ML17334B507

Submits Response to NRC 940422 Ltr,Supplying Sets of Summary Tables Containing Pressurized Thermal Shock & Upper Shelf Energy Info for Plant,Per GL 92-01

ML17334B507
Person / Time
Site:	Cook
Issue date:	05/20/1994
From:	Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AEP:NRC:1173C, GL-92-01, GL-92-1, TAC-M83453, TAC-M83454, NUDOCS 9405310207
Download: ML17334B507 (27)

Text

-ACCELERATED DISTRIBUTION DEMONSTION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9405310207 DOC.DATE: 94/05/20 NOTARIZED: NO DOCKET FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH. NAME AUTHOR AFFILIATION FITZPATRICK,E. Indiana Michigan Power Co. (formerly Indiana !!'ichigan Ele RECIP.NAME RECIPIENT AFFILIATION RUSSELL,W.T. Document Control Branch (Document Control Desk)

SUBJECT:

Submits response to NRC 940422 ltr,supplying sets of summary tables containing pressurized thermal shock 6 upper shelf energy info for plant,per GL 92-01. D DISTRIBUTION CODE: A028D TITLE: Generic Letter 92-01 COPIES RECEIVED:LTR Responses ENCL (Reactor Vessel SIZE. $0 P tructural Integrity

~ 1 NOTES:

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RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D "PD3-1 PD 1 1 HICKMAN,J 2 2 INTERNAL: NRR/DE/EMCB 2 2 NRR/DORS/OGCB 1 1 D

NRR/DRPE/PDI-1 1 1 NRR/DRPW 1 1 NUDOCS-ABSTRACT 1 1 OC/LPJ3CB 1 0 OGC/HDS2 1 0 M322h Ol 1 1 RES/DE/MEB 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 D

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NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 15 ENCL 13

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Indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216

à IMEMSlMA SNCMMWN PWM AEP:NRC:1173C Donald C. Cook Nuclear Plant Units 1 and 2 Docket Nos. 50-315 and 50-316 License Nos. DPR-58 and DPR-74 RESPONSE TO REQUEST FOR VERIFICATION OF INFORMATION SUPPLIED FOR GENERIC LETTER 92-01, REVISION 1 (TAC NOS. M83453 and M83454)

U. S, Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555 Attn: W. T. Russell May 20, 1994

References:

1) NRC letter from John B. Hickman to Mr. E. E. Fitzpatrick dated, April 22, 1994.

2) Letter No. AEP:NRC:1173 from Mr. E. E. Fitzpatrick to T. E. Murley (NRC) dated, July 13, 1992.

3) Letter No. AEP:NRC:1173A from Mr. E. E. Fitzpatrick to T. E. Murley (NRC) dated, November 29, 1993.

4) Letter No. AEP:NRC:1173B from Mr. E. E. Fitzpatrick to T. E. Murley (NRC) dated, January 24, 1994.

This letter is submitted in response to your April 22, 1994, letter which supplied two sets of summary tables containing pressurized thermal shock (PTS) and upper shelf energy (USE) information for each of the Donald C. Cook Nuclear Plant Units. Those summary tables were developed by your Staff based on our responses to Generic Letter (GL) 92-01, Revision 1, "Reactor Vessel Structural Integrity", and on previously docketed information. Specifically, in your April 22, 1994 letter, you requested that we verify that the information we supplied in response to GL 92-01, Revision 1, has been accurately entered in the subject summary tables.

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Mr. W. T. Russell - 2 AEP:NRC:1173C As requested, we have reviewed and verified, with the exceptions noted in Enclosure 1 to this letter, the accuracy of the PTS and USE summary tables. For clarity, a copy of the PTS and USE tables, annotated with our comments, is included as Enclosure 2 to this letter.

Sincerely,

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~ Fitzpatrick Vice President dr Enclosures cc: A. A. Blind G. Charnoff J. B. Martin - Region III NFEM Section Chief NRC'Resident Inspector J. R. Padgett

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ENCLOSURE 1 TO AEP:NRC'1173C COMMENTS ON PRESSURIZED THERMAL SHOCK AND UPPER SHELF ENERGY

SUMMARY

TABLES to AEP:NRC:1173C Page 1 Donald C. Cook Unit 1 Summary File for Pressurized Thermal Shock The 32 EFPY/EOL ID neutron fluence of 1.41 x E19 n/cm~ is high for nozzle shell plates B4405-1, -2, and -3, and the nozzle/

intermediate shell axial and circumferential welds 1-442 A/C, and 8-442 respectively since the plates and welds are located above the core affected height. The Donald C. Cook Unit 1 ID neutron fluence at 32 EFPY/EOL for the above mentioned plates and welds has been estimated by Westinghouse at 1.10 x E18 n/cm~. Westinghouse used the 1.1 x E18 n/cm~ fluence value in developing the Donald C. Cook Unit 1 USEs presented in WCAP-13587 "Reactor Vessel Upper Shelf Energy Bounding Evaluation for Westinghouse Pressurized Water Reactors." WCAP-13587 has been reviewed by the NRC Office of Nuclear Reactor Regulation, and is the subject of their safety assessment, dated April 21, 1994 (see Attachment 1 to Enclosure 1).

