ML20203L193
| ML20203L193 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 02/25/1998 |
| From: | Dunn B, Swindlehurst G, Taylor Z DUKE POWER CO., FRAMATOME |
| To: | |
| Shared Package | |
| ML20203L159 | List: |
| References | |
| NUDOCS 9803050389 | |
| Download: ML20203L193 (29) | |
Text
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DUKE ENERGY CORPORATION DUKE ENGINEERING & SERVICES,INC.
YANKEE ATOMIC SMALL BREAK LOCA TECHNICAL REVIEW REPORT Technical Review Conducted For:
Duke Ennineerine & SenIces. Inc, Location:
Duke Ennineerine & Sen ices Inc. - Bolton. 51assachusetts j
i Scope:
Annlication of RELAP5YA in Small Break LOCA Analyses at hiaine Yankee Dates:
February 2 throuch Februnn 6.1998 PERFORh1ED BY:
Z. L. Taylor Duke Power Company G. B. Swindleburst Duke Power Company B. bl. Dunn Framatome Technologies L. E. Hochreiter Pennsylvania State University l
DATE: February 25, 1998 PREPARED BY:
G. B. Swindlehurs W
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A DATE: February 25, 1998
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Z. L Taylor 9903050389 900227 PDR ADOCK 05000029 W
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APPENDIX C TABI E OF CONTENTS Eut I. ' EX ECUTIVE S Uhihi ARY................................................. C-1
- 11. IN VESTI O ATI ON....................................................... C-2 A.
List Of Technical Review Team hiembers................................ C 2 i
B.
List Of Personnel Contacted During The Investigation....................... C 3 C.
Entrance hiceting And Exit Debriefing................................. C-3 D.
Technical Re view Process............................................ C-4 E.
B ac k g ro u n d....................................................... C 5 III. REPORT DETAILS...................................................... C 6 A.
Yankee Atomic SBLOCA hiethods Development History.................... C-6 B.
Maine Yankee SBLOCA Licensing Basis History.......................... C-6 C.
hiaine Yankee SBLOCA Analysis Comparisons........................... C 8 D.
Review Team Observations and Conclusions
............................C10 IV. TECHNICAL REVIEW TEAhi'S CONCLUSIONS ON ISSUES IDENTIFIED IN NRC'S DEhf AND FOR INFORhf ATION................................... C-17 A.
Summary of Apparent Violation as Stated in Demand wr Information.......... C-17 B.
Summary of Apparent Violation as Stated in Demand for Information.......... C-19 C.
Summary of Apparent Violation as Stated in Demand for Information.......... C-21 D.
Summary of Apparent Violation as Stated in Demand for Information.......... C-22
= = = = =.
C-li Y
L EXECUTIVE SUhthf ARY During the period of February 2 through February 6,1998, a Duke Engineering & Services, Inc.
l (DE&S) sponsored technical review of activities at the DE&S (formerly Yankee Atomic) Bolton office was conducted by a team composed of industry experts in the area of LOCA analysis. It was the purpose of this review to provide an independent technical review of Yankee Atomic's small-break LOCA evaluation model(RELAP5YA SBLOCA) as applied to hiaine Yankee and to determine compliance with 10CFR50.46 and 10CFR50 Appendix K requirements. The Review Team also evaluated Yankee Atomic's cognizanec of the potential technicalissues with the model, adequacy of corrective actions that may have been required, and effectiveness of communication of these with hiaine Yankee and the Nuclear Regulatory Commission.
During the course of the review numerous technical documents and memoranda were examined.
in addition, interviews were conducted with key Yankee Atomic staff members involved in the development of RELAPSYA and its application to hiaine Yankee. Overall, based on the information available to the Review Team, including other analysis results, it was concluded that the results of Yankee Atomic's SBLOCA calculations for hiaine Yankee are consistent with resuhs of some other calculations by other codes and organizations. However, Yankee Atomic failed to fully appreciate NRC expectations, as understood within the nuclear industry's LOCA community, for 10CFR50.46 model acceptance. Although these expectations are not completely documented in specific references, such as the NRC Standard Review Plan or Regulatory Guides, they have been clearly communicated to the industry's LOCA community through extensive and numerous interactions since the Interim Acceptance Criteria of 1971. The interactions have made it clear to what degree the community was to keep the NRC informed, what types ofissues needed NRC approval, and what the NRC expectations were for approved evaluation models. Yankee Atomic would have benefited from the experience of companies or individuals who more fully participated in the process. The Review Team recognizes that it remains Yankee Atomic's responsibility to understand NRC's expectations. Yankee Atomic would also have benefited from improved, continuous and direct technical communications with the NRC Reactor Systems Branch, which could have eliminated many of the issues of concern as identified by the Review Team.
Overall observations and concluPns of the ter' are summarized as follows:
1.
There was a failure of the Yankee its organization as well as the LOCA Group in fully understanding NRC expectations un particular the NRC's Reactor Systems Branch) for an Appendix K evaluation model, as understood within the industry LOCA community, for 10CFR50.46 model acceptance. Plant specific calculations were undertaken too late to change the direction of the development project following the identification of significant code limitations.
2.
During the development of the hiaine Yankee specific SBLOCA model and analyses there were a number of organizational barriers and a culture which impeded the LOCA Group from completing the project in a manner which met NRC expectations.
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3.
The Yankee Atomic organization was aware of performance problems with the RELAPSYA code. These code problems were addressed by the LOCA Group from a technical perspective, however the organization did not adequately maintain communications with the NRC technical staff. This resulted in a failure to resolve all technicalissues necessary to satisfy NRC expectations prior to implementation for the Cycle 14 reload.
4.
The technical capabilities of the Yankee Atomic LOCA Group are acceptable for performing LOCA analysis. The LOCA Group was qualified and knowledgeable l
regarding SBLOCA computer models and phenomena. Interviews and documents did l
not indicate any evidence of deliberate violations of NRC regulations.
5.
Based on the information available, including other analysis results. it can be concluded j
4 that the results of Yankee Atomic's SBLOCA calculations for hiaine Yankee are consistent with results of some other calculations by other codes and organizations.
Therefore it is concluded that the SBLOCA PCTs for hiaine Yankee meet the 2200*F criterion of 10CFR50.46, and that SBLOCAs remain bounded by LBLOCAs for this plant.
6.
Yankee Atomic's direct interface with the NRC on the application of the RELAPSYA model appears to have ended once the SER was received in 1989. Not maintaining a direct interface between the LOCA Group and the NRC technical staff resulted in a loss of communications regarding NRC's expectations for SBLOCA calculations performed with RELAP5YA to support the hiaine Yankee license.
7.
Continuous communications with the NRC staff could have eliminated many of the issues of concern.
A more detailed discussion on each of the above observations and conclusions is provided in Section Ill.D of this report.
II. INVESTIGATION A.
List Of Technical Review Team hiembers Z. L Taylor Duke Power Company 18 years experience with Duke Power Company as (Team Leader)
Engineering Supervisor and as Licensing hianager at 4
Catawba Nuclear Station Currently Nuclear Support hianager in Duke Power's Nuclear Assessment and Issues Division G. B. Swindlehurst Duke Power Company C2
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l 20 years at Duke Power Company in Safety Analysis area
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which includes non.LOCA accident analysis Currently Manager of Duke Power's Safety Analysis Group in the Nuclear Engineering Division 4
B. M. Dunn Advisory Engineer at Framatome Technologies (formerly B&W) 29 years of LOCA analysis experience J
t L E. Hochreiter Pennsylvania State University Greater than 25 years experience in LOCA analysis with Westinghouse Currently Professor of Nuclear Engineering at Pennsylvania State University Assisted By:
J. M. McGarry Attorney at Law with the firm of Winston & Strawn M. J. Wetterhahn Attorney at Law with the firm of Winston & Strawn B.
I Ist Of Personnel Contacted Durine The Investication Michael Scott Engineer DE&SNAEC Steve Schultz Mgr N&F Serv.
