending stress = Thickness = = 22.15 Ksi (Figure 4-5(b)) 13.15 Ksi (Figures 3-1 & 4-5(b)) 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 = Key Elements = Membrane stress = Appendix E Maximum (pp. 74, 75) # 7461 in Finite-Element Model 18.9 Ksi (page 75) 41 Bending stress Thickness = 22.1 Ksi (page 75, 97) = 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: Appendix E (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: ( 3 ) da/dT = 5.435 x 10-lO K 4.0 I Units for da/dT are inches/year. Incremental time, dT, in which the crack will grow through the indicated dimensionless crack depth 42 a/t Interval (fraction} 0.08 --0.10 0 .10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 O.&i5 --0.50 Table 5-1 St. Lucie 1 & 2 Crack Growth Rates At KI Vane Number 8 Level D (KSI mo 49.98 60.08 68.76 76.76 84.17 91.69 99.35 108.3 118.4 da/dT (IN/YEAR} 3.39 x 10-3 7.08 x 10-3 1.21 x 10-2 1.89 x 10-2 2.73 x 10-2 3.84 x 10-2 5.29 x 10-2 7.48 x 10-2 0.107 {am

• 20.95, ab* 19.87, t
• 4.75") t>.Time (YEARS} 28.0 45.6* 24.4* 12.5 8.7 6.2 4.5 3.2 2.2
• 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 (fraction} (KSI /IN) (IN/YEAR}

0.08 --0.10 38.9 1.25 x 10-3 0.10 --0.15 46.6 2.56 x 10-3 0.15 --0.20 53.0 4.30 x 10-3 0.20 --0.25 58.9 6.55 x 10-3 0.25 --0.30 64.3 9.28 x 10-3 0.30 --0.35 69.7 1.28 x 10-2 0.35 --0.40 75.2 1. 74 x 10-2 0.40 --0.45 81.6 2.41 x 10-2 0.45 --0.50 88.9 3.39 x 10-2 .. (am* 17.4, ab* 21.1, t

• 3.3") Appendix E (YEARS} 52.8 64.5 38.4 25.2 17.7 12.8 9.5 e 6.8 4.8 44 a/t Interval {fraction}

0.08 --0.10 0 .10 --0 .15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Appendix E Table 5-3 St. Lucie 1 & 2 Crack Growth Rates At Suction Nozzle -Level C KI da/dT {KSI Im} {IN/YEAR} 34.92 8.08 x 10-4 42.37 1. 75 x 10-3 48.93 3.12 x 10-3 55.08 5.00 x 10-3 60.93 7.49 x 10-3 66.88 1.09 x 10-2 73.05 1.55 x 10-2 80.20 2.25 x 10-2 88.25 3.30 x 10-2 t.T {YEARS} 74.2 85.6 48. l 30.0 20.0 13.8 9.7 6.7 4.5 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 {aLt

• l {KSI mo {INLYEAR}

0.08 --0.10 32.27 5.9 x 10-4 0 .10 --0 .15 38.99 1.26 x 10-3 0.15 --0.20 44.85 2.2 x 10-3 0.20 --0.25 50.29 3.48 x 10-3 0.25 --0.30 55.41 5.12 x 10-3 0.30 --0.35 60.61 7.33 x 10-3 0.35 --0.40 65.97 1.03 x 10-2 0.40 --0.45 72.20 1.48 x 10-2 0.45 --0.50 79.21 2.14 x 10-2 (um* 17.8, ub

• 13.2, t
• 3.375) Appendix E {YEARS} 114. 134. 77. 48.5 33. 23. 16.4 e 11.4 7.9 46 From -To (a Lt ,.. l 0.08 --0.10 0.10 --0.15 0.15 --0.20 0.20 --0.25 0.25 --0.30 0.30 --0.35 0.35 --0.40 0.40 --0.45 0.45 --0.50 Table 5-5 St. Lucie 1 & 2 Crack Growth Rates At Hanger Bracket #1 Vicinity KI da/dT (KSI mn (INLYEAR) 43.40 1.92 x 10-3 51.90 3.94 x 10-3 59.10 6.65 x 10-3 65.80 0.0102 71.80 0.0145 77.90 0.02 84.10 0.0272 91.40 0.0379 99.50 0.0533 (um
• 18.9, ub m 22.1, t
• 3.6") l1T (YEARS}" 37.4 61.3* 33.5* 17.6 12.4 9.0 6.6 4.7 3.4
• 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, 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. Appendix E 5.6.1 Emergency Condition Stresses (1) Diffuser Vane 8--Level D: Key elements Condition = # 3828 & 3832 = # 505 48

• Appendix E 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 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. 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 under the influence of the conservatively defined stress cycles in design specification. The hypothesized crack will then grow larger until it 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, 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 Field Name 29 MIN T25 30 MIN JIC290 31 MIN KJC290 32 MIN T290 33 MAT PAR 34 c 25 35 c 290 36 N 25 37 N 290 38 FERR CALC2 39 CRM_EQUIV 40 NIC_EQUIV 41 CRNI_EQUIV 42 AFS 25 43 AFS 290 44 UAFS_25 Appendix F DATABASE FILE STRUCTURE Content in Database Minimum Room Temperature Average Tearing Modules Minimum 550°F JIC Minimum 550°F JJC Minimum 550°F Average Tearing Modules Material Parameter from Chopra J-R Curve coefficient .at Room Temperature J-R Curve Coefficient at 550°F J-R Curve Exponent at Room Temperature J-R Curve Exponent at 550°F Calculated Ferrite Content from Hull's Factors Chromium Equivalent from Hull's Nickel Equivalent from Hull's Factors Rate of Chromium + Nickel Equivalents from Hull's Factors Aged Flow Stress at Room Temperature Aged F_low Stress at 550°F Unaged Flow Stress at Room Temperature LISTING F-1 (Continued) 5 "'* 1 2 3 4 5 Field Name MATERIAL COMP ID PLANT MAT SPEC MAT TYPE 6 HEAT NO 7 c 8 MN 9 SI 10 s 11 p 12 CR 13 NI 14 MO 15 CB 16 N 17 FERR_MEA$ 18 FERR_CALCl 19 FERR SPAC 20 FERR_WELD 21 YIELDSTR 22 TENSILSTR 23 ELONGATION 24 REDINAREA 25 HARDHRB 26 MIN_CVNE 27 MIN_JIC25 28 MIN_KJC25 Appendix F DATABASE FILE STRUCTURE Content in Database Casing Material Case/Scroll, Hub or Weld Pump Casing Serial Number RCP Plant or Utility Location Material Specification Material Type Heat Number of Material Carbon Content Manganese.content Silicon Content* Sulfur Content Phosphorus Content Chromium Content Nickel Content Molybdenum Content Columbium/Niobium Content Nitrogen Content Measured Ferrite Content Calculated Ferrite Content from Schaefer Equation Ferrite Spacing Calculated Ferrite Content of Weld Metal from Schoef ler Diagram Yield Strength Tensile Strength Elongation Radiation in Area Hardness Rockwell B Minimum Charpy Impact Energy Minimum Room Temperature JIC Minimum Room Temperature JJC LISTING F-1 4

• -CLEAR SET STATUS OFF SET TALK OFF USE USE CAST SS GOTO TOP-COMPUTER dBASE PROGRAM CASEINFO.PRG 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
• UNAGFS25 * (YIELDSTR*+

TENSILSTR)/2 REPLACE CRM EQUIV WITH CRMEQUIV REPLACE NIC-EQUIV WITH 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 -END DO USE CLOSE DATABASES CLOSE ALL SET STATUS ON Appendix F LISTING F-2 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
• 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 @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 END DO DO TEARMOD USE CLOSE CLOSE ALL END Appendix F LISTING F-3 (Continued) 8 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
• 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 END DO 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 ENO DO 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

• I * ) SECTION 5 RELIEF REQUESTS 33
• CONSUMERS POWER COMPANY PALISADES NUCLEAR POWER PLANT THIRD; INTERVAL INSERVICE INSPECTION PROGRAM RELi EF REQUESTS RELIEF SECTION REQUEST XI ITEM COMPONENT NUMBER CATEGORY . NUMBER DESCRIPT.

RR-1 B-J B9.ll PCS B9.12 PIPING RR-2 B-A Bl.40 RV HEAD-TO FLANGE WELD RR-3 C-A Cl.10 SG SHELL-TO-CONE WELDS *--RR-4 B-D 83 .130 *sG NOZZLE-TO-SHELL WELDS RR-5 s.:.D B3 .150 REGEN. HX WELDS RR-6 8-8 82.11 PZ. HEAD, 8-D 82.21 SHELL*& 82.22 NOZZLE 83 .110 WELDS . RR-7 C-A Cl.10 SOC HX C-8 Cl.30 SHELL, C2.21 FLG. & TS WELDS RR-8 B-A Bl. 21 RV CLOSURE HEAD WELD 34 RELIEF AL TERN. REQUESTED EXAM INACCESS. 100% UT FOR O.D. FROM I.D. SURFACE .& EXAM VOL. EXAM INACCESS. ACCESS FOR 100% VOLUNE &

• VOL. EXAM SURFACE EXAM INACCESS.

ACCESS FOR 100% VOLUME

• VOLUME EXAMINE EXAM INACCESS.

ACCESS. FOR 100% VOLUME VOLUME EXAMINE . EXAM INACCESS. ACCESS. FOR 100% VOLUME VOLUME EXAMINE EXAM INACCESS. ACCESS. FOR 100% VOLUME VOLUME EXAMINE EXAM WITH ALL MERID. WELDS . INACCESS. ACCESS. FOR 100% VOLUME AND VOLUME ALL EXAM: SURFACE EXAMINE INACCESS. ACCESS FOR 100% VOLUME VOLUME EXAMINE EXAM

• 35 RELIEF SECTION ! COMPONENT REQUEST XI ITEM RELIEF AL TERN. NUMBER CATEGORY NUMBER DESCRIPT.

REQUESTED EXAM 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 RELIEF REQUEST NUMBER -RR-1 COMPONENT IDENTIFICATION Code Class Code Reference Examination Category Item Number Component Description . 1 IWB-2500 Table IWB-2500-1 B-J 89 .11, 89 .12 . 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 PCS-30-RCL-2A-15LU-l, 15LU-2, 15 PCS-30-RCL-2B-15LU-l, 15LU-2, 15 CODE REQUIREMENT Table IWB-2500-1 requires a surface and volumetric examination of the circumferehtial and longitudinal welds. 36 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. I. PROPOSED ALTERNATE EXAMINATION 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 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

• 37 implemented for thts .relief request. In addition to performing 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 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 -

COMPONENT IDENTIFICATION Code Class Code Reference Examination Category Item Number Component Description CODE REQUIREMENT 1 IWB-2500 Table IWB-2500-1 B-A Bl.40 Reactor Vessel Head to Flange Weld, 6-118A 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 Code Class Code Reference Examination Category .Item Number Component Description CODE REQUIREMENT 1 IWB-2500 -Table IWC-2500-1 C-A Cl.10 Generator Upper Shell to Shell Cone Welds, 1-101-221, 2-101-221 Table IWB-2500-1 requires a volumetric examination of welds at.gross discontinuities also.includes essentially 100% of the weld length. BASIS FOR RELIEF 39 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 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 Code Class Code Reference Examination Category Item Number Component Description 1 IW8-2500 Table IWB-2500-1 B-D B3.130 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 . 40 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. .:*:

• ) RELIEF REQ\IEST NUMBER -RR-5 COMPONENT IDENTIFICATION Code Class Code Reference Examination Category Item Number Component Description CODE REQUIREMENT 1 IWB-2500 Table IWB-2500-1 B-D 83 .150 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 Table IWB-2500-1 requires all nozzle to vessel welds to be volumetrically

-examined during' each inspection interval. 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 . 41

• ** .. J RELiEf REQUEST NUMBER -RR-6 COMPONENT IDENTIFICATION Code Class Code Reference Examination Category Item Number Component Description CODE REQUIREMENT 1 IWB-2500 Table IWB-2500-1 B-B B-D B2 .11, B2. 21, B2. 22 B3 .110 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 42 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 . PROPOSED ALTERNATE EXAMINATION The accessible volume of the welds, including meridional welds, will be examined in lieu of the 100% volumetric examination requirements of the welds.

• * ) RELIEF REQUEST NUMBER -

COMPONENT IDENTIFICATION Code Class Code Reference Examination Category Item Number Component Description I IWC-2500 Table IWC-2500-1 C-A C-B Cl.10, Cl.30 C2.21 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 43 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 Code Class Code Reference Examination Category Item Number Component Description CODE REQUIREMENT 1 IWB-2500 Table IWB-2500-1 B-A Bl. 21 Reactor Vessel Closure Head Circumferential Weld 6-1188 Table IWB-2500-1 requires the accessible length of all welds to be volumetrically examined each inspection interval.

BASIS FOR RELIEF 44 .) 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 Code Reference Examination Category Item Number Component Description CODE REQUIREMENT 1 IWB-2500 Table IWB-2500-1 B-0 83.90, 83.100 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 45 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 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. 46 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 .

• RELIEF.

-RR-10 COMPONENT IDENTIFICATION Code Class Code Reference Examination Category I tern N.umber Component Description CODE REQUIREMENT 1 and 2 IWB-2500, IWC-2500 Table IWB-2500-1, IWC-2500-1 8-J, C-F-1, C-F-2 89.12, C-5.12, C5.52 Pressure Retaining Piping NPS 4 and larger*longitudinal welds 47 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.*

• SECTION 6 VERIFICATION OF SECtION XI COMPLIANCE 49
  • • **
• j ' / 50 6. VERIFICATION OF SECTION XI COMPLIANCE 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 2 . The minimum and maximum number of components to be 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. Second; Period lhe minimum number 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 .

• D. INTERVAL COMPLIANCE

,,,.,_,,_.,,., .. Third period cumulative percentage totals which equal or exceed 100% verifies 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. 51 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 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.

• VERIFICATION OF COMPLIANCE THIRD INTERVAL_, .. BY CATEGORY ATTACHMENT:

-Database Population Summary -Category and Item Number Designation, Legend -Category Summary, Page(s) 1-2 52 PALISADES NUCLEAR PLANT DATABASE POPULATION

SUMMARY

* (includes all records, regardless of schedule)

Class 1 components B-A components = B-B components = B-D components = B-E components = B-F components = B-G-1 components = B-G-2 components = B-H components = B-J components = B-K-1 components = B-L-1 components = B-L-2 components = B-M-1 componen11s- =

• B-M-2 components B-N-1 components I B-N-2 components B-N-3 components B-0 components B-Q components
• ) Report: file _pop File: dcprogma Project: file_pop = = = = =

= 28 36 42 0 47 175 152 6 746 20 4 4 2 14 1 2 1 233 0 Class 2* Components Class 3 components C-A components = 18 D-A components = 0 C-B components = 22 D-B components = 56 c-c components =* 63 D-C components = 0 C-F-1 components = 414 C-F-2 components = 174 component Supports Non-sec. XI Components F-A components = 892 N/A components = 929 Inservice Inspection -Database Totals Total Class 1 Components =1513 Total Class 2 Components = 691 Total Class 3 Components = 56 Total Component Supports = 892 Non-Section XI Components = 929 Database Total: 4081 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 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 or C-f-2*-NA as applicable.