For the above plates and welds, the value of 1.41 x E19 should be replaced with 1.10 x E18.

2. As identified on page 2-9 of the Donald C. Cook Unit 1 PTS evaluation, dated March 1990 (see Attachment 2 to Enclosure 1), the 32 EFPY/EOL ID neutron fluence for the longitudinal welds 2-442A/C and 3-442A/C is 0.95 x E19 n/cm~. The Unit 1 PTS report was included as Attachment 15 to the Generic Letter 92-01 response, AEP letter No. AEP:NRC:1173 dated July 13, 1992.

3. As identified on page 2-8 of the Donald C. Cook Unit 1 PTS evaluation, dated March 1990 (see Attachment 2 to Enclosure 1), the percent Cu for axial welds 1-442A/C, 2-442A/C, and 3-442A/C and circumferential weld 9-442 is estimated at 0.28 percent. Again, the Unit 1 PTS report was also included as Attachment 15 to the Generic Letter 92-01 response, AEP letter No. AEP:NRC:1173 dated July 13, 1992. The 0.28 percent Cu value may change as additional information becomes available; however, the percent Cu value of .27 should be replaced with .28 for the above welds.

4, The Unirradiated Upper Shelf Energy values are not included in the PTS Table, and as such, the footnote should be removed.

I to AEP:NRC:1173C Page 2 Donald C. Cook Unit 2 Summary File for Pressurized Thermal Shock

5. As identified on page 8 of the Donald C. Cook Unit 2 PTS evaluation dated February 1993 (see Attachment 3 to Enclosure 1), the 32 EFPY/EOL ID neutron fluence for the lower shell plates C5540-2 and C5592-1 is 1.71 x E19 n/cm~. For the above plates, the value of 1.71 x E18 should be replaced with 1.71 x E19. The Unit 2 PTS evaluation was not included with the Generic Letter 92-01 response; however, it was submitted per AEP letter No. AEP:NRC:0561F dated April 12, 1993.

As identified on page 7 of the Donald C. Cook Unit 2 PTS evaluation dated February 1993 (Attachment 3 to Enclosure 1), the percent Ni value for intermediate shell plate C5556-2 is 0.57, the percent Cu value for intermediate shell plate C5521-2 is 0.125, and the percent Ni content for lower shell plate C5592-1 is 0.59. The respective percent Cu and Ni values for these plates should be replaced with the values specified in this paragraph.

D.C. Cook Unit 1 Summary File for Upper Shelf Energy The change in the percent Cu content of the axial and longitudinal welds, identified in Comment No. 3, results in the need for the NRC to re-calculate 1/4T USE at EOL. Using either the 0.27 or 0.28 percent Cu value, the revised 1/4T USE exceeds the minimum USE Energy criteria of 50 FT-Lbs.

The percent Cu values do not appear in the USE Tables, and as such the footnote should be removed.

9. It should be noted that, with the correction of the ID neutron fluence values for the nozzle shell plates and the nozzle/intermediate shell axial and circumferential welds as indicated in comment No.l above, the 1/4T neutron fluence at EOL should also be re-calculated by the NRC. Westinghouse has determined that the 1/4T neutron fluences at EOL for the nozzle shell plates and the nozzle/intermediate shell axial and circumferential welds are 5.77 x E17 n/cm~.

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Hr. William H. Rasin Vice President and Director Technical Division Nuclear Hanagement and Resources Council 1776 Eye Street, N.W., Suite 300 Washington, DC 20006-3706

SUBJECT:

SAFETY ASSESSHENT OF REPORT WCAP-13587, REVISION 1, "Reactor Vessel Upper Shelf Energy Bounding Evaluation for Westinghouse Pressurized Water Reactors," September 1993

Dear Hr. Rasin,

By letter dated Harch 16, 1993, the Nuclear Hanagement and Resources Council (NUHARC) submitted a report prepared by the Westinghouse Electric Corporation for the Westinghouse Owners'roup (WOG) entitled, "Reactor Pressure Vessel Upper Shelf Energy Bounding Evaluation For Westinghouse Pressurized Water Reactors," WCAP-13587, Revision 0.

The report was provided to NRC as supplemental information in response to Generic Letter (GL) 92-01. No review and approval was requested. The report is intended to demonstrate through fracture mechanics analyses that there exist margins of safety against fracture equivalent to those required by Appendix G of ASHE Code Section III for beltline materials having upper shelf energy (USE) values below the screening criteria of 50 ft-lb.

By letter dated July 1, 1993, the NRC issued a request for additional information (RAI) concerning the report. The reply to the RAI was received by NRC on August 20, 1993 and the NRC was notified that the report would be revised based on the RAI. WCAP-13587, Revision 1, was subsequently submitted to NRC through NUHARC by letter dated September 30, 1993.

