ML20203L155

Provides Response to Demand for Info to Yankee Atomic Electric Co & to Duke Engineering & Services,Inc,Re Inadequate Engineering Analyses & Materially Incomplete & Inacurate Info to NRC License

ML20203L155
Person / Time
Site:	Yankee Rowe
Issue date:	02/27/1998
From:	Norris J DUKE ENGINEERING & SERVICES
To:	Collins S NRC (Affiliation Not Assigned)
Shared Package
ML20203L159	List: ML20203L155 ML20203L162 ML20203L193 ML20203L227 ML20203L245 ML20203L269
References
NUDOCS 9803050378
Download: ML20203L155 (120)

Text

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gDE55OuktE%iMeriq& Services Prescent & fE ve Chartone, NC 282011004 704 382-7448 Fan 704 382-7969 February 27,1998 Mr. Samuel. J. Collins Director, Office of Nuclear Reactor Regulation i

U.S. Nuclear Regulatory Commission j Washington, D. C. 20555 Sub;*ect: DEMAND FOR INFORMATION TO YANKEE ATOMIC ELECTRIC

{ COMPANY (YAEC) AND TO DUKE ENGINEERING & SERVICES,1NC.

(DE&S)- RE: PROVIDING INADEQUATE ENGINEERING ANALYSES AND MATERIALLY INCOMPLETE AND INACCURATE INFORMATION TO AN NRC LICENSEE i

Reference:

NRC Letter dated December 19,1997 from Samuel J. Collins to Messrs. Donald K.

, Davis ('YAEC) and John F. Norris, Jr. (DE&S) t

Dear Mr. Collins:

As requested by the referenced letter, issued pursuant to 10CFR2.204, the enclosed rcport provides the Duke Engineering & Services, Inc. (DE&S) response to the subject Demand for Information (Demand).

The Demand specifically requests DE&S and/or Yankee Atomic Electric Company (YAEC) to provide information as to: (1) why the NRC should permit NRC licensees to use the services of DE&S and/or YAEC to perform LOCA or safety-related analyses, and (2) why the NRC should

. not consider the inadequate analyses described in the Demand to be the result of willfulness on

- the part of DE&S and/or YAEC personnel. Additionally, the Demand identifies four general NRC concerns and four specific NRC technicalissues. The enclosure provides the DE&S response to each of these NRC concerns and technical issues.

DE&S, a wholly-owned affiliate of Duke Energy Corporation, prides itself on quality work and compliance with NRC requirements. DE&S, a current provider of technical services to the nuclear industry, is committed to bring its nuclear professionalism and experience to the acquired

- YAEC organizations. In this regard, DE&S conducted a number ofindependent assessments of the acquired YAEC organizations and the products and services they provide. The results of l

these assessments were then reviewed by DE&S senior management. Throughout this process, . i*f, 2

DE&S did not identify any areas where the current performance of, or the prodtets and services / D, provided by, the acquired YaEC organizations is unacceptable. Meaningful improvements had been made, since the events described in the Demand, in the quality of YAEC procedures and the

. emphasis these procedures place on NRC requirements and NRC reporting. The technical quality of work products currently being produced, as well as the professionalism and technical 9803050378 980227 -

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l I U.S. Nuclear Regulatory Commission Page 2 February 27,1998 competence of the workforce, are consistently high. Another common finding of the DE&S assessments was the high degree ofintegrity and openness of the employees. DE&S has found no evidence that would question the sincerity and dedication of this workforce, or would otherwise prevent DE&S activities to be performed in full compliance with NRC requirements.

Additionally, DE&S has strengthened the acquired YAEC Bolton office leadership with a proven nuclear industry executive, William H. Rasin. Effective March 1,1998, Mr. Rasin will become the DE&S Vice-President of Nuclear, Fuel, and Quality Assurance Services. DE&S also has underway a systematic transitior, of the acquired YAEC organizations into the DE&S project planning and work execution programs and procedures. The DE&S project planning process emphasizes clear definition and documentation of project-specific requirements, both technical ,

and organizational, and training of project personnel on these requirements. "

It should be noted that the DE&S programs and procedures described within the enclosure are revised on a routine basis to ensure that management expectations for continuous improvement are met. These revisions also maintain effective DE&S programs and procedures that are in line with cunent industry practices and compliant with NRC requirements. Full transition and integration of the acquired YAEC organizations into DE&S will place the acquired YAEC work practices, products and services under this continuous improvemen: process.

