ML20236W950
| ML20236W950 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 08/03/1998 |
| From: | Callan L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Seale R Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-GENERAL, NUDOCS 9808060333 | |
| Download: ML20236W950 (7) | |
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y & UNITED STATES s* j NUCLEAR REGULATORY COMMISSION WASHING)ON, D.c. 20555-0001 i
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\,...,/ August 3, 1998 Dr. Robert L. Seale, Chairman f2-3 Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission l Washington, D.C. 20555-0001 ;
Dear Dr. Seale:
Thank you for your third interim letter of June 15,1998, on the views of the Advisory Committee on Reactor Safeguards (ACRS) regarding the safety aspects of the Westinghouse Electric Company application for certification of the AP600 design. The Office of Nuclear Reactor Regulation (NRR) has reviewed your concems related to the test and analysis program and environmental qualification testing for passive autocatalytic recombiners (PARS) as they relate to the AP600, and your recommendations for future evaluation of leak-before-break (LBB) issues conceming feedwater piping and in-vessel retention. The staff has the following responses to your comments:
Test and Analysis Proaram
. Request for Specific information In the June 15 letter, the ACRS identified four items that required additional information from Westinghouse. The staff believes that the Westinghouse commitments were fulfilled, based on information provided during the ACRS Advanced Reactor Designs ,
Subcommittee meeting on July 7,1998, and supplemented by two Westinghouse letters to the ACRS dated July 1 and July 2,1998.
. Acceptability of Test and Analysis Documentation I
in the June 15 letter, the ACRS stated that it was concerned about the quality of test and analysis documentation related to information needed to certify the AP600 design.
In its letter of March 23,1998, to the ACRS, the staff stated that the acceptance of documentation is based on three factors: (1) that the documentation is technically sound, (2) that the documentation contains no known technical errors, and (3) that the documentation is sufficient for the staff to make a technical determination of ccmpliance.
At the time the letter was sent to the Committee, some of the documentation had either ;
not been submitted or the staff's review had not been completed. Since then, the expected submittals have been received from Westinghouse, and the staff has completed its review. The staff has concluded that although voluminous and difficult to review, the
, test and analysis documentation meets these three criteria. The staff has documented c00 the acceptability of these analytical tools for application to the AP600 in its Advance Fmal Safety Evaluation Report (AFSER). The results of the staff's review have been ,
discussed during recent meetings with the Thermal-Hydraulic Subcommittee. The staff I notes that two typographical errors were recently identified by the Committee in the /
NOTRUMP verification and validation report. Westinghouse has committed to correct I these errors, and the staff will confirm that these corrections are properly incorporated i into the report. $)I g A PDR 78 "/N
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Dr. Seale 2 As stated during the ACRS Advanced Reactor Designs Subcommittee meeting on July 7, ,
1998, the AP600 analytical computer programs have been developed specifically for I application to support the AP600 design-basis analyses and design certification. These computer programs are not being approved for generic analyses and licensing. Any j significant changes to the AP600 design would require re-approval by the NRC. Any new evaluation by the NRC would necessitate a review of the applicability of both the computer programs and testing programs to the new design.
In aid! tion, the ACRS further recommended that the staff determine whether the quality of the A P600 documentation could withstand an NRC design-basis inspection. In November 1997, the staff performed a 5-day,9-member team quality assurance inspectan at Westinghouse that focused on the quality of documentation associated with computer program qualification, testing, and design-basis analyses. The inspection identified numerous documentation errors, which are discussed in Section 17.3 of the staff's AFSER. As a result of this inspection, Westinghouse conducted a comprehensive design assurance review to establish the extent of the problem and to ensure that the documentation complied with its quality assurance requirements. Westinghouse concluded that tha documentation error rate was limited and did not affect any design-bases conclusions. Westinghou.,e also corrected known errors or demonstrated that the errors resulted in conservative calculational results. On the basis of design assurance review findings and the Westinghouse corrective actions, the staff has concluded that the ,
AP600 design documentation complies with Appendix B of 10 CFR Part 50 and is 1
' acceptable.
