NRC 2005-0117, Request for Withholding of Proprietary Information from Public Disclosure

Request for Withholding of Proprietary Information from Public Disclosure

ML052580275
Person / Time
Site:	Point Beach
Issue date:	09/07/2005
From:	Koehl D Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2005-0117
Download: ML052580275 (34)

Text

NMCA Committed to Nuclear Excellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC September 7, 2005 NRC 2005-0117 10 CFR 2.390 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units I and 2 Dockets 50-266 and 50-301 License Nos. DPR-24 and DPR-27 Request for Withholding of Proprietarv Information from Public Disclosure

References:

1. Letter from NMC to NRC dated May 13, 2005 (NRC 2005-0063)

2. Letter from NMC to NRC dated June 9, 2005 (NRC 2005-0072)

3. Letter from NMC to NRC dated June 20, 2005 (NRC 2005-0079)

4. Letter from NMC to NRC dated July 24, 2005 (NRC 2005-0094)

5. Letter from NMC to NRC dated July 1, 2005 (NRC 2005-0084)

In Reference 1, Nuclear Management Company, LLC (NMC), submitted Westinghouse Report, "Assessment of Reactor Vessel Head Drop, Point Beach Unit 1 and Unit 2",

(LTR-RCDA-05-428, Revision 1, dated May 13, 2005). Reference 2 submitted Sargent & Lundy Calculation, "Analysis of Postulated Reactor Head Load Drop onto the Reactor Vessel Flange", dated June 8, 2005, and Automated Engineering Services Corporation Calculation, "Finite Element Analysis of Reactor Vessel Head Drop on Reactor Vessel Flange", dated June 8, 2005. Reference 3 submitted Sargent & Lundy Report, "Analysis of Postulated Reactor Head Load Drop onto the Reactor Vessel Flange", Revision 1, dated June 19, 2005. Reference 4 submitted Sargent & Lundy Report, "Analysis of Postulated Reactor Head Load Drop onto the Reactor Vessel Flange", Revision 3, dated July 22, 2005. Reference 5 submitted Westinghouse letter WEP-05-168, Revision 1, dated June 20, 2005, containing Westinghouse Report, "Point Beach Unit 1, Cycle 29 - Reactor Vessel Upper Closure Head Volume Best-Estimate Mean Fluid Temperature", (LTR-RCDA-05-379, Revision 1). All these submittals contained proprietary information.

Westinghouse Electric Company ("Westinghouse"), the owner of the information, has informed NMC that LTR-RCDA-05-428, Revision 1, which was submitted as a proprietary document in Reference 1, was reclassified by Westinghouse as non-proprietary. Enclosure 1 provides Revision 2 of this document, which is non-proprietary and may be released for public disclosure by the NRC. No changes to the content of this document were made.

6590 Nuclear Road

• Two Rivers, Wisconsin 54241

'P Ct)

Telephone: 920.755.2321

Document Control Desk Page 2 References 2, 3 and 4 contained the same set of certified material test reports (CMTRs) for the Point Beach Unit 2 reactor vessel. These CMTRs contain information that is proprietary to Westinghouse. The specified pages (marked "Proprietary") of the documents containing the CMTRs were initially provided in Reference 2 and are supported by the authorization letter and affidavit submitted therein. Enclosure 2 to this letter provides a non-proprietary version of the CMTRs.

Westinghouse has reissued LTR-RCDA-05-379, Revision 1 (submitted in Reference 5),

in both proprietary and non-proprietary versions. Enclosure 3 provides the reissued Westinghouse reports, "Point Beach Unit 1, Cycle 29 - Reactor Vessel Upper Closure Head Volume Best-Estimate Mean Fluid Temperature", dated August 30, 2005, (Proprietary) and "Point Beach Unit 1, Cycle 29 - Reactor Vessel Upper Closure Head Volume Best-Estimate Mean Fluid Temperature", dated August 30, 2005 (Non-proprietary). Also provided in Enclosure 3 are a Westinghouse authorization letter, accompanying affidavit, Proprietary Information Notice and Copyright Notice for the reissued reports.