The enclosed report consists of five sections. Section 1.0 provides an introduction to the report, including an overall discussion of the DE&S response methodology and background information on events described in the Demand. Section 2.0 describes the DE&S assessment of the acquired YAEC organizations and the products and services provided by these organizations. Section 3.0 addresses the four genera! concerns identified in Section IV of the Demand. Section 4.0 addresses the four specific technical issues described in Section III of the Demand. Section 5.0 provides the DE&S response to the two information requests stated in Section V of the Demand.

Also prosided, as Appendices to the enclosed report, ar< copies of various assessment reports and other detailed information that support and suppl <.nent the main report. Appendix A contains background information pertaining to DE& S, the DE&S Design Control and Quality Assurance Programs, and an overview of the DE&S acquisition of certain YAEC organizations.

Appendix B provides a sununary of the actions taken by YAEC in response to the safety allegations involving inadequate safety analyses performed for Maine Yankee Atomic Power Company (MYAPCo). Appendices C through G contain the reports of the assessment teams chartered by DE&S to independently evaluate the acquired YAEC organizations and the issues and concerns raised by the Demand. Summary reports ofindependent assessments of DE&S/YAEC safety analyses performed for Vermont Yankee and Seabrook are provided in Appendices H and I, respectively.

i l

l U.S. Nuclear Regulatory Commission Page 3 February 27,1998 In summary, DE&S independent assessments of the acquired YAEC organizations and DE&S

- management evaluation of these assessment results did not identify any areas where the current

- performance of, or work products and services provided by, these organizations is unacceptable.

Consistent with previous DE&S acquisition experience, these assusments did highlight areas that require DE&S management attention to ensure a successful traasition and integration of the acquired YAEC organizations into DE&S. Actions that address there areas have been ir.itiated.

For these reasons, combined with the nuclear commitment and experience represented by DE&S, the NRC should have a high degree of confidence that safety-related analyses, prodnets and services provided by DE&S to NRC power reactor licensees comply with the principles of: (1)-

adherence to NRC requirements and appreciation for NRC expectations, (2) effective management control of safety-related activities, (3) accurate and complete communication with licensees and the NRC, and (4) conduct of work in accordance wnh NRC requirements.

Therefore, DEAS concludes that there is no reason to preclude DE&S from continuing to provide the full range of nuclear services to NRC licensees, including safety-related analyses.

Furthermore, DEAS concludes that there is no evidence that the concems and issues identified in the Demand were the result of willfulness on the part of DE&S and/or YAEC personnel.

- We look forward to meeting with you and your staff to discuss the enclosure in more detail and our plans for going forward. If there are questions about the information in this response, please

- contact Bill Rasin at (978) 779-6711.

Very truly yours, hn F. Norris, Jr.

JFNjr/fgh Enclosures

= __

l U.S. Nuclear Regulatory Commission

! Page 4 l~ February 27,1998 l

J. F. Norris, Jr., being duly sworn, states that he is President and Chief Executive Officer of Duke Engineering & Services, Inc., a wholly-owned affiliate of Duke Energy Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission this response to the Demand for Information pursuant to 10CFR2.204; and that all statements and matters set forth herein are true and correct to the best of his knowledge.

c242wfaa oldt F. Norris,'Jr. '

/'

Subscribed and sworn to me dil&%t+ M, /f#

Date Notarypublic My commission expires: g E JM /

(Sea!)

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cc: James Lieberman Director, Office of Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Lawrence J. Chandler Associate General Counsel for Hearings, Enforcement, and Administration U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Huben J. Miller Regional Administrator U.S. Nuclear Regulatory Commission, Region 1 475 Allendale Road

. King of Prussia, PA 19406-1415 Donald K. Davis, YAEC Michael Meisner< MYAPCo LOCA Group Manager LOCA Principal Engineer

bb PD/t RESPONSE TO THE NUCLEAR REGULATORY COMMISSION'S DEMAND FOR INFORMATION (NRC OI REPORT NO.1-95-050) by Duke Engineering & Services, Inc.