3. If excluded by Table 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 Item No. 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 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-C, Integral Attachments where the line is not exempt based on line thickness per note (4), but is 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 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 CATEGORY ACTIVE REC SCHEDULED AUG 44 B-A 28 B-B 36. B-D 42 B-F 47 B-G-1 175 B-G-2 152 B-H 6 B-J 746 B-K-1 20 B-L-1 4 B-L-2 4 B-M-1 2 B-M-2 14 B-N-1 B-N-2 2 B-N-3 Report: i3catpop File: sch_cat Index: key/Exam Category 44 27 22 38 47 171 152 2 217 0 0 0 2 PERCENT TOTAL POP 100.0% 96.4% 61.1% 90.5% 100.0% 97. T'!. 100.0% 33.3% 29.1% 0.0% 25.0% 0.0% 50.0% 0.0% 100.0% 100.0% 100.0% PALISADES.GRAM PLAN THIRD INTERVAL ATEGORY

SUMMARY

*PERIOD 1 PERIOD 1 PERIOD 1 SCHEDULED COMPLETE PERCENT 0 0 0% 0 0 0% 7 0 32% 8 0 21% 13 0 28% 56 0 33% 41 0 27% 0 50% 38 0 18% 0 0 0% 0 0 0% 0 0 0% 0 0 0% 0 0 0% 0 100% 0 0 0% 0 0 0% PERIOD 2 PERIOD 2 PERIOD 1+2 SCHEDULED COMPLETE PERCENT 0 0 0% 0 0 0% 6 0 59% 12 0 53% 9 0 47'/. 56 0 65% 40 0 53% 0 0 50% 82 0 55% 0 0 0% 0 100% 0 0 0% 0 0 0% 0 0 0% 0 200% 0 0 0% 0 0 0% PERIOD 3 PERIOD 3 PERIOD 1-3 SCHEDULED COMPLETE PERCENT 0 0 0% 27 0 100% 9 0 100% 18 0 100% 25 0 100% 59 0 100% 71 0 100% 0 100% 97 0 100% 0 0 0% 0 0 100% 0 0 0% 0 100% 0 0 0% 0 300% 2 0 100% 0 100%

• 02/28/95 TOTAL PERCENT COMPLETE COMPLETE 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% Note: All totals .based on Active records. Period totals must also be Code Credit "Y". Page nunber 1
• \ ., *-EXAM TOTAL .INTERVAL 3 CATEGORY ACTIVE REC SCHEDULED B-0 233 C-A 18 C-B 22 c-c 63 c-c# 11 C-C* 56 C-F-1 414 C-F-1# 4 C-F-1* 1304 C-F-2 174 C-F-2# 24 C-F-2* 138 D-B 56 F-A 892 N/A 578 TOTALS: 5311 Report: i3catpop Fi le: sch_ cat Index: key/Exam Category 6 9 16 57 0 0 215 0 0 86 0 0 56 229 0 1400 PERCENT TOTAL POP 2.6% 50.0% 72. 7"-' 90.5% 0.0% 0.0% 51.9% 0.0% 0.0% 49.4% 0.0% 0.0% 100.0% 25. 7"-' 0.0% PALISADES.OGRAM PLAN THIRD INTERVA[*-*

CATEGORY

SUMMARY

PERIOD 1 PERIOD 1 PERIOD 1 SCHEDULED COMPLETE PERCENT 0 0 0% 2 0 22% 4 0 25% 11 0 19% 0 0 0% 0 0 0% 68 0 32% 0 0 0% 0 0 0% 16 0 19% 0 0 0% 0 0 0% 13 0 23% 67 0 29% 0 0 0% 346 0 PERIOD 2 PERIOD 2 PERIOD 1+2 SCHEDULED COMPLETE PERCENT 0 0 0% 4 0 6 7"-' 6 0 63% 25 0 63% 0 0 0% 0 0 0% 69 0 64% 0 0 0% 0 0 0% 27 0 50% 0 0 0% 0 0 0% 21 0 61% 74 0 62% 0 0 0% 433 0 PERIOD 3 PERIOD 3 PERIOD 1-3 SCHEDULED COMPLETE PERCENT 6 0 100% 3 0 100% 6 0 100% 21 0 100% 0 0 0% 0 0 0% 77 0 100% 0 0 0% 0 0 0% 43 0 100% 0 0 0% 0 0 0% 22 0 100% 88 0 100% 0 0 0% 578 0

• 02/28/95 TOTAL PERCENT COMPLETE COMPLETE 0 O.D% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 Note: All totals based on Active records. Period totals must also be Code Credit "Y". Page m.ntier 2
• 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 53

• * .) Palisades Nuclear Power Plant Third Inservice Inspection Interval ASME Section XI Category and Item Number Oesianation Legend 1. An 11 R 11 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 Item No. 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 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 A 11 Designates one directional "B" Designates multi-directional 11 C 11 Designates thermal movement
• "--"-EXAM ITEM CATEGORY NUMBER AUG MP-4.0 OVERLAY RG-1.14 TS-4.12 CATEGORY TOTAL: B-A B1 .11 B1 .12 B1.21 B1.21R B1.22 B1 .30 B1.40 B1 .40R CATEGORY TOTAL: Report: i3it""°p File: sch_item TOTAL ACTIVE 10 2 4 28 44 3 9 12 0 28 Index: key/Exam Category THIRD INT 3 PERCENT PERIOD 1 PERIOD 1 SCH ED TOTAL POP SCHEDULED COMPLETE 10 100.0% 0 0 2 100.0% 0 0 4 100.0% 0 0 28 100.0% 0 0 -----44 0 0 3 100.0% 0 0 9 100.0% 0 0 100.0% 0 0 0 0.0% 0 0 12 100.0% 0 0 100.0% 0 0 0 0.0% 0 0 100.0% 0 0 27 0 0 PLAN
• 02/28/95 INTERVAL --,TEM NUMBER

SUMMARY

PERIOD 1 PERIOD 2 PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3 TOTAL PERCENT PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE 0% 0 0 0% 0 0 0% 0 0.0% 0% 0 0 0% 0 0 0% 0 0.0% 0% 0 0 0% 0 0 0% 0 0.0% 0% 0 0 0% 0 0 0% 0 0.0% --------*------**--*- 0 0 0 0 0 0% 0 0 0% 3 0 100% 0 0.0% 0% 0 0 0% 9 0 100% 0 0.0% 0% 0 0 0% 0 100% 0 0.0% 0% 0 0 0% 0 0 0% 0 0.0% 0% 0 0 0% 12 0 100% 0 0.0% 0% 0 0 0% 0 100% 0 0.0% 0% 0 0 0% 0 0 0% 0 0.0% 0% 0 0 0% 0 100"1, 0 0.0% 0 0 27 0 0 Note: All totals based on Active records. Period totals RlJSt also be Code Credit 11 Y 11* Page nllJlber 1 .*.:

• EXAM ITEM TOTAL CATEGORY NUMBER ACTIVE B-B B2.11 B2.11R 82.12 82.21 82.21R 82.22 B2.22R B2.31 82.32 B2.40 B2.51 82.80 CATEGORY TOTAL: 8-D 83.100 83.100R 83.110 Report:

File: sch_item 0 2 4 0 8 4 10 2 2 2 36 0 6 0 Index: key/Exam Category INT 3 SCH ED 0 2 2 0 8 2 2 2 22 0 6 0 PERCENT TOTAL POP 0.0% 100.0% 50.0% 100.0% 100.0% 0.0% 100.0% 50.0% 10.0% 50.0% 100.0% 100.0% 0.0% 100.0% 0.0% PLAN TH I RD INTERVAL ----dEM NUMBER

SUMMARY

PERIOD 1 PERIOD 1 PERIOD 1 SCHEDULED COMPLETE PERCENT 0 0 0% 0 50% 0 0 0% 0 0 0% 0 100% 0 0 0% 4 0 50% 0 0 0% 0 100% 0 0 0% 0 0 0% 0 0 0% 7 0 0 0 0% 0 0 0% 0 0 0% PERIOD 2 PERIOD 2 PERIOD 1+2 SCHEDULED COMPLETE PERCENT 0 0 0% 0 100% 0 50% 0 0 0% 0 0 100% 0 0 0% 4 0 100% 0 0 0% 0 0 100% 0 0 0% 0 0 0% 0 0 0% 6 0 0 0 0% 0 0 0% 0 0 0% -----------

---

PERIOD 3 PERIOD 3 PERIOD 1-3 SCHEDULED COMPLETE PERCENT 0 0 0% 0 0 100% 0 0 100% 0 0 100% 0 0 0% 0 0 100% 2 0 100% 0 0 100% 0 100% 2 0 100% 2 0 100% 9 0 0 0 0% 6 0 100% 0 0 0%

• 02/28/95 TOTAL PERCENT COMPLETE COMPLETE 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0 0.0% 0 0.0% 0 0.0% Note: All totals based on Active records. Period totals must also be Code Credit ,"Y". Page number 2 L
• EXAM ITEM TOTAL INT 3 CATEGORY NUMBER ACTIVE SCHED B-D B3.110R B3.120 63.130 B3.130R B3.140 B3. 1SO B3.150R 63. 160 B3.160R B3.90 B3.90R CATEGORY TOTAL: B-F BS .130 B5. 130R BS .140 85. 150 Report: i3itrf'4'0p Fi le: sch_ item 6 6 0 6 6 0 4 0 2 0 6 42 23 0 13 5 Index: key/Exam Category 6 6 0 6 6 0 2 0 0 0 6 38 23 0 13 s PERCENT TOTAL POP 100.0% 100.0% 0.0% 100.0% 100.0% 0.0% S0.0% 0.0% 0.0% 0.0% 100.0% 100.0% 0.0% 100.0% 100.0%

PLAN THIRD INTERVAL ----,:TEM NUMBER

SUMMARY

PERICO 1 PERIOD 1 PERIOD 1 SCHEDULED COMPLETE PERCENT 0 1 ?"" 0 0 0 0% 3 0 50% 3 0 50% 0 0 0% 0 0 0% 0 0 0% 0 0 0% 0 0 0% 0 0 0% 8 0 s 0 22% 0 0 0% 7 0 S4% 0 0 0% PERICO 2 PERIOD 2 PERIOD 1+2 SCHEDULED COMPLETE PERCENT 5 0 100% 5 0 100% o-0 0% 0 0 50% 0 0 50% 0 0 0% 2 0 100% 0 0 0% 0 0 0% 0 0 0% 0 0 0% 12 0 4 0 39% 0 0 0% 0 0 S4% 0 0 0% PERIOO 3 PERICO 3 PERIOD 1-3 SCHEDULED COMPLETE PERCENT 0 0 100% 0 0 100% 0 0 0% 3 0 100% 3 0 100% 0 0 0% 0 0 100% 0 0 0% 0 0 0% 0 0 0% 6 0 100% 18 0 14 0 100% 0 0 0% 6 0 100% 5 0 100%

• 02/28/9S TOTAL PERCENT COMPLETE COMPLETE 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0.0% 0 0 0.0% 0 0.0% 0 0.0% 0 0.0% Note: ALL totals based on Active records. Period totals must also be Code Credit "Y". Page number 3
• PLAN
• D2/20/95 THIRD INTERVAL * -EM NUMBER

SUMMARY

EXAM ITEM TOTAL INT 3 PERCENT PERIOD 1 PERIOD 1 PERIOD 1 PERIOD 2 PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE 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

---

• PAL I SADE.:OGRAM PLAN
• 02/28/95 ,....,...., THIRD INTERVAL EM NUMBER

SUMMARY

EXAM ITEM TOTAL INT 3 PERCENT PERIOD 1 PERIOD 1 PERIOD 1 PERIOD 2 PERIOO 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3 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 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: 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 --------------_ __J EXAM ITEM CATEGORY NUMBER B-J 89.32 B9.40 CATEGORY TOTAL: B-K-1 B10.10 B10.20 CATEGORY TOTAL: B-L-1 B12.10 8-L-2 812.20 8-M-1 B12.40 8-M-2 812.50 Report: i3itrrpop File: sch_item TOTAL ACTIVE 5 351 746 4 16 20 4 4 2 14 Index: key/Exam Category INT 3 PERCENT SCH ED TOTAL POP 5 100.0% 103 29.3% 217 0 0.0% 0 0.0% 0 25.0% 0 0.0% 50.0% 0 0.0% THIRD PLAN INTERVAL NUMBER

SUMMARY

PERIOD 1 PERIOD 1 PERIOD 1 PERIOD 2 PERIOD 2 SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE 0 0 0% 5 0 18 0 17"-' 49 0 38 0 82 0 0 0 0% 0 0 0 0 0% 0 0 0 0 0 0 0 0 0% 0 0 0 0% 0 0 0 0 0% 0 0 0 0 0% 0 0

• 02/28/95 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3 TOTAL PERCENT PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE 100% 0 0 100% 0 0.0% 65% 36 0 100% 0 0.0% 97 0 0 0% 0 0 0% 0 0.0% 0% 0 0 0% 0 0.0% 0 0 0 100% 0 0 100% 0 0.0% 0% 0 0 0% 0 0.0% 0% 0 100% 0 0.0% 0% 0 0 0% 0 0.0% Note: ALL totals based on Active records. Period totals rrust also be Code Credit "Y". Page m.rnber 6
• \;__,,-' EXAM ITEM TOTAL INT 3 PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP B*N-1 B13.10 100.0% B-N-2 813.50 100.0% !?13.60 100.0% *CATEGORY TOTAL: 2 2 B-N-3 813.70 100.0% 8-0 814.10 233 6 2.6% C*A C1. 10 4 2 50.0% C1.10R 4 2 50.0% C1.20 6 3 50.0% C1.30 2 50.0% C1.30R 2 50.0% Report:

File: sch_item Index: key/Exam Category ---- PLAN

• 02/28/95 THIRD INTERVAL *--'dEM NUMBER

SUMMARY

PERIOD 1 PERIOD 1 PERIOD 1 SCHEDULED COMPLETE. PERCENT PERIOD 2 PERIOD 2 PERIOD 1+2 I PERIOD 3 PERIOD 3 PERIOD 1-3 SCHEDULED COMPLETE PERCENT I SCHEDULED COMPLETE PERCENT TOTAL PERCENT COMPLETE COMPLETE ' 1 0 100% 0 200% 0 300% 0 0.0% 0 0 0% 0 0 0% 0 100% 0 0.0% 0 0 0% 0 0 0% 0 100% 0 0.0% 0 0 0 0 2 0 0 0 0 0% 0 0 0% 0 100% 0 0.0% 0 0 0% 0 0 0% 6 0 100% 0 0.0% 0 0 0% 0 50% 0 100% 0 0.0% 0 0 0% 0 50% 0 100% 0 0.0% 2 0 67% 0 100% 0 0 100% 0 0.0% 0 0 0% 0 100% 0 0 100% 0 0.0% 0 0 0% 0 0 0% 0 100% 0 0.0% Note: All totals based on Active records. *Period totals must also be Code Credit "Y" *. Page nll!lber 7 \

• '--' EXAM ITEM CATEGORY NUMBER C-A CATEGORY TOTAL: C-B C2.21 C2.21R C2.22 CATEGORY TOTAL: c-c C3.10 C3.20 CATEGORY TOTAL: c-c# CC#-NA C-C* CC*-NA Report: i3itTr4J0p File: sch_item TOTAL ACTIVE 18 12 4 6 22 12 51 63 11 56 Index: key/Exam Category INT 3 PERCENT SCH ED TOTAL POP 9 8 66. 7"!. 2 50.0% 6 100.0% 16 6 50.0% 51 100.0% 57 0 0.0% 0 0.0% ------

PLAN THIRD INTERVAL --dEM NUMBER

SUMMARY

PERIOD 1 PERIOD 1 PERIOD 1 PERIOD 2 PERIOD 2 SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE 2 0 4 0 0 13% 4 0 2 0 100% 0 0 0 17% 2 0 4 0 6 0 0 0 0% 0 0 11 0 22% 25 0 11 0 25 0 0 0 0% 0 0 0 0 0% 0 0

• 02/28/95 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3 TOTAL PERCENT PERCENT SCHEDULED COMPLETE PERCENT CCJ4PLETE COMPLETE 3 0 0 63% 3 0 100% 0 0.0% 100% 0 0 100% 0 0.0% 50% 3 0 100% 0 0.0% 6 0 0 0% 6 0 100% 0 0.0% 71% 15 0 100% 0 0.0% 21 0 0 0% 0 0 0% 0 0.0% 0% 0 0 0% 0 0.0% Note: ALL totals' based on Active records. Period totals must also be Code Credit "Y". Page mmber 8
• PALISADE.WGRAM PLAN
• 02/28/95 '-....-*'