The Hateri als and Chemical Engineering Branch has completed a safety assessment of Revision 1 of the report. The enclosure contains the safety assessment. The staff finds the methodology employed and the analysis performed for the WOG to be acceptable except as noted in the enclosure. The report did not contain an equivalent margins analysis based on weldment properties because the WOG concluded that all of the beltline welds would be above 50 ft-lb at end of license (EOL). As part of the GL 92-01 review, the staff will be evaluating each beltline weld for the WOG plants to determine the screening criterion will be met at EOL. Individual licensees desiring to if reference WCAP 13587, Revision 1, as the basis for addressing the USE requirements of 10 CFR 50, Appendix G must confirm the specific plant applicability by either demonstrating that all beltline welds will have EOL CONTACT: E. Hackett, 301-504-2751

USE above 50 ft-lb or that the bounding plate used in the report has a lower J-R curve than any other beltline material in the vessel. Further, individual.

licensees must request approval in accordance with 10 CFR 50, Appendix G. The staff also concluded that comparison of applied fracture driving force with material fracture resistance was overly conservative for some of the cases analyzed. This issue. was addressed for a limiting plant specific case by letter dated December 21, 1993, and the WOG is preparing a revision to the report which addresses all of the cases considered.

'f you have any questions, please contact Ed Hackett of my staff at 301-504-2751.

H. Wayne Hodges, Acting Director Division of Engineering Office of Nuclear Reactor Regulation

Enclosure:

As stated

ENCLOSURE SAFETY ASSESSHENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION REACTOR VESSEL UPPER SH F BOUNDING EVALUATION FOR WESTINGHOUSE PRESSURIZED MATER REACTORS MCAP-13587 REVISION 1 1.0 REVIEW SUHHARY The staff has evaluated the equivalent margihs analysis presented in report MCAP-13587, Revision 1, submitted by the Westinghouse Electric Corporation for the Mestinghouse Owners'roup (WOG). The staff finds the methodology employed and the analysis performed for the WOG to be acceptable except as noted herein. The report did not contain an equivalent margins analysis based on weldment properties because the WOG concluded that all of the beltline welds would be above 50 ft-lb at end of license (EOL). As part of the GL 92-01 review, the staff will be evaluating each beltline weld for the WOG plants to determine if the screening criterion will be met at EOL. Individual licensees desiring to reference WCAP 13587, Revision 1, as the basis for addressing the USE requirements of 10 CFR 50, Appendix G must confirm the specific plant applicability by either demonstrating that al.l beltline welds will have EOL USE above 50 ft-lb or that the bounding plate used in the topical report has a lower J-R curve than any other beltline material in the vessel. Further, individual licensees must request approval in accordance with 10 CFR 50, Appendix G. The staff also concluded that comparison of applied fracture driving force with material fracture resistance was overly conservative for some of the cases analyzed. This issue was addressed for a limiting plant specific case by letter dated December 21, 1993 [17], and the MOG is preparing a revision to the report which addresses all of the cases considered.

2.0 INTRODUCTION

By letter dated Harch 16, 1993, the Nuclear Management and Resources Council (NUHARC) submitted a report prepared by the Mestinghouse Electric Corporation for the MOG entitled, "Reactor Pressure Vessel Upper Shelf Energy Bounding Evaluation for Westinghouse Pressurized Water Reactors," WCAP-13587, Revision 0 [1]. The report was provided to NRC as supplemental information in response to GL 92-01 [2]. No review and approval was requested. The report is intended to demonstrate through fracture mechanics analyses that there exist margins of safety against fracture equivalent to those required by Appendix G of ASNE Code Section III for beltline materials having USE values below the NRC screening criteri a of 50 ft-lb in 10 CFR 50, Appendix G. By letter dated July 1, 1993, the NRC issued a request for additional information (RAI) on the report [3]. The reply to the RAI was received by the NRC on August 20, 1993 and the NRC was notified that the report would be revised based on the RAI.

WCAP-13587, Revision 1 [4], incorporating the responses to the RAI, was subsequently submitted to the NRC through NUMARC by letter dated September 30, 1993.

3.0 APP CA R GU ATORY R U R M NTS AN UID N S Regulatory requirements for upper shelf safety margins are contained in Appendix G of The Code of Federal Regulations, Title 10, Part 50 (10CFR50).

Appendix G requires that the initial unirradiated Charpy V- Notch (CVN) upper shelf energy at the start of vessel life be no less than 75 ft-lb and that the vessel maintain an upper shelf energy level of no less than 50 ft-lb throughout the service life. If it is anticipated that a vessel might fall below 50 ft-lb before EOL, an analysis must be submitted which demonstrates "margins of safety against fracture equivalent to those required by Appendix G of the ASME Code." This analysis is subject to the approval of the Director, Office of Nuclear Reactor Regulation.