Charlotte, North Carolina February 27,1998 l

l l

Executive Summary On December 1,1995, the Union of Concerned Scientists (UCS) sent a letter to the State of Maine Nuclear Safety Advisor stating that they had receivea anonymous documentation alleging that the management at Maine Yankee deliberately falsified reports to the NRC (Nuclear Regulatory Commission). NRC received these allegations on December 4,1995 and sent an assessment team to Yankee Atomic Electric Company (YAEC) to investigate the allegations. On January 3,1996 NRC issu:d a Confirmatory Order limiting power operation and containment pressure at Maine Yankee and a Demand for Information to Maine Yankee Atomic Power Company (MYAPCo). The order was based in part on NRC allegations that Maine Yankee could not demonstrate that the RELAP5YA computer code would reliability calculate the peak cladding temperature (PCT) for all Small Break Loss-of-Coolant Accidents (SBLOCAs) for Maine Yankee).

As a result of these allegations, YAEC and MYAPCo chartered three teams to look into these issues.

YAEC and MYAPCo management formed a Response Team, composed of managers and technical specialists with responsibilities for the analyses in question, and an Independent Review Team, composed of managers and technical specialists with no prior responsibilities for the analyses in question, to investigate the allegations. In February,1996, MYAPCo and YAEC executive managemera chartered an Assessment Team to determine the underlying cause(s) of the allegations. YAEC subsequently developed a schedule for and implemented corrective actions in a wide range of areas including organizational changes to clarify the relationship between Maine Yankee and YAEC, new engineering procedures, revisions to the Employees Concerns Program, e vised training programs, revised commitment tracking, and a new Condition Reporting system.

On December 19,1997, the NRC sent a letter to Yankee Atomic Electric Company and Duke Engineering and Services, Incorporated (DE&S) which formally transmitted a " Demand for Information to Yankee Atomic Electric Company and to Duke Engineering & Services - Re: Providing Inadequate Engineering Analyses and Materially Incomplete and Inaccurate Information to an NRC Licensee (NRC OI Report No. 1-95-040)." As a result of the issues described within the Demand, the NRC requested YAEC and/or DE&S to provide the following information:

A. "An explanation why, in the view of the matters set forth above [in the Demand), the NRC should permit any NRC Licensee to use the services of YAEC LOCA Group and/or DE&S, to the extent that YAEC LOCA Group was transferred to DE&S, to perform LOCA analyses or any safety related analyses to meet NRC requirements."

l B. "An explanation why the NRC should not consider the inadequate analyses, which apparently I caused MYAPCo (Maine Yankee Atomic Power Company] to be in violation of NRC

! requirements, to be the result of willfulness, whether deliberateness or careless disregard, on the l part of YAEC and/or DE&S personnel."

Also,in December,1997 DE&S announced the acquisition of the Nuclear Services Division of YAEC.

DE&S was aware of the situation between YAEC and NRC concerning Maine Yankee prior to this acquisition. With the issuance of the NRC Demand for Information (Demand) ongoing reviews and assessments were accelerated and several additional assessmentt were cha tered in order to provide the requested response to NRC. The scope of these DE&S reviews and assessments was:

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l 1. To independently review the findings. recommendations, and corrective actions taken from the

[ three teams previously formed by YAEC, j 2. To review the engineering and technical work processes and Quality Assurance programs at j YAEC,

3. To independently review the technical issues cancerning SCLOCA analyses,

4. To determine, from a legal perspective, whether any of the personnel actions taken or decisions made related to the SBLOCA analyses were the result of willfulness,

5. To review a sampling of analyses performed by YAEC for other NRC licensees, specifically, Vermont Yankee and Seabrook.

The resuks of these reviews and assessments are summarized in the main body of this response. Copies of the various reports are provided as Appendices to this response. Ultimately, the information from these reports provided the necessary information to respond to the two specific itc ns in Section V of the Demand. Additionally, in Section IV of the Demand, NRC raised " serious questium" in four areas to which DE&S provided responses.