Since the document listed above as Proprietary contains information proprietary to Westinghouse Electric Company, it is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may. be withheld from public disclosure by the Commission and addresses with specificity, for each, the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR 2.390.

Correspondence with respect to the copyright or proprietary aspects of the above documents, or the supporting Westinghouse affidavit, should reference the appropriate authorization letter (CAW-05-2050) and be addressed to B. F. Maurer, Acting Manager of Regulatory Compliance and Plant Licensing, Westinghouse Electric Company, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

This letter contains no new commitments and no revisions to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 7, 2005.

Dennis L. Koehl Site Vice-President, Point Beach Nuclear Plant Nuclear Management Company, LLC

Document Control Desk Page 3 cc:

Project Manager, Point Beach Nuclear Plant, USNRC Regional Administrator, Region 111, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC

ENCLOSURE 1 WESTINGHOUSE REPORT, "ASSESSMENT OF REACTOR VESSEL HEAD DROP, POINT BEACH UNIT I AND UNIT 2",

(LTR-RCDA-05-428, REVISION 2, DATED SEPTEMBER 1, 2005)

(NON-PROPRIETARY)

(16 pages follow)

Westinghouse Non-Proprietary Class 3 g Westinghouse To: Kerry B Hanahan Date: 9/1/2005 cc: David H. Roarty John Ghergurovich From: Reactor Component Design & Analysis Your ref:

Ext: 412-374-6253 Our ref: LTR-RCDA-05-428, Rev. 2 Fax: 412-374-6647

Subject:

Assessment of Reactor Vessel Head Drop, Point Beach Unit 1 and Unit 2 Ref. 1 WEP-82-584, 'Wisconsin Electric Power Company, Point Beach Nuclear Plant, Reactor Vessel Head Drop Analysis," W. J. Johnson, November 15, 1982, (Westinghouse Proprietary).

Ref. 2 CN-RCDA-05-46, Rev. 2, "Comparison of Original and Replacement Head Drops for Point Beach Unit 1 and Unit 2," George J. Demetri, May 2005, (Westinghouse Proprietary).

At the request of Mr. Joe McNamara, Point Beach Engineering, Westinghouse's Mr. David Roarty recently performed an analysis of the Point Beach Reactor Vessel Head Drop Analysis (Ref. 1) to confirm its accuracy and adjust for the different weights of the new RRVH/HAUP package. Mr. Roarty confirmed the correctness of the analysis and, as expected, the results of that analysis showed similar analytical results with the new heavier RRVH/HAUP package (Ref.

2). Based upon continued discussion and dialog between Westinghouse and Nuclear Management Company, Westinghouse contracted with Dr. William LaPay to perform an additional review of the 1982 analysis that would consider if inelastic structural behavior would change the conclusions reached in 1982.

Attached is Dr. Lapay's white paper response to that request. This has been reviewed by the undersigned and we agree with the technical judgments expressed by Dr. LaPay. Dr. LaPay's resume is also attached for your reference.

The purpose of Revision 2 was to re-classify this document as non-proprietary. No changes to the content were made. Revision 3 corrected the reference letter number in the attachment.

David H. Roarty' Reactor Component Design & Analysis Attachments Official Record Electronically Approved in EDMS 2000.

Page 1 of 16 A BNFL Group company

Westinghouse Non-Proprietary Class 3 Assessment of Reactor Vessel Head Drop Point Beach Unit 1 and Unit 2 May 13, 2005 (reclassified 9/1/2005)

William S. LaPay, PhD Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355

© 2005 Westinghouse Electric Company LLC All Rights Reserved LTR-RCDA-05-428 Rev. 2 Page 2 of 16

1.0 PURPOSE...........................................................................

4

2.0 REFERENCES

4 3.0 SUNMNIARY OF REVIEW AND CONCLUSIONS............................................................

5 4.0 SIGNIFICANCE OF INELASTIC BEHAVIOR DURING IIEAD DROP............................... 6 4.1.