Guidelines that the staff finds acceptable for conducting equivalent margins analyses are contained in ASME Code Case N-512 [5] and draft Regulatory Guide DG-1023, "Evaluation of Reactor Vessels with Charpy Upper Shelf Energy less than 50 ft-lb"[6]. DG-1023 incorporates the criteria of Code Case N-512 and .

provides additional guidance on material properties and transient selection.-

DG-1023 was released by NRC for public comment on October 1, 1993. NUREG/CR-.

6023 [7] presents the results of generic bounding equivalent margins analyses conducted .according to Code Case N-512 guidelines for both PWRs and BWRs.

As the majority of licensees do not have fracture toughness"information for their limiting vessel materials, CVN data is typically used to estimate the fracture toughness. Regulatory Guide 1.99, Revision 2 [8] provides a procedure for estimating the decrease in CVN upper shelf energy as a function of copper content and fluence. NUREG/CR-5729 [9] provides empirically derived models for predicting material fracture toughness (J-R curves) from CVN data or chemical content and fluence. The NUREG/CR-5729 models are applicable to the majority of RPV materials.

NRC Branch technical position MTEB 5-2 [10] provides criteria for estimating the CVN toughness for the transverse orientation from longitudinal data for plate materials. The 50 ft-lb criterion in 10CFR50, Appendix G is for CVN data from transversely oriented specimens in plate or for welds. The transverse orientation in plate material has lower impact properties than the longitudinal orientation. The impact properties of the weld metal are not differentiated based on orientation.

4. DMILJPII The WOG employed the procedures and criteria of ASME Code Case N-512 [5] to perform the equivalent margins analysis. For service levels A and B, the WOG performed an analysis for the limiting normal operation cooldown rate (100'F/hr). The analysis is consistent with the guidelines provided in Code Case N-512 [5] and DG-1023 [6]. In accordance with both the Code Case and Draft Regulatory Guide, the WOG assumed a quarter thickness depth flaw for Levels A and B and a flaw depth of up to 1/10 of the base metal wall thickness

't plus cladding for Levels C and D. The flaw geometry considered for all service conditions was a semi-elliptical surface flaw with a 6: 1 aspect ratio.

A question raised in the RAI [3j concerned the issue of whether both axial and circumferential flaws had been addressed as required by the Code Case. The response to the RAI t3] cited that both cases were considered and the applied fracture driving force for the axial flaws was in all cases greater than those, for the circumferential flaws. Consequently, the axial flaw cases were considered limiting. The RPV beltline base materials covered by the WOG included:

(1) A302 Grade A (Plate)

(2) A302 Grade B (Plate)

(3) A533 Grade B, Class 1 (Plate)

(4) A508 Class 2 (Forging)

(5) A508 Class 3 (Forging)

The RPV beltline welds covered by the WOG include those produced using the following fluxes:

(1) Linde 1092 (2) Linde 1091

,(3) Linde 0091 (4) Linde 124 (5) UH89 (6) ARCOS B5 (7) Grau Lo (8) SMIT 89 Linde 80 weldments were not included in the WOG report. The WOG maintains that "weldments fabricated with fluxes other than Linde 80 will not exhibit less than 50 ft-lb during service." Weldments will be discussed further in Section 4.2. Table 1 provides the WOG overall minimum acceptable USE values by plant type (2, 3 or 4 loop) from the equivalent margins analysis. The values in Tab'le 1 are based on fracture properties for plate and forging materials. No analysis was performed based on weldment properties since the WOG maintains that all of their'weldments will exceed 50 ft-lb at EOL ~

The equivalent margins analysis compared the EOL USE value for each plant with the minimum acceptable USE values shown in Table 1:

TABLE 1 Minimum Acceptable USE Values From WCAP 13587, Revision 1 CASE LOWEST USE ft-lb 2 Loo 29 3 Loo 42 4 Loo 43

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.'he EOL USE values for each plant were determined in accordance with the methodology provided in Regulatory Guide 1.99, Revision 2 [8]. The adequacy of the equivalent margins analysis can be determined from the following considerations: (I) selection of mechanical properties, (2) the determination of predicted EOL USE value, (3) selection of the model for generating the fracture toughness data (J-R curves), (4) selection of the limiting transients, (5) the calculation for applied fracture driving force (J,~~)

and, (6) the bounding nature of the analysis.

4. I Mechanical Pro erties Plant specific values of tensile mechanical properties for all of the WOG vessel materials were requested in the RAI [3]. The WOG instead chose to bound the mechanical properties for all of the plants by using the ASHE Code minimum .allowable mechanical properties at 600'F. The stress-strain curve used in the equivalent margins analysis was generated using typical data for carbon steel and then adjusting the curve to match the ASHE Code minimum values. The staff finds this methodology to be acceptable.