5 environmental Qualification Testina of PARS in me June 15 letter, the ACRS recommended that environmental qualification tests for passive autocatalytic recombiners include r acquirements for timing of exposure and exnosure to pyrolysis products. This matterwas discusse:1 further during the July 7 meeting of the Advanced Reactor Design Subcommittee to clarify that tae Committee was concemed about non-fire-related pyrolytic byproducts created by the interaction of organic material (such as hydrocarbons or sulfur from the decomposition of cabling)in the containment as the result of radiolytic processes as -
I well as heat and steam.
The staff recognized in the AFSER that the PARS may be susceptible to catalytic poisons or other effects that could inhibit recombination. Some of the chemical constituents included in the NUREG-1465 source term assumptions (e.g., tellurium), as well as silicone oils and phosphates, were specifically identified as chemicals that have the potential to be present within containment in significant-enough quantities during a loss-of-coolant accident (LOCA) to affect the PARS
- recombination rate. Significant quantities of hydrocarbons and sulfur from pyrolytic l decomposition of cabling, due to radiolytic processes combined with heat and steam, have not been identified as expected chemical constituents in the containment environment following a design-basis accident in any information provided by Westinghouse, or in publicly available
, literature on this topic independently reviewed by the staff. In addition, the staff has not l
p.eviously required environmental qualification for pyrolytic byproducts of organics due to design-basis accident containment conditions on other safety-related components within containment, l which might be subject to similar adverse reactions to such chemicals (such as charcoal adsorbers in various filtration systems or seals for containment sump recirculation systems).
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Dr. Seale 3 The combined operating license (COL) applicant is committed to test the PARS against the criteria for a harsh environment that a mechanical component would be exposed to in containment, including the chemical elements assumed in NUREG-1465. The staff has accepted this level of environmental qualification for the PARS, in part, because of the excess capacity provided by the two full-size PARS and two quarter-size PARS. Analyses performed by the applicant show that one PAR, the performance of which is degraded by 90 percent is capable of maintaining the hydrogen concentration inside the containment below the lower flammability limit of 4 percent following a design-basis LOCA. In addition, the AP600 has been provided with 64 igniters to control hydrogen if the concentration of hydrogen inside containment exceeds the lower flammability limit of 4 percent.
Therefore, the staff concludes that this hydrogen control design meets the requirements of 10 CFR 50.44.
Other Recommendations in addition to the preceding AP600-related issues, the ACRS recommended that, although the following items have been adequately addressed for the AP600 design, the staff should further evaluate the following items to support review of future license applications or licensing actions:
Leak Before Break of Feedwater Pipina The ACRS requested that the staff re-examine its position on the likelihood of the conditions assumed in calculating the load used in its bounding analysis for water hammers in feedwater piping. As noted in the June 15 ACRS letter, the staff has not attempted to estimate the probability of a damoging water hammer in the AP600 feedwater piping. The hypothetical water hammer scenario presented by the staff was developed as a scoping tool to show that the loads generated by such an event could threaten the integrity of the feedwater piping. The staff found that Westinghouse AP600 design may be effective in reducing the occurrence of water hammer in the feedwater lines but cannot totally eliminate its potential for occurrence. In addition, the staff had concems about the lack of operating experience with the APG00 design, as well as difficulty with defining and incorporating water hammer load in the design of feedwater lines. The staff has not previously accepted LBB for feedwater piping in operating pressurized-water reactors. To establish such a precedent, the staff requires substantial technical justification to demonstrate that either (1) the probability of water hammer in the feedwater piping is extremely low, in accordance with the requirements of General Design Criterion 4 of 10 CFR Part 50, Appendix A, or (2) the maximum hypotheticalloads are within the design basis of the piping. For the AP600, the staff found that Westinghouse did not meet either of these two conditions, and thus concluded that LBB should not be applied to the AP600 feedwater piping. However, the staff would be willing to consider application of LBB to feedwater lines in any future application, provided sufficient technicaljustification is submitted by the applicant to demonstrate that design features address water hammer concerns and that the applicable rules and regulatory guidance
! are met.