IMPACTSUPPORTSCENARIOS............................................................

6 4.2.

EVALUATION OF SCENARIOI............................................................

6 4.2.1.

Design Characteristics of Reactor Vessel Vertical Support............................................... 6 4.2.2.

Energy Fonnulation and Evaluation.................................................................................

7 4.3.

EVALUATIONOFSCENARIO2........................................................

10 4.3.1.

Design Characteristics of Primary Coolant Loop Piping................................................. 10 4.3.2.

Energy Fornludationz and Evaluation.............................................................

11 5.0 RECOMMENDED FURTHER ANALYSES...........................................................

16 LTR-RCDA-05-428 Rev. 2 Page 3 of 16

Purpose To review the Westinghouse head drop analyses performed in 1982 to confirm the validity of the postulated sequence of events leading to damage to the attached piping which could potentially prevent the removal of decay heat from the core. If this worst case scenario is determined not to be reasonable, then define what is a realistic postulated sequence of events.

Refcrcnccs

1. Westinghouse Original Point Beach Nuclear Plant Reactor Vessel Drop Analysis.

Westinghouse Letter WEP-82-584, PT-SSD-527, "Wisconsin Electric Power Company Point Beach Nuclear Plant Reactor Vessel Head Drop Analysis," November 15, 1982, and Supporting Documentation. (Westinghouse Proprietary)

2. Westinghouse Calculation Note CN-RCDA-05-46, Rev. 2 "Comparison of Original and Replacement Head Drops for Point Beach Unit 1 and Unit 2," approved 4127/05. (Westinghouse Proprietary)

3. Point Beach Nuclear Plant Drawings, Bechtel Job Number 6118.

a.

Bechtel Drawing C-135, "Containment Structure INT Reinforcing Sections Sh. 2," Rev.

11.

b. Bechtel Drawing C-325, "Containment Structure Biological Shield Liner Plate Penetrations," Rev. 7.

c.

Bechtel Drawing C-326, "Containment Structure Biological Shield Liner Plate," Rev. 3.

d. Bechtel Drawing C-2320, "Reactor Steel Supports," Rev. 4.

e. Bechtel Drawing C-2322, "Location Plans, Major Component Support Structures," Rev.

3.

f.

Bechtel Drawing C-2325, "Containment Structure Biological Shield Liner Plate Penetrations," Rev. 3.

g. Bechtel Drawing C-2326, "Containment Structure Biological Shield Liner Plate, Rev. 2.

4. Smith, J.O., O.M. Sidebottom, Elementary Mechanics of Deformable Bodies, The Macmillan Company, Toronto, Ontario, 1969.

5. Roark, Raymond J., Formulas for Stress and Strain, McGraw-Hill Book Company, Inc., New York, 3 rd Edition, 1954.

6. Bechtel Specification 0852, Job # 6118, "Structural Design Criteria for the Point Beach Nuclear Plant, Revision 2, 1967.

7. WCAP-14439, 'Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the Point Beach Nuclear Plant Units 1 and 2 for the Power Uprate Program," Revision 1, March, 2002.

8. ACI Standard 349, "Code Requirements for Nuclear Safety Related Concrete Structures (ACI 349-85)."

9. Gaylord, Edwin H. Charles N. Gaylord, and James E. Stallmeyer, ed., Structural Engineering Handbook, 4h ed., McGraw-Hill, New York, 1997.

10. Manual of Steel Construction, American Institute of Steel Construction, Inc., 7th Edition, 1970.

LTR-RCDA-05-428 Rev. 2 Page 4 of 16

Summary of Revicew and Conclusions A review of the calculations identified in References I and 2 was performed. The analyses used accepted energy formulations to define the impact loads. The basic conclusion in Reference I is:

The results of the analysis for the Point Beach units show that the reactor vessel nozzle stresses are well within allowables. However, the support loads caused by the imnpact arre greater than the critical buckling load of the reactor vessel support columns. These supports cannot be relied upon to absorb enough of the energy of impact to prevent severe damnage to the safety injection lines attached to the reactor vessel or to the primary coolant loop piping.