4.2 Predicted S glue Initial upper shelf energy values were determined from material certification and surveillance test reports. GL 92-01 responses were also used to obtain supplemental CVN data. The EOL USE values were estimated using the Regulatory Guide 1.99, Revision 2 methodology [8]. Using this approach three plants were found to have EOL USE values below 50 ft-lb as shown in Table 2:

TABLE 2 WOG PLants with USE Less Than 50 ft-lb at EOL and Comparison with Minimum Acceptable USE Values From The Equivalent Ha'rgins Analysis PLANT CASE SHELL HATERIAL PREDICTED MINIHUH I.D. No. COURSE EOL USE ACCEPTABLE USE ft-lb ft-lb 3 Loo Intermed. A302 Gr A 42 3 Loo Lower A302 Gr B 37 4 Loo U er A508 Cl 2 49 43 4 Loo Intermed. A508 Cl 3 The limiting material for all three plants was either plate (A302, A533) or forging (A508). A comparison of the predicted EOL USE values with the WOG minimum acceptable USE values in Table, I, shows that the predicted plant data is equal to or bounded by the minimum acceptable values for all three plants.

The subject report also noted an additional case (Plant 16) where EOL USE

. above 50 ft-lb was not demonstrated for an A533 B plate. It was noted by the WOG that the USE values for the upper shell course of Plant 16 were based on

- data from unirradiated longitudinal specimens with less than 100K shear

6 fracture indicating that the tests did not represent full upper shelf behavior. The licensee for Plant 16 has submitted an additional analysis of the Charpy data for their beltline plates [18j. This submittal shows that the upper shell course plates will achieve an EOL fluence of only 4.85 E+17 n/cm and therefore the percent drop in USE with irradiation is low (approximately 10K). With the small predicted drop in USE an initial value of only 56 ft-lbs is required for the upper shell course plates to remain above 50 ft-lb to EOL.

The submittal demonstrates that, although full shear performance was not achieved in the CVN tests for two of the upper shell course plates, the extrapolation to an EOL USE greater than 56 ft-lbs for full shelf performance is reasonable. The staff finds this analysis to be acceptable.

As the MOG did not provide a listing which identified the WOG plants and their corresponding USE data by name, confirmation of the input data (chemistry and fluence) and calculations on a plant-by-plant basis was not possible.

As stated previously in Section 4.0, the WOG maintains that "weldments fabricated with fluxes other than Linde 80 will not exhibit less than 50 ft-lb during service." Table 2-2 in the subject report confirms this statement by showing a listing with initial USE and projected EOL USE values

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for each plant. The lowest EOL USE value reported was 50 ft-lb for Plant 17 with ARCOS B5 flux. As was the case with the base materials, the WOG did not provide a listing which identified the WOG plants and their corresponding USE data by name. Therefore, confirmation of the input data (chemistry and fluence) and calculations on a plant-by-plant basis was not possible.

Further, since the minimum acceptable "SE values from the equivalent margins ana'Iysis were based on fracture toughness data for plate or forging materials, these values will not apply to weldments. However, calculations performed for NUREG/CR-6023 [7j with the Linde 80 weld model have shown a minimum acceptable USE of 41 ft-lb for a PMR. The staff considers that, while weldments for the WOG plants are not likely to be limiting for USE, the subject report does not provide sufficient information to make this determination conclusively. As part of the GL 92-01 review, the staff will be evaluating each beltline weld for the WOG plants to determine if the screening criterion will be met at EOL.

4.3 Selection of the Fracture Tou hness J-R Curve Model The J-R curve information required for the equivalent margins analysis was generated from the bounding EOL USE values calculated for each plant category (2 loop, 3 loop and 4 loop). The J-R curves for RPV base materials were obtained from CYN data using the Charpy model of NUREG/CR-5729 [9). As stated previously, the WOG maintains that all of the WOG welds will be above 50 ft-lb at EOL and therefore did not develop J-R data for welds. For levels A and 8 the J-R curves for a temperature of 390.5'F were used. This value represents the greatest temperature at the crack tip during the 100'F/Hr. cooldown for a 1/4T flaw based on peak stress results from a 2-D finite element analysis.

For Levels C and D the temperatures used for the J-R curves were from the appropriate transients (in the range of 400'F 500'F). The staff finds, the approach to developing the J-R curve data for all of the WOG base materials to

. be acceptable with the exception of the method used for A302 B. For A302 8 plate, the WOG considers the 6T C(T) J-R curve data of NUREG CR/5265 be a lower bound for a 50 ft-lb A302 8 plate.

[ll] to The MOG adjusted the 6T data

for lower USE values by decreasing J , < by 3X per ft-lb of CVN energy.

This rate of decrease corresponds to Tee rate determined from the Charpy model of NUREG/CR-5729 [9]. The staff also consider the 6T A302 B data of NUREG/CR-5265 to be a lower bound for A302 B. However, the staff finds the overall J-R Curve adjustment procedure for A302 B to be unacceptable for the following reasons:

(1) The percent decrease in J , ;, calculated to EOL for the limiting A302 5 material assumed tRaf. the material in NUREG/CR-5265 [11] had a USE of 50 ft-lb. However, the actual USE for the A302 B material in NUREG/CR-5265 [ll] is 53 ft-lb. Therefore the X decrease in J <,< cited for A302 B in the subject repor't should be approxImately 33X as opposed to 24X.