DE&S/YAEC l
Bob Harvey Engineer DE&SNAEC Ramu Sundaram Engineering Consultant DE&SNAEC Liliane Schor Project Manager II DE&SNAEC Bruce Slifer Engineer DE&S/YAEC Jim Chapman Gen Mgr Nuclear Engr DE&S/YAEC l
Mark Robeck Attorney at Law Baker & Botts Glenn Swanbon Engineer Northeast Utilities l
Pete Anderson Maine Yankee Interface DE&S/YAEC l
Mike Lennon Attorney at Law Baker & Botts C.
Entrance Meetine And Exit Debriefine l
l Entrance Meetinc-February 2,1998 L
i Z. L. Taylor Nuclear Support Mgr NAID Duke Power Co.
[
G. B. Swindlehurst Mgr Safety Analysis Duke Power Co.
J. M. McGarry Attorney Winston & Strawa M. J. Wetterhahn Attorney Winston & Strawn L E. Hochreiter Professor of Nuclear Engr.
Penn State University B M. Dunn Advisory Engineer Framatome Technologies Greg Hudson Project Manager DE&S cmmu n.
C-3
Jim Chapman Gen higr Nuclear Engr DE&S/YAEC Paul Bergeron higt THSAG DE&S/YAEC Steve Schultz higt N&F Serv.
DE&S/YAEC Bob Harvey Engineer DE&S/YAEC Chris Hickey Senior QA Engr BECo.
Liliane Schor Project hianager 11 DE&S/YAEC Ramu Sundaram Engineering Consultant DE&S/YAEC C. D. Nunzio Senior QA Engr DE&S/YAEC Wayne hierritt Project higr DE&S/YAEC hiike 1.cnnon Attorney Baker & Botts (Houston) hiike Scott Engineer DE&S/YAEC Exit Debriefing: February 6,1998 Z. L. Taylor Nuclear Support higt NAID Duke Power Co.
G. B. Swindlehurst higt Safety \\nalysis Duke Power Co.
Greg Hudson Project hianager DEES Steve Schultz higt N&F Serv.
DE&S/YAEC Pete Richardson Independent Assess.
NAES Co.
Jim Chapman Gen higr Nucler Engr DE&S/YAEC hiike Scott Engineer DE&S/YAEC Frank Sabadin QA DE&S/YAEC Bob Harvey Engineer DE&S/YAEC Paul Bergeron higt THS AG DE&S/YAEC Rich Faix Engr. Supervisor NAES Co.
hiark Robeck Attorney Baker & Botts Liliane Schor Project hianager II DE&S/YAEC Wayne hierritt Pioject hianager DE&S/YAEC Camillo A. DiNunzio Sr, QA Engr.
DE&S/YAEC Steve C. White Ops QA higr.
DE&S/YAEC D.
Technical Review Process Copies of technical documents were reviewed prior to the team's arrival at the DE&S Bolton office.
An entrance meeting was held with appropriate DE&S management to discuss the technical review plan for the week. Refer to Section C of the report for list of attendees.
Inteniews were conducted with key personnelinvolved in the development of the RELAPSYA SB LOCA evaluation model, its application to hiaine Yankee, and the best estimate approach for addressing NUREG-0737 and related 10CFR50.59 issues. These inteniews were held with alllevels of the Nuclear Engineering Department organization in place at the time of this work, including director level, LOCA Group laanagers, lead mammaa C-4
engineers, as well as other engineers involved with the technical aspects of the work.
Technical work reviewed included (but was not limited to):
The RELAPSYA SBLOCA (YAEC-1300P)
The best-estimate approach being considered for addressing NUREG 0737, items !!.K.3.30 and II.K.31 Evaluations performed at the request of Maine Yankee for a reduced steam generator pressure Maine Yankee-specific SBLOCA application of RELAP5YA (YAEC-1868)
Upon completion of the review and interviews, the team discussed a number of questions drafted by the law firm of Winston & Strawn to ensure that specific concerns identified in the NRC's Demand For Information (DFI) letter had been addressed.
Overall observations and conclusions were then formulated by the team rCating to the development of the RELAPSYA model and subsequent work on and application of the evaluation model to Maine Yankee.
At the conclusion of the review, an exit debriefing was conducted by the Review Team Leader to present preliminary results of the review. Refer to Section C of the report for list of attendees.
t E.
Backgmund in the NRC's DFI letter addressed to Duke Engineering & Services, inclYankee Atomic Electric dated December 19,1997, the NRC states that engineering services provided to Maine Yankee for SBLOCA analysis were not in compliance with regulatory requirements. The NRC also states that Yankee Atomic, acting as a vendor for Maine Yankee, performed the inadequate analyses. This technical review was chartered to provide an independent review of the adequacy and compliance with 10CFR50.46 and 10CFR50 Appendix K of the subject evaluation model at the time of development and application to Maine Yankee, evaluate the cognizance of Yankee Atom!c personnel of the modellimitations, determine the appropriateness and timeliness of corrective actions that may have been taken, and to determine whether or not the modellimitations were effectively communicated to Maine Yankee and the NRC.
The Review Team worked in conjunction with two attorneys from Winston & Strawn in providing input for DE&S's response to paragraph IV.A of the NRC's DFl. The Review Team also assisted the Winston & Strawn attorneys in their investigation in support of their response to paragraph IV.B of the DFl.
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j I11. REPORT DETAILS A.
Yankee Atomic SBLOCA hiethods Development Historv Yankee Atomic initiated the RELAPS/hiODI, now called RELAP5YA, SBLOCA methods development project in 1980 to broaden its analytical capabilities to support their utility customers and Yankee Atomic's Rowe plant. Yankee Atomic was already performing non LOCA analyses for Yankee Rowe, Vermont Yankee, and hiaine Yankee. LBLOCA analyses based on the WREhi package were also being performed as a result of a collaborative effort with Exxon Nuclear (now Siemens) as part of the transition to Exxon fuel supply for hiaine Yankee. The RELAPS program was a natural extension of the existing Yankee Atomic analysis capability, and was an attempt to upgrade to what Yankee Atomic considered the state-of the-art SBLOCA code.
An additional consideration for developing these analytical capabilities in house, rather than to continue to purchase analyses from vendors, was the uniqueness of the plant designs. Rowe in particular was a one of a-kind plant, and consequently there existed imique analysis requirements. Obtaining anilyses from vendors over the life of the plant was also recognized as a significant economic consideration.
i Yankee Atomic's decisions involving the development ofin house analysis capability were similar to those of many nuclear utilities. However, undertaking Appendix K LOCA methods was a very aggressive and challenging undertaking for any utility or engineering service organization. The decision by a utility to take SBLOCA evaluation model development responsibilities in house has only been successfully completed by one other utility in the industry.
B.
h1ainc Yankee SBLOCA Licensine Basis Historv Original SBLOCA Licensing Basis
- The original Final Safety Analysis Report (FSAR) Chapter 14 SBLOCA licensing basis t
cakulations for hiaine Yankee at 2440 htWt were performed by Combustion Engineering (CE). The LOCA break spectrum analyzed included large breaks, which 2
were defined as the largest double ended break down to 1.0 ft, and a different method for intermediate and small breaks. The SBLOCAs (intermediate and small breaks) were i
simulated with the BARF II code for a break spectrum which included 0.55,0.2,0.1, 2
0.049, and 0.0218 ft break sizes. The BARF-II code represented the Reactor Coolant System (RCS) primary with a two node model. For break sizes below approximately 2
0.07 ft no core uncovery was predicted. For the larger break sizes the specific peak cladding temperatures (PCTs) were not given, but were summarized as not exceeding 2000*F.
Revised SBLOCA Licensing Basis (1971)
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The original SBLOCA analysis was revised in 1971 with analyses performed by CE with the CEFLASH-4, PERC, and STRIKIN codes for the licensed power level of 2440 MWt. The methodology uses a multi-node representation of the RCS. In the document describing these re-analyses, the small break spectrum is described as " breaks with an area less than 0.5 ft:," and the break spectrum analyzed consisted of 0.5,0.2 and 2
0.1 ft breaks. The PCTs were 1702*F for the 0.5 ft break and 985'F for the 0.2 ft2 break. The core did not uncover for the 0.1 ft: break size. For the break sizes which result in core uncovery the accumulator injection refloods the core and limits the PCT to acceptable values. Along with the LBLOCA analyses, these results met the requirements of the AEC Interim Acceptance Criteria for Emergency Core Cooling Systems.