The 1982 analysis limited itself to elastic behavior of the structures that are impacted. This is conservative. No attempt was made to reflect inelastic material behavior. There is significant energy loss when inelastic behavior is considered. In order to confirm the validity of the conclusion that the impact of the reactor vessel head can cause severe damage so that the core cannot be cooled, a head drop evaluation needs to be performed considering inelastic structural behavior. An assessment was performed to determine if inelastic structural behavior would change the conclusions reached in 1982.

It was concluded from this assessment that the 1982 conclusion is not realistic, inelastic material behavior will absorb significant energy such that there will be no structural failure that would cause loss of coolant to the core. This evaluation has used energy principles to demonstrate that gross plastic failure of the support columns is not likely to occur. The maximum strain levels, for head assembly drop of approximately 45 feet above the flange through air, is 2%, and below 1% for a 20' drop. From this evaluation, it can be concluded that the columns can absorb significant energy and still remain functional.

However, if the columns do fail, there is still reserve strength from the primary coolant loop piping that will support the reactor vessel. This is the second scenario that is discussed below.

A second drop scenario is evaluated in which the columns and other potential restraining structures are assumed to fail and only the main coolant pipes are available to restrain the vessel from further drop. This is conservative since this scenario assumes that the reactor vessel safety injection piping fails, columns fail, and the concrete below the vessel ring girder fails, without providing energy absorption. In this evaluation plastic hinges are assumed to form in the loop pipes and the resulting displacement of the reactor vessel assembly is determined. For a 50 foot drop through air, a maximum deformation of less than 7 inches is predicted, after the loop pipes contact the bio-shield wall. For a 15' drop the maximum deformation is less than 3 inches. Assuming that the available flow area is reduced by these deformations, the loss in flow area is under 20%. That is for a 50 foot drop through air, about 80% of the flow area remains, and for a drop of 15 feet the remaining flow area is 95% or the original flow area.

This assessment also confirmed that the concrete supporting the steel structures would not fail considering the maximum loads determined.

Therefore it can be concluded that if a potential head drop was to occur (maximum drop height of 50' considered in analysis), the reactor vessel will be adequately supported either by the supports, or by the primary coolant piping such that there will be no damage to the attached primary coolant piping which would prevent the removal of decay heat from the core. Further, the Reactor Vessel integrity is maintained since it is adequately supported so that it doesn't break and leak out water, except for the safety injection lines that potentially may be lost.

LTR-RCDA-05-428 Rev. 2 Page 5 of 16

It is recommended that a detailed analysis be performed for the head drop analysis using inelastic nonlinear methods (see Section 5.0). This analysis will provide the permanent documentation supporting the conclusions reached herein that used valid expert opinion and simplified analyses, and provide the documentation that meet NUREG-0612. This evaluation should consider the following:

• Nonlinear time history analyses using different drop heights

• Finite element model that includes gaps (e.g., Ag)

Actual stress strain curves for the material inelastic behavior for the column support and shear pin, and the hot and cold leg piping.

• The reactor head should impact the vessel with no assumption of inelastic impact considered (i.e.,

head and vessel are one body after impact).

• The model should also address potential rebound of the vessel and vessel head after impact.

Significance of Inelastic Behavior during Head Drop An evaluation is performed to determine if there is sufficient energy loss due to inelastic behavior of the impacted structures to prevent the loss of critical components needed to cool the core.

The following assumptions are made that are consistent with Reference 1.

1. A concentric drop of the reactor vessel head onto the reactor vessel is assumed.

2. The vessel head is assumed to be rigid and deflection of the head at impact is neglected.

3. The reactor vessel and vessel head remain together after impact.

4. Seventy-five percent of the available mass of the fuel, water, internals, and remainder of the vessel is helping to dissipate the energy of the drop.

5. The head drop is through air.

Impact Support Scenarios Two scenarios are evaluated to determine if the impact load can be adequately supported.

Scenario 1: The Reactor Support Columns, Shear Pins, and Concrete Floor supports maintain their structural integrity during impact.