Section III of the Demand described four specific technicalissues concerning information supplied by YAEC to MYAPCo which was related to six apparent violations received by MYAPCo. DE&S addressed these technicalissues specifically as part of this response since they dealt with the development and use of RELAP5YA SBLOCA computer code by YAEC which goes to the heart of the Demand.

The DE&S response to each of these items is briefly summarized below.

Demand (Section V, Item A):

An explanation why, in view ofihr ; natters setforth [in the demand], the NRC shouddpennit any NRC licensee to use the services cf YAEC LOCA Group and or DE&S, to the extent that YAEC LOCA Group was transferred :o DE&S, to perfonn LOCA analyses or any safety-related analyses to meet NRC requirements Ruponse:

DE&S has found, based on its assessments, that performance of, and the products and services provided by, the acquired YAEC organizations are acceptable (e.g., compliant with NRC requirements). Meaninf ful improvements have been made during the last two years in the quality of YAEC procedures and the emphasis these procedures place on NRC requirements and NRC reporting. The technical quality of the work products being produced, as well as the professionalism and technical competence of the workforce, are conistently high DE&S is systematically transitioning the acquired YAEC organizations to the DE&S project planning and work execution programs and procedures. The DE&S project planning process emphasizes clear definition and documentation of project-specific requirements, both technical and organizational, and training of project personnel on these requirements. DE&S has strengthened the Boiton office leadership with a proven nuclear industry executive, William H. Rasin. DE&S is also currently mmmon. ES-2

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jointly engaged with each of the nuclear clients formerly supported by YAEC to ensure that roles, responsibilities, and expectations are mutually understood and documented. For these reasons, and the nuclear commitment and experience presented b. DE&S, the NRC should have a high  !

degree of confidence that safety-related analyses, prodacts and services provided by DE&S to NRC power reactor licensees comply with the principles of: (1) adherence to NRC requirements, (2) effective management control of safety-related activities, (3) accurate and complete communication with licene.;es and the NRC, and (4) conducting work in accordance with NRC requirements.

Demand (Section V, Item B):

An explanation why the NRC should not consider the inadequate analyses, which cpparently caused MYAPCo to be in violation of NRC requirements, to be the result of willfulness, either deliberateness or careless disregard, on the part of YAEC and for DE&S personnel.

Response

DE&S *ias found, throughout its assessment process, that employees of the acquired YAEC organizations display a high degree of professionalism and technical competence. Allindividuals are open, honest, and communicative. While in certain instances, there may have been inadequate analysis, there was neither deliberateness nor careless disregard resulting from the deficiencies described in the Demand. DE&S found no willfulness on the part of the two individuals -

mentioned in the Demand and believes that these individuals are capable of conducting their activities in conformance with NRC requirements. DE&S also found no evidence that would question the sincerity and dedication of the acquired YAEC work force, or would otherwise prevent DE&S activities from being conducted in full compliance with NRC reqcirements.

Demand (Section III, Item A):_

It uns not possible to confinn that the ihniting break had been identified and that the emergency core cooling system uns capable of mitigating the most severe postulated breakfor Maine Yankee Cycles 14 and 15.

Response

DE&S agrees that standard industry practice, as conducted by experts in LOCA analyses, is to utilize a code or set of codes with the capability to analyze all points within the prescribed break spectrum. As described within the Demand, the codes utilized by YAEC (i.e., RELAP5YA and WREM) did not have this demonstrated capability. YAEC had taken the position that analyses combined with an understanding of the physical phenomena occurring throughout the break

- spectrum provided the basis for compliance wah the technical requirements of 10CFR50.46.

DE&S believes that the reading of 10CFR50.46 by YAEC was understandable. DE&S also notes that the results of Maine Yankee SBLOCA analyses performed by YAEC are similar to results of SBLOCA analyses performed by other organizations. Thus, it is possible that YAEC's understanding of 10CFR50.46 may have received FRC approval had the Maine Yankee SBLOCA analyses and supporting documentation been submitted to the NRC for review. The error in judgement was not in the interpretation of the technical requirements of 10CFR50.46, but in the

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manner in which compliance with 10CFR50.46 was demonstrated, or assumed to have been demonstrr.ted. DE&S also notes that ineffective communication between YAEC, MYAPCo. and the NRC played an important role in the assumptions of all parties regarding the demonstration of comphance with the technical requirements of 10CFR50.46.

Demand (Section III, Jtem B):

Infonnation provided to Maine Yankee bt support of Cycles 14 and JS reload analyses was not complete and accurate in all unaterial respects regarding compliance with 10CFRSO.46(a)(1).

Response

DE&S believes that the SBLOCA analysis document trcnsmitted to Maine Yankee (i.e.,

YAEC-1868) was sufficiently accurate and complete with respect to its intended audience (i.e.,

one kr.owledgeable in the LOCA analysis field such as an NRC reviewer). DE&S does note that YAEC 1868's summary is potentially misleading in that it refers to an analysis of the complete break spectrum. However, the body of the report is sufficiently detailed to accurately communicate the details of the SBLOCA analyses that were perfor med.

Demand (Section III, Item C):

Incorrect calculation ofpenetrationfactors, mis application of the Alb Chambre correlation and inadequate Quality Assurance (QA) review of report YAEC 1868 produced an SBLOCA evaluation model that overpredicted core cooling and overstated the conservatisen of the evaluation model usedfor Maine Yankee Cycles 14 and 15.

Response

DE&S believes that the modeling approach utilized by YAEC was reasonable and consistent with industry practicec. However, DE&S does agree that a large change in penetration factors is more appropriately treated as a model revision, rather than being considered a change in model inputs as was done by YAEC Nonetheless, DE&S believes that differing views on the modeling of physiccl phenomen,and the treatant ofinputs would have been satisfactorily resolved had an ongoing dialogue on Maine Yankee SBLOCA issues occurred with the NRC technical staff.

DE&S has no reason to believe that the core cooling capability of Mane Yankee was overpredicted.

Demand (Section III, Item D):

A ~Best Estimate" SBLOCA analysir prepared by YAEC was subsequently used inappropriately byMaine Y rce as part of a 10CFR50.59 analysis conceming the effects of a reduction in steam j . atorpressure on SBLOCA analyses.

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Response

DE!. oelieves that methods other than the approved Evaluation Model could have been utilized foi se work undertaken, however, the limitations of use of such methods should be stated. DE&S

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notes that YAEC memoranda initiauy mischaracterized the Maine Yankee Best Estimate model as the approved Maine Yankee Appendix K Evaluation Model. YAEC did subsequently provide MYAPCo with a draft 50.59 evaluation that was based on analyses utilizing the Appendix K Evaluation Model. DE&S also notes, that during the time frame of the referenced analyses, significant changes were underway within the NRC and industry regarding the expectations for an effective 50.59 cvaluation as evidenced by pablication of NRC Generic Letter 91-18 in late 1991.

Nonetheleu, errors occurred in mis stating the Best Estimate model as the licensing basis analysis and in failing to state that these analyses were performed using non NRC approved methods.

Finally, DE&S expects that the effect of reduced steam generator pressure for the magnitude of steam generator tube fouling and plugging that was being evaluated would not be significant.

Demand (Section IV, Concern 1):

Regardfor and adherence to NRC requirements

Response

DE&S is 100% committed to providing products and services to NRC power reactor licensees that are unquestionably compliant with all applicable NRC requirements and the licensing basis of the nuclear station for which the work is provic'ed. Independent assessments by DE&S of the acquired YAEC organizations show that significant improvements have been made in the treatment of NRC requirements within workplace procedures. Training 1,as beer provided to affected personnel on NRC requirements. Additionally, current work products, primarily calculations and analyses are compliant with NRC methodology guidelines and exhibit knowlec'ge of relevant NRC requirements. A strong bias to establish safety margins through use of conservative assumptions is present. DE&S has strengthened its senior management in the Bolton office and commitment to compliance with NRC requirements by naming Mr. William H. Rasin Vice-President of Nuclear, Fuel, and Quality Assurance Services, effective March 1. Additionally, licensing functions associated with the acquired YAEC organizations will be fully integrated with existing DE&S licensing functions to strengthen the level of regulatory support provided to the acquired YAEC organizations. To provide additional assurance of quality m.k products that are in accordance with DE&S management expectations and in compliance with applicable NRC requirements throughout the transition and integration of the acquired YAEC organizations into DE&S, additional emphasis will be place on the DE&S independent assessment process during the transition and integration period.

DE&S firmly believes that the actions described above, in combination with ac: ions previously implemented by YAEC, will assure that products ar.d services provided to NRC power r actor licensees by DE&S comply with all applicable NRC requirements.

Demand (Section IV, Concern 2):

bfanagement control and supervision over licensed activities

Response

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DE&S is 100% committed to maintaining effective management control and supervision of its safety related and NRC licensed activities. Independent assessments by DE&S of the acquired YAEC organizations show that a strong proceduralized work process that reflects NRC guidelines is in place. Significant improvements have been implemented in the deficiency reporting system and its active utilization is indicative of a workforce that openly identifies potential deficiencies. Additionally, the technical quality of work products demonstrates the effectiveness of the existing YAEC work processes in producing high technical quality. All are evidence of good management control. "E&S is verifying that its regulatory and organizational interface f

requirements are clearly defined and formally documented with each nuclear client formerly supported by YAEC, preventing recurrence of the organizational uncertainties that contributed to the events described in the Demand. The DE&S independent assessment function will provide DE&S management with direct feedback on the quality of work products, their conformance with DE&S management expectations, and their compliance with applicable NRC requirements tiuoughout the transition and integration of the acquired YAEC organizations into DE&S. Other DE&S actions underway include the integration of acquired YAEC work processes and organizations into DE&S.

Corrective actions implemented by YAEC have produced a good infrastructure. Additional DE&S management actions to integrate this infrastructure into DE&S, to clearly define organizational responsibilities, emphasize strong technical oversight in the workplace, and strengthen technical leadership will provide assurance that DE&S is exerting effective management control over all safety related and potentially safety related activities.

Demand (Section IV, Concern 3):

Willingness of DE&S in titefuture to provide complete and accurate infonnation to licensees and to the NRC.

Response

DE&S is 100% committed to providing accurate and complete information to its power reactor licensee clients and to the NRC. The success of DE&S as a business mandates that all products and services provided to NRC power reactor licensees be accurate and complete. Independent assessments by DE&S of the acquired YAEC organizations show that the work products produced are consistently of a high technical quality and consistent with NRC guidelines and requirements. The assessments also found documentation to be adequate. In some cases, recourse to a technical document's originator was required to provide clarifications or respond to review questions. The only clearly identified trend by the DE&S assessment teams was a noticeable improvement in the quality of more recent documentation packages. DE&S has initiated a number ofjoint client DE&S technical teams to review client specific documentation for adequacy to ensure that their individual business and regulatory needs are met. Additionally, DE&S is transitianing ongoing work performed by the acquired YAEC organizations to the DE&S project planning process, which requires clear definition and documentation of client requirements, methods used to ensure effective communication with the client, and training of project personnel on the requirements and provisions of the project plan.

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DE&S believes that quality and accuracy of recent technical documents shows that the acquired YAEC work practices and procedures are producing quality documentation. However, older documentation may require recourse to the document's originator to provide clarincations or respond to review questions. DE&S believes that accurate and complete communication is currently being provided. Actions have been initiated with clients formerly supported by YAEC to ensure that each client's documentation needs and expectations are clearly understood and that existing documentation meets their business and regulatory needs. These actions, combined with the transition of work to the DE&S planning process and the dennition of organization responsibilities described earlier, ensure that accurate and complete documentation is provided to ,

licensees and the NRC.

Demand (Section IV, Concern 4):

Willingness and ability of DE&S to conduct activities in accordance with the Commission's requirements.

Response

DE&S is 100% committed to conducting its wc-k activities in a manner that complies fully with all applicable NRC requirements and the licensing basis of the nuclear station for which the work is provided. Independent assessments by DE&S of the acquired YAEC organizations show that:

(1) a strong proceduralized work process that reflects NRC guidelines is in place, (2) the work products produced are consistently of a high technical quality, and (3) the work products produced are consistent wS NRC guidelines and requirements. Additionally, the assessments noted that the workforce consistently demonstrates a high level of competence and knowledge within their technical areas of expertise , with an evident knowledge ofindustry practices and requirements. As previously described, DE&S also has underway a number of actions to further strengthen work practices and formality of operations. These actions include: (i) integration of the acquired YAEC work processes into the DE&S work processes, (ii) formally documenting organizational roles, responsibilities, and communication requirements, and (iii) strengthening management leadership. Additionally, the DE&S independent assessment function provides DE&S management with direct feedback on the compliance of work process, practices, and products with DE&S management expectations and NRC requirements.

DE&S firmly believes that using its existing work processes as a foundation, in combination with the DE&S transition actions and the corrective actions implemented by YAEC, provide assurance that DE&S work practices and conduct of work activities will comply with all applicable NRC requirements.

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TABIE OF CONTENTS l' arc 1.0 INTRO D U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 1 1.1 Demand for Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 - 1 1.2 Respo nse Appro ach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 1.3 S u mmary Re sponse . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 1.4 B ac k gro u nd . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -3 1.5 DE&S Pe rspec tive . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 6 2.0 DE&S EVALUATION AND ACTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 1 2.1 Evalu ation Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 2.2 Evaluation Result s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 2.3 Follow up A ctions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 10 3.0 DE&S RESPONSE TO NRC CONCERNS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1 Adherence to NRC Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 1 3.2 M anage ment Co ntrol . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 3.3 Accurate and Complete Information . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 3 3.4 Conduct o f Work . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4 4.0 DE&S RESPONSE TO NRC ISSUES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . : . . . . 4 1 4.1 LOC A B reak S pect rum . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4- 1 4.2 Materially Accurate and Complete Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2 4.3 Evaluation Model Conservatism . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-4 4.4 Best Estimate Analysis Utilization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 5 5.0 DE&S RESPONSE TO NRC DEMAND FOR INFORMATION . . . . . . . . . . . . . . . . . . . . . . 5 1 5.1 Continued Performance of Safety-Related Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 5.2 Willfulness of Personnel Behavior . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-4 l

APPENDICES i A DESCRIPTION OF DUKE ENGINEERING & SERVICES, INC.

B DESCRIPTION OF YAEC RESPONSE TO SAFETY ALLEGATIONS l

C RELAPSYA SBLOCA TECHNICAL ISSUES ASSESSMENT D PERSONNELBEHAVIOR ASSESSMENT E ROOT CAUSE AND CORRECTIVE ACTION ASSESSMENT F ENGINEERING PROCESS ASSESSMENT G QUALITY ASSURANCE ASSESSMENT H- VERMONT YANKEE SAFETY ANALYSIS ASSESSMENT

SUMMARY

I SEABROOK SAFETY ANALYSIS ASSESSMENT

SUMMARY

J- RESUME OF WILLIAM H. RASIN KL LIST OF ACRONYMS wa-u i s

loo INTRODUCTION 1.1 Demand forInformation On December 19,1997, the Nuclear Regulatory Commission (NRC) issued a letter to Yankee Atomic Electric Company (YAEC) and Duke Engineering & Services, Inc.(DE&S) formally transmitting a " Demand for Information to YAEC and to Duke Engineering & Senices - RE:

Providing Inadequate Engineering Analyses and Materially Incomplete and Inaccurate Information to an NRC Licensee (NRC 01 Report No. 1-95-050)." The Demand for  !

Information (Demand) was issued to obtain information the NRC considered necessary to determine whether YAEC and/or DE&S should continue to provide engineering analyses, and in particular Loss of Coolant Accident (LOCA) analyses, to NRC power reactor licensees. As a result of the issues described in the Demand, the NRC requested that YAEC and/or DE&S provide the following information:

A. "An explanation why, in view of the matters set forth above (in the Demand), th: NRC should permit any NRC Licensee to use the services of YAEC LOCA Group and/or DE&S, to the extent that YAEC LOCA Group was transferred to DE&S, to perform LOCA analyses or any safety-related analyses to meet NRC requirements."

B. "An explanation why the NRC should not consider the inadequate analyses, which apparently caused MYAPCo [ Maine Yankee Atomic Power Company] to be in violation of NRC requirements, to be the result of willfulness, either deliberateness or careless disregard, on the part of YAEC and/or DE&S personnel."

1.2 Resnonse Aporoach DE&S review of the Demand indicates that a number of technicalissues and concerns are identified. This report provides the specific information requested from DE&S, as well as a DE&S response to each of the concems and technicalissues expressed by the NRC within the Demand. DE&S has structured this report to provide a systematic and thorough response to each of these concerns and technical issues. Section 2.0 of this report describes: (1) the evaluation approach utilized by DE&S, (2) the results of the DE&S evaluation, and (3) additional follow-up actions being taken by DE&S. Section 3.0 of this report provides the DE&S response to each of the NRC concerns identified in Section IV of the Demand. These concerns are summarized as:

a. Regard for and adherence to NRC requirements,

b. Management control and supervision over licensed activities,

c. Providing complete and accurate information to licensees and to the NRC, and

d. Willingness and ability to conduct activities in accordance with NRC requirements, s=mowa n 1-1

Section 4.0 provides a DE&S response to each NRC technical issue identified in Section 111 of the Demand regarding the development and use of RELAP5YA. These technicalissues are summarized as:

a. It was not possible to confirm that the limiting break had been identified and that the emergency core cooling system was capable of mitigating the most severe postulated accident for Maine Yankee Cycles 14 and 15. (Demand Section Ill.A)

b. Information provided to Maine Yankee in support of Cycles 14 and 15 reload analyses was not complete and accurate in all material respects regarding compliance with 10CFR50.46(a)(1). (Demand Section Ill.B]

c. Incorrect calculation of penetration factors, misapplication of the Alb-Chambre correlation ed inadequate QA review of YAEC-1868 produced an SBLOCA evaluation model that over predicted core cooling and overstated the conservatism of the evaluation model for Maine Yankee Cycles 14 and 15. [ Demand Section Ill.C]

d. A "Best Estimate" SBLOCA analysis, that was subsequently relied upon by Maine Yankee in connection with a 10CFR50.59 analysis conceming the effects of a reduction in steam generator pressure, was inappropriately used to determine the effects of a reduction in steam generator pressure on LOCA analyses. (Demand Section III.D]

Finally, Section 5.0 provides a DE&S response to the two NRC information requests quoted earlier. Appendix A provides a detailed description of DE&S, including a description of: (1) the company history,(2) the current scope of nuclear services DE&S provides to NRC power reactor licensees, (3) the DE&S December 1997 acquisition of YAEC assets, and (4) the DE&S quality assurance, design control, and project planning programs. Appendix B provides a summary of actions taken by YAEC in response to the Maine Yankee safety allegations, including a description of: (1) technical assessments performed, (2) the root cause assessment, (3) initial corrective actions taken, (4) longer term process improvements implemented, and (5) audits and assessments performed. Appendices C I provide the reports of the independent assessment teams utilized by DE&S to assess the effectiveness of(l) the acquired YAEC organizations and (2) the corrective actions taken by YAEC in response to the safety allegations.

1.3 Summary Response The DE&S assessments and management evaluation of these assessments found that the performance of, and work products and services provided by, the acquired YAEC organizations are acceptable. Consi"ent with previous DE&S acquisition experience, these assessments did highlight areas that require DE&S management attention to ensure a

successful transition and integration of the acquired YAEC organizations into DE&S. The four most important areas highlighted are: (1) organizational roles, responsibilities and communications, (2) personnel training, (3) nuclear regulatory support, and (4) accurate and complete documentation. DE&S follow up actions are underway in each of these areas.

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The DE&S programs and procedures described within this repon are revised on a routine basis. These revisions are part of routine DE&S process improvements to maintain