Revised SBLOCA Licensing Basis (1977)
A revised SBLOCA spectrum analysis was performed in 1977 by CE using the 1974 l
CEFLASH 4AS model at a power level of 2630 MWt and a peak linear heat rate of 16.5 kW/ft. The purpose of these reanalyses was to establish a break spectrum which was consistent with the technical specification kW/ft limits. Prior analyses had used a lower kW/ft value for SBLOCAs. Break sizes of 0.5,0.3, and 0.1 ft: were analyzed. The PCT results were 1348'F (0.5 ft ),1222 (0.3ft:), and 1176 (0.1 ft ). T2 0.5 ft: case is 2
mitigated mainly by the injection from the accumulators. The 0.3 ft case is mitigated mainly by the HPSI pump, with the termination of the analyses justified by accumulator 2
injection occurring. The 0.1 ft case is mitigated solely by HPSI pump injection. The-analyses were implemented with the Cycle 4 reload.
Maine Yankee SBLOCA Licensing Analvses Performed Bv Yankee Atomic Electric Comnany Yankee Atomic submitted YAEC-1300P to the NRC for review in January 1983. This topical report describes the models and correlations in the RELAP5YA code used for LOCA analysis per 10CFR50 Appendix K. One of the objectives of this submittal was to respond to NUREG 0737, Item II.K.3.30 for Maine Yankee. While the NRC review of YAEC-1300P was in progress, the RELAP5YA code was applied by Yankee Atomic in an April 1984 submittal to the NRC to resolve the SBLOCA Reac:or Coolant Pump (RCP) trip issue for Maine Yankee. The NRC issued an SER on the RCP trip study in April 1986. The SER on the YAEC-1300P topical report was issued on January 30, 1909. This SER included twelve conditions, among which was a requirement to submit future plant specific licensing applications. By letter dated May 8,1989, the NRC recognized that its SER closed out II.K.3.30. Although no submittal of plant applications had been made, the same letter closes out II.K.3.31. The letter stated that implementation may be the subject of a future inspection by NRC.
Application of the RELAP5YA code as described in YAEC-1300P to Maine Yar kee was completed in l993. The results of these analyses are summarized in YAEC-1868 2
dated June 1993. A break spectrum of 0.10,0.15,0.20,0.25,0.30, and 0.35 ft was C-7 sumow m r---r.-m y.
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analyzed. The 0.35 ft case was not able to be run through to the final PCT. The final PCTs givea in the report for the 0.35 ft: breaks are estimated based on a 58*F APCT obtained from a 0.30 ft break run, which was added to the calculated PCT at the time of-safety injection tank actuation. The limiting PCT result was determined to be 1887*F for 2
the 0.15 ft break. PCTs for smaller and larger breaks were lower, thereby indicating a local PCT peak within the small break range. This report was not submitted to the NRC for review. Maine Yankee and Yankee Atomic have taken the position based on the May 8,1989 NRC letter that submittal was no longer required, and that having the documents on file for NRC inspection was sufticient.
Implementation of the YAEC-1868 analysis results occurred with the Cycle 14 reload.
The Core Performance Analysis Report for Cycle 14, which describes the RELAP5YA-SBLOCA analysis results, was submitted to the NRC. This reload submittal replaced the 1977 Cycle 4 CE SBLOCA analyses with the YAEC analyses as the licensing basis am. lyses of record.
C.
Maine Yankee SBLOCA Analvsis Comnarisons Licensing Basis Analyses Bv CE
' The SBLOCA analyses that established the licensing basis prior to Cycle 14 were performed by CE. The original licensing basis analyses were performed with a two node model and are not considered to be relevant. The revised SBLOCA spectrum analysis performed in 1977 used the 1974 CEFLASH-4AS evaluation model at a power level of 2630 MWt and a peak linear heat rate of 16.5 kW/ft. Break sizes of 0.5,0.3, and 0.1 ft2 were analyzed. The PCT results were 1348*F (0.5 ft:),1222*F (0.3 ft:), and ll76*F (0.1 ft:). These analyses were implemented with the Cycle 4 reload. The PCT vs. break size trend increases with break size and does not indicate a local PCT peak within the small break range.
1996 Analyses Bv Si m-ns Power Cornoration Siemens was contracted by Maine Yankee in 1996 to supply a SBLOCA analysis with the intent oflifting the NRC restriction to 90% full power operation. Siemens uses an l
NRC-approved Appendix K evaluation model that is based on RELAP5/ MOD 2.
Siemens ran break sizes of 0.05,0.10,0.15,0.20,0.25, and 0.61 ft. The resulting PCTs 2
2 were 1704*F(0.10 ft ),1522*F(0.15 ft ),1363*F(0.20 ft ), i157*F(0.25 ft:), and 2
1382*F (0.61 ft ). The core did not uncover for the 0.05 ft: case. Compared to the Yankee Atomic results in YAEC-1868, the Siemens results show a shift in the worst break size from 0.15 to 0.10, and a decrer.se in the limiting PCT from 1887'F to 1704*F. The Siemens results also show one local PCT peak, whereas the Yankee Atomic results show a second smalllocal PCT peak. No comparison can be made at the 0.61 ft: break size since ti!e Yankee Atomic model was unable to analyze that case. It is ummow.co.
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l noted that the f o r a prediction for 0.61 ft: does indicate an increase in the PCT relative to the 0o Y break.
1996 Analyses By ABB I
ABB was contracted by Maine Yankee in 1996 to supply SBLOCA analyses for Cycle 15 at a power level of 2440 MWt and 14.1 kW/ft. ABB used their 1977 Supplement 1 H
NRC-approved SBLOCA evaluation model. The break spectrum analyzed included 0.10,0.15,0.20, and 0.50 ft: break sizes. This range of break sizes is consistent with the ABB ECCS performai.ce evaluation methodology. The PCT results are 1134.7'F (0.10 2
2 ft ),1405.0*F (0.15 ft ),1290.9'F (0.20 ft ), and 1520.7*F (0.50 ft ). The limiting break size remains the same as the 1977 CE analysis, but a local peak at 0.15 ft is new to this reanalysis. It is noted that many of the assumptions in the reanalysis have changed and contribute to the differences with the 1977 analysis. The latest ABB methods do not agree with either the Yankee Atomic or the Siemens analyses in terms of the limiting break size.
WREM-Based Analysis By Yankee Atomic (1977)
In support of Cycle 5 Yankee Atomic performed an analysis of the 0.5 ft break with the NRC-approved WREM based LBLOCA evaluation model. This analysis was submitted to the NRC witt. ne Cycle 5 technical specification change package, which was approved by the NRC. The analysis was performed to show that the transition to Exxon (Siemens) fuel did not adversely affect the SBLOCA analysis of record. The 0.5 ft break was the limiting break size for the CE licensing basis analysis of record. The WREM based tnalysis is characterized as a " demonstration analysis" in the licensing documentation, and was not intended to replace the CE analysis. The result of the analysis at 16.0 kW/ft was a PCT of 1229.9*F, therefore indicating that the larger SBLOCA break sizes were less limiting.
RELAP5/ MOD 2 & MOD 3 Analyses By Yankee Atomic Yankee Atomic has analyzed SBLOCAs for Maine Yankee with the more advanced RELAP5/ MOD 2 and MOD 3 codes. In 1990 the Yankee LOCA staff gained experience by running the Siemens Appendix K SBLOCA evaluation model version of RELAP5/ MOD 2. This experience provided modeling guidance related to the ECCS bypass pnenomenon. The Siemens model did not calculate the excessive ECCS bypass as was occurring in the Yankee Atomic model. The RELAP5/ MOD 3 version used by Yankee Atomic does not include the Appendix K models and has not been submitted for NRC review, but the results are useful for comparison. These analyses were important mainly in that they demonstrated that the PCT occurred shortly after accumulator injection, and then steadily decreased. This information supported the Yankee Atomic interpretation that the unstable RELAP5YA behavior following accumulator injection samowuos C-9
i was non-physical, and that the unstable PCT behavior following accumulator injection was not significant.