Scenario 2: The Reactor Support Columns are lost during impact along with the safety injection lines.

The reactor is supported by the Primary Coolant Loop Piping Evaluation of Scenario I Design Characteristics of Reactor Vessel Vertical Support Reactor Vessel Vertical Column Design Characteristics per Reference 1:

Six columns support Reactor Columns are 12" Sch 120 pipe Area = 36.9 in2 Radius of Gyration = r = 4.17 in LTR-RCDA-05-428 Rev. 2 Page 6of 16

• Material A53 Yield Stress = Fy = 30 ksi (assumes type/grade E/A, F, S/A)

E = 3OE3 ksi Column Length = L = 229"

• Pin-Fixed Boundary Condition (K = 0.7)

KL / r = 38.4 < 40; Region of compressive yield and not buckling

• Capacity based on compressive yield = Pc = 36.9 x 30 = 1107 kips

• Capacity of all six Columns = PT = 6 x Pc = 6,642 kips The original load capacity of the columns was reported to be 1157 kips. The difference between Pc and the value of 1157 kips is due to the use of a lower yield stress (30 ksi is used in this evaluation, and 35 ksi was previously used), and it is assumed herein that the column experiences compressive yield failure and not buckling.

Reactor Vessel Column Shear Pin per Reference 1:

• 4" diameter pin Area = 12.56 in2

• Shear Stress = = 32 ksi (working stress)

Faulted Capacity = 1.7 x 32 = 54.4 ksi

• Shear pin support is double shear Shear Pin Capacity = Pv = 2 x 12.56 x 54.4 = 1367 kips Since Pv is larger than Pc, failure of the column will occur before failure of the pin. The shear pin will not see a higher load than Pc since failure of the column will not have a load greater than that defined by a compressive yield failure. Therefore, the shear pin will support the column load Pc, and there is no need to evaluate the shear pin further.

Concrete Supporting Column Support per Reference 6:

• Concrete Compressive Yield Stress = fc' = 4,000 psi Energy Formtulation and Evaluation From the design characteristics given in Section 4.2.1 it is seen that column will fail before the shear pin, therefore, the column is the critical component in the reactor vessel vertical support structure. The concrete under the support will be evaluated separately.

The column is made of ductile steel, and is assumed to fail in an elastic perfectly plastic behavior as shown below. No credit is taken for strain hardening.

LTR-RCDA-05-428 Rev. 2 Page 7 of 16

Cy Cn/

E = cry/

Strain st Figure 1-Stress-Strain Curve The energy loss due to the impact of a mass onto another mass follows References I and 2 that uses energy loss formulations given in Reference 5. This formulation is given below for the energy loss factor K.

A moving body of mass M strikes axially one end of a bar of mass Ml, the other end of which is fixed. There is a body of mass M2 attached to the struck end of the bar. Then MI M2 K

3M M

(1) 2M M)

Weights are used instead of masses. From Reference 2 the following weights are used:

W = 194,000 lb = weight of replacement head drop weight.

WI = 137,600 lb = weight of vessel (from flange to nozzles) plus nozzles W2 = 0.75 x 570,000 lb = 427,000 lb = 0.75 x (weight of vessel plus water in vessel - WI);

based on assumption 4 in Section 4.0.

Solving Equation 1, K = 0.272.

The vertical stiffness (Kv) of the vessel, nozzles, and supports is used. Per Reference 2:

Kv = 17.38E3 kip/in per support.

The energy formulation including elastic-perfectly plastic behavior follows Reference 4. This formulation is given below where h is the height of head drop, and A is the amount of deformation that includes elastic displacement (A,) and total displacement (A) that includes plastic deformation.

W (K x h + A) = I/2 PT A, + PT (A -A)

(2)

Solving Equation (2) for A:

LTR-RCDA-05-428 Rev. 2 Page 8 of 16

A = [(K h W + PT A, /2) / (PT - W)]

(3)

Where, X = PT/ Kv = 6,642 / 17.38E3 = 0.382 (4)

Equation (3) is solved for drop heights above the vessel mating surface (El 40' 7 1/8") from 15' to 50'.