THIRD INTERVAL NUMBER

SUMMARY

EXAM ITEM TOTAL INT 3 PERCENT PERIOD 1 PERIOD 1 PERIOD 1 PERIOD 2 PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3 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

• PLAN
• 02/28/95 TH I RD INTERVAL **--.rEM NUMBER

SUMMARY

EXAM ITEM TOTAL INT 3 PERCENT PERIOD 1 PERIOD 1 PERIOD 1 PERIOD 2 PERIOO 2 PERIOD 1+2 PERIOD 3 PERIOO 3 PERIOD 1-3 TOTAL PERCENT CATEGORY NUMBER ACTIVE SCH ED TOTAL POP SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE 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. 1 OB 29 8 27.6% 3 0 38% 3 0 75% 2 0 100% 0 0.0% F1. 1 OC 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 *'*'*

• PLAN
• 02/28/95 ... .... TH I RD I MTERVAL -*-1 TEM NUMBER

SUMMARY

EXAM ITEM TOTAL INT 3 PERCENT PERIOD 1 PERIOD 1 PERIOD 1 PERIOD 2 PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3 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 CATEGORY NUMBER GRAND TOTAL: Report: i3itmpop File: sch_item TOTAL ACTIVE --5311 Index: key/Exam Category THIRD INT 3 PERCENT PERIOD 1 PERIOD 1 SCH ED TOTAL POP SCHEDULED COMPLETE 1400 346 0 PLAN INTERVAL ---, fEM NUMBER

SUMMARY

• 02/28/95 PERIOD 1 PERIOD 2 PERIOD 2 PERIOD 1+2 PERIOD 3 PERIOD 3 PERIOD 1-3 TOTAL PERCENT PERCENT SCHEDULED COMPLETE PERCENT SCHEDULED COMPLETE PERCENT COMPLETE COMPLETE 433 0 578 0 0 Note: All totals based on Active records. Period totals must also be Code Credit "Y". Page number 12
• KEY: Red =Class 1 Blue = Class 2 Green = Class 3 .) 54 SECTION 7 COLOR CODED PIPING AND INSTRUMENT DIAGRAMS

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INSTALLED ON L T-0105

• 2. MV-PC-112145A OPERATOR IS INACCESSIBLE DUE TO BEING ENCASED IN LEAK SEALANT PER SC-c:i4-068.

THIS DWG. JS IS! COLOR CODED. REV[SED PER APP. R CIRCUIT ANAL'fSlS Cl OCR-"14-1087, ALB wu: NONE DfWlo'N PG BROWN PALISADES PLANT CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM PRIMARY COOLANT SYSTEM 12195121 M-201 SH.2 . -( H G F E D c B A .. H G F E D c B ... 7 PT-021212 )M-202 SH. (F-7) LOCV' s )M-202 SH. tE-6l OPCT M-210 SH 2 CF-Bl NSSS SAMPLE FLUSH 5 HC-21-11 rq:;:] @PURIFICATION ION EXCHANGERS 3 1 -NX2ciMZDR-2032 3 @ DEBORAT ING ION EXCHANGER 2 ------ 1s>--_ .. , .. _ __,_ __ -I LC WASTE GAS SURGE TANK M-211 SH. 2 <F-6l ( P-N130M3-2:320). 8 7 6 HC-19-3 1 3u -NX29M2DR-2068 3 1 -NX26M30R-101) 5 C32 4 CG-4l HC-32-3 11 CLEAN RESIN TRANSFER TANK M-210 SH LB <B-1) 1 -224g }--."'4-SERVICE AIR Hc-13-3u SPENT RESIN STORAGE TANK T-69 *__i _____ 1=== d 2HC-1-2' LO (3 1 -NX2CJM2DR-21C!7) (3" -NX2CJM2DR-203'3) HC-13-%11 YS-0205 M-202 SH. lA CG-71 9509120015 -03 51 THIS DWG. IS ISI COLOR CODED. 1-17 q5 AODED PREVlOUSL'I' UNTAGGED VALVE MY-cvc5q5 AS "314' NX-176YR-5q5' PER DCR-q'5-"18. PALISADES PLANT CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM CHEMICAL & VOLUME CONTROL SYSTEM 0950 M-202 *:\;H. 1 ALB 51 M202-l.DGN 3 2 1 H G F E D c B A H G F E D c B .. 8 R VDT M-21!21 SH. 2 HC-2-3 1 CH-3-;l ______ ,_ 3 1 -N23BM2R-2137 7 HCO-11218-'%I CHARGING PUMP ACCUMULATORS M-21212 SH. 18 CF-21 PURIF. FILTERS M-21212 SH. 1 (8-7) (p;\ HC-17-1 1 1 n -N276Y-212 6 FROM POS-21558 <E-71 5 VGCH M-211 SH. 3 IF-ll DILUTE DJ LUTE T1 r---V:" -17SY-2120 DEDT HC-7-3 1 "' "" N 4 +/- i I "' HC-8-4 1 2 1 -NS 176R-2127 3 1 -N29M2DR-2 126 HC-7-3 1 <r "" N "' I HC-7-3 1 2 1 -NS 176R-2133 TO 4 3 u __; MV-CVC-615 2 HC-6-3 1 HC 1 HCC RECYCLED BORIC ACID STORAGE TANK HCC-55-2' M-650 SH IA IF-8) P -N 176YR-2175 FLOOR 121713 C33 Also on Card 9509120015 -oLf THIS DWG. IS ISI COLOR CODED. 24 REV. VALVE 76'(-2J29A PER FES-94-36q &. OCR q4-rn1e JGO H G F E D c B @.-CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT COVERT, MICHIGAN PIPING & INSTRUMENT DIAGRAM CHEMICAL & VOLUME CONTROL SYSTEM H G F E D c B A 8 I 7 PRIMARY MAKE-UP \!(ATER M-202 SH. 1A (f-4) DEOT M-2l0 SH 1 <B-7l HC-1-!I.,' LEAKAGE TEST CC-7-2" * -V2276 V2277) V2278 5 @ CHARO I NG PL !MP SEAL p S5A, p-558, P-55C !DENT I CAL EXCEPT TAG. NO!!-ARE IN PARENTHESIS> 4 MV-CVC-591 OBA & NORMAL SHUTDOWN SEOUENCER (START P-5581 i 2* -N7M3-2195 H.P. SAFETY INJECTION M-204 SH. lA IC-Bl WELL CHARGING PUMP 4 L. o. I _ WATER FOR O ISCHARGE MA INF OLD FLUSH AND HYDROS TAT IC TEST CONNECTION CHARGING PUMPS <REFER TO M-209 FOR COOL ING) DEDT TRIP PUMP ON LOW' SUCTION -PRESS WHEN P55B IS POWERED pmflR ps 5 tf tf RcE DEDT WATER FOR DISCHARGE MA INFOLD FLUSH AND HYDROSTATIC TEST CONNECT fON OEOT HC-33-%." THIS DWG. IS !SI COLOR CODED. 3 DEDT WATER 7 VCT M-202 SH. IA CC-7l l 9509120015 1 g i;f 1 -05 JGD H G F E D c B 8 * ... H G F E D c B 8 NJTROGEN M-222 SH. I <G-ll VCVH M-211 SH. 2 (0-8l SIRW TANK M-204 SH 1 CG-Bl CLOSES REACTOR COOLANT LOOP lA < M-201. SH. l"E <F-3) REACTOR COOLANT LOOP 18 M-201. SH. 1 CE-41 REACTOR COOLANT LOOP 2A M 201, SH. 1 <E-4l REACTOR COOLANT LOOP 28 M 201. SH. 1 CF-5l I_ 7 GC-7-1' HB-1-1 1 :.. .'. 7 HB GC CC-4-12° 12 1 -ES 3131 12° -ES 3146 7 6 5 4 3 HB-1-1' * :.. CC-4-12" CC 4 12P THIS DWG. IS ISI COLOR CODED. 6 5 4 3 2 cc-4-sn cc 4 6 11 CC-4-6° SAFETY INJECT ION HEADER M-203. SH. 2 <D-Bl SAFETY INJECT ION HEADER M-203. SH. 2 tC 81 SAFETY INJECT !ON HEADER M-203. SH. 2 CB-Bl SAFETY INJECTION HEADER M-203, SH. 2 CA-Bl 1 9509120015 -Db 3 7 5 qr CHANGED CV-306q STATUS TO CLOSED PER OCR Cli-404 H G F E D c fP' 2 ".::::+/-:) COVERT, MICHIGAN 1' PIPING & INSTRUMENT l=i§ill!i!ODlli<qSAFETY INJECTION. CONTAINMENT ,;_PRAY & SHUDOWN COOLING SYSTEM M203 37 1 ....... H G F E e D c B

• A @-0 "' "' ..;. ti <C-2l T-829 M-203, SH. 1 <8-2l T-B2C M-203. SH. 1 CB-2l T-820 M-21213. SH. l CB-21 8 2 1 -ES z '4* ti ['.: +/- TEST w 8 7 FA! 7 OBA SEQUENCER I --1 I COL) "'"' :i"'

T '"""' I C§E I e-50A e -----!-----"3FM-209 SH. 1 >uh ----1 I T C-33 El I "' I I ----t I I I Y,* I P-50C "'-----!--

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L) D ---1 CA-4> GC-1-6" 6 I I t I J BCA-1-2 1 GC-1-12" @ 0303 @ 0301 7 e3 @ 5 y FLOOR DRAIN C-33 C-03 at\-33 LOW PRESSURE INJECTION M-204-SH 1 3088 LOOP a 1 HOT LEG ::i>M-201.

SH. I> <B-3l 2 1 -CA 739 A/S <A-B> C8-3J I I A/S ,._ ___ _.

i: ZS EVENT RECORDER (QPENI CONTAINMENT SPRAY PUMP TEST INTERLOCK IOPEN & CLOSE) I -CA 738 +-------@@ CONTAINMENT SPRAY IOUAL ITY GROUP 21 _d33 -B GC-10-IP 1 1-ES 3227 1 RECIRCULATION TO SIRW TANK l'-ES3217 GC-10-1" GC-1-8 1 GC-1-8 1 LC GC-10-1 1 RECIRCULATION TO SIRW TANK M-21214 SH 1 SHUTDOWN COOLING CG-Bl HEAT EXCHANGER E-608 M-204 SH 1 C0-8l SHUTDOWN COOLING HEAT EXCHANGER E-S0A M-204 SH 1 CE-Bl THIS DWG. IS ISI COLOR CODED. *' .. H G F E D c B HMW CIC APP A CONSUMERS POWER COMPANY PALJSAOES NUCLEAR PLANT MICHIGAN 9509120015 01 PIPING & INSTRUMENT.-

D!AG::tAM pWUll!il!;:;io;:::jSAFETY INJECTION. CONTAINMENj:' SPRAY & SHUTDOWN SYSTEM M203 .""' 2 16 4 3 2 1 H G F E D c B -ij HC-23-6 1 1 1 -Nlc:IM3-3236 <D-1l B TRAIN M-203 SH. 2 CONTAINMENT SPRAY RECIRCULATION A TRAIN M-21213 SH. 2 <c-1> CONT A rNMENT SPRAY M-203 SH. 2 (8-ll CONTAINMENT SPRAY ( M-2"3 SH. 2 ( <C-1l LP SAFETY INJECTION M-21213 SH. 2 m-4l SHUTDOWN COOLJNG CROSSOVER TO SHUTDOWN COOL ING LINE THIS OWG (E-D 8 GC-10-P GC-10-1' 6 1 -N26M3DR-3225 REFER NOTE 5 GC-1-10 1 1 1 -N 130M3-3378 1 1 -N1312JM3-3380 WEST SAFEGUARDS ROOM SUMP M-211 SH. l A GC-1-8 1 8° -N138M3DR-3214 T I 3ZAZ4 c L.C. CONNECTION 32H 8 24 C TO SPENT FUEL POOL GC-1-8 1 t5 C33 ELO C33 COMPONENT COOLING WATER ---------------h 2 ( E-608 ) ----------------i---*' 7 CA 233A CA 233 WEST SAFEGUARDS ROOM SUMP M-211 SH. 1 :. It. ... JD-2-Y.:: 1 --1ijf/ C33 GC-1-12 1 _6_ GC-3-10' C33 @SHUTDOWN COOLING HEAT EXCHANGERS u --' w CONTAINMENT SPRAY VALVES TEST INTERLOCK t 0 u A c GC-1-12a GC-4-4 1 u 0 u 0 HP SAFETY INJECT ION PUMP .E=..Qfill_ @ LP SAFETY INJECT ION PUMP .E.::...6.Til. GC-3-10' OC-2-2" 0 GC-10-3a SIS CONTAINMENT SPRAY PUMPS I OR OBA SEQUENCER DC-1-3" LO (4' CHP ' u "' ' '" u "' OBA SEQUENCER J ::: 0.52-111 I -t:l 52-206 o..:ao ::i 9g:; lil:;;;: g " GC-L-12 1 5 5 s 5 o N ' :::c ::k <r-co-,....V> <.DV> ov 4 WEST ROOM 1--EAST ROOM NOTES1 L REFER TO DWG M-209 FOR SAFEGUARDS PUMPS COOL ING 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. SAFETY INJECTION PUMPS VENT AND DRAIN PIPING. 4, REFER TO DWG N555-6CJ5-1 <CE mm 2966-0-512182> FOR H.P. SAFETY INJECTION PUMPS VENT ANO DRAIN PIF'INO. 5. REFER TO P&I DIAGRAM M-225 FOR CONTROL AIR SYSTEM DETAILS. 6.HCC-101-2 1 & HCC-102-2 1 SHALL BE TIED INTO EXISTING VALVES ADJACENT TO TIE 7. ARE q ES" AND TAGGED AS BUILT UNLESS 3 l <D u "' ELC SPRAY REC IRC TO S IRW' M-204 SH. 18 CH-Bl RECIRC TO SIRW M-204 SH. 18 CG-Bl SIRW M-204 SH. LB CE-7l P668 SUBCOOL ING M-204 SH. 18 <0-7) SIRW (M-204 SH. IB( CC-7> LHT 42-2 c 8 1 -N26M3DR-3205 C33 GC-8-8 LC TEMPORARY CONNECTION FOR SPENT FUEL POOL C03 L 00 2--23 9 cs COOLING M-221 <A-5> SIRW (M-204 SH. IB( CA-7l -* f' LC -------TR-0351 Also Avsneola on Aperture Card 9 5 0 9 12 0 0 1 5 '"CJ{) REV DAT THIS DWG. IS ISI COLOR COOED. CHANGED GATE VALVE 3235 TO GLOBE VALVE PER OCR 94-613 DESCRIPTION CONSUMERS POWER COMPANY HMW 8'1' APP H G F E D c B PALISADES NUCLEAR PLANT COVERT. MICHIGAN M21214-1.0GN A PREVlOUS RECORD ISSUE PIPING & INSTRUMENT DIAGRAM INJECTION CONTAINMENT SPRAY AND SHUTDOWN COOLING SYSTEM A I M204 2 1 H G F E D c B A ---t3 ... (Q-4) CHARGING PUMPS )M-202 SH.

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j;oo*---'

CB-4l I ENG. SFG. SUMP M-210 SH. 1 CA-8) DC-1-4' CC-6-3' POS 312136 C33 HC-1-1' !. '-' "' RV 3286 I ' '-' D GC-4-4" N "' @ >JEST ROOM _J --EASTRoOM -i --'-' :! "' "' CONTAINMENT 1 14* -N2qM2DR-322s SPRAY PUMP P-54A LO L_OR OBA SEQUENCER 2 1-NL10M3-3188 @HP SAFETY INJECT ION PUMP .E=..fil?B C03 MO I I 3189 ---r GC HC J 6 1 -N234M3R-31B3 "" ;;:; V> w I ELC C33 I WEST ROOM C33 1 CONTAINMENT SUMP ... =; :'::: ::::: j N DIRTY WASTE ORA IN HEADER <F-8l NOTESo 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. 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.