Summarv of Comparisons All of he SBLOCA analyses confirm that the LBLOCA PCTs are limiting. The Yankee t
Atomic, Siemens, and 1996 ABB methods all predict a local peak in the range of 0.10 to 2
0.15 ft. The ABB analyses and the CE analyses with both the 1974 and 1977 evaluation 2
models predict that the PCT is highest for the 0.50 ft break. The Siemens model 2
predicts that the PCT increases as the break size approaches 0.61 ft, but that it is non-limiting relative to the local PCT peak at 0.10 ft. This trend cannot be confirmed from the Yankee Atomic RELAP5YA analyses due to no breaks being analyzed from 0.35 to 0.50 ft: due to either codc execution difficulties or no cases having been run.
2 The Yankee Atomic position is that the 0.5 ft break PCT is bounded by the 0.15 ft PCT of 1887'F. Given the information provided and based on the trends of other analyses, the Review Team was unable to draw w. definitive conclusion regarding the RELAP5YA PCTs for the unanalyzed portion of the Maine Yankee SBLOCA spectrum.
However, based on all the available calculations, it is concluded that the SBLOCA PCTs for all of these analyses meet the 10CFR50.46 2200*F criterion, and that SBLOCAs remain bounded by LBLOCAs.
D.
Review Team Observations and Conclusions I.
There was afailure of the Yankee Atomic organi.ation as wellas the LOCA Group in fully understanding NRC expectations (in particular the NRC's Reactor Systems Branch)for an Appendix K cvaluation model, as understood within the industry LOCA community,for 10CFR50.46 model acceptance.
Plant specific calculations were undertaken too late to change the direction of the developmcnt projectfollowing the identification of significant code limitations.
The YAEC-1300P RELAP5YA SBLOCA code was approved by the NRC in 1939.
However, the development and NRC approval process was not without risk in that no plant applications were performed in parallel with the development of the code and model. It is the standard industry practice that the plant applications be performed to identify any problems prior to submitting the code and evaluation model for NRC review. Limiting the testing and validation of the model to scaled test facilities does not challenge the method sufficiently, since some code and model problems willinvariab!y arise with the first plant applications. Any applications should be completed utilizing a process that meets NRC's expectations prior to any implementation in the plant's licensing basis The Review Team's understanding of NRC's expectations is based on many years of LOCA-related licensing interactions.
These expectations are evident in the Demand for Information Letter and in the C-10 s m.mmu ms
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transcript of the December 18,1995 meeting between the NRC and Yankee Atomic.
The NRC's intent regarding the conditions attached to the YAEC-1300P SER was to ensure that plant applications of RELAP5YA would be reviewed and approved I
prior to implementation. This is a standard process for approving LOCA methods applications. Variations on this process are possible, but are subject to the approval of the NRC. Yankee Atomic should have realized that this step in the approval process was the standard industry practice despite the NRC letter dated May 8,1989.
Yankee Atomic failed to bring to the attention of the NRC several significant problems that were encountered in the application of the approved YAEC-1300P method to Maine Yankee. The code results were unstable following accumulator injection for much of the break spectrum. The model could not run through the historical break spectrum up to 0.5 ft, in particular when considering that the 2
limiting PCT case for the current licensing basis was 0.5 ft. The NRC expects that the approved evaluation model should be capable of running any case should the need arise. The Review Team acknowledges and agrees with the NRC position.
Partial break spectrums are intended to reduce the number of required LOCA calculations when agreement can be reached that all of the critical challenges to the ECCS system have been covered. The reduced spectrum is not intended to provide for inadequacies in tha maly:is methods. An exception could be made but only with NRC concurrence Yankee Atomic asserts that its break spectrum was sufficient to meet the licensing requirements. The Review Team acknowledges, that partial break spectrums are licensable; however, the method must be capable of calculating all break sizes, in the spectrum. Assuming rather than confirming that the Yankee Atomic break spectrum was acceptable, was a failure in meeting NRC expectations on the part of Yankee Atomic.
Excessiu ECCS bypass (a code deficiency) was occurring in the SBLOCA analyses. To address t!c excessive ECCS bypass, Yankee Atomic used a large non-physicalloss coefficient for the junction between the two reactor vessel downcomer volumes. Yankee Atomic considers this large non-physical loss coefficient to be an input to the code rather than a model change. The Review Team disagrees with Yankee Atomic's rationale. The change comprises a manipulation of the intended use of the input parameter in that it is altered sufficiently to nullify the original intent of the parameter as described in the code documentation. The Review Team considers this to be a modeling change that should have been discussed with the NRC before application.
Application of the unapproved best-estimate RELAP5YA methodology.o the steam generator pressure reduction issue without NRC concurrence was improper.
The Yankee Atomic LOCA Group staff applied the best estimate methodology i m m o w nras C-Il
since it was the only available in-house analysis tool. This was done without a LOCA Group staff or Yankee Atomic management awareness or understanding of the NRC approval status of the best-estimate model, the Maine Yankee licensing basis, or NRC expectations.
The above shortcomings in Yankee Atomic's program for performing and implementing the Maine Yankee SBLOCA application are all related to an unfamiliarity with the NRC's expectations for the LOCA analysis approval process as understood by the industry LOCA community. Although these expectations are not completely documented in specific references, such as the NRC Standard Review Plan or Regulatory Guides, they have been clearly communicated to the industry's LOCA community through extensive and numerous interactions with the NRC (the Reactor Systems tranch, the Advisory Committee on Reactor Safeguards (ACRS), and others) since the Interim Acceptance Criteria of 1971.
These interactions have made it clear to what degree the community was to keep the NRC informed, what types ofissues needed NRC approval, and whr :he NRC expectations were for approved evaluation models. Yankee Atomic would have benefited from the experience of companies or individuals who more fully l
participated in the process. The Review Team recognizes that it remains Yankee l
Atomic's responsibility to understand NRC's expectations. Yankee Atomic would also have benefited from improved, continuous and direct technical communications with the NRC Reactor Systems Branch, which could have eliminated many of the issues of concern as identified by the Review Team. This situation could have been avoided with sustained communications and developing a continual working relationship with the NRC technical staff, or by consulting with industry LOCA experts. The Review Team does note that Yankee Atomic has apparently been successfulin licensing LOCA analysis methods for Yankee Rowe and Vermont Yankee.
2.
During the development of the Maine Yankee specific SBLOCA model and analyses there were a number of organi:.ational barriers and a culture which knpeded the LOCA Groupfrom completing the project in a manner which met NRC expectations.
Root causes for the problems that occurred during the development of the Maine Yankee specific SBLOCA model and analyses are associated with organizational barriers and the culture in the Yankee Atomic organization. Although Yankee Atomic upper management was aware of the RELAP5YA SBLOCA development project and technical and licensing issues, there was insufficient upper management involvement and direction. Due to its high visibility and legal status in the NRC regulations, a LOCA development program require.s a very significant resource commitment which includes a sustained critical mass of expertise and continuity of key personnel. In the Review Team's opinion, Yankee Atomic management did not obtain sufficient experience and depth in the staffing of the RELAP5YA project emmo ucs C-12 U
team, in particular at the beginning of the project in the early 1980's. Few of the staff had any prior LOCA analysis experience. This comment is bas:d on the knowledge of the resources that other LOCA analysis organizations employ on similar projects. The Review Team notes that Intermountain Technologies Inc. was contracted with to install many of the Appendix K models into RELAP5YA. The REL A P5YA project team was working in relative isolation from the LOCA community with a code that only one other organization was applying for licensing-basis SBLOCA analysis. The culture at Yankee Atomic, which appears to prefer to be self sufficient and independent in providing engineering services, contributed to the isolation of the LOCA Group and the failure to r.ppreciate the LOCA licensing implications. No benchmarking to other industry LOCA organizations occurred. A final contributing factor was computer resource limitations. Prior to approximately 1992 the LOCA engineers were unable to run all of the desired cases to fully develop and test the code and models, and the execution time was too slow for production work.
With the benefit of hindsight, the original RELAP5/ MODI code limitations presented too great a technical challenge to the Yankee LOCA group in achieving an Appendix K model that could be used with confidence. The project was further challenged by the resource limitations, ineffective management, and the independent culture at Yankee Atomic. These problems appear to be limited to the RELAPSYA SBLOCA application to Maine Yankee, as these problems did not surface on the RELAP5YA licensing for Rowe and Vermont Yankee. This comment is based on the Review Team's understanding of the thorough inspections that have occurred following the allegations in December 1995, and the absence of any other significant problems.