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Dr. Seale 4 i
1 i In-Vessel Retention
, The June 15 ACRS letter recommended that since in-vessel retention is widely considered to be l an isnportant accident management strategy for operating reactors, the staff should assess the impact of intermetallic exothermic reactions on this strategy.
l In response to this recommendation on intermetallic exothermic reactions, the staff is making a proposal to the RASPLAV (Russian Research Center /Kurchatov institute) that small scale tests be performed to explore this issue.
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Please feel free to contact me or Theodore Quay of the NRR Standardization Project Directorate at 301-415-1118 for any further information conceming this letter.
Sincerely, L. J ph CWan l Executive Director for Operations cc: Chairman Jackson Commissioner Diaz Commissioner McGaffigan SECY OGC OCA OPA CFO ClO l
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a Dr. Seale 4 in-Vessel Retention The June 15 ACRS letter recommended that since in-vessel retention is widely considered to be an important accident management strategy for operating reactors, the staff should assess the impact of intermetallic exothermic reactions on this strategy, in response to this recommendation on intermetallic exothermic reactions, the staff is making a proposal to the RASPLAV (Russian Research Center /Kurchatov Institute) that small scale tests be performed to explore this issue.
Please feel free to contact me or Theodore Quay of the NRR Standardization Project Directorate at 301-415-1118 for any further information conceming this letter.
- il Sig22/j 'Oy L. .l. Callatl
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L. Joseph Callan Executive Director for Operations 1
l cc: Chairman Jackson Commissioner Diaz Commissioner McGaffigan SECY OGC OCA OPA CFO CIO DISTRIBUTION: See next page DOCUMENT NAME: A:GT980388.GTK To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy *SEE PREVIOUS CONCURRENCE l OFFICE PM:PDST:DRPM l D:PDST:DRh A l ' (A)D:DRPM/ D:DS4A n j 1 e^ {
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DISTRIBUTION: Letter to Dr. Seale. Dated: August 3. 1998 Docket File PUBLIC PDST R/F EDO GT980388
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HThompson,0-16 E15 AThdani,0-16 E15 PNorry,016 E15 JBlaha,0-16 E15 SCollins/FMiraglia,0-5 E7 BBoger,0 5 E7 BSheron 0-5 E7 WTravers,0-14 D4 JRoe DMatthews TRQuay TKenyon WHuffman JSebrosky JNWilson DScaletti MDusaniwskyj ACRS Bums,
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AUlfdfil Adp, EDO Principal Correspondence Control FROM: DUE: 07/16/98 EDO CONTROL: G980388 DOC DT: 06/15/98 R. L. Seale, ACRS FINAL REPLY:
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C211an, EDO FOR SIGNATURE OF : ** GRN **
CRC NO:
Collan, EDO DESC:
ROUTING:
THE SAFETY ASPECTS OF THE WESTINGHOUSE ELECTRIC Callan COMPANY APPLICATION FOR CERTIFICATION OF THE AP600 Thadani PLANT DESIGN - INTERIM LETTER 3 Thompson Norry Blaha DATE: 06/16/98 Burns Morris, RES Knapp, NMSS ASSIGNED TO: CONTACT: Martin, AEOD NRR Mitchell, OEDO Collins ACRS File SPECIAL INSTRUCTIONS OR REMARKS:
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! ADVISORY COMMITTEE ON REACTOR sAFEJUARDS e,, WASH WGTON, D. C. 3066s June 15,1998 l
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l Mr. L Joseph Callan Executive Director for operations
! U.S. Nuclear Regulatory Commission
! Washington, D.C. 20555-0001
Dear Mr. Callan:
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SUBJECT:
THE SAFETY ASPECTS OF THE WESTINGHOUSE ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE AP600 PLANT DESIGN - INTERIM LETTER 3 Durin0 the 453rd meeting of the Advisory Committee on Reactor Safeguards, June 3 5,1998, we reviewed the AP600 test and analysis program, various chapters of the AP600 Standard Safety Analysis Report (SSAR), the Level 2 and 3 AP600 Probabilistic Risk Assessments (PRAs), severe accidents, regulatory treatment of non-safety systems, and the associated chapters of the NRC l staffs advance Final Safety Evaluation Report (FSER). Our Subcommittees on Thermal Hydraulic Phenomena and Advanced Reactor Designs reviewed these items on May 11 12 and May 1315, 1998, respectively. During these reviews, we had the benefit of discussions with representatives of the NRC staff and the Westinghouse Electric Company. We also had the benefit of the documents referenced.