The results are given in Table 1. Also provided in this table is strain (£ = A / L) in the support column for each drop height. As seen from this table the strain in the column is 2% and under for heights as high as 45', and under 1% for heights below 20'. It is noted that dead weight is not considered in this evaluation, but it will not have significant effect on the results.

Table 1 - Strain in Vertical Support Columns from Head Drop h

A Strain feet inches C

50 5.10 0.022 45 4.61 0.020 31.4 3.28 0.014 26.4 2.79 0.012 25 2.65 0.012 20 2.16 0.009 16 1.77 0.008 15 1.67 0.007 Concrete Evaluation The concrete must support the column load Pc (load based on column yield). Failure of the supporting concrete structure will be local bearing failure under a support column. The supporting floor will be more than adequate to support the total load PT.

The evaluation follows Reference 8, paragraph 10.15, Bearing Strength.

Design Bearing Strength on concrete shall not exceed 4(0.85 fc' Al), except as follows.

When the supporting surface is wider on all sides than the loaded area, design bearing strength on the loaded area may be multiplied by A2/A1, but not more than 2.

After a review of the Reference 3 drawings, it is concluded that the supporting surface is wider on all sides than the loaded area. Therefore, the ratio of A2/Aj is taken as 2. The concrete bearing stress (SB) from the load Pc is:

SB=Pc/2A, = lOQOx 1107/(2x484)= 1144 psi LTR-RCDA-05-428 Rev. 2 Page 9 of 16

Column base plate is 22"' x 22" (Reference 3d)

Al = 22 x 22 = 484 in2 Allowable bearing stress = 0.85 c fc' = 0.85 x 0.7 x 4000 = 2380 psi Note, per Reference 8, ¢ = 0.7 for bearing.

The concrete is acceptable since SB < 2380 psi.

Evaluation of Scenario 2 This scenario is based on the assumption that the columns fail and that only the loop pipes are available to carry the drop load. This scenario also assumes that the direct vessel safety injection piping fails and the concrete below the vessel ring girder fails, without providing energy absorption.

Design Characteristics of Prinary Coolant Loop Piping Hot leg Per Reference 1:

Outside Diameter = OD = 34" Inside Diameter = ID = 29" Thickness = t = 2.5" Area = A = 7t (OD2 - ID2)/4 = 247.4 in-Shear Area = A, = 247.2/2 = 123.7 in2 Inertia = I = nt (OD4 - ID 4)/64 = 30,879 in4 Per Reference 7:

• Material A376, Type 316 (ASME Material List Yield Stress = 30 ksi, room temperature)

Note that per Reference 7 the actual material property is higher than minimum yield stress of 30 ksi.

Unit I yield stress = 32,300 psi Unit 2 yield stress = 41,800 psi The minimum yield stress value is used.

The modulus of elasticity at room temperature is 28.3E3 ksi.

Cold leg Per Reference 1:

Outside Diameter = OD = 32.25" Inside Diameter = ID = 27.5" Thickness = t = 2.375" Area = n (OD 2 - ID 2)/4 = 222.9 in2

• Shear Area = 222.9/2 = 111.45 in2 Inertia = I = n (OD4 - ID 4)/64 = 25,026 in4 LTR-RCDA-05-428 Rev. 2 Page 10 of 16

Per Reference 7:

• Material A376, Type 316 (ASME Material List Yield Stress = 30 ksi, room temperature)

Note that per Reference 7 the actual material property is higher than minimum yield stress of 30 Wsi.

Unit I yield stress = 31,700 psi Unit 2 yield stress = 47,000 psi The minimum yield stress value is used.

The modulus of elasticity at room temperature is 28.3E3 ksi.

Concrete Supporting Column Support per Reference 6:

• Concrete Compressive Yield Stress = fc' = 4,000 psi Energy Formulation and Evaluation It must be determined if the pipes fail in shear or by the formation of a plastic hinge. It is assumed that the primary coolant loop piping is supported by the shield wall. The piping is assumed to support the reactor vessel as a cantilever beam. The length of the cantilever beam is assumed to be the distance from the nozzle safe-end to approximately half the distance through the biological shield wall.