3.
The l'ankee Atomic organization was aware ofperformanceproblems with the RELAPS1'A code. These codeproblems were addressed by the LOCA Group from a technicalperspective, however the organization did not adequately maintain communications with the NRC technical staff. This resulted in a failure to resolve all technicalissues necessary to satisfy NRC expectations prior to implementation for the Cycle 14 reload.
The RELAP5/ MODI code, which was the starting point for the RELAP5YA version, has inherent limitations which complicate the application for an Appendix K SBLOCA evaluation model suitable for NRC approval. Long-standing code deficiencies were identified during the project. The Review Team considers completely addressing these code limitations to be a significant challenge to the Yankee Atomic LOCA Group. RELAP5YA's 5-equation formulation does not appropriately represent non-equilibrium behavior. This formulation lags the state-of-the-art and can result in excessive condensation and vaporization. To compensate for this deficiency an artificially high (200*F) ECCS water temperature was introduced. Excessive prediction of interfacial drag caused non-physical ECCS
-mmas C-13 J
bypass that severely hindered SBLOCA modeling. The drag model was improved l
b " the problems persisted for the larger break sizes. In a further a' tempt to address l
- ae ECCS bypass problem a very large non-physical loss coefficient was introduced between the two volumes representing a split reactor vessel downcomer. This approach was supported by Yankee Atomic by verification t experiments with relevant data. The Review Team considers this approach to i m acceptable compensation for a RELAPSYA code deficiency, and the ust a value of 600 l
obtained an amount of ECCS penetration that was consistent with industry experience. The preferred approach would have been to implement the ECCS penetration modeling as a code change. The Review Team also considers this modeling approach, whether termed a model change or an input change, to be significant and should have been discussed with the NRC, Yankee Atomic also experienced large changes in RELAP5YA results with only small changes in inputs. The code was excessively sensitive and PCTs were inconsistent. The code and model development project was concluded without completely addressing the underlying code limitations. It was decided at least annually to continue with the licensing of RELAP5YA given these code and modeling limitations, and then to implement the results documented in YAEC-1868 without NRC review. The Review Team questions the advisability of these decisions given the expectations of the NRC in the LOCA arena. It is noted that there was no realinappropriate use of YAEC-1300P prior to the implementation c f the methodology for Cycle 14. Continuction of the communications with the NRC technical staff that were established and functional during the review of YAEC-1300P could have been a path towards acceptable resolution of these issues.
Since implementation for Cycle 14 proceeded without the capability to analyze the full SBLOCA break spectrum, NRC expectations as understood by the industry LOCA community were not met. Yankee Atomic's position that their ana!.
break spectrum was complete in that it bounded the limiting SBLOCA PCT u understandable, but is not consistent with standard industry practice since all break sizes could not be calculated and did not address the limiting break size for the analysis of record.
4.
The technical capabilities of the Yankee Atomic LOCA Group are acceptablefor performing LOCA analysis. The LOCA Group was qualified and knowledgeable regarding SBLOCA computer models andphenomena. Interviews and documents did not indicate any evidence ofdeliberate violations of NRC regulations.
Based on interviews and documentation reviews, the Review Team concludes that the Yankee Atomic LOCA Group is knowledgeable about SBLOCA computer models and phenomena, and that the staff are technically qualified. The project included extensive assessment of RELAP5YA to scaled test facility SBLOCA data.
. mm,,ues C-14 l
The LOCA Group staff interviewed are sincere, honest, open, and cooperative.
Throughout the history of the RELAP5YA project the staff has been conscientious l
and hardworking, and had good intentions of establishing independent analysis capabilities to broaden their support of their customers. It was also apparent to the Review Team that the LOCA Group staff was not expert in LOCA analysis-related his'ory, precedent, and licensing processes. No evidence of deliberate violations of NRC regulations were identified.
S.
Based on I 'se information available, including other analysis results, it can be concluded that the results of l'ankee Atomic's SBLOCA calculationsfor Maine l'ankee a e consistent with the results of some other calculations by other codes and orga.ti:.ations. Therefore it is concluded that the SBLOCA PCTsfor Maine Yankee meet the 2200 *F criterion of10CFR50.46, and that SBLOCAs remain bounded by LBLOCAsfor thisplant.
The results of the Maine Yankee RELAP5YA analyses documented in YAEC-1868 include instabilities related to the limitations of the code. These instabilities mainly occur following accumulator injection, which occurs after a significant uncovery and heatup of the cladding. Yankee Atomic has concluded that the predicted PCTs are conserutive, and that the instabilities do not affect the prediction of the PCT for the hniting break size (1887'F for the 0.15 ft break). Yankee Atomic has also concludeo that the inability to calculate the larger break sizes up to 0.50 ft is not significant since the larger break sizes are non-limiting. The results of other calculations have been evaluated by the Review Team. The prior SBLOCA analyses performed by CE in 1977 with the 1974 SBLOCA evaluation model have lower PCTs, with an increasing PCT with break size trend (1348'F for 0.50 ft breakt The 1996 comparison calculations performed by ABB with the 1977 SBLOCA evaluation model predict a PCT of 1520.7'F for the 0.50 ft break. The ABB 2
analysis also showed a local PCT peak of 1405*F at 0.15 ft. It is noted that the 2
analysis inputs for the 1977 CE and 1996 ABB analysis do not match the Yankee Atomic or the Siemens analyses, but are still useful for comparison purposes. The 1990 comparison calculations performed by Siemens show a PCT of 1704*F for the 2
2 0.10 ft break, and a PCT of 1382*F and an increasing PCT trend at 0.61 ft. Other Maine Yankee SBLOCA analyses were performed by Yankee Atomic in 1977 i; sing the WREM-based model, in 1990 using the Siemens RELAP5/ MOD 2 model, and in 1995 using RELAP5/ MOD 3. The WREM-based demonstration calculation indicated a low PCT relative to LBLOCAs. Experience gained with the Siemeas model helped address the ECCS bypass problem. The RELAP5/ MOD 3 best estimate analyses, although not NRC approved, predicted a SBLOCA transient response which is similar to RELAP5YA following accumulator injection. The comparison of all of the SBLOCA results for Maine Yankee, including the Yankee Atomic RELAP5YA results, indicates that the PCTs for SBLOCA are less than 2200 F, and that SBLOCAs are bounded by LBLOCAs. Therefore there was no reduction in the safety of the plant.
mamms C-15 l
6.
l'ankee Atomic's direct interface nith the NRC on the application of the l
RELAPSl'A modelappears to have ended once the SER uas received in 1989.
l Not maintaining direct a interface betu'een the LOCA Group and the NRC technical staff resulted in a loss of communications regarding NRC's expectationsfor SBLOCA calculations performed uith RELAPSl'A to support the Maine l'ankee license.
Communications with the NRC during the NRC review of the YAE-1300P topical report appear to be adequate, since NRC approval was eventually obtained. The Review Team notes that the large number of conditions in the SER and the substantial nature of the condition; are not positive or desirable results following a multi-year NRC review, and indicates, in part, code weaknesses. Yankee Atomic then developed responses to the conditions and documented them internally. The Maine Yankee applications completed in 1993 and documented in YAEC-1868 were not submitted for NRC review. This decision was based on the May 8.1989 NRC letter which closed out NUREG-0737 Item II.K.3.31. Although this NRC letter conflicted with the YAEC-1300P SER conditions requiring plant applications to be submitted, no Yankee Atomic communication with the NRC occurred. The Review Team views this as a major breakdown in the licensing function, and an absence of awareness of NRC expectations. The major issues that have resulted evolved, in part, from this lack of communication and coordination with the NRC.