Based on our review to date, no additional issues were identified that would prevent the certification of the AP600 design. Our assessment is based,in part, on agreement by Westinghouse to improve its documentation of the test and analysis program. In addition, we identified several issues related to NRC staff assessment of accident phenomena. Our comments are provided below.
TEST AND ANALYSIS PROGRAM in our interim letter dated February 19,1998, we identified a list of outstanding thermal-hydraulic issues related to the documentation of the reactor coolant system and containment designs. The issues related to the containment were discussed by our Thermal Hydraulic Phenomena Subcommittee on June 1112,1998. Westinghouse responded to the issues related to the reactor coolant system at the May 11-12,1998 Thermal Hydraulic Phenomena subcommittee meeting, and .I committed to perform additional analyses and studies, and to provide additional explanations. Based on our assessment of the Westinghouse responses, we are satisfied that Westinghouse has fulfilled several commitments by: 1 Performing a sample analysis of the small-break loss-of-coolant accident (LOCA) involving automatic depressurization system (ADS) activation through initiation of in-containment refueling water storage tank (IRWST) flow to show the relationship between the IRWST level l
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. .a 2-penalty and flow resistances in the ADS piping. This analysis provided assurance that the level penalty Westinghouse takes in the NOTRUMP small-break LOCA code is an appropriate and conservative compensation for neglecting the momentum flux terms in the blowdown equation.
. Amplifying the Westinghouse scaling analysis to include the relationships between core inventory and the multiple flow paths. This permitted evaluation of the usefulness of relevant data obtained from the Oregon State University and the SPES-2 test facilities during the ADS actuation phase of an accident.
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! . Explaining the difference in timing for the minimum reactor vessel water level between the value calculated with the NOTRUMP code and the test data.
l Westinghouse still needs to submit the following additional information:
l l . H.s results of the break area sensitivity study for one of the severe small-break LOCAs to l c.ns,tre that the process for compensating for exclusion of momentum flux terms in the NOTl? UMP code is robust for a range of blowdown rates.
l . A discussion of the implications of the sensitivity of the results to the assumed heat loss ;
l distribution in the SPES-2 test facility in validating both the LOFTRAN and NOTRUMP codes, c
. A description in SSAR Chapters 4 and 15 of the interrelationships among the LOFTRAN.
THINC-IV, and WESTAR codes and the test data. Clear identification of channel-to channel 1 mixing coefficients to be used. I
- Clearidentification in the SSAR of the inadequacies in the NOTRUMP code and the steps taken to compensate for them.
STANDARD SAFETY ANALYSIS REPORT ANDTHE ADVANCED FINAL SAFETY EVALUATION REPORT We reviewed the Level 2 and 3 PRAs, severe accidents, regulatory treatment of non-safety systems, the following SSAR chapters, and the associated NRC staff's advanced FSER:
. Chapter 3 - Design of Structures, Components, Equipment, and System l . Chapter 6 - Engineered Safety Features l . Chapter 14 - InitialTest Program
. Chapter 16 - Technical Specifications
- Chapter 17 - Quality Assurance Based on our review of the above, we offer the following comments: l l
Reaulatory Treatment of Non Safety Systems l The active systems in the AP600 are designated as non safety whereas,in existing plants many of these active systems are designated as safety related. The regulatory treatment of these non-safety systems, which are relied upon for defense-in-depth and to meet plant investment protection goals, is an excellent example of a good risk-informed and ,
performance-based regulatory approach. l
Chapter 9 - Auxiliarv Systems. Includina Anoendix 9A - Fire Protection Analysis Since issuing our second interim letter concoming the AP600 plant design on April g,1998, we have completed our review of the fire protection system design and the fire protection analysis. The NRC staff has agreed with the Westinghouse proposal that the AP600 design should be govemed by 10 CFR 50.48, " Fire protection." The AP600 fire protection analysis used the Fire Induced Vulnerability Evaluation (FIVE) screening methodology. The NRC staff review of this analysis identified that the original design dio not provide separate water supplies for the fire fighting capabilities. Although Westinghouse did not agree that an additional water supply was needed, Westinghouse modified the design by Ialocating the diesel-driven fire pump from the turbine building to a prefabricated enclosure to be located in the yard. This proposed modification by Westinghouse will provide a separate water
! supply for fire fighting.