7. ALL VALVE PREFIXES ARE 1 ES 1 AND TAGGED AS BUILT UNLESS OTHER WI SE NOTED. 9so9i2oo1 s -OCf H F E D c THIS DWG. IS ISI COLOR COOED. REV, 8'-N1"M30R-3187 TO 3' PER OCR qJ-638

@ CONSUMERS POWER COMPANY

• PALISADES NUCLEAR PLANT COVERT, MICHIOAN H G F E D c B A . -t. SPRAY RECIRC. ) M-204 SH. 1)-.., IH-2J REC I RC M-204 SH. 1 CH-2> 8 ------r I SPENT FUEL POOL M-221 SH. 2 <C-Bl P668 M-21214 SH. 1 P668 SUBCOOLING M-204 SH. 1 <E-2) P54C M-204 SH. I <0-2) HC-4-6 1 GC-4-sn GC-4-4 1 l @l?' 7 EVENT RECORDER <CLOSED> P54B M-204 SH. 1 IB-2l L C33 HC-3-14n ::! a. :; 8 N :t 5 ES 33718 HC-4-6 1 HCC-102-2 1 J' )I H H HC-20-3' LO HCC-102-2 1 HCC-102-2 1 ---------1 --==========c='-=
• SEE NOTE *q OEDT 1" -N17B'(R-ES 502 RADWASTE M-210 SH 18 <C-3> FUEL POOL COOLING PUMPS SUCTION M-221 SH. 2 NOTES1 <C-4> eves M-202 SH lA CD-6> l .. Also AvaU£ibfte on Aperture Card 1. REFER TO DWG M-20CJ FOR SAFEGUARDS PUMPS COOLING. 2. REFER TO DWG M-211 FOR SAFEGUARDS ROOMS SUMPS. 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. 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 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. H F E D c 9509120015 -10 THIS DWG. IS ISI COLOR CODED. 3 2 CHANGED N.C.VALYE ES-3372 TO N.O.PER DCft q'!l-038. HMW CE6CRIPTJON 81' Ck N>P A CONSUMERS POWER COMPANY PALlSADES NUCLEAR PLANT COVERT, MlCHIGAN PIPING & INSTRUMENT u1!AGRAM INJECTION CONTAINME":;r-M204 1B 25 1 H G F D c B A :i "' "' -., N'\' STACK MONITOR I I t 11-K 11 ffi T--1 M-218 SH. 2 L--1 CG-5l I I e: 2; I I !-<H-5l I EE] HS ft" 1 11 -I M-2185.2 CG-5l 1'E-°'m 7 l.J.JWC) W oi5 "' ..,. "' 8 t;;w <M-226 SH. 1= CG-11 8 ---Wtn "'"'"" "'"' -<E " """' --' N " EB-6-Bn E-CJC M-205 SH. !B (A-3> EB-3-8 1 ::J <';' 'f w " "' " !ti ::C@ w :i t5"' 0"' I OS> 7 STOP VAC\7'ES 5 GOVERNOR VALVES EB-6-12 11 "' [jj ffi@ ffi 8 "' 5 OPENS Cl 80 PSIG m '" . WI (/) -::. w"' ' '=' "'m w ---; N "' z "'. w "'u.. §1 8 "' 4 3 I -<f1-207 SH. 1c< ((3-4) MAIN STEAM RD* K-7A & B HP SIDE -205 SH. 2A EB-5-6 1 EB-6-12 1 EB-s-ea GLAND SEAL STEAM SUPPLY M-205 SH. lA (0-6) MSR E-<::is 3>'.M-205 SH. 11§> CH-51 EB-s-au GB-2-16' MSR E-gs -205 SH. 1A CG-2l MSR E-go M-205 SH. 18 CG-51 MSR E-CJD -205 SH. 18 CH-31 E-5A M-2!216 SH. 1 0-7 J <DJ E-58 M-206 SH. l 0-3 ili rnrn "'"' <no '"' "'"' "'r;l .. ,, "' m LPA & B GLAND SEAL EXHAUST -205 SH. lA (F-7) 9 5 0 9 12 0 0 15 -l / THIS DWG. IS ISi COLOR CODED. MSR E-CJA 1A> 58 4gi3 ADDED TG-177 PER OCR G4-327 ITT H F E D c A PALISADES f\l.ICLEAF? PLANT COVERT, MICHICAN M20'3-1.ClGN 2 PIPING & INSTRUMENT DIAGRAM MAIN M205 1 58 H G F E D c B A _ .. PS-0741A B. DD, 2/3 LOW SUCT lON PRESSURE SH. 2 CG-7l AUX. F. P. 8 1 I OPENS ON LOSS _ _j _ _J OF DC POWER P-88 >M-207 SH 2 >----<H-6l <M-207 SH CG-Sl DRAIN JB-14-4 1 CK FW' 416 N ' .... <D .., ATMOSPHERE 7 AUX. FEEDWATER PUMP TURBINE DRIVER K-8 1-1 I ' F. P. TURBINE TRIP THRUST r I I 153 MAIN STEAM E-50A M-21217 SH 1 CG-Bl <VJ '-130-FW 557 1 1 F"-' 559 <DJ 5 CONDENSER M-206 SH 18 <C-4l 4 1 q BLEED VAL YE MS 500A rn -,,_.. _" "!"oo --_ j !RuD\_JJ' 2 1-130-FW 177 38 _I ,. 0 -' "-"' w > 0 1'-19-FW 561 4 3 4 1-14-FW 151 $ :.. ' "' ... !l! m I 2 I PBB RUNNING M-207 SH. 2 SPEED CONTROLLER HlC-0526 -7 M-207 SH.1 > (8 3l HB-40-1 1 M-207 SH.1< CA-2l P-lA SEAL WATER -21217 SH lA CA-5l EK 14 31 -2q-57q CONDENSER -206 SH 18 75 IC-5) CC,D-4l LS-0547 M-205 SH. 2A <C-3l 9509120015 \ I EB-5-4P 1 MAIN STEAM H.P. SIDE M-205 SH. 2A CG-Bl FT-0527 -<f'!-205 SH. 2A< lF-6l FT FW 617 4 GB 4-12 1 MSR E-GB LP STEAM M-205 SH. LA CG-7) FOR NOTES SEE M-205 SH. 1 THIS DWG. IS !SI COLOR CODED. 10-25 REV'D. YS-0520 & ID'D. YS-0513 4 1 94 PER DCR'-94-909 JGD ltAL.£ NONE P. C.. BROWN PALISADES PLANT CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM MAIN STEAM ANO AUXILIARY TURBINE SYSTEMS M-205 SH. 2 41 M205-'i. OGN H F E d i c _J

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7 6 5 4 3 2 1 .. H G F E 0 c 8 TURBINE BYPASS VALVE MAIN STEAM --r---------------

-!_ ----------

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_J TAVE <H-Sl I l ----------

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@TURBINE AU)( FW TURBINE K-8 < M-21Zl5 SH. 2 C CG-6l TO S. G. TEST COOLER ---SS)( TRIP I I AST INPUT I l __ _,, HB-GEN. LEV. / F. W. FLOW ----------1 FLASH TO S. G. TEST COOLER TANK T-2qA DBB-q-4* > M-226 SH. 1> lF-Sl I I I I 1 I I I I I I I I I 1 I 1 I 1 I NOTE 0781 0780 r 1 B B I T rE7P\ E/P 0790 I TO S. G. TEST COOLER FLASH TANK T-2qA <M 226 SH. 1 CF-Bl NOTE 1 L--------AUX FW TURBINE K-8 > M-205 SH. 2> CH-3l STEAM GEN. RECIRCULATION

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<( M-226 SH. 1< CB-Bl -----< TO S. G. TEST COOLER .

NOTES* 1. SHOWN FOR CV 07811 CV 078121. CV 077g & CV 121782 ARE IDENTICAL.

2. SX-IZl77g THAU SX-0782 ARE HYDROSTATIC TEST TAPS ..,, --,.--THIS DWG. IS ISI COLOR CODED. HIGH LEVEL OVERRIDE FROM LIA-0702 I TO LIC-121701 I DLs-1 {JFE\ 'eY) I I .L -DLS I TURBlN TRIP CAUTO-STOP I OIL PRESS> O..> N NONE HMW J.L. STERRETT PALISADES PLANT CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM FEEDWATER

& CONDENSATE SYSTEM ;i=::.:: 8 y '--/---' 0'l50 M-207 SH. 1 71 I U? N A @*,-, J 0 oCD N -N 9 5 0 9 12 0 1 5

k :k M207 ...

8 7 5 I_ 5 4 I 3 2 1 H G F E c B A 8 ... H G F E "' PS 0711 L D I I +--I I L_ c B A _B C42 <M21'l SH. 10< <H-7l PRIMARY PLANT DATA PROCESSOR t 7 E-68 REC!RC M-207 SH. lC F-46 [Hsl ELD CA-41 1 1 -SPtci-1588 A/S M-212 SH. 2 <G-7> 7 "' -"' '° "' ' w ,;; I 0 %* -FW 746 @(: LC <Vl "' z N ¢::t: "' 5 P -lG-FW 81C3A 1h -1G-FW 819 0708 t 0-ffi 6 ffi --,-DB-1-P THIS DB-1-1 1 DWG. <G-1) 5 :.. .'. SEAL WATER TO DRAIN TANK T-2SA M-21215 SH. 2 <B-2, 3l 5 z-:C i53 z" 8 s; "'"' "' !il .'. gs HB-40-1 ' SEE NOTE 1 T-15 & T-1GA M-220 SH. 2 ND SV 0744 A 4 L ELD 4 W N >-* "'"' "'"' -lli ':'1 '-' :l: l.i... '";J s: x :.. <D w E6A RECIRC. M-21217 SH. IC <A-4> (St * ...... * "':r: -l5 55 (!;:! V') V _J I ..JN ...._ N ' N ' " " --I I -+ I 3 '-' :l: :.. cb w :i "'"'-r;' b' 2 FILL 1 1' -FW 723 CHEMICAL ADDITION T-35 CORROSION PRODUCT MON !TOR M-q7 A/S M-212 SH. 2 <H-71 NOTES1 THIS DWG. IA-Bl MAIN FEEDWATER LOCAL ff-26Al 1 l. REFER TO P& IO M-225 FOR CONTROL AIR SYSTEM DETAILS. I 1---- SV-07 lla ;;i>M-207 SH. 1c> <A-8) FS 0708-GB-c:i-3 1 3 E58 M-207 SH. lC (H-6) SV-0711 -----ZM-207 SH. ic> CB-Bl AN STEC APERTURE CARD Also Awansbte on Apertu;-e Card THIS DWG. IS ISI COLOR CODED. 9509120015 CONDENSATE M-207 SH. lC ll E5A -207 SH. lC !H-7, Sl ? NONE PALISADES PLANT M207-1A. DGN CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM FEEDWATER & CONDENSATE SYSTEM M-207 SH. lA H F E D c A 8 H G F E

• D c B P-llZIA & B SEALS M-21217 SH. lC CO, E-2. 3) BYP TO E3A M-207 SH. lC CG-61 WATER BOX WEST OUTLET CHEMICAL INJECTION HYDRAZINE M-653 SH. 1 (8-7) ) M-220 SH. 2) CA-81 08-54-31 1 P-10A & 8 SUCTION (M-207 SH. IC ( (8-2) 7 " ,, -s l/) UJ Q? GB-5-14 1 ;'.'; ;?_ w Cl 1!l .,. "' .,. --' "' <l: f-y 6 u " c::i -:
• E9 " " 5 18 1 CD 724A GB-3-1' p -l'l co 15121 C42

{l-Fl <%* -13121-CD i!l_,_ __ __, ffl HOT WELL CD 151 1 I -130-CD 530 4 C-208 PANEL M-q03 CG-7l 6 1 -2CJ-CD 130 3 ffi Q;i

• ::i: > HB-26-12a ---. ---6 N y :.. J EVENT -RECORDER

&DLS 6 1 CD 737 EVENT >RECORDER &DLS 33 136) I I I I I I I I I I I I I I I I I I I I I I I I I 2 MAIN STEAM M-205 SH. 1 <A-B> CV 05228 STEAM TRAP M-21215 SH. 2 IF-Sl PLANT HEATING STEAM TRAP M-206 SH. 1 CH-Bl c:S"' u_, g;"' """ a.."' =>"° !;!1o0 <<_, :.:_, ;;: ' "' "' "' " 93 Bj ii1 L1J :1: 1'-1CjCQ142 LP TURBINE HOOD SPRAY -205 SH 18 CG-21 B/P TO E-38 GB-5-14" M-207 SH. LC <G-31 1 ANSTEC APERTURE CARD Aleo Card 9509120015-(S-TK. > FAST MAKE-UP FOR MA IN STM. DUMP NOTES1 1. REFER TO P&IO M-225 FOR CONTROL A lR SYSTEM OETA ILS INGERSOLL-RAND CONN. NO.' S CF. P. 5q35-\::5' M4-4. I-R DWG: N4 210RET501Xll THIS DWG. IS ISI COLOR CODED. 22 q!4 CORRECTED C0-646 CONNECTION PER OCR q4-823 PALISADES PLANT CONSUMERS POWER COMPANY *PIPING & INSTRUMENT DIAGRAM FEEDWATER & CONDENSATE SYSTEM IZJ"l51ZJ M-207 SH. lB 22 H F E D c 8 A "' -H G F E D c 8 8 I 7 5 ,_ {0 --LOW FLOW LOGIC TO START P-BB 0-OR P-BC I 4 PUMP P-8A RUNNING I I I I y I ISEE NOTE 11 1-filTI !i ili --! >= I <SEE NOTE 1l I -lE.ruJ I <SEE NOTE [) I I -lE.ruJ I [TOCFMlCll - I -,._ I I I 75 c I I I T 3 I' ,-----1-I CSEE NOTETl I -filTI I <SEE NOTE ll I T Cll <@;} I I TO CFM ._ 79 T _<NDTE4l I c .-H t(@q ' I 74 c T I DBB --t---EBB I 1 I c I cu I I I I 2 +---LH I _500 E-50B ISOLATED CII I E-'50B ISOLATED Cll _i ---ANY ISOLATION I CU 141 VALVE CLOSED OR C SG ISOLATED -t I E-50A ISOLATED -H I T ----50A E-50A ISOLATED I 1 NOTES ._., 1* SEE P & lD M-20710> SKEET J, !FOGG A I MEANS STEAM GENERATOR E-500 IS ISOLATED. !FOGG B I MEANS STEAM GENERATOR E-50A IS [SOLATED.

2. REMOVE STRAINER lNTERNALS DURING NORMAL OPERATION

3. REMOVE STARTUP STRAINER DURING NORMAL OPERATION

• • 4. ALL AUXILIARY FEEO'lilATER SYSTEM SIGNAL MONJTOR fCFMl, EXCEPT H.O.