I -

-,d See Below RPV

/

Resisting Load Pp Pipe with Effective Cantilever Length of 4' Figure 2 - Piping Modeled as a 4' Cantilever The plastic moment, when in the flat plateau shown in Figure I with a stress equal to ay, is defined by Equation (5), per Reference 9, Section 9-1:

MP = (TY Z (5)

where, LTR-RCDA-05-428 Rev. 2 Page 11 of 16

MP = Plastic Bending Moment y= Yield Stress Z = Plastic Section Modulus d

{d Figure 3 - Piping Cross Section Z =(d3 / 6) ( 1 - [1 - (2t/d)]3 )

(6)

Hot Leg From Equation (6), the plastic section modulus for the hot leg is:

Z = (343 / 6) {1 - [I - (2 x 2.5/ 34)]3)

Z = 2485.8 in3 Per Equation (5):

Mp = 2485.8 x 30 = 74,575 kip-in Pi = MP / Lp = 74,575 / 48 = 1,554 kips The octahedral shearing stress theory is used to define the shearing yield point.

y= ay/ 3 = 30 / 3= 17.32 ksi P, =

x A, = 17.32 x 123.7 = 2,143 kips Comparing Pp and P, it is obvious that the pipe will experience a plastic bending moment before it yields.

Cold Leg From Equation (6), the plastic section modulus for the cold leg is:

Z = (32.253 / 6) { I - [1 - (2 x 2.375 / 32.25)] 3j Z = 2124.2 in3 Per Equation (5):

Mp = 2124.2 x 30 = 63,726 kip-in Pp = MP / LP = 63,726 /48 = 1,328 kips LTR-RCDA-05-428 Rev. 2 Page 12 of 16

The octahedral shearing stress theory is used to define the shearing yield point.

'y = cry/ 3 = 30/ 3 = 17.32 ksi Pv =

xy X A, = 17.32 x 111.45 = 1,930 kips Comparing Pp and P, it is obvious that the pipe will experience a plastic bending moment before it yields.

Energy Formulation The same energy formulation as used in Section 4.2.2 is used. The pipe is assumed to follow elastic-perfectly plastic behavior with plastic behavior represented by the yield stress. This assumption is not representative of the expected behavior of stainless steel, but is conservative since stainless steel generally does not have a significant flat region of perfectly plastic behavior at its yield stress, but has higher capacity due to strain hardening effect that is not being considered. The elastic displacement (A,) is defined below:

AlPp / K' (7)

K' =3 E I / L3 (8)

For the Hot Leg:

K' = 3 x 28.3E3 x 30,879 / 483 = 23,705 kip/in 4 = 1,554 / 23,705 = 0.0655" For the Cold Leg:

K' = 3 x 28.3E3 x 25,026 / 483 = 19,212 kip/in A,= 1,328 / 19,212= 0.0691" As seen above, the elastic displacement is similar for both the hot leg and cold leg; therefore, the average value is used in this evaluation.

X = (0.0655 + 0.0691)/2 = 0.0673" The same K value as defined by equation (1) is used, K = 0.272.

There is a 6.5 inch gap, Ag, between the pipe and the shield wall. It is assumed that the total weight of the reactor head assembly, vessel, nozzles, and all contents are assumed to also drop through this gap that adds to the energy of the head drop.