Such a modification brings the AP600 design into compliance with the requirements of 10 CFR 50.48 and the enhanced fire protection criteria approved by the Commission.
Consequently, we conclude that the AP600 fire protection system design is adequate.
Environmental Qualification Tests for Passive Autocatalvtic Recombiners Supported platinum or palladium catalysts will be used to control hydrogen concentrations in the AP600 reactor containment following design-basis accidents. Such catalysts are known to be fully capable of providing hydrogen recombination sufficioni to meet regulatory and safety requirements. Catalytic recombiners are susceptible, however, to deactivation during protracted use due to:
4 e poisoning of the catalytic surface, e cok!ng that occludes catalytic surfaces, e
surface diffusion and sintering of catalytic materials that result in a loss of active surface area, and e
interactions of noble metalwith the substrate.
The effect of these processes is cumulative as the time of recombiner operation increases.
Some short-term tests have examined the susceptibility of hydrogen recombiners to poisons and coking. Some of these tests are of questionable utility. The tests first exposed the catalysts to the poisoning material and then, in separate tests, measured the capacities of the exposed catalysts to recombine hydrogen. Any synergistic effects of poison and recombination activity would not have been revealed by this procedure. Similarty, effects of radiolytically generated ozone and nitrous oxides were not examined in the tests. On the other hand, the tests have examined a wide range of materials that might be expected to adversely affect catafytic activity and only modest (<20%) reductions in catalytic activity were found in the short-term tests.
Tests that simultaneously examine prototypic environments of temperature, radiation field, and catalytic activity for appropriate service times do not seem to have been done. The j adverse effects shown in short-term tests may well become more significant as service continues. Synergistic effects of radiation may exacerbate effects that are small under thermal conditions. .
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To increase confidence that the passive autocatalytic hydrogen recombiners will perform their intended functions effectively, there is a need for better environmental qualification tests. This may well be the responsibility of the Combined License (COL) applicant if, for no other reason, catalysts can be expected to be improved between now and the time a license is sought to operate an AP600 plant. We recommend that environmental qualification tests for passive autocatalytic recombiners include requirements for timing of exposure and exposure to pyrolysis products.
ITEMS FOR CONTINUED STAFF EVALUATION FOR LICENSING ACTIONS Although the following items have been adequately addressed for the AP600 design, additional evaluation of these items is needed to support efficient review of future license applications or licensing actions:
Leak Before-Break Evaluation of Feedwater Pinino The leak-before-break (LBB) criteria require that piping have high fracture toughness and not suffer from modes of degradation such as flow-assisted corrosion or stress-corrosion cracking that could result in significant loss of strength before detectable leakage occurs.
The piping must also not be subject to large loads that were not accounted for in the original design, such as those which might result from a large water hammer. The NRC has rieveloped guidelines and procedures (NUREG-1061, Vol. 3) that can be used to demonstrate that piping will exhibit LBB behavior.
The AP600 design makes more extensive use of the LBB concept in the design of reactor system piping than current reactors. In the advanced FSER, the staff has concluded that Westinghouse has been able to demonstrate through the choice of materials for the piping, stress and fracture mechanics analysis procedures, and the controls placed on water chemistry, that the piping for safety-related systems meets the LBB guidanne in NUREG-1061, Vol. 3.