APERTURE CARD Also on Aperture Card 9509120015 -lb 22 THIS DWG. IS ISI COLOR CODED. 0-1 94 GI FLAGS PER OCR 94-9 02 HMW H G F E D c B A 22 "-*o* ____ __!__ ___ __:7'.'.__ ___ _L ___ ___

=:=::::

<VJ H G F E D c 8 SERVICE WATER PUMPS HiJB CR IT I CAL SERVICE _______ E WATER <E-Bl -------------------

---

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---

J§"C-i-24'

-1 ESS PUMPS SEAL COOLING :-------------------< M-209 SH.2 ( I CA-7> l I i ! ! i j ' ' ! I ! ! ' 8 7_ _t-h.so . W _L_ PS

• 0876 -<sw Js ! HB :.i cbi :r:: I ! i 1 :3-:::::c:

HB-23-24' JB HB 9509120015-\ 4 I PLANT AIR COMPRESSORS NOTES1 1.ALL VALVES ON THIS SHEET HAVE I sw* PREFIX. UNLESS 3.EXPANSION JOINT HAS HEIGHT PER SC ct0-153 . _3___ 7* UNEXTENDED REV DATE 7 [ __ JB-1-3 1 0-t JsH' HB I I L HB-23-24' JB-1-16" COOL ING TOWER PUMP STRUCTURE

• -? M-653 SH. 3) <G-8) SW RETURNS ( M-208 SH.I ( IE,F-11 CC-Bl SW RETURNS M-208 SH.l <B-1) NSTEC ERTURE CARD en THIS DWG. IS ISI COLOR cdorn. HMW DESCRIPTION BY APP CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT COVERT, MICHIGAN M208-!A.DGN PIPING & INSTRUMENT DIAGRAM SERVICE WATER M208 H G I I E D c A

---r *--,.. THIS DWG n C-8 n 0 0 r "' H "' THIS DWG D-8 G F E D c B STEAM GENERATOR TEST COOLERS E-71 CLOSE VALVE ON SIS r--_ 1'33 EB---t v 67 '10ij ......... '.-"'to 8 o_'------; VAL'fttsibLED C33 MV-SW 85C! MV-SW 860 MV-SW 861 ELO EEl--r 7 _J_ 6 5 4 CONTAINMENT AIR COOLERS SERVICE WATER RETURN 3 2 1 MV-SW 5qz (M-208 SH. 1Af-BC-44-4n <Vlc -suPPLY Hsc-44-3* r1

• _

D ----)M-200 SHa 1A)-.:..:-- ---...c.-J ..-LO._.. : 1 --------- 1-ilJ-----iJJ i %' ((j-41 "l;/ w l' -MV-2C! 1-SWS 1' -MV-294-SWS 3* -MV-291-SWS HB-2}-16' L!'.___,,.,,r:,, --::..* j (1

• MV 2s5 sws) :_!_'____

____ (P -MV-284-SWS) MV-SW 836 A/S DPS 1681 )M-218 SH. 7) <F-41 OPS 1681 ?M-218 SH. 7) CF-4l CONTROL ROOM H. V. A.C. CONDENSERS

---

)M-208 SH. IA) ------>---

CA-Bl I -CA748 9509120015-THIS DWG. IS IS SHEET HAVE I sws* 2' l<D ISI COLOR CODED. 1. SUFFIX. UNLESS OTHERWJSE NOTED. POWER COMPANY PIPING & INSTRUMENT DIAGRAM SERVICE WATER SYSTEM M208 is'-20 1 H G F 8 7 6 5 4 3 2 1 .* __ r--_ ____:::___-----"-- _ ___:___ __ L____Q___ _ _ __1_ _

• --H G F E D c B A *. p _, PC 1133A CVl & CQ) PC 1132A <Vl & CDJ M-2l0SH.2 SEAL LEAKAGE DPCT C0-5l -210 SH. 2 HC-27-*n 7 J:Y:if:l --l:D-t THIS OWG. CF-6) SpT-01318 SpT-01328

--LOCALLY MOUNTED _J REMOVE L ALARM 0g20 Z M-214 > '---SEE NOTE 1 C0-21 --m PC 1144A CVJ PC THIS DWG. CH-61 _J ..L THIS DWG. CO-Bl SpT-121131A SpT-0132A '--LOCALLY --J MOUNTED L ALARM 0ci1g z M-214 > THIS OWG. <F-Sl _J ..L THIS mm. CB-Bl K H (D DANGER' ) THIS DWG. C0-2> -ffi--.::-.t6'

---

SpT-01418 --01 1 SpT-1211428 i A --J L .;LARMM-:1:21 > REMOVE 113A 6 @PRIMARY COOLANT PUMPS INSTRUMENTATION <REFER TO M-214 FOR LUBE OIL SYSTEM JNSTRUMENTATIONl I ______ J t SEAL FLUSH M-203 SH 2 C0-5) lll -0.!i -------j L SEAL FLUSH M-203 SH 2 <F-5l _____ ___,, ______ J L ______ J SEAL FLUSH M-203 SH 2 CC-Bl 4 L 10' -418-CC 105 -156-CC 526 L ____ _ JB-7-3' JB-7-3' tl .:, JB-7-UZI JB-7-3 1 "' f*= CE LTYP. FO 45) REACTOR SHIELD COOL I NG SYSTEM M-221 SH 1 CE-8) !0 JS B28 3 1 X2. 5' RED. NOTE: 1. THE AMPLIFIERS & LIGHTS THAT ARE I.IN PLACE WILL BE REMOVED DETAIL "A" LOWER MOTOR BRG. HT. EXCH. 3n X2. 5 1 RED. 3n X. 75' RED. 1. 51 1----PCP MECH SEAL---------' 7 HT. EXCH. r;-i!l OUTLET I I A/S A/S Also Aval!able on Aperture Card THIS DWG. IS ISI COLOR CODED. STERRETT PALISADES PLANT CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM COMPONENT COOLING SYSTEM SH. 47 9509120015-1q I 3 2 ',;' H G F E D c B A _J 8 7 5 5 4 3 H F 1--------------{::}--- ___ .. ---- 1' -cc 517 ! '-r! s

'

I / ... :-c..i 1'*v "' .I SHUWGWN rnou NG <>* -" '" c*-'""-.--C. .. if >\! HEAO EXCHANGERS ....... ,,,.-.. ---*-*-.. w "'I ! I . '.,! sis ! +""' Y ['--;+/-:-* U -cc 518 ) J ' W 'J A : ,._,-------__:.___ ) r--------------------- ---J:i.'!:.24-E' C 12*-cc 104 ) *t _ E-54 ! _ ......

• '"' '

'"' < i ->----------------- ..._ -... : <G-Sl .,: 0g13 i '--' -:I I

• iJi '-' -! -E-+-J l v :i , ___ 120* t:.c I i CW-cc s1g / -------

j j \ ,_.,, >1 ' _ .. !'" --JJ, I * "' ! I ' "'! I i ' ,._" ,,. . WO i [_______ )p------*-- 12' -cc 102 ) i -ffi=J.< .. ! ---------------G E 0 c B A .,* 8 7 6 5 4 3 jg C2i a.L e ' "' i I : ! _L -, 2 CPERMIT ALARM WHEN v-0q13. cv-0e7q DR CV-0880 ARE OPENED Alsso on Aperture C6rd 1 9509120015 18 THIS DWG. IS ISI COLOR CODED. TO P-52 PERC-PAL-'l4-0HlB & *ONE J. L. STERRETT PIPING & INS COMPONENT DIAGRAM SYSTEM M-20g SH. 2 H G F E 0 c B A 18 2 l ____ _ G F E D c B < M-20G SH IG-5> RADWASTE EVAPORATOR M-655 CF-1l __ 8 7 m -!Iill I 6 5 MV-CC-565 _l_ 4 <B-!l 3 WASTE GAS COMPRESSOR 2 l. l M'l-CC-154 THROTILEO . ' RADWASTE EVAPORATORS M-655 CF-ll THIS DWG. IS IS! COLOR CODED. REV'D, CC-162 &164 TO L PER C-PAL-94-04l7A & DgR-94-990 JGD H G F E D c B A ':-+>_ SUMERS POWER COMPANY PALJSADES NUCLEAR PLANT COVEfl:T, MICHIGAN PIPING & INSTRUMENT DIAGRAM COMPONENT COOLING SYSTEM < M20CJ 3 1 27 ! 8 .. _ . ._ ... H G F E 0 c B A 3v -CAW 407 DEGASIFIER PUMP DISCHARGE M-210 SH 2 CH-2> 7 HB t-2" -CRW 508 HC Hc-3q-2* HC-38-2U J:Q;J I T I c1"38 A1s I 1_ 1 1 -CRW 514 L'-J DEDT y 2' 4' "' N ,', I.BIP 3' t-HC GI::f40 C40 RECEIVER -TANK CIRC. 4 PUMP P-70 5 HC-2-3 11 HS-1-2" HB-43-2" CLEAN WASTE RECEIVER TANKS HC-2-6 1 3 1 -CRW 765 "' N t-HC y lQiEJ A/S I 4 3 HC-2-3 3 1-CRW 121 3 11 -CRW 123 2"-CRW 510 2 3 1-cRw 125 A/S I c 1 M-222 SH 1 < <G-!l HBC-23-1" S. I. TANK-VENT M-211 SH 2 CE-8l 9509120015-2.{,. THIS DWG. IS ISI COLOR CODED *_1' M210-1A.OON PALISADES PLANT CONSUMERS POWER COMPANY tillI.E;.

1. ALL VALVES ON THIS PRINT ARE CAW PREF IX UNLESS OTHER'<<' ISE NOTE0.-2. RUD-1018 RMOVED TD ALLOW CONTAINMENT VENT PATH PIPING & INSTRUMENT DIAGRAM ADIOACTIVE WASTE TREATMENT SYSTEM CLEAN 0950 M-210 SH. lA 12 H G F E 0 c B A H G F E D c B A CLEAN 'f/ASTE F-57 A/8 M-210 SH 1 CE-ll CLEAN WASTE F-57 A/8 M-21121 SH 1 co-ll 8 -----I 3 z l ...... A/S HC-2-3 1 HC-2-3 1 I HC-2-3° HC"-f-HC

_______________________________________________________ Hc-2-30 HCC-45-3' 3 1 -CR'W 174 8 PURIFICATION DEM I NERAL I ZERS DRAINS M-21212 SH. 1 <B-1) DEDT M-210 SH. 1 CB-81 Pl 11ZJ64 IOl 6 Ll "' \ . 1 3' -CRW 751 CLEAN WASTE TRANSFER PUMP HCC-45-3' M-650 SH. 18 (A-8) HC-2-3' JB-2-HB-1-tYP HC-32-3 1 3 1 -CR'N 147 31 -CAW 13 ID> "' IDl HC-13-3P 4 u r 3 1 -CR'J 13 "'" l' [ a. "' . :ir:::i :c: sa.. Cf) r;-i 9 Ll ' ii!"' "' 3" 3 1-CRW 411 E=..li9ll I \ C-40@\ 42-I 4 I C-40 */ E::::..0fl \&(' ' 3° -CRW 128 I I 3" 2' 3 1 -CRW 3" 3" 3 1 -CAW HC-32-3' AN STEC APERTlJRE CA,RD Also At:aUable en C:trd 210 CLEAN RESIN TRANSFER TANK L:fil PURIFICATION & DEBORATING ION EXCHANGERS M-202 SH. 1 9509120015 -1' r.i 3 3 1 -CRW 15 tillI.E.;. ALL VALVES ON THIS PRINT ARE t;.Bil PREFIX UNLESS DTHERW I SE NOTED. T-Sg M-657 tB, C-Bl T-Sg M-657 CO-B> 2 THIS DWGw IS ISi COLOR CODED NONE PALISADES PLANT CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM RADIOACTIVE WASTE TREATMENT SYSTEM CLEAN @J 0"350 M-210 SH. lR 11 M2 10-lB. OGN 1 P&ID H G F E D c H G F E D c B 3 1-CRW 175 J. _ -I I 31 -CAW 748 Hc-2e-1* TREATED WASTE MON !TOR TANK HC-2-3' HC-2-2 1 LC z* -CAW 170 31 -CAW 167 1-----L -TRIP CLOSED --+ ON HIGH RADIATION LAUNDRY DRAIN PUMP cJ & FIL TEREO WASTE M-211 SH. 1 CB-1> HC-2-3 1 31 -CRV 173 HC-2-3 1 TREATED WASTE MONITOR TANK LOCAL C-101 c-<@}c DI -650 SH. lB <G-2> HC-2-2 1 2: .... 8 42' DILUTION LINE '------<-- 3 > HC0-23-2' 3' L.!.J 3 1-cRw 151 LCI HJO...W;!CC HCC-58-3' 168 HC-2-3 1 tiQIE_;_ HC-2-3* RADllASTE FILTER F-57A OR T-5q M-210 SH 18 CB-31 153 CLEAN WASTE DISTILLATE SYSTEM H-650 SH. 1 IB-2> Also AwaUab!e on Aperturta Card CILL VALVES ON THJS PREF IX UNLESS OTHERlllSE

• 9509120015 THIS DWG. IS ISI COLOR CODED. 10-18 REMOVED SEISMIC FLAGS PER OCR q4-q0z g q4 REV DATE DEStRJPTION HMW BY H G F E o, I c B Ck APP CONSUMERS POWER COMPANY A PALISADES NUCLEAR PLANT COVERT* MICHIGAN MZ10-1C.DGN PIPING & INSTRUMENT DIAGRAM RAD IOACTIVE WASTE TREATMEN;,;