The work done by the plastic hinge is:

Work = Mp 0 (9)

Assuming small angle deflection, theta is defined as:

LTR-RCDA-05-428 Rev. 2 Page 13of 16

O=AI

(/1L (10)

And the relationship between Mr and Fr is:

MP= PpL Therefore, Work = MP 0 = Pp L A / L = Pp A (I11)

This is the same relationship as used in Section 4.2.2. Therefore, the formulation as given in Equation (2) may be used modified by the added energy caused by the drop through the gap A.. The value of PT in Equation (2) is defined as the sum of all of the Pp values associated with the Hot and Cold legs:

PT = 2 x (1,554 + 1,328) = 5764 kips W(Kx h+A)+ (V+ WI +W2)xA, = '2PTA4+PT(A-A4)

(12)

Solving Equation (12) for A:

A= {[(K hW+PT4/2)+ (W+WI +W2)xAg]/(PT-W)}

(13)

Equation (13) is solved for drop heights above the vessel mating surface (El 40' 7 1/8") from 15' to 50'.

The results are given in Table 2. Considering that the flow area of the pipe is reduced by the deformation, a measure of remaining flow area can be obtained. The formulation using Reference 10, page 6-22, is shown in Figure 4. As seen from this table, at a drop of 50' about 80% of the flow area remains, and for a drop of 15' the remaining flow area is 95% or the original flow area.

l r = ID/2 Remaining Flow Area = AR nt r2 - 0 r2 - r (r-A) Sin 0 Lost Flow Area =

0 r2 - r (r-A) Sin 0 Figure 4 - Remaining Flow Area after Deformation LTR-RCDA-05-428 Rev. 2 Page 14of 16

Table 2 - Remaining Flow Area from Head Drops h

A Hot Leg Cold Leg feet inches AR/It r2 AR/XT r2 50 6.77 0.82 0.81 45 6.20 0.84 0.83 31.4 4.65 0.90 0.89 26.4 4.09 0.91 0.91 25 3.93 0.92 0.91 20 3.36 0.94 0.93 16 2.90 0.95 0.94 15 2.79 0.95 0.95 Concrete Evaluation The Shield Wall concrete must support the pipe loads Pp. Failure of the supporting concrete structure will be local bearing failure under a cold or hot leg pipe at the shield wall interface. The shield wall will be more than adequate to support the total load from the coolant piping.

The same evaluation as done in Section 4.3.1 is used. The evaluation follows Reference 8, paragraph 10.15, Bearing Strength.

The ratio of A2/A, is taken as 2. Al is defined as follows using the pipe OD and shield wall dimensions based on Reference 3 (6' for hot leg and 5' for cold leg):

A, (Hot Leg) = 34 x 6 x 12 = 2448 in2 A, (Cold Leg) = 32.25 x 5 x 12 = 1935 in2 The concrete bearing stress (SB) from the load Pp is:

Hot Leg:

SB = Pp / 2A, = 1000 x 1,554 / (2 x 2448) = 317 psi Cold Leg:

SB = Pp / 2A, = 1000 x 1,328 / (2 x 1935) = 343 psi Although the actual bearing will likely result in local deformation of a portion of the shield wall concrete, this evaluation supports the contention that the shield wall will not grossly fail.

Allowable bearing stress = 0.85 ¢ fc' = 0.85 x 0.7 x 4000 = 2380 psi Note, per Reference 8, ¢ = 0.7 for bearing.

The concrete is acceptable since SB < 2380 psi for both the hot and cold leg piping.

LTR-RCDA-05-428 Rev. 2 Page 15 of 16

Recommended Further Analyses A head drop evaluation needs to be performed considering inelastic structural behavior. This analysis will provide the permanent documentation supporting the conclusions reached herein that used valid expert opinion and simplified analyses. This evaluation should consider the following:

• Nonlinear time history analyses using different drop heights

• Finite element model that includes gaps (e.g., Ag)

• Actual stress strain curves for the material inelastic behavior for the column support and shear pin. and the hot and cold leg piping.

• The reactor head should impact the vessel without the assumption of inelastic impact considered (i.e., head and vessel are one body after impact). Impact shall consider more realistic material behavior and dynamic structural response.

• The model should also address potential rebound of the vessel and vessel head after impact.

LTR-RCDA-05-428 Rev. 2 Page 16of 16

ENCLOSURE 2 POINT BEACH UNIT 2 REACTOR VESSEL CERTIFIED MATERIAL TEST REPORTS (CMTR)

(NON-PROPRIETARY)

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