The staff dsnied the request to apply the LBB concept to the feedwater piping design. The staff agrees that the present design meets all the LBB guidelines in NUREG-1061, Vol. 3, except for susceptibility to water hammer. The staff also agrees that the feedwater piping and steam generator designs for the AP600 have incorporated the " lessons loamed" from 1 operating plants for avoiding water hammers and that the piping design meets all the design guidelines for reducing susceptibility to water hammer. The staff argues, however, that there is no operating experience applicable to the AP600 design to demonstrate that the probability of a large water hammeris sufficierdly low, and proposes a bounding water hammer load that is 10 times as large as that proposed by Westinghouse. Since Westinghouse determined 1 that it was impractical to design the piping for a pressure pulse this large, Westinghouse l agreed to drop the request to apply the LBB concept to the feedwater piping.
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The bounding water hammer load proposed by the staW is based on the assumption that the main feedwater line fills with steam and then a large slug of cold water at high velocity is introduced into the piping. The staff concedes that the sequences of events that might lead to such a water hammer would require misalignment of several valves, but did not attempt to estimate the probability of such an event. According to Westinghouse, in order to establish the initial conditions assumed by the staff, the steam generator water level would have to be l st a point that would trip the reactor. All procedures for refilling a steam generator following i
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, e a reactor trip require using the auxiliary feedwater system, which injects water through a separate auxiliary feedwater injection line. The bounding water hammer based on ir.jection of cold water into the auxiliary feedwater line results in much smaller loads than those calculated by the staff for the main feedwaterline.
We believe that the stan should reexamine its position on the likelihood of the initial conditions assumed in calculating the load used in its bounding sinalysis for water hammers in feedwater piping. The staff has stated that it feels that some operational experience should be obtained with the AP600 feedwater system before approving the application of the LBB concept to the feedwater piping, it is completely impractical to demonstrate by operational monitoring the degree of assurance against large water hammers sought, which is <104 events / year. The degree of assurance could, however, be demonstrated by PRA techniques, which could be benchmarked by comparing the results of such analyses for current feedwater piping systems with operational experience.
, in Vessel Retention
, An AP600 strategy for mitigating the consequences of severs accidents is in-vessel retention of molten debris through extemal cooling of the reactor vessel. The reactor cavity is flooded with water to provide cooling of the lower head. A substantial experimental program using scaled models and sections of the lower head to support the heat transfer analyses has been used to evaluate the retention of the core melt. These tests, however, have not used prototyp!c materials.
The analysis ofirevessel tstention performed for the AP600 fails to demonstrate convincingly that vessel failure during a core melt is extremely unlikely. This analysis relies on a specified melt geometry in the lower head and considers only decay heat and stored energy. The possibility of a zirconium-iron exothermic interaction leading to vessel failure has not been adequately considered. The existence of such intermetallic exothermic reactions could alter l the severe accident picture for future analyses and should be further investigated.
The models and analyses used to develop this core degradation scenario have not been validated against experiments involving large volumes of molten metals and molten oxides.
The deficiencies of the core degradation modeling afflict both the likelihood of in-vessel retention of core debris and the susceptibility of the reactor to in-vessel steam explosions. -
The RASPLAV experimental activities supported by the NRC are not likely to resolve the most important issues of materialinteractions involving in-vessel retention. Results of these experiments will not be usefulin studying the effects of mixing a large volume of molten metal with hypostoichiometric reactor fuel.
Based on the results of the analysis, Westinghouse concludes that it is " physically unreasonable" for the vessel to be penetrated by molten core debris. The NRC staff, on the -
other hand, has concluded that the possibility of reactor vessel penetration cannot be excluded. We agree with the staffs conclusion.
Even discounting retention within the vessel and assuming containment vulnerability, the AP600 poses low risks to the public relative to existing reactors because the AP600 has quite a low core damage frequency and because the cavity will be flooded.
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Since in-vessel retention is widely considered to be an important accident management strategy for operating reactors, the impact of intermetallic exothermic reactions on this strategy should be assessed by the staff.