(2) The resulting J ,. data were not temperature corrected. The data from NUREG/CR-5F63 [(1] were generated at a temperature of 180'F.

The staff maintains that a temperature correction is necessary. One method that has been proposed for determining the decrease in J , ,, with temperature for NUREG/CR-5265 [11] material is contained in reference [12]. Using the equation provided in this reference, the decrease in J < with temperature from 180'F to 390'F is approximately 34'.

Adjusting the analysis to account for the above yields a J , , ~ value at 0. 1 inches of crack extension (Jo ) which is significantly less ~Chan that obtained by the WOG. A'ethodology for adjusting A302 B J-R curve data which is similar to that in reference [12] and has been found acceptable by the staff is contained in the BMR Owners Group Topical Report NEDO-32205 [13]. It should be noted that the A302 B plate tested for NUREG/CR-5265 [11] was a high sulphur plate which received a minimal amount of cross rolling. The resultant fracture properties were severely degraded by the presence of large manganese-sulfide (HnS) inclusions. For these reasons, it is questionable whether the plate can be considered representative of A302 B which would have been used in RPV construction. However, until recently, NUREG/CR-5265 represented the only large scale fracture data that were available for A302 B. Hence, the staff have considered these data to be a lower bound for A302 B plate. Recent preliminary fracture test results from the Oak Ridge National Laboratory (ORNL) on a potentially mor e representative A302 B plate indicate vastly superior properties to those obtained in NUREG/CR-5265 [11]. The st'aff will consider the ORNL data and other available fracture data in establishing future positions for A302 B plate fracture resistance.

44 Sele tion i iti Transient The subject report used the 100'F/hr heatup/cooldown case as bounding for Levels A and B. This is consistent with the evaluation performed in NUREG/CR-6023 [7]. For Levels C and D the WOG used both peak stress and the overall magnitude of the through-wall stress as the criteria to determine the bounding transients. The resulting temperatures at the crack tip (10K of wall thickness + cladding) were 511 F for Plant 17 and 487'F for the remaining plants. Using these criteria, it was judged that a small steam line break (SLB) was the limiting Level C transient and large loss of coolant accident

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I (LOCA)'and large SLB were the limiting Level D transients. For Level D, the large SLB was ultimately selected as the limiting Level D transient as a stress analysis conducted by the MOG showed that the large SLB produced slightly larger stresses at the crack tip for a 1 inch deep flaw. In response to the RAI, the WOG provided the pressure and temperature histories for all of the transients considered in their assessment. Based on an examination of all of the transients the staff concludes that the limiting C and D transients selected by the WOG are acceptable.

4.5 Calculation of A lied Fracture Drivin Force J The MOG employed the procedures of Code Case N-512 [5] and DG-1023 [6] in the determination of the fracture driving force (J . ) for Levels A and B. The staff performed an independent evaluation of tfie 7 ,.~ calculations for the limiting Plant 7 case and were able to verify the S( calculations. For levels C and D the pressure and temperature histories for all of the transients considered were input to a 2-D finite element model of the nuclear steam supply system (NSSS) using the WECAN [14] computer code. The resulting stress distributions for the limiting transients, which included the contributions of the cladding to the thermal stress, were used to calculate J , ~ using the PCFAD [15] computer code. Consistent with the results of tlute founding analyses performed for NUREG/CR-6023, the MOG J, , ~ values for Levels A and B were found to be controlling. The staff finds the methodology employed by the WOG for calculation of the fracture driving force to be acceptable. However, the staff discovered that the final analysis for equivalent margin; was overly conservative when comparing the fracture driving force (J <,~) with the material fracture resistance (J ,-). This will be discusseP>n the next section.

4.6 Boundin Nature of the Anal sis The results of the WOG equivalent margins analysis are presented in Tables 3, 4 and 5 for Levels A, 8, C and D. These Tables show a comparison between the applied driving force and material resistance values for the three generic cases (2, 3, and 4 loop) and two plant-specific cases. The MOG concluded that in all cases the criteria of Code Case N-512 were met (J , ,, > J, ~).

However, in examining the Plant 7 case (A302B plate) in Petal), the staff concluded that the material value for Jp determined by the WOG was not acceptable (see Section 4.3). The material value that the staff estimated for q

J~, for Plant 7 was well below the applied value calculated by the MOG.

Further examination of this case revealed that the actual reason for exceeding the Code Case N-512 criteria was that the MOG had conservatively compared the applied driving force for the axial fl~w with the material resistance for the weak orientation which corresponds to the circumferential flaw (i.e., meeting the criteria under these conditions assures that the criteria will also be met when the values are compared as suggested by the Code Case.) A conference call [16] with the Mestinghouse Electric Corporation and the licensee for Plant 7 confirmed that this conservative analysis was performed for all of the cases evaluated. The staff considers this analysis to be overly conservative.