The root cause of this lack of communication and coordir.ation with the NRC is the apparent transfer of responsibility for NRC interface on LOCA matters from Yankee Atomic to Maine Yankee following the issuance of the YAEC-1300P SER in 1989. Due to the high visibility and regulatory significance of LOCA analysis, it is essential to maintain an interface betweer the LOCA analysis group and the NRC technical staff. This did not occur. At the major vendor LOCA analysis organizations this NRC interface is given a high priority and is maintained because the analytical modelis the responsibility of the vendor. In discussions with the Yankee Atomic Project Manager for Maine Yankee it was stated that Maine Yankee directed Yankee Atomic not to directly interface with the NRC. It has become the standard industry practice for a licensee to insist on vendor interactions with the NRC, associated with the licensee's docket, to be through the licensee.
However, this constraint, as understood by Yankee Atomic, contributed to the licensing process breakdown in this highly technical subject area.
7.
Continuous communications uith the NRC could have climinated many of the issues ofconcern.
Subsequent to initial NRC approvals, issues have arisen and errors have been discovered in industry LOCA evaluation models and plant applications. As issues or errors are identified within the industry or by a LOCA analysis organization, it is the sumom ms C-16 l
recognized responsibility of each organization to respond and communicate with the NRC. These communications are essential for problem identifi:ation, initiating appropriate corrective actions, and maintaining good technical credibility with the NRC. Due to the complex technical nature of LOCA analysis, it is necessary that these issues be communicated and discussed by the experts. It is necessary that the organization appreciate and recognize the significance of staying witnin the LOCA regulations and in understanding the NRC's expectations in that regard.
Historically, good communications with the NRC has enabled reasonable and fair regulation as LOCA issues are dispositioned The Review Team concludes that successful communications between the Yankee Atomic LOCA Group and the NRC technical staff ould have eliminated many of the issues of concern related to Maine Yankee SBLOCA analyses.
IV. TECHNICAL REVIEW TEAM'S CONCLUSIONS ON ISSUES IDENTIt"DE NRC'S DEM AND FOR INFORM ATION A.
Summarv of Annarent Violation as Stated in Demand for Information During Cycle 14 operations, and in the Cycle 14 and Cycle 15 CPAR analyses, l'AEC caused hiaine l'ankee to use apparently unacceptable evaluation models which could not calculate or reliably calculate ECCS performance. The models used were in apparent violation of10 C.F.R. f 50.46(a)(1), because there was a region of the small break spectrum between break sizes of 0.35ft' and at least 0.6ft'for which no acceptable evaluation model could either calculate or reliably calculate ECCS performance. The Afanager and the Lead Engineer knew of thiz in addition the oscillations and instability in the analysis became more severe at larger break si. es, increasing the risk that the limiting breaks had not been identified.
Review Team's Conclusion on Item A The Maine Yankee SBLOCA analysis performed by Yankee Atomic with the RELAPSYA code consisted of a spectrum of break sizes ranging from 0.1 ft to 0.35 ft.
2 Yankee Atomic was not able to run the 0.35 ft break size through to the final PCT. The 2
final PCTs given in YAEC-IS68 for the 0.35 ft breaks are estimated based on a 58'F APCT obtained from a 0.30 ft break run, which is added to the PCT calculated at the time of safety injection tank actuation. Yankee Atomic concluded that the limiting SBLOCA PCT was determined to be within this range of break sizes, and determined a 2
limiting PCT value of 1887*F for the 0.15 ft break. The rationale for this conclusion was documented as being based on the decreasing trend of PCT for smaller and larger break sizes. Yankee Atomic maintained that this scope of SBLOCA analyses met the 10CFR50 Appendix K and 10CFR50.46 regulations. This interpretation was consistent with Yankee Atomic's understanding of the regulations. In response to questions from the Review Team, the LOCA staff provided additional verbaljustification consisting of an explanation of SBLOCA phenomena which supported their conclusion. Additional
.mmmwas C-17
t r
l analysis results fcom other codes have been presented by Yankee Atomic which tend to support the Yankee Atomic position. It is noted by the Review Team that some of the trends in the additional ana'yses are not consistent with the Yankee Atomic position.
l The prior licensing basis SBLOCA analysis performed by Combustion Engineering in 1977 established a trend ofincreasing PCT with break size, with the 0.5 ft: break having a PCT of 1348'F. Other non-licensing basis analyses performed later by Siemens and ABB indicate both a local PCT peak near the limiting Yankee Atomic break size, and a trend ofincreasing PCT at the largest break size analyzed. Given the information provided and based on the trends of other analyses, the Review Team was unable to I
draw a definitive conclusion regarding the RELAP5YA PCTs for the unanalyzed pertion of the Maine Yankee SBLOCA spectrum. However, based'on all of the calculations, it is concluded that the SBLOCA PCTs for all of the,e analyses meet the 10CFR50.46 2200*F criterion, and that SBLOCAs remain bounded by LBLOCAs.
The Review Team agrees that the industry standard practice is consistent with the NRC's position that an Appendix K SBLOCA evaluation model must be capable of analyzing any break si e within the plant's SBLOCA licensing basis. This does not mean that all break sizes ne a to be analyzed, but rather that the model must be capable of analyzing them. The RELAP5YA evaluation model has not demonstrated the capability to analyze the historical Maine Yankee SBLOCA break spectrum. To meet the expectations of the NRC, sound engineering arguments can be used but should be communicated to the NRC and agreed upon prior to implerientation.
The Review Team agrees with the NRC's position that the Yankee Atomic SBLOCA model produced oscillatory and unstable results. This behavior is evident following accumulator injection and in particular for the larger break sizes. These code problems were long-standing and widely known within the Yankee Atomic organization. The model is also considered by the Review Team to be unreliable in that an unexpected large change in results can occur with only a small change in the input to the code.
Yankee Atomic considered the oscillations and instability to be non-physical, and that the PCTs predicted were sufficient. The Review Team believes that the oscillations and instability may, in part, be non-physical and due to fundamental limi ations in the t
RELAP5YA code. Based on a broader knowledge of SBLOCA phenomena and results trom other codes, Yankee Atomic was confident that SBLCCA was bounded by LBLOCA. Based on this expectation Yankee Atomic accepted the results from RELAP5YA as adequate for showing compliance with the regulations. The Review Team understands that this may have been a correct conclusion. However, Yankee Atomic's conclusion is based, in part, on information beyond the demonstrated results of runs of the RELAP5YA code for Maine Yankee. The Review Team concludes that this situation should have been communicated to the NRC prior to implementation.
The Yankee Atomic report YAEC-1868, which summarizes the results of the application of the SBLOCA evaluation model to Maine Yankee, discloses the results of amowncs C-18
the analyses, irluding the break spectrum analyzed, that the 0.35 ft' break case would not run through, and Yankee Atomic's explanativn as to why the results meet the reguhtions. The illustrated results show the unstable and oscillatory behavior of the code. This report was reviewed by Yankee Atomic management and approved. The Review Team understands that YAEC-1868 vas not submitted for NRC review based l
on a communication from the NRC Project Manager for Maine Yankee and an expectation that NRC review would be by a future inspection. Not submitting YAEC-1868 for NRC review was e error in that it should have been recognized that NRC review was necessary prior to implementation.
As a consequence of the breakdown in the LOCA licensing process which resulted in implementation of the SBLOCA analyses for Maine Yankee Cycle 14 prior to obtaining
(
NRC review and approval, Yankee Atomic did provide Core Performance Analysis Reports (CPARs) to Maine Yankee with the deficiencies described above. This situation was not a result of any deliberm action to avoid compliance with the regulations, but rather failure of the Yanker. Atomic organization to understand NRC expectations. The Review Team does not consider this organizational failure to have resulted in a reduction in the safety of the plant, since LBLOCA is limiting and sets the core operating limits. Had the SBLOCA analyses been completed it is likely that the core operating limits for Cycle 14 would have renained the same.
B.
Summary of Apparent Violation as Stated in Demand for Information As a result of l'AEC's preparation and review cf l'AEC-1868, l'A ECprovided h1l'APCo with information that was not complete and accurate in all material respects, and thus caused h1l'APCO to be in apparent violation 10 C.F.R. f 50.9(a).