SYSTEM CLEAN r-,

G F E D c B A 8 7 _r-r-HC-12-4' >M-218 SH,2>--CH-4> PUMP LEAKOFFS BELOW EL 5q0*-0* AUX. BLDG. FLOOR DRAINS BELOW EL. 5q0*-0* PUMP LEAKOFFS <BELOW EL. 5q0'-0'> AUX. BLDG. FLOOR DRAINS <BELOW EL.5q0'-0'> RW 114 RW 603 T-78A <EAST ROOM> ENGINEERED SAFEGUARD ROOMS SUMP PUMPS SPENT RESIN STRG. TANK <T-100> CF-71 HCD XCD HCD-17-3" AUXIUARY BUILDING SUMP PUMPS 8 7 HCD-127-2* 3"-DRW 788 VENT HCC-80-3' VGCH M-655 6 HC-1-2' 6 5 ---------- EMERGENCY SHOWER -""=-------- WASHDOWN AREA DRAINS 4 LAUNDRY DRAIN TANK T-70 3*-DRW 130 3 CHEM. I LAB, DRAIN TANK DRAINS <F-6> HC-18-4" HC-1-3' 2 1 -RW'-142 VCT RELIEF -202-SH 1 CF-SJ VACUUM DEGAS FIER RELIEF M-210 SH 2 CF-5l TURBINE BLDG. HC-1-2* LC He-1q-2* DlSCH. HC-l-.HB <C-3> 1' -DRW 510 2 y ODWH WASHING MACHINE SUMP PUMPS C40 @7----1 HB HBO l'-ORW 582) ? LAUNDRY EQUIP. 1 DIRTY WASITmTER F-53 2"-0RW 79 5 4 HC-1-2" MISCELLANEOUS Rl>OWASTE SYSTEM <M-851 SH. J E <D-BJ z FILTERED !i ..h'.ASIE MONITOR PUMP HC-1-2' HCC-5-3" I -DRW 124 9509120'015-l-b THIS owe;. IS lSI COLOR CODED 3 2 VGCH M-211 SH. 3 53 L CLEAN Wl>STE RECEIVER TANKS Al.. on pertMre Card I ( 2 1 -DRW 137) RADWASTE DISCHARGE HEADER -210 SH lC CA-6> HCC-10-3 1 MISCELLANEOUS WASTE DISTILLATE SYSTEM M-651 SH, IB IG-1> -CONSUMERS POWER COMPANY PALISADES PLANT PIPING & INSTRUMENT DIAGRAM DIRTY WASTE & GASEOUS WASTE 0950 M-211 Sl'i.l H G E D c B A H G F E D c B CLEAN AW TANK VENT HEADER -210 SH lA 2"-WG 120 QUENCH TANK VENT M-201 SH 2 CH-SJ HBC-23-1' LOCATE PILOT !E-2> 1'-WG 724 IN AIR ROOM SAFETY INJECTION TANK VENT M-203 SH 1 !H-8> 8 HB-1-1' HB-1-1' 7 L!!-fU HB-1-4' 1'-WG 723 7 -1 RV-1111 THIS DWG IF-4) HCC-1-3/4' DT-1113 THIS OWG IF-3l F-58 DRAIN M-218 SH 2 <C-Bl 6 1' HBC 22 1" 5 5 WASTE GAS SYSTEM HB-1-1 112' VCT VENT M-202 SH lA DIRTY WASTE DRAIN TANK RELIEF M-210 SH 1 CE-4l EQUIPMENT DRAIN TANK RELIEF M-210 SH I <E-Bl DEGASIFIER VACUUM PUMP DISCHARGE M-210 SH 2 <G-8> HB-1-1 1/2' HB-1-1 112' DEDT T-80 M-211 SH 1 (Q-8) 5 WG 516 HCC-125-11/2' 4 PURFICATION & DEBORATING ION EXCHANGERS VENT M-202 SH 1 <D-2l RADWASTE EVAPORATOR VENT M-650 WASTE GAS SURGE TANK I M-211 SH 3 THIS OWG (E-Gl HB-20-2' 2"-WG 706 HCC-125-1" 4 3 RE-1113 AND STACK M-211 SH 3 <G-9) 22 THIS OWG CD-Bl 2 1 T-68'S & T-101'5 HB I HBC HBC-35-2' RELIEF VALVES M-211SH 3< 112' WG 513 112" WG 514 FROM COMPRESSOR CONTROLLER I I I CF-81 I I I I I WASTE GAS DECAY TANKS M-211 SH 3 __ __ BY VENDOR _______ _ _I (C-8l WASTE GAS COMPRESSORS 517 9509120015 -21 THIS DWG. IS IS! COLOR CODED 3 ? HCC-123-1' HBC-21-l" GASEOUS WASTE SURGE TANK M-211 SH 3 <F-4) OT 1164 DISCHARGE M-211 SH 3 CB-7l HMW -CONSUMERS POWER COMPANY PALISADES PLANT MZll-2,0GN PIPING & INSTRUMENT DIAGRAM RADIO-ACTIVE WASTE TREATMENT SYSTEM GASEOUS WASTE ,"'{ 0'150 M-211 SH.2 H -1 G F: I E D c B A H G F E D c B A PS-1203 18-7) >THIS DWG >---52-1107/CS <C-7) < TH IS OWG $----PS-1202 IC-7l >THIS DWG >-----52-1207/CS l0-8) >THIS DWG >--1,============ C-35 COMPRESSOR CONTROLLER CSEE E-167) I I i------1 I f I I Pl L 1200A*._ ___ _ I I I I FILTER I I s ILENCER I I I l I I I -£E-C-13 -s L ___ _ C-35 < THIS DWG IG-7> ---* C-35 >THIS DWG CG-7l PI > -1202A*._ ___ _ FILTER SILENCER < <-_j CG-7) R 7 PS 1213 STBY START ------< THIS DWG < <C-4l PS-1214 C ---< THIS DWG < <C-Bl ' -CA 8;: g ... I DRAIN 6 5 *__[j:ji] C-35 )> THIS OWG > CG-SJ <Dl 4 II I I 1 I 1 I 1 I 1 I 1 I 1 I 1 I 1 NOTES1 1. ALL VALVE PREFIXES ARE n CA 1 UNLESS OTHERWISE NOTED. 2. ALL PLUG VAL YES ARE 3; 4u UNLESS OTHERWISE NOTED. '--------'-----'---------_JI I L. --_-_-_-_-____ :JJ INSTRUMENT A IR 1 I -263 M-212 SH. 2 IH-Sl N D -, INSTRUMENT AIR AUX IL I ARY BLOG. M-212 SH. 3 <F-31 (314' } t3/4' -MV-CA803) x ,,c""o"'o"'L'-'I'-'-N,_,G,,__T!..!O"'-W!!..=E!..!R_,_P_,,UccM.!!.P__,H_cO"'-U'=!.>!.S=.E TURB. BLDG

• INST. AIR TO 'N'EST SAFEGUARDS M225 SH 1 <H-8> DELUGE SYSTEM PS-1341 & PS-1344 M216 SH 2 3 CONTAINMENT BLDG I_ 2 THIS OWG. IS ISI COLOR CODED REMOVED CA-705 PER OCR g3-1185 ITT CONSUMERS POWER COMPANY PALISADES NUCLEAR PLANT COVERT,MICHJVAN PIPING & INSTRUMENT DIAGRAM SERVICE & INSTRUMENT AlR SYSTEM 4. M-212 1 H G F E D Cl H G F E D c B RE 1917 CV 3043 CV 3005 CV 3084 NEAf!. Y..!

CV 0867 CNEAR VHX-2> 8 CV 5018 CV 5020 I CV 5143 CV 5140 AUX BLQG AODJTION HEADER 7 CV 5138 CV ___ I_ IVl cv*s 1100A&8 SJ TANK VENT CV' S CTO STACI( OR WGSTl f---- CV 0148 CV 0157 M-212 SH. 5 CG-6> EVAPS E/P CV 5025 CCV CV 5007 5136 ID> 6 CV 5135 CV 112133 CV 1031 CV 1063 CV 101q CV 1018 5 CV 3003 CV 1058 CV 1028 CV 1030 CV ll'Zl24 CV 1023 CV 1056 CV 1013 CV 1012 602' P'lPEWm:::iY EVAP DIST _TANK .fil:L CV CV 200a2117 -<B CV CV 2001 3044 CV 3040 CV 1007 CV 1038 CV 1045 CV 200q CV 1002 CV 1036 CV 1044 CV 1001 CV 1037 CV 5128 <CV-3046> CV 3065 L_ ---602 IS' PIPEWAY CV 1004 cv 1102 CV 5133 CV 5132 CV 5122 CV 1101 2004 2115 CV 2003 CV 2113 13 CV *' TO CA-2'qq EAST ESS M-225 SH. I CH-1> I 112L3 CV 1114 HlC 1123 3 MIDDLE CORRIDOR N2 M-222 SH. 2 CE-61 CA-221 M-212 SH. 3 <G-8> l .*. CV CV CV CV CV CV I 121D 1121C 11218 lll'lC lllCJD lllCJA CV CV CV CV I 120A I 1208 I 120C I 120D 9509120015 -L..C\ 2 Tl-US DRAWING JS ISl COLOR CODED. 11T CONSUMERS POWER COMPANY Plll.JSADES l'tlJCLEM PLANT CD'IERT, HICMIGAN HZlZ-4.DGN PIPING & INSTRUMENT DIAGRAM INSTRUMENT Al.fl WALK DOWN M-212 4. 1 H G F E D c B A H G F E D c B A ', NON-CRITICAL SERVICE llATER M-208 SH.I <H-5> CIRC. llATER LINE HYPOCHLORITE INJECTION DIFFUSER M-663 SH. ( -4> IEAOER B L. 58 FILL <OUTSIOEI 11--1111---------, I *-HH 4 I "-CHM RTD-1-2" 2" -CHM 752 T-lSA S. B HYPOCHLORITE STORAGE TANKS 7 -----, 3* I I I';! INTAl<E STRUCT. !l!l'iJ UPSTREAM TRAVELING SCREENS I I I --, CHLORINATOR CONTROL PANEL C48 I *-CHM 103 I -, TO FLOOR DRAIN BIO BOX 3"-54-SW 10'5 .:re-21-4* CORROSION TEST ASSEMBLY L.C. *-s11 !101 H11-Sll 342 HY-Sii 821

• -sw s10 HB-23-24" INTAkE STRUCTURE

' SERVICE WATER I ( LOCAL LI I I 301 I._ ) SCREEN WASH SYSTEM 9509120015 -30 /' PUMPS @C-13 HB-23-12" ... f A/S STAND-BY PUMP START ----*--> Q) 0 ---g SW 332/ P 7 c PACKING / _-__ LEAl<-OFF / TO FLOOR / / / / / HY-SW 662 820 Also A vaULPib!c on Aperture Card A/S SCREEN WASH PUMP P-4 THIS DWG. IS ISi COLOR CODED. CONSUMERS POWER COMPANY E 0 c B P.tL.16AE:JES N.J:U:AR FUNT tmm::;iii*t---;:p:-:-1:::P::1 A UC RE, M213 H G F E D c B A *: LOCAL ALARM <lll'll 0 CP-'165 H-C!l07 IG-ll I I I I I K-!'I VENT <E-ll HS 1415 & 1452 lTHIS DllG. I T-24 16 SH 1 <G-8> T-40 H216 SH l I0-81 te-5-2' 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 l !---F0-402 .. HS 1419 & 1453 <THIS DllOI HS 5353B & 1417 lM-2161 HS 1416 lM-2151 HS 5600B <M-6551 DIESEL OIL STORAGE TANK Em @-----@ START-STOP COMPRESSOR t t 114881 INSIDE DIESEL 8 7 T-24 H-216 SH 1 <F-91 T-40 H-216 SH 1 C0-7) SYSTEM T-3'1 H-655 <H-8> T-26 ./1!, M-215 <E-91 t' --FO 111 :. HB-5-2' LO JB-24-1 :. y , .... LO *.. 0130-0E I + I .. y y *THI'S 1?twa.) .... _.,_,__,._,._ <THrd 8 Cwo> I -0130-0E I + P18A lTHIS D\IGI I . A ,. __ j I'. -1 EL. 615' LO te-5-2' 2' -01'1-DE 102 EMERGENCY PIESEL GENERATORS PAY TANKS LOCAL YLL K-SB OIL RESERVE lSEE TYPICALI ENG AIR HANIFOLO PRESSU>E IEAMP-21 I I JACKET WATER COOLER JACKET WATER SYSTEM

• '++'++'++'+'+'+

TO MAIN BEARINGS ********* , TO PISTONS 12 PER BEARING!,. + INTERNAL SUMP QETAIL '8' 4 FILL 3 LOCAL <3014) i BECHTEL ---+. PRE LUBE TIMER JB-23-20" TO INJECTORS ttttttt!t 2 RETURN* +++++++1+ TO INJECTORS DETAIL 'A' FUEL OIL SYSTEM ml.ES.; I. PLATE WITH "6" HOLE 2* RUNS ATOP DIESEL HOUSING. 3. TWO TU>BO CHARGER INLET LltES ANO OUTLET LINE TO DIESEL ARE FLEXIBLE CONNECTIONS. 46 '44 THIS DWG. IS ISI COi.OR CODED. REVISEO OFF SHEET REF. !CONDENSATE MAKE-UP TO T*BI. M220 SH. I. 0*81 PER OR F-CG-'12-021 & OCR '14-102" DESCRIPTION HMW-BY' CIC WP CONSUMERS POWER COMPANY H G F E D c B H214-J.OGN A 2 PIPING & INSTRUMENT DIAGRAM LUBE OIL, FUEL OIL & DIESEL GENERATOR SYSTEMS*-* M214 1 -"*** 46 8 7 .... _ H G Ii" F E 0 c I I I--_l l_J I I_ _______ _ B CG*41 JBD*IS*B" A . <l\o 8 7 8'f VDCJOft . JB-U-1* JBD*lq..12" -+ 6 .JB-11-r M¥"Jl'""Y H-6" IG-6> JB*U*IB" THIS OVG. -** THIS DvG. 5 CONDENSATE STORAOE TAN< t£AT EXCHAOOE;R E*27 CONQENSATE .Il!e!l2EB ..IatlS. 4 3 SEE NOTE 2 CARD* i Also Avsi!abtQ on Aperture Ca1*d 1* 115& YA .. PRIMARY SYSTEM MAKE*UP STORAGE TAN< HEAT EXCHANGER E*2S DO!£STIC WAJtR STORAGE TANK HEAT EX&t'@iGER E*2' 3 I __ H G F E 0 c 09120015 B .tllliiL PS.ID PLANT HEATING SYSTEM M-215 2 1 H G F E 0 c B (C-1231 <1854) * / fTiC\*c-124 <C-1231 B ............ --..... _f.!ct " '(V <1854-Bl "r"Jh* C1854-Cl H2 RECOMBINER <INS IDE CONTA INMENTl M-sqs Is rDENT I CAL EXCEPT TAG NUMBERS IN PARENTHESIS FRC-2316 M-223 <E-Sl

• FLASH TANK & SLOWDOWN TANK M-226 SH. l HB-18-12" IH-Sl STEAM JET AIR EJECTOR M-206 SH lC CF-6) STACK GAS HB-13-1' HC-43_ 11 <B-1l I __ INTAKE FROM COMP COOL ING 5q0* ELEV. CONTAINMENT PENETRATION COOLING SEE M-201 CH-ll SEE M-205 SH. 1 MSIV' S RADWASTE & FUEL HANDLING AREA AUX. BLDG. ADO. RADWASTE & FUEL HANDLING AREA AUX. BLOG. ADO. <t '° .. " .... ' "' >* " CONTAINMENT ISOLATION ANO SAFETY INJECTION SIGNAL 2/4 LOGIC I I I I I I I I I I I I 11 L 1== II_ I I I I_ A/S s ,,; iD .!. 0: UJ UJ "' ; 8 "' "" "' ,_ z UJ > "' "' N' ... :>: VENTS FROM CLEAN WASTE RECEIVER TANKS <F-11 (MV-WG-533) 10-61 HB-lG-2" j HB-1-1 112' HB-1-1 112* HB-28-1 112' HB-39-2" HB-1-2' 8 TEMP DRAIN VALVE <Dl CONTROLLED CHEM LAB. M-21121 SH.1 TANK VENTS CH-8l M-21121 SH. 18 RADWASTE DEMINERALIZER VENTS <E-7l M-211-l FIL TEAED WASTE MONITOR 10-ZI M-210 SH.1C CE-7l M-20'=1 SH. 3 IH-3l M-202 SH. IA <A-5l M-202 SH. 1A CA-7l M-655 SH IA CD-7l M-222 SH. lA <E-4l M-202 SH. lA CH-5l TANK VENTS TREATED WASTE MONITOR TANK VENTS COMPONENT COOLING SURGE TANK VENT CHARGING PUMPS WELLS VENTS CHARGING PUMPS SEAL LUBE TANKS VENTS YCT Hz S. N2 SUPPLY RELIEF 7 "':C ol:I U);:::::; N !61 -. > " "' c "'"'-' "'"'"' "'"' C<tX fE "' 506 508 6 JBB-2-8 1 JBB-1-8 1 .E!fil...