In a revision letter dated December 21, 1993 [17] which specifically addresses the Plant 7 case, the WOG demonstrated that the appropriate comparison of the applied and material values shows that the Code Case N-512 [5] criteria were

met. The letter [17] also addressed the USE adjustment and temperature correction concerns for A302 B plate that were discussed previously in Section 4.3, The MOG is preparing a more comprehensive revision to the report which will address all of the cases considered. The staff concludes that, in the interim, the subject report and the alternative evaluation provided in the revision letter of December 21, 1993, demonstrate equivalent margins of safety as per Code Case N-512 [5] for WOG plate and forging materials.

For the case of welds, as stated previously, the WOG maintains that "weldments fabricated with fluxes other than Linde 80 will not exhibit less than 50 ft-lb during service." However, the WOG did not provide a listing which identified the WOG plants and their corresponding USE data by name. Therefore, confirmation of the input data (chemistry and fluence) and calculations on a plant-by-plant basis was not possible. Further, since the minimum acceptable USE values from the equivalent margins analysis were based on fracture toughness data for plate or forging materials, these values will not apply to weldments. Calculations performed for NUREG/CR-6023 [7] with the Linde 80 weld model have shown a minimum acceptable USE of 41 ft-lb for a PWR.

Therefore the staff considers that, while weldments for the WOG plants are not likely to be limiting for USE, the subject report does not provide sufficient information to make this determination conclusively. As part of the GL 92-0I review, the staff will be evaluating each beltline weld for the MOG plants to determine if the screening criterion will be met at EOL.

In the subject report, the MOG cited a minimum acceptable USE value of 29 ft-lb for plates and forgings in 2 loop plants. Mhile numbers this low can result from the calculations and have been cited elsewhere ['6], [7], their significance is not clear at this time and they represent an extrapolation of the data used to develop the correlations in NUREG/CR-5729 [9]. There is not at present a technical consensus on the CVN energy lower limit for equivalent margins analyses. However, values below 30 ft-lb for reactor pressure vessel steels are highly suspect as such low energies are likely to be associated with ductile tearing in combination with other fracture modes (e.g.,

intergranular, incipient cleavage) not considered in the development of the low upper shelf criteria.

5.0 CONCLUSION

S The staff has evaluated the equivalent margins analysis presented in report WCAP-13587, Revision 1, submitted by the Westinghouse Electric Corporation for the WOG. The staff evaluation supports the following conclusions:

(1) The methodology employed for the WOG report was consistent with the guidelines in Code Case N-512 [5] and DG-1023 [6] and is therefore acceptable to the staff.

(2) The subject report demonstrates margins of safety equivalent to those of the ASHE Code for WOG beltline plate and forging materials.

(3) An equivalent margins analysis based on weldment properties was not performed because the WOG concluded that all of the WOG beltline welds

'ould be above 50 ft-lb at EOL. As part of the GL 92-01 review, the

l4 10 staff will be evaluating each beltline weld for the WOG plants to determine if the screening criterion will be met at EOL.

(4) Individual licensees desiring to reference WCAP 13587, Revision 1, as the basis for addressing the USE requirements of 10 CFR 50, Appendix G must confirm the specific plant applicability by either demonstrating all beltline welds will have EOL USE above 50 ft-lb or that the 'hat bounding plate used in the report has a lower J-R curve than any other beltline material in the vessel. Further, individual licensees must request approval in accordance with 10 CFR 50, Appendix G.

(5) The staff finds the WOG approach for modeling the J-R curve behavior of the A302 B plate in WCAP-13587, Rev. 1, to be unacceptable (see Section 4.3). However, the A302 B J-R curve methodology employed in the plant specific revision letter of December 21, 1993 [17] includes appropriate adjustments for USE and temperature corrections and is acceptable to the staff.

The staff finds that the final comparison of applied fracture driving force with material fracture resistance was overly conservative for all cases analyzed. This issue was adequately addressed for a limiting plant specific case by letter dated December 21, 1993 [17], and the WOG is preparing a revision to the report which addresses all of the cases considered.

(7) Calculated minimum. acceptable USE values below 30 ft-lb for reactor pressure vessel steels are highly suspect as such low energies are likely to be associ ated with ductile tearing in combination with other fracture modes (e.g., intergranular, incipient cleavage) not considered in the development of the low upper shelf criteria.

6.0 REFERENCES

WCAP-13587, Revision 0, "Reactor Vessel Upper Shelf Energy Bounding Evaluation. for'estinghouse Pressurized Water Reactors", S. Tandon and T.R. Hager, Westinghouse Electric Corporation, February, 1993.

[2] Generic Letter 92-01, Revision 1, Reactor Vessel Structural Integrity,. 10 CFR 50.54(f), U.S. Nuclear Regulatory Comnission, March 6, 1992.