CONCLUSION As noted above, we have identified no additional issues that would prevent the certification of the AP600 design. We plan to complete our review of the AP600 design, including resolution of our previous concoms, at the July 1998 meeting. We continue to be concemed about the quality of test and analysis program documentation related to information needed to certify the AP600 design. The l staff should evaluate whether the quality of the AP600 documentation could withstand an NRC design-basis inspection.
Sincerely, R. L Scale Chairman
References:
- 1. Letter dated April 9,1998, from R. L Seale, Chairman, ACRS, to L Joseph Callan, Executive Director for Operations, NRC, subject: The Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Plant Design interim Letter 2.
- 2. Letter dated February 19,1998, from R. L Seale, Chairman, ACRS, to L Joseph Callan, Executive Director for Operations, NRC,
Subject:
Interim Letter on the Safety Aspects of the Westinghouse Electric Company Application for Certification of the AP600 Plant Design.
- 3. U. S. Department of Energy Report DE-AC03 90SF18495 dated June 26,1992, prepared by Westinghouse Electric Corporation, "AP600 Standard Safety Analysis Report," updated through Revision 22 (issued April 6,1998).
- 4. U. S. Department of Energy report DE AC03 90SF18495 dated June 26,19g2, prepared by Westinghouse Eiectric Company, "AP600 Probabilistic Risk Assessment," updated through Revision 11 (issued March 1998).
- 5. U.S. Nuclear Regulatory Commission," Advance Final Safety Evaluation Report Related to the Certification of the AP600 Design," dated May 1998 (Predecisional information). o
- 6. Westinghouse Electric Company, WCAP-14807 Revision 4, dated February 27,1998,
- NOTRUMP Final Validation Report for AP600"(Proprietary).
- 7. Memorandum dated March 13, 1998, from Brian A. McIntyre, Westinghouse Electric Corporation, to U. S. Nuclear Regulatory Commission, transmitting errata pages to WCAP-14807, NOTRUMP Final Validation Report for AP600, Revision 4.
- 8. Set of page changes to WCAP-14727: " Scaling and PIRT Closure Report", Volumes 1 and 2, to update the report to Revision 2. -
- 9. Memorandum dated March 2,1998, from Brian A. McIntyre, Westinghouse Electric Corporation, to U. S. Nuclear Regulatory Commission, transmitting additional material for incorporation into WCAP-14727, AP600 PXS Scaling and PlRT Closure Report, Revision 2, RAI Responses for Appendix A.
- 10. Westinghouse Electric Company, WCAP 14305, Revision 2, dated April 7,1998, "AP600 l Test Program ADS Phase B1 Test Analysis Report * (Proprietary).
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- 11. Westinghouse Electric Company, WCAP-14171, Revision 2, dated March 1998 "WCOBRA/ TRAC Applicability to AP600 Large Break Loss-of-Coolant Accident" (Propdetary). -
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- 12. Memorandum dated April 20, 1998, from Brian A. McIntyre, Westinghouse Electric Corporation, to Robert Seale, Chairman, ACRS, transmitting "Roadmap" of Westinghouse Responses to ACRS Concoms.
- 13. Letter dated April 28,1998, from Westinghouse Electric Corporation, to U. S. Nuclear Regulatory Commission,
Subject:
Revised Response to FSER Open item 440.796F, Part E, on the NOTRUMP Final Validation Report.
- 14. Westinghouse Electric Ccmpany, WCAP-14305, Revision 3, dated April 1998, "AP600 Test Program, ADS Phase B1 Test Analysis Report"(Proprietary).
- 15. Westinghouse Electric Company, WCAP-14776, Revision 4, dated March 1998, "WCOBRA/ TRAC OSU Long-Term Cooling Final Validation Report"(Proprietary).
- 16. Letter dated April 9,1998, from T. H. Essig. Office of Nuclear Reactor Regulation, NRC, to N. J. Uparulo, Westinghouse Electric Corporation,
Subject:
Documentation of Topical Report WCAP-12945(P)" Westinghouse Code Qualification Document for Best Estimate Loss of Coolant Analysis."
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