HANDLING ilBf.B_ f I TO PENE. 17A. SEE I DETAILS ON M232-SH. 2 PENE. 52A. [£ill I -r=-$ --------/ (4-N2gM2DR-VA 100) 5 LPIR-0382 <M 211 SH. 1 CF 7) ----4 JBD-82-8u JBD-81-sn STEAM >- E-508 RIA-1817 <M-223 SH 1A< CONTAINMENT ISOLATION RADIATION MONITORS SEE M-223 . . *' ,,-( ,,-( STEAM GENERATOR E-50A I 71 CONTAINMENT AIR COOLER RECIRCULATION I I I J ___ _ I TE 1812 1 r;:r-Also Av3Habie on -Aperture Cerd TE 1813 IRLT FILL LINE --9-LPIR-0383 --)> M 211SH.1> IF-7) --9-LPIR-0382 -f I -04 8 1 46 -----3F M-211 SH.1> I !F-7l - NOTES: 1. ALL VALVES ARE 1 VA' PREFl)( UNLESS OTHERWISE NOTED. 2. DAMPER P0-1546 HAS ITS POSIT!ONERS REMOVED AND IS LOCKED PER SC-87-322. THIS DWG. IS ISI COLOR CODED 3 2 SUPPLY THIS DRAWING IS !SI COLOR CODED CIUllN P.G, BROWN PALISADES PLANT CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM HTG. VENT. & AIR COND. CONTAINMENT BUILDING M-218 SH.2 MW M218-2. DGN 1 H F E 0 c B A H G F E TRAIN "A" TRAIN "B" D c B A ... R i--1 I I MER ROOF 7 l

• CNOM. O.D.) I I I I I __ I I I M "' D <D JDJ-5-1 I JOJ* 12-1 I I I I I I I §I <H-2> L ___ _ 1' <NOM .. 0.0.) I I J I I I I I I I I I L 7 NOTE £:? ..., D ..., <F-2 l 6 JDJ-5-1 I I ( NOM. o.o. l JOJ-6-' 2 ____ ____JJN....!.!._

OUTLINE _J JOJ-12-1 ' I <MON. o.o.) '----r-1-- l 2 * -Mv-s20-vAs-----<1G-" FIL TEA I DRYER -----J-N a-0 ___ J JOJ-4-JDJ-13-I 6 JOJ-1-2 JDJ-1-2

• 4 I I I? 3 TWO-TWO STAGE TROUBLE JI -------------

INDICATOR Ir-----_________ _ I __________ J IG87 _ -111---__ I ! t;-;,-o;;;-J-'--{::ofq--7"--S-U-CT_l_ON---l----;'1-i I FILTER VALVE I I : ' "-MV-23121-VAS i _: --------<M-208 SH.I< ____ _J ! <F-6J * -MV-606-VAS L-------------<M-208 SH.1 < REFRIGERATION CONDENSING UNIT VC-10 I ,-------CE-Gl 1 1 osA srnuENCER rl I 1. .r::J§J < C-11 Al I _ _j _ _J_J_ -@ -<_M ___ 2_l_B_S_H_6_<_ J-1089 N +/- i;i COMPRESSOR CRANKCASE OIL SYSTEM CH-4) TWO-TWO STAGE I __________ J 1' OBA SEOUENCERf-1 I I IHsl <C-llA J I _j _ _J_J_ I I ____ J COMPRESSOR CRANKCASE OIL SYSTEM -<M-218 SH s< (C-4> \[§7 J-112188 2 l. INDICATES COMPRESSOR TRlPS ON HIGH DISCH. PRESS. LOW OIL PRESSURE, OR LOW SUCTION PRESSURE. ALSO INDICATES HIGH VIBRATION S. HIGH SERVICE WATER PRESSURE DROP ACROSS THE CONDENSER. INDICATES HIGH OIL TEMPERATURE TRIP AS WELL. 2, SET AT 350 PSIG. 3. UNIT STARTS S. LOADS ONE CYLINDER Q TEMPERATURE AS DETERMINED BY PLANT (TB). LOADS THIRD S. FOURTH CYLINDER Q TB*S' +/- l" l 4. LOADS SECOND CYLINDER Q TB*3" F( +/- l" ) 5. FOR SERVICE WATER PIPING CONNECTIONS REFER TO SHEET M-208< 0 >. 6. THE FOLLOWING LISTED ITEMS ARE NON-0* PI-1678, PI-1688. VS-16':::H21. PI-1676. PI-1877. PI-1687. VS-1689. PI-1675. 7. REFERENCE BY ELLIS S. WATTS DRAWING 41F35 <BECHTEL DRAWING 12447-54-M-91-66).

8. BACKSEATING OF THE FOLLOWING VALVES WILL ISOLATE COMPRESSOR FAULT PROTECTION1 MV-233-VAS MV-230-VAS MV-234-VAS MV-231-VAS

9. CLOSING OF "*MV-613-VAS AND * -MV-606-VAS WILL PREVENT COMPRESSOR SHUT OO'WN AFTER THE AUTO PUMP* DOWN CYCLE 1121. ALL LOCALLY MOUNTED INSTRUMENTATION SHOWN IS MOUNTED IN Jl12188 OR J1089 UNLESS OTHERWISE SPECIFIED.

9509120015 PALI SADES PLANT CONSUMERS POWER COMPANY M2tB*7,0GN 1 I THIS DWG. IS !SI COLOR CODED. PIPING & INSTRUMENT DIAGRAM HTG. VENT. & AIR COND. CONTROL ROOM 0950 M-218 SH.7 --j 6 _J

• JOJ-11-3/8 I 5 4 3 2 H F E D c A H G F E D c B A .. 8 PRESSURIZER VAPOR PHASE 8 <}i-219 SH. <C-8> LC <M-21g SH. IC-Bl PRESSURIZER LIDUID PHASE <SURGE LINE> I PRIMARY COOLANT LOOP 2 HT LEG I L -------1lr -L I QUENCH TANK LIQUID PHASE -Ir---O>M-219 SH. 2> CC-Bl A/S OUINCH TANK VAPOR PHASE -< r ---+ 5 -, g N ' PURIFICATION ION EXCH. INLET v-1 PURIFICATION FIL TEAS OUTLET LOW PRESSURE INJECTION PUMPS DISCHARGE sx 3336 --+/--e N u "' V-5 PURIFICATION ION EXCH. OUTLET SI TANI< DRAIN CONTAINMENT SPRAY PUMPS DICHARGE -V--U "' HC-42-1 4u v-e GC-14-V.P e ' SIRW TANK RECIRC v-13 ----Ir---------

--1 r -------o; " q -N210M3 I 1-I I I I HC-42-1,4 1 I .c J .l .l --;>* .,.. ..,.--1 'r' :------------t

---

4---------:1---L-----o--!6.

__________ , l ! I I I i

---

' : .;f' j I HC-42-1/.1 V-24 SAMPLE COOLER (r"j\ i 1 J SAMPLE ;=-=h !_:._l-, iL 1 1 , 4 1 _j : l i COOLERS i i : i i t j i i r I i-i 1-----1 i--1 I l_ __________________________________________________________

'---u l---u V-21 V-58 L _____ _ I CC-16-l n 7 MV-PC-161 V-57 MY-PCS-100 V-27 ----_J V-28 I NSSS DEGAS PAT. H r----_I 6 A,3 V-32 MV-PCS-10g V-31 V-37 -, I V-35 V-33 PASM PANEL M-21G SH. (8-Sl TO BURET V-36 V-34 BOMB BY-PASS SX-2022 TO BOMB INLET V-3G BOMB OUTLETS <PASSJ M-211SH.1 5 Ln $: CSI ;i; <iJ " CE-51 M-21q SH. 2 <B-7> 5 NSSS SAMPLING STATION C-32 4 V-40 V-17 V-2 SAMPLE SINK V;-4 u "' V-6 POST ACCIDENT SAMPLE MONITORING SYSTEM PANEL DRAIN V-8 M-21g SH. 2 CB-Bl HC-5-1%1 1'!ll.IEL

1. Hs-1g01 IS 6 POSITION SELESTOR SWITCH GE TYPE 581. 2. THE DESIGN TEMPERATURE RATING rs 11Zl5' F AT THE OUTLET OF THE SAMPLE COOLERS PER FC-676. 9509120015-3 I 1 g v-10 V-12 v-14 CONTROLLED CHEM. LAB. DRAIN TANK HC-5-1%1 M-210 SH. 1 CH-Bl 2 HIGH PRESSURE INJECT ION PUMPS DISCHARGE (s;\ J -!--v-15 .. ' ----V-4G --, 0 Avanab!e on perture Card s IG14 i-168 I -_ [ 2 -168 s I _I I v-16 V-5121 V-51 16 GRAB SAMPLE CTYP. l / SAMPLE SINK FLUSH RING THIS DWG. IS ISI COLOR COOED. 1-22 " ADDED CK VLV

• PER OCR C!3-043 ITT PALISADES PLANT CONSUMERS POWER COMPANY"'"--" H G E D; -1 I PIPING & INSTRUMENTATION DIAGRAM A PROCESS SAMPLING SYSTEM 1 H G F "' "' "' __, :>: w 0 8 E D c A 8 CONTROLLED INSTRUMENT SHOP & CHEM. LABS. ROOM 112 PRIMARY SYSTEM MAKE-UP STORAGE TANK T-q0 M-652 <G-8> (\'2"-PMUl43)

D +/- --1D<<L'::J-'-H"'C"'D'--q-'2,,_--' 1..._*---1 c: (Z?"-PMU141) LHC0-Cj2-Y-2' w (:z"-PMU142)

"'"' i= :J SHIELD COOLING SURGE TANK M-221 SH.1 <0-1) DIESEL GEN. JACKET WATER MAKE-UP TANK M-214 SH.l' CC-2l NENT WATER TANK SH.3 CONDENSATE REC TK T-20 M-215 <D-8) CHEM ADD TANKS T-15. T-16 & T-lgA M-220 SH.2 <D-5) HB-26-l' CHEM ADO TANKS T-198 & T-19C M-220 SH.2 tC-3) HB-26-l" <T w I HB-26-1' HB-26-2' JB-11-2' JB HB-26-2" HB-26-1' CD710 ... I 8 CD167 HB CDl68 LC ! ! : ! 1 I I i <Vl CVl l ________ _ 7 PRIMARY SYSTEM MAKE-UP STORAGE TANK HEAT EXCHANGER 7 UJ N ;?-. "f V) LP u e :>:: 1-

4 L ---t:- HBD-33-2' CVl DMW1686 CVl u,,. I DMW1677 CDC SYSTEM I < M-904 <E (8-5) BACKWASH WATER PUMPS P-926 A & 8 M-919 I DMW 631 DMW 458 C-203 L --i I C-203 fTiS\_ EK L ...., L !LI;\_ EK R020L I PEMJNERALIZEP WATER STORAGE TANK I HEAT EXCHANGER ..E.=..':liiU ! I : L --i ;--------------* I i rt+ i ANS1bEC APERTUR CAflD Also on AperturG C!ud I DMW1341 DEMINERALIZED WATER RECIRCULATION PUMPS A/S OMW 632 WATER TANI< T-918 M-912 <D-5) *--"o"M"w"'2"0"4"'q" HCD-2' CE-6)

• NOTE: CIA 202121 INTERLOCKS SV2008 AND SV2010 ON HIGH CONOUCTIVlTY.

<E-8) HEATING BOILER FEED WATER PUMPS P20A S. 8 THIS DWG. IS !SI COLOR CODED . 3 2 ALB NONE aUlGNm E. P. LET)(E iwoWN )(. A. DELELL[S PALISADES PLANT M220-l. DGN CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM MAKE-UP DOMESTIC WATER & CHEMICAL INJECTION SYSTEMS 0"l50 M-220 1 , 52 I H' G F E Di c B A .. * .... H F E D OVERFLOW HYDRAZINE c B A / WELL "1 CBUFFALO BOXl RV 044 <LEFT HANO VALVE> ACP---t-JB ;:;i;;-OW171 -1 r------------------------------------------------------>----4 I I I I PS 2013 OPENS & CLOSES I MO 2013 & STARTS & STOPS I WELL PUMP P-66 I I LED I ,-I LS-200c:i cp & T I ADDITION PUMPS I FS 2030 f..i -JB-a-2Y,z* 1 a. . HYDRO PNEUMATIC ACCUMULATER I=l lY, "-DWZ39 COLO CHEM LAB JDG-801-2' SHOWER S. EYEWASH DOMESTIC WATER TO CONDENSATE Sc MAKE-UP OEMIN. BUILDING DOMESTIC WATER STORAGE TANK (0)521) EMERGENCY SHOWER & EYE WASH Yz" CD> 1"-FW228 1'-FW275 EBO ! GBD !'-FW335 '" i T GBD SUPPORT BUILDING .liB..I.EB. I.fl..!::iK I..=.ll9.. 11 11 v !'-FW342 EBO ! GBO SUPPORT BUILDING 1'-FW268 l"XY.2" P-15A .. 1'-FW330 ,.. :': EBO 1'-FW341 t GBO EBD-14-Y.!:' T l'-FW338 l'Xl1a' GBO STEAM GENERATOR BLOW DOWN M-226 SH. 1B m-31 1'-FW339 STM GEN N0.1 CE-50Al M-207 SH. lA (G-71 OEMIN. INLET M-220 SH. l ([-Bl ! I I ----------------' 1"-FW227 1'-FW331 EBD-13-1' _J_EBD tEB STM GEN N0.2 CE-508> M-207 SH. lA <G-4l AUX FEEDWATER M-207 SH.1A DEMIN WATER M-220 SH. 1 <C-8l !H-3l VENT

., 2010 6 L.., I I OVERFLOW ! -----------/ I I L ---------

-1 I I I ----, L --------- .... SERVICE BLDG DOMESTIC 'WATER SYSREM M-959 SH. 9 <0-7l tJlif HYDROPNEUMATIC TANK CONTROL PANEL IN SERVICE BLDG CHEMICAL ADDITION SYSTEM THIS DWG. IS ISI COLOR COOED. Also on Aperturn NONE CJmfJf'GJ E. P. LETKE PALISADES PLANT M.220-2.ooN CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM MAKE-UP DOMESTIC WATER & CHEMICAL INJECTION SYSTEMS 0'l50 M-220 SH.2 @ ,.. y '" CHEMICAL ADDITION SYSTEM ... ,;_,. R 7 5 t:='5 _I 4 3 H F E D c A H G F E D c B A 8 COMPONENT COOLING WATER JB-7-3 1 M-20Cl SH. 1: (8-3l IFS-0g32 fc -IH-31 BLOCK ALARM WHEN CV-0G32 IS CLOSED CVl CV> r.-J r.-J EL N "' tl ' -7 CV> I FS-2104 fc -IE-4l BLOCK ALARM WHEN CV-0CJ34 IS CLOSED NO JB-6-3 1 31 -418-CC 11216 CV> 6 L COOLING COILS EMBEDDED I 1 1 -156-SC IN REACTOR SHIELD <A> --I I I c 18 L COOL ING COILS EMBEDDED I IN REACTOR SHIELD (8) u:r---=r ---ri::u::i::1 "'='-'--I I I JB JB-6-P CTYP FOR 44l 513 Y: I -156-SC 506 SHIELD COOL ING HEAT E><CHANGER N 51 "' "' 'f' "" CD> 1 N N "' w w CD> m COMPONENT COOL ING /:i.JBC-l-31 %' CC-2l] t 30 CC JB JBC $ JB .. 8 7 r;;\ JB-6-3 1 P -130-SC 105 6 FLOOR SLAB COOL ING CO I LS CBl RECIRC. FOR CHEMICAL 3° -257-SC40 l 2q1 C-08 P-77A cs SH IELO COOLING .E.UM..E.s. 5 2 1-156-SC llCJ 404 JB-6-4 1 JB-6-4' ... ' "' 4 3 1 -257-SC 4!213 4 -1Hs-0g32 I tH-8> RV 2107 POS APERTURE Also on Aperture Ctard (*."* 0ci3q >--------.... H-----{;; ___ _,,H"'Bc:_-:26"'-"-"11(,,2:_' --'----f:X'I----'----< l {1}2 1 -223-CD 401) CA-3l 1 1 -130-CD 536A 516) :J 4 1 x3 1 THIS OWG. IS !SI COLOR COOED. 3 2 PALISADES PLANT CONSUMERS POWER COMPANY PIPING ANO INSTRUMENT DIAGRAM SHIELD COOLING SYSTEM 0950 M-221 SH.l -1 H G F E D c B A H G F E 0 COMPONENT COOL lNG I WATER HB-24-uz1n ______ ___! < M-209 SH, 3 ?>-*---* ... -------------------------

• IE-ll c B *, . ! ... *.