[3] Request for Additional Information on WCAP-13587, "Reactor Vessel Upper Shelf Energy Bounding Evaluation for Westinghouse Pressurized Water Reactors," U.S. Nuclear Regulatory Commission, July 1, 1993.

[4] WCAP-13587, Revision 1, "Reactor Vessel Upper Shelf Energy Bounding Evaluation for Westinghouse Pressurized Water Reactors", S. Tandon M.J. Halone and T.R. Mager, Westinghouse Electric Corporation, September, 1993.

[5] Code Case N-512, "Assessment of Reactor Vessels with Low Upper Shelf Charpy Impact Energy Levels,"Section XI, Division 1, ASHE Boiler and Pressure Vessel Code, February 12, 1993.

[6] Draft Regulatory Guide DG-1023, "Evaluation of Reactor Pressure Vessels with Charpy Upper Shelf Energy Less than 50 ft-lb," U.S.

Nuclear Regulatory Comaission, August, 1993.

NUREG/CR-6023, "Generic Analyses for Evaluation of Low Charpy Upper Shelf Energy Effects on Safety .Margins Against Fracture of Reactor Pressure Vessel Materials," T.L. Dickson, USNRC,'ashington, DC, July 1993.

[8] Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Comnission, Hay, 1988.

[9] NUREG/CR-5729, "Multivariable Modeling of Pressure Vessel and Piping J-R Data," E.D. Eason, J.E. Wright and E.E. Nelson, USNRC, Washington, DC, 1991.

[10] NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," Branch Technical Position MTEB 5-2, Fracture Toughness Requirements, Section 5.3.2, USNRC, Washington, DC, 1981.

NUREG/CR-5265, "Size Effects on J-R Curves for A302-B Plate," A.L.

Hiser and J.B. Terrell, USNRC, Washington, DC, 1989.

[12] Memorandum, A.L. Hiser and S.N.M. Halik of USNRC to K.R. Wichman, USNRC, "J-R Curves for Low Toughness A302 B Plate," USNRC, September 9, 1993.

12

[13] NED0-32205, ".10CFR50 Appendix G Equivalent Hargin Analysis for Low Shelf Energy in BWR/2 Through BWR/6 Vessels," Licensing 'pper Topical Report for the BWR Owners'roup, GE Nuclear Energy, April, 1993.

[14] Westinghouse Electric Computer Analysis (WECAN) Code, File 87-1J7-WESAD-Rl, dated December 1987.

[15] Bloom, J.H. and D.R. Lee, "Users Guide for the Failure Assessment Diagram Computer Code FAD," Babcock and Wilcox, Rev. 4, April 1990.

[16] Conference Call, NRC, Carolina Power E Light and Westinghouse, December 13, 1993.

[17] Letter from D.B. Waters, Carolina Power 5 Light Company to USNRC, of December 21, 1993, "Carolina Power 5. Light Company, H.B. Robinson Steam Electric Plant Unit No. 2, Docket No. 50-261/License No.

DPR-23, Reactor Vessel Upper Shelf Energy For Reactor Vessel Core Region Plate Haterials."

[18] Letter from E.E. Fitzpatrick, Indiana Hichigan Power to T.E. Hurley, USNRC, AEP:NRC: 11738, with Attachment entitled, "American Electric Power Service Corporation, Donald C. Cook Nuclear Plant Unit 2, Upper Shelf Charpy Values," January 24, 1994.

13 TABLE 3 J <.~ vs. J , for Levels A 5. B from MCAP 13587, Revision 1, (Pnhs on J are >n-lbs/in )

APPLIED MATERIAL MET CASE dJ da dJ da CRITERIA?

2 Loo 384 318 702 2925 Yes 3 Loo 500 321 585 527 2140 599 Yes 4 Loo 590 345 614 2330 Yes Plant 7 525 222 527 599 Yes Plant 17 548 197 614 2330 Yes

( )

Calculations based on J-R Curves from Reference 11 TABLE 4 J ~.~ vs. J <, for Le~el C from MCAP 13587, Revision 1 (Pnhs on J are sn-lbs/in )

APPLIED MATERIAL MET CASE dJ/da dJ/da CRITERIA?

2 Loo 311 225 702 2925 Yes 3 Loo 310 252 585 527 2140 599 Yes 4 Loo 311 225 614 2330 Yes Plant 7 308 250 527 599 Yes Plant 17 319 240 614 2330 Yes

( )

- Calculations based on J-R Curves from Reference ll

14 TABLE 5 J <,.~ vs. J ,, for Level D from WCAP 13587, Revision 1 (Pn>ts on J are 1n-lbs/in )

APPLIED MATERIAL MET CASE dJ da dJ da CRITERIA ?

2 Loo 447 2925 Yes 3 Loo 2140 599 Yes 4 Loo 447 2330 Yes Plant 7 599 Yes Plant 17 468 2330 Yes

( )

Calculations based on J-R Curves from Reference ll