CPARs maintainedfor infonnation and submitted to the NRC by ofl'APCo, in support of Cycle 14 and Cycle 15 reload applications were apparently not complete and accurate in all material respects. Afl' relied on l'AEC-1868 to prepare Cycle 14 and Cycle 15 l A Rs in order to demonstrate compliance with 10 C.F.R. f 50.46.
l'AEC-1868, n.,s entirety, conceals the lack of an acceptable evaluation model to calculate ECCS performancefor a portion of the b>cck spectrum between 0.35ft' and at least 0.6ft'. As a result of the 01investigatim...,ppears that no Afaine l'ankee personnel reali cd that the RELAPSl'A coaefailed at 0.35ft'or that there might be a portion of the break spcts nfor which there was no acceptable evaluation model to calculate ECCSyrformance, and that no one at l'AEC informed ofl'APCo personnel that RELA PSl'A hadfailed at 0.35ft'.
Review Team's Conclusion on Item B YAEC-1868 documents the results of the SBLOCA analyses performed by Yankee Atomic using the NRC-approved YAEC-1300P SBLOCA evaluation model. The Review Team concludes that this document was sufficiently complete and accurate as a summary of the SBLOCA calculations that were performed. However, the downcomer wansom ms C-19
modeling changes that were not discussed in YAEC-1868 should have been communicated to the NRC in YAEC-1868, as a revision to YAEC-1300P, or in some other communication. It is clea that Yankee Atomic did not appreciate the regulatory significance of the downcomer mcJsng changes. YAEC-1868 was understandable to its intended audience, and was suitable for a licensing submittalin support of Maine i
Yankee. It is noted that the abstract is potentially misleading in the use of the word
" complete", but that the scope of the analysis as contained within the report is characterized correctly. The amount of technical information included was appropriate for any knowledgeable person to understand the results of the analysis. There was no intentional concealment ofinformation which would have identified any non-compliance with NRC regulations. The statements in the report regarding compliance with the NRC's regulations were consistent with the Yankee Atomic LOCA Group's understanding of the regulations. The Review Team concludes that Yankee Atomic's understanding of the regulations was not consistent with the understanding of peers in the industry LOCA community. The compliance statements and the supporting analyses in YAEC-1868, as interpreted by a knowledgeable engineer not trained in the LOCA licensing process, could be understood as a logical basis for compliance. Therefore, YAEC-1868 could be understood by a knowledgeable engineer not trained in the LOCA licensing processes to be complete and in compliance with the regulations. It is likely that the compliance statements and the supporting analyses in YAEC-1868 would be understood by an NRC reviewer and would have led to interactions with Yankee Atomic, in particular since the limiting break size in the analysis of record was 0.5 ft and was not analyzed with RELAP5YA.
Based on discussions with Yankee Atomic personnel, the problems with applying the RELAP5YA code to Maine Yankee were a topic of discussion over several years with cognizant Maine Yankee personnel. Although utility staff are generally not LOCA experts, people assigned to interface with LOCA organizations generally have sufficient knowledge to understand the subject, to review associated documents, and to make appropriate decisions on behalf of the utility. However, utility staff generally rely on the LOCA organization to manage the technical details associated with compliance with Appendix K and 10CFR50.46. The Review Team concludes that Maine Yankee staff, similar to typical utility staff, should have relied on the Yankee Atomic LOCA organization for LOCA licensing support. Yankee Atomic understands that Maine Yankee directed it not to independently interface with the NRC staff. As a result, there was no Yankee Atomic interface with the NRC on LOCA issues. The Review Team maintains that the code and its models are Yankee Atomic's responsibility. Had Yankee Atomic management been more aggressive and achieved the continuation of communication with the NRC, Yankee Atomic personnel would have been better prepared to make decisions relative to NRC expectations with the likely result that communications on the deficiencies of YAEC-1868 would have occurred.
.mmom m C-20 l
C.
Summary of Apparent Violation as Stated in Demand for Information During Cycle 14 operations and in the Cycle 14 and Cycle 15 CPAR, l'AEC caused hfl'APCo to use an apparently unacceptable SBLOCA evaluation model which over predicted core cooling. l'AEC-1868 apparently did not satisfy the requirements of10 C.F.R. f 50.46(a)(1) because as a result ofincorrect calculations of tin penetrationfactors, which arosefrom misapplication of the Alb-Chambre penetration correlation, the analysis provides no basis to assumefullpenetration of the emergency core cooling system injection andprovides no basis to derive the loss coefficient of 600 usedfor the split downcomer nodali ation. These deficiencies M
resulted in over-prediction of core cooling and overstatement of the conservatism of D
the model. If the Alb-Chambre correlation had been applied correctly, penetration factors would have been calculated in the range of-0.6657 to -0.77G7, which is a meaningless result because the calculations would have been leis than ero. Such calculations also indicate otherpossible errors in application of the Alb-Chambre correlation. Adequate QA review would have revealed the errors and the unacceptability of the RELAPSl'A SBLOCA analysis described in l'AEC-1868.
Review Team's Conclusion on Item C Early applications of the RELAP5YA SBLOCA evaluation model to Maine Yankee a
identified excessive ECCS bypass relative to what was expected based on scaled test facility data and the results of other codes, Revisions to the interfacial drag model were only partially successful in addressing this non-physical prediction. Various modeling approaches were attempted to make the ECCS penetration into the vessel lower plenum more physical. Eventually an artificially large loss coefficient was introduced in the junction connecting the two volumes representing a split reactor vessellower downcomer. The value of this loss coefficient was varied to obtain a balance between the expected ECCS penetration and the effect on steam venting via the break. A value of 600 was selected as an appropriate value. The amount of ECCS penetration obtained with this modeling approach was justified, in part, on the Alb-Chambre correlation. This correlation was applied to confirm that the amount of ECCS penetration predicted by RELAP5YA was conservative. The Review Team considers this approach to be an acceptable compensation for a RELAP5YA code deficiency, and the use of a value of 600 obtained an amount of ECCS penetration that was consistent with industry experience. The preferred approach would have been to implement the ECCS penetration modeling as a code change. This modeling approach is not expected to result in an overprediction of core cooling, but since the calculations were not completed for all break sizes, it cannot be definitively.onfirmed.
An error was made in the application of the Alb-Chambre correlation. This arithmetic error was not identified during the quality assurance process. This is a failure of the quality assurance process. Closer investigation of this failure offers a reasonable explanation as to why it occurred. The arithmetic error did not skew the result of the C-21 wmmmm ms
calculation, which was in the range of the expected result that complete penetration was predicted. The correlation can produce results, in excess of the value of 1 (in this case a value of 8), which have the meaning of complete ECCS penetration. A person performing a quality assurance review is influenced by the resuh based on experience and expectations. These are the most difficult errors to identify. A correct application of the Alb-Chambre correlation without the arithmetic error would have produced negative values indicating complete ECCS bypass. This result would have been immediately recognized by Yankee Atomic as non-physical for the SBLOCA conditions ofinterest.
The cause of the non-physical result would have been traced to excessively conservative input values. More reasonable values would then be inpu' to the correlation, and g
reasonable and valid results indicating significant ECCS penetration would have resulted. Therefore, although the application of the Alo-Chambre correlation was in error, the basis for incorporating the loss coefficient with a value of 600 remains valid.
Thus the results of the error in applying the Alb-Chambre correlation did not result in invalid input to the SBLOCA analyses.
The Review Team concludes that the use of the Alb-Chambre correlation as a confirmation of the modeling approach which includes the junction loss coefficient of 600 in the reactor vessel downcomer is reasonable given the available data and the deficiency of the code. The Review Team recognizes that the PCT results of the SBLOCA analyses for some of the break sizes are very sensitive to the value of the loss coefficient, tis entire modeling approach was not presented to the NRC for review.
Yankee Atomic considers the value of the downcomerjunction loss coefficient to be an input to the evaluation model. This is literally true, since alljunction loss coefficients and most of the plant applications model are in the form of inputs. However, the Review Team concludes that due to the non-physical value used and the significance of this input parameter, it is in reality an important model change requiring NRC approval. The Review Team believes that this model could have been approved by the NRC in this form or with some revision.
D.
Summary of Anparent Violation as Stated in Demand for Information YAu caused Maine Yankee to apparently violate 10 C.F.R. f 50.46(a)(1) by relying on an unacceptable SBLOCA evaluation model(Best Estimate RELAPSYA SBLOCA evaluation model) to calculate ECCS coolingperfor nance in preparing a Section 50.59 analysis used to determine if a decrease in steam generatorpressure involved an unreviewed safety questi ~n. Additionally, the proposed BE RELAP5YA evaluation model was not the approved SER computer code and did not comply with 10 C.F.R. Part 50, Appendix K requirements. The NRC indicates that is also reasonable to conclude that the Manager knew that the analysis which YAEC performed regarding the effects of a reduction in steam generatorpressure on LOCA analyses was a safety analysis which would be used by Maine Yankee in c Section 50.59 analysis or other safety analysis.
. mte.ncs C-22
In view of the intended use of the 1%EC analysis, the Manager should have provided Maine Yankee with an analysis which met NRC requirements, Review Team's Conclusion on Item D Due to difficulties in applying the YAEC-1300P Appendix K RELAP5YA SBLOCA code to the Maine Yankee plant, the Yankee Atomic LOCA Group initiated a parallel effort to develop a "best-estimate"(BE) LOCA evaluation model, and indicated this in a memoraadum. The BE modeling approach is an alternative approach to the traditional Appendix K approach. Since the YAEC-1300P topical report did not include this BE
]
approach, separate NRC approval of this method would be necessary prior to n
implementation. The Yankee LOCA Group Manager clearly understood that the BE L.J approach required further NRC review. The possibility of using the proposed BE approach to satisfy NUREG-0737 Item II.K.3.31 for Maine Yankee had been discussed with the Maine Yankee staff. The Yankee Atomic LOCA Manager was under the impression that the BE SBLOCA methodology description had been submitted to the NRC by Maine Yankee for review.
In 1990 Maine Yankee initiated a service request with Yankee Atomic to perform a scoping analysis of the effect of reduced steam generator pressure on the licensing basis transients and accidents. Steam generator pressure was decreasing steadily due to fouling of the steam generator tubes. The Yankee Atomic LOCA Group was responsible for addressing the LOCA aspects of this issue. In order to be responsive to the customer's request, Yankee Atomic used the only available LOCA analysis tool at that time, the BE model, to assess the effect of reduced steam generator pressure on the analysis of record. Memorandum LOCA 91-04 dated January 25,1991 documented the use of the BE SBLOCA model. In this memorandum the BE model is erroneously stated to be the " licensing basis SBLOCA analysis," misrepresenting the BE analysis. The LOCA Group Manager was apparently out of the office when this memorandum was signed off, and the Department Manager signed off on the approval. It is apparent that the LOCA Group staff was confused about what the actual SBLOCA licensing basis was at this time. The analysis of record at this time was the 1977 Combustion Engineering malysis.
'lE. analysis documented in LOCA 91-04 was eventually transmitted by reference to Maine Yankee as part of two other memoranda (NED 91-18 dated January 28,1991, and TAG-MY-92-035 dated May 29,1992). The LOCA Group Manager did sign his approval on the TAG-MY-92-035 memorandum. It is noted that the SBLOCA discussion in this memorandum is a very small part of the technical content and there is no mention of the referenced analysis as using the BE model. In April 1992, Yankee Atomic noted on a service request form (92-42) related to the reduced steam generator pressure that the Appendix K SBLOCA analyses were nearing completion, and that Yankee Atomic would provide the reanalysis oflowered steam generator pressure using the RELAP5YA evaluation model to Maine Yankee when completed. However, the BE mtmoswm C-23
SBLOCA analysis memorandum was referenced by Maine Yankee in a 50.59 evaluation dated April 12,1993 as part of the justification for operation with reduced steam generator pressure. The Appendix K SBLOCA results were forwarded to Maine Yankee on April 12,1993, along with a draft 50.59 for Maine Yankee's use. These results were not referenced by Maine Yankee until January 13,1994, when the 50.59 was revised.
In the various memoranda sent to Maine Yankee by Yankee Atomic that applied the BE SBLOCA model there were never any statements to the effect that the model was not NRC-approved. Memorandum LOCA 91-04 actually stated the contrary, which was iacorrect. This indicates a lack of administrative control and communication within the Yankee LOCA Group and within the Yankee Atomic Maine Yankee Project Group. It also indicates a failure to appreciate the disclosures that should be made regarding application of non-NRC approved LOCA analysis technology ir. responding to plant support requests. There was a period of time available to correct the situation when the results of non NRC approved methods were found to have been released to the customer.
The Review Team disagrees with the NRC position that only approved Appendix K evaluation models can be used in performing some scoping safety evaluations, including input to 50.59 evaluations. Models such as BE models can be appropriately used provided that the r.pplication does not replace the analysis of record, and provided that the use of the analysis method is clearly stated and justified. If there is any doubt regarding the appropriateness of such an application, then the NRC should be consulted prior to implementing the results of the analysis. Yankee Atomic's failure in this situation was in incorrectly characterizing the BE SBLOCA analysis as the licensing basis analysis, and furthermore not stating that the analysis used non-NRC approved methods and had restrictions on its use.
The Review Team notes that the effect of reduced steam generator pressure on the SBLOCA results for the magnitude of steam generator tube fouling and plugging that was being evaluated would not be expected to be significant. This is particularly true given that the analyses of record showed the SBLOCA PCTs to be lower than the LBLOCA PCTs. An evaluation could have beenjustified without any SBLOCA analysis.
An evaluation could also have been justified using the BE analysis methodology provided that sufficient qualification was included, and provided that the analysis of record was not replaced.
V.
ATTACHMENTS Technical Review Plan a
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TECHNICAL REVIEW PLAN OBJECTIVE:
The purpose of this team is to provide an independent technical review of Yankee Atomic's l
small-break LOCA evaluation model(RELAP5YA SBLOCA) for Maine Yaakee to determine compliance with 10CFR50.46 and 10CFR50 Appendix K requirements. If non-compliance issues are identified, the team will assess Yankee Atomic's cognizance of the problems, adequacy of the corrective actions taken, and communication of these issues with Maine Yankee and the NRC.
The results of this review will be presented to Duke Engineering & Services, InclYankee Atomic to assist in formulating their response to the NRC's December 19,1997 Demand For Information Letter.The Review Team will specifically look at inadequacies discussed in the NRC letter. Based on technical reviews and interviews with key Yankee Atomic staff, the team should arrive at some conclusion as to whether the technicalissues (if present) were the result of personnel errors, process breakdowns, or failures wnhin the quality verification process.
ON. SITE DATES:
February 2 - 6,1998 ENTRANCE MEETING:
February 2 EXIT MEETING:
February 6 REVIEW TEAM:
Z. L. Taylor
- Duke Power Company - Nuclear Assessment & Issues Division G. B. Swindlehurst Duke Power Company - Safety Analysis (Nuclear Engineering Division)
B. M. Dunn Framatome Technologies L. E. Hochreiter Pennsylvania State University
- Team Leader DETAILS:
In a Demand For Information letter addressed to Duke Engineering & Services, Inc./. ".kee Atomic Electric dated December 19,1997, the Nuclear Regulatory Commission states that engineering services provided to Maine Yankee for small-break LOCA analysis were not in compliance with regulatory requirements. The NRC also states that the Yankee Atomic, acting as a vendor for Maine Yankee, knowingly performed the inadequate analyses. This technical review was charted to provide an independent review of the adequacy and compliance with 10CFR50.46 and 10CFR50 Appendix K of the subject evaluation model at the time of origination, evaluate the cognizance of Yankee Atomic personnel of the mod? weaknesses (if present), appropriateness and timeliness of
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corrective actions that may have been taken, and whether or not the model weaknesses (if present) were effectively communicated to Maine Yankee and the NRC.
CONDUCT:
An entrance meeting will be held with appropriate DE&S management to discuss the Review Team's activities.
The review will consist of a technical review of the evaluation model used for the Maine Yankee mall-break LOCA analysis. Interviews will be conducted with key personnel involved in the preparation, review and approval of the evaluation model. Communication with Maine Yankee and the NRC of any potential model weaknesses will also be assessed.
At the conclusion of the review and prior to departure, the team will hold an exit meeting with appropriate DE&S management to discuss the results of the review. A final report will be prepared and submitted to DE&S within two weeks following the exit meeting.
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APPENDIX D PERSONNEL BEHAVIOR ASSESSMENT
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