.... '.;*" : : Also on C1.ud L .... Sk IMMERS <2> -REACTOR CAVITY I I "' oi>*"" P.G, PALISADES PLANT --648' EL HMW-BROWN I HI G F E 0 c CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM SPENT FUEL POOL COOLING SYSTEM SH. 2 17 H G F E 0 c B A 8 7 BULK N2 SUPPLY NITROGEN BOTTLES I L ______ _ ASDV N2 BACKUP M-212 SH 3 CD-Sl MV N 164 SURFACE SLOWDOWN E-50A M-226 SH 1 (8-7) BOTTOM SLOWDOWN E-50A M-226 SH 1 CH-7> MV MV N ( MV N2 600 ) WASTE GAS SURGE TANK M-211 SH 2 HB-45-Id <F-4) WASTE GAS DECAY TANl<S M-211 SH 3 HB-45-P <A-5> VACUUM DEGAS I FI ER M-21121 SH 2 HB-45-P <H-Sl T-80 EDT M-210 SH l HB-45-lP <0-8> T-60 D'NDT M-211 SH l HB-45-P (0-4) 5 2' 15l MV N MV N 153 1" 5 P -N 208 CONTROLLED INSTRUMENT SHOP CROOM 112l SURFACE SLOWDOWN E-508 M-226 SH 1 <B-61 BOTTOM SLOWDOWN E-508 M-226 SH 1 CH-?l 4 \"2P -130 N 2 1 77 P -N2 1 SC! HB-45-1P -N lg121 3 11-N2 188 HBO-12-1" HB HBO 1'-N2 172 MV-N SYSTEM 18 P-N2704 PRESSURE SETTINGS: PCV-2281 ***** 30PSIG PCV-2282 ***** 30PSIG PCV-2283 ***** 25PSIG HB-43-111 HB-43-1' :.. 2 1'-N2 184 ) I JI -N 183 32 HB-43-2 1 HB-43-P HBD-12-3 I 1*. CLEAN WASTE RECEIVER TANKS M-210 SH 1A <F-1, 2> PRIMARY SYSTEM ORA IN TANK M-210 SH 2 <G-5, 61 SAFETY INJECTION TANKS M-203 SH 1 CH-Bl VOLUME CONTROL TANK SH 1 CH-5l EVAPORATOR M-5qA M-650 SH 1 CD-7> EVAPORATOR M-5CJB M-651 SH 1 C0-2l CHANGED NO VALVE MV-3373 TO NC PER OCR C!5-038. P. G. BROWN HMV PALISADES PLANT M222-t. DGN CONSUMERS POWER COMPANY H G F E 0 c B PIPING & INSTRUMENT DIAGRAM A MISCELLANEOUS GAS SUPPLY SYSTEMS 1 4 .. _3 2 32 H G F E D c B SLOWDOWN TANK VENT M-226 SH 1 lH-Sl CG-5, Gl RR-2301 PT. 2 SERVICE WATER DISCHARGE MONITOR !SEE M-208, F-1l HB-18-12° RE 2320 RR-2300 PT. CJ .-BLOWDOWN TANK VENT MONITOR T-2CJ M-226 SH 1 CF-6> P-CJBB M-226 SH 1 (8-4) STACK 8' MS 171 B-11-1a MS 172 L ,---------- RR-2301 PT. 5 DfSCHARGE STRUCTURE M-653 SH 3 <8-7) HC-43-<8-2) 8 + cw 511 / / CW 107 DISCHARGE STRUCTURE fB-6> COLLECTION TANK '°' CIRCULATING WATER MONITOR 7 HC-43-11 I I J 6 !B D "' -+/- I I STM. GEN. BLOWOOWN TANK T-298 MS 173 MS 804 <VJ l 0 L CHECK OURCE RR-2301 PT. 1 A 4 --+ --RE 0707 I I ________ J TRIP CLOSED VALVES c v-0704. 0738, 073CJ, 0770 & 0771 CKT. I l' STEAM GENERATORS BLOWDOWN MONITOR !SEE M-226 F-5) COMPONENT COOL ING CIRC. PUMPS DISCHARGE > M-20g SH 3 >----_. -------: (C-4l CRW 571 ,------------------ 1 RR-2301 PT. 3 + ,-----I CHECK ,_I/souRcE ,---., TRIP CLOSED VALVE CV-0CJ15 -;i> M-20g SH 3> <H-4> I -1' I I I I 1 _____ L ____________ L __ ! HB-24-1' ! ! ". H 8_24_ 1, I

.---------------------l f--->M-20CJ SH 3 > . ! , I ; "" !Dl COMPONENT COOLING WATER MONITOR I I I I I CAW 570 (OJ r I t RR-2301 PT. 4 * ---f$rH_ CHECK SOURCE J I I I TRIP cv-104q & 1051 -k M-210 SH IC> CRW 5q9 IDl CB-4) 1 RAD\i/ASTE I -r TREATED WASTE MON !TOR PUMPS DISCHARGE HC-2-3" HC-2-311 I D!SCHARGE HEADER f----'-"'--=-=--1---1 M-210 SH 1C t8-6l (8-6) I I _J RADWASTE DISCHARGE MONITOR L----------------------------------------------

5 4 3 ffe\lso on Card l J& H G F E D c B 9 5 0 9 120o"1 5 -L-f THIS DWG. IS ISI COLOR CODED. 2 PALISADES PLANT CONSUMERS POWER COMPANY" 2 ,,_10*00" PIPING & INSTRUMENT DIARGRAM A RADIATION MONITORING AND SAMPLING SYSTEMS @ IZJ951ZJ M-223 SH. 18 _ 7 1 8 7 5 5 4 3 2 1 J,;-* -!.* '-H H G F E 0 c B A REFER. DRAWINGS 950Y328*Ml68 SH 4 FOR SPLASH BOX IN / PLACE ---r---> M-202 SH.lA <G-7,81 VOLUME CONTROL TANK T54 I I I J @i M-211 SH.2 <E-41 WASTE GAS SURGE TANK T67 sx 1113 I I I J J-175 '----------H-C_D ___ 11_0 ___ 1/_4_'__,c____ HCD-113-L/ 4' COMPONENT ___ ti_§-24-1' <F-ll EQUIPMENT DRAIN TANK -210 E-74 ccuci CC120 WG145 TEMPORARY MOO. DO NOT REMOVE CLOUDS M-21g SH.lA CG-2l WASTE GAS DECAY TANI< T-688 sx 1120 HCD-112-3/8" M-219 SH.2A 3/S*-HcD-GL-626-WGS @-175 s 3 _[ SAMPLING PANEL 9so912001s-L{L THIS DWG. rs ISI COLOR CODED SCILE NONE D!IWN P,G, BROWN CONSUMERS POWER COMPANY PALISADES PLANT PIPING & INSTRUMENT DIAGRAM GAS ANALYZING SYSTEMS 0950 M-224 SH.1 22 _______________ fji'24-!.DGN < __

• _ 8_ h. 4 ? *pp, m-*, G F E 0 c B H G 2 F E D c B
• A 8 CONTAINMENT AIR FILTER F818 !-=._ -=---=---=--I I Ir=--=---=-

I I 11 11 11 11 11 11 11 J II ffi-11 1:L-I/ -JJ 1 I -WG 530A RELAY LOGIC C-11A CONTROL PNL. HCC-1!31-1S" NOTE 6 = = ::3F M-219 SH. 2> CG-Bl C-103 l I ET PASM HC-47-7'8 1 CF. G-8) °"-..Y-2 BY CPCO BY COMSlP. LEFT CHANNEL I i 5 RA Ck i CALIBRATION GAS i 9. 5% Hz IN N 2 Ll -1 I I I I L ____________ _ C-103 PASM M-21G SH.2 RIGHT CHANNEL RACK I i i CALIBRATION GAS i 9.5% Hz IN Nz Ll L ________ i--l OPERATING DATA FLOW (CC/MIN> (PSIGl ("'F> 1 28800 I 90 2 150 0 270 *Jt!A ANALYZER CELL LOW TEMP Cl6l 1121% H2 RANGE 20% H2 RANGE 3 40 25 90 250 25 90 I I HTR ON I I e---REAGENT OR CAL GAS LOW PRESSURE _j HTR ON LOW ANAYZER --@---§!IA l I I I L

• Jt61 *Jt61 SAMPLE ZERO *Jt!A *Jt!A SAMPLE ZERO -1 I I ---11 CELL FAILURE I I ASLl I CllA l -;r:::_C-161 I

-1 CONTNMT 1 HIGH HYDROGEN H2 MONT. 1 C161 HIGH* I r-1 i _Jj ' VT.23 I CELL FAILURE I LEFT CHANNEL GAS BOTTLE RACK MG !G28 REAGENT GAS 100% 02 I i i I *2 REAGENT GAS I 100% 1212 '_1 _____________________ j I I I I I I I *Jt61 SPAN *Jt!A SPAN NOTES; 4 5 7G00 I 150 6 500121 " 285 7860 " 285 DESIGN DATA NORMAL RANS I ENT PRESS EMP PRESS TEMP PSIG <"'F> PSIG , .. F) 1 55 2 60 IC11A SAMPLE MOOE LOCAL/REMOTE SEL. SWITCHES 300 55 300 300 l. HEAVY LINES INDICATE PRIMARY SAMPLE PATH. 2. TUBING TO BE SERIES 3121121 STAINLESS STEEL 3. HYDROGEN DETECTOR COMSIP INC. DELPHI SYSTEMS DIVISION MODEL K-111 4. L !NE DOES NOT PASS THRU HOT BOX 5. INSTRUMENT TAGS MARKED WITH AN <*l ARE COMSIP TAG NUMBERS USED FDR COMPARISON W [TH VENDOR DRAWINGS.

6. THREE (3) LOGIC SIGNALS ARE WIRED TO ILLUMINATE THREE (3) LIGHTS ON PANEL C-168. A. SV-2412 A & B & SV-2414A & B OPEN. 8. SV-2424A & B OPEN SV-2412A & 8 & SV-2414A & B & SV-2424A & B CLOSED. H G F l CllA I ANALYZER CELtC162 LOW TEMP L *Jt!A 10% H2 RANGE Also n Aperture Card 20% H2 RANGE DX _ 1------HTR ON HTR ON :

I REAGENT OR CAL GAS LOW PRESSURE LOW ANAYZER --@---11 CONTNMT H2 MONT. I HYDROGEN SYSTEM I : '9"" I L -T ----<>IA

I *Jt62 SAMPLE *Jt!A SAMPLE I _________

j_l I I CONTAINMENT I H 2 I L_ _1_ -----1 I I 242 Cl!A L ___ I I IL I _J_J I I CELL FAILURE C!IA H ASL 2401 2401 __ _ I C162 FAILURE ALARM RESET PUSH BUTTONS *Jt62 *Jt62 ZERO SPAN *Jt1A ZERO *Jt!A SPAN 1 HIGH HYDROGEN C 162 HIGH' *i162 CELL FA I LURE 9509120015-L.13 REAGENT GAS 100% 0 2 i i I *2 REAGENT GAS I 100% 1212 THIS DWG. IS 151 COLOR CODED PALISADES PLANT M224-2. OGN CONSUMERS POWER COMPANY PIPING & INSTRUMENT DIAGRAM FSO-CONTAINMENT HYDROGEN MONITORING SYSTEM SYMBOLDGY1 IZJ951ZJ M-224 SH. 2 14 E:LECTRIC HEAT TRACING D c B A G F E D c B 8 RIA 0707 >M-223 SH. 1a>--7 BOTTOM SLOWDOWN N2 SUPPLY E-508 M-222 SH. 1 C0-5) BOTTOM SLOWDOWN N2 SUPPLY E-50A M-222 SH. 1 m-al .---I TJiwIGS I I (Q-Sl <F-3> I .! -_GHiSl I I A/S S. G. SLOWDOWN E-50A OBB-'=t-4a M-207 SH. l CC-5l qas E§ S. G, BLOW DOWN EB M-207 SH. 1 CC-4l 0_ _ l _ fCHPl C-13L I --e-STM. GEN. RECIRC. E-50A <M-207 SH. IC (0-5J STM. GEN. REC I RC. E-508 M-207 SH. 1 <E-3> NOTES CE-3) I I I D88-10-4P DBB-12-4u i-C 42 CE-31 M21q SH. 1C CH-7l C-13L M-212 SH. l THAU 4 FOR CONTROL 8 7 EB-11-2" -1 <E-Sl RIA-0707 L <M-223 SH. 1a< CG-31 EB-11-2 1 6 CV-0771 CH-71 CV-0770 <D-SJ cv-073q co-71 CV-0738 <C-61 HB-28-2 1 2 1 -MS 164 2 1 -MS 165 5 EQUIPMENT DRAIN TANK T-80 M-210 SH. 1 <A-SJ HBD-26-8° RADIATION MONITOR M-223 SH. 18 CE-8l HB-28-4 1 TI 2626 5 HBD-31-12 1 T-2'=t8 TANK HB0-28-2° \ HB0-24-%1 5 STM. GEN. REC I RC. RETURN M-226 SH. 18 CC-11 HBD-28-10 1 CVl 4 2' -MS 167 I I _L ___ C-105 L A LL HIA-0707 ---.J._ -<fi-223 SH. 18< <G-3l 3 -----t I -<p------- 1 t---1 TD I TJ'wb 5 I MA l N CONDENSER M-206 SH. 18 CH-Bl MAIN CONDENSER 2 M-206 SH. 18 20 cc, 0-4) HB0-50-2n 1 ST 0571 > M-205 SH. 1 :>>-------I CE-Bl ST 056G M-205 SH. 1 CC-Bl HB0-1* HB-17-2n 2° -MS 168 TD 1 CV-0771 (G-71 CV-0770 CO-B> CV-073c::t CD-7> CV-0738 (C-7l _!HS\-105 I Also on Aperture Card 4 RE 0707 RADIATION MONITOR M-223 SH. 18 (E-SJ HBD-27-6 1 STEAM GENERATOR BLOWDOWN PUMPS 9!5"09120015 l 2 LCV 6002 BLOWDO'NN HEAT 3n EXCHANGER M-226 SH. lA CA-7) I 7 DAW 573 ) THIS DWG. IS ISI COLOR CODED. Bl' APP CONSUMERS POWER COMPANY PALJSACIES NUCLEAR PLANT COVERT, MICHlVAN M226-1.DGN PIPING & INSTRUMENT DIAGRAM STEAM GENERATOR BLOWDOWN MODIFICA T!ON I:' M226 42 **, H G F E I I D c _j H G F E D 8 8 7 A__ E:R\/ICE WATER (COOLING TOWER JB JBD DISCHARGE f-f MAKE-UP> JB-t-24 , >M-208 SH.14>----

• --------<G-ll i v ' i SLOWDOWN L __ TANK DISCH, HB-28_6 ,,

C8-4l \ \ -----36"-CW733 SLIDING GATE VALVES --. -A .. 8 7 5 5 COOLING TOWER PUMPS C126 CMUL TIPOINTl = I I I '-SW573 1-r1 ___ _ II L_ II I L I l /2'-SW721 .... *.: .. .'.'>. PROHIBIT START P-3C!B TRIPS MAKE-UP BASIN (TYP OF 2l 36" REDUNDANT SERVICE WATER SUPPLY RADIATION MONITOR SEE M-223 6 -I --1 ---------------------- [__ __ _l L A K E 5 -653 SH. 4 tF-7) 18"-CW101 CNORTH> RECIRCULATION TO INTAKE STRUCTURE 4 t8'-CW102 (SOUTH> 4 CVl 3 z "' 0 0 "' 0 __, "' a: w " cl 8 lJ: Ilea z 6 0 6 al a: w " "' z __, § 1'> CLOSE VALVE ON P-3c:':IB TRIP CLOSE VALVE ON P-39A TRIP 5 C\</736 "" £9 +/- :z "' 0 0 "" 0 al a: w " "' :z __, 0 0 '-' 4' x ., "' '\' !il "' i CW631 2'-CW552 I I I L L_ (Q} THIS DWG. IS ISi COLOR CODED. 2 1 1'-CW567 a: w 0=> w :J:>: § 2 r L ffi MN w " "'"' c.p Z<r "' §z 2"-CW556 II I ?' M-653 SH 2 > CB-Sl ---7> M-653 SH 2 > {A-8l C126 (V} 2'-CW501 INTERLOCK -TO SV-1054 (Q} P-4IZIA (NORTH> I 0-1 REMOVED SEISMIC FLAG PER 22 q4 OCR q4-c:l0'3 HMW CONSUMERS POWER COMPANY PALISADES PLANT PIPING & INSTRUMENT DIAGRAM COOLING TOWER SYSTEM M-653 SH 3 1 22 H G F E D c 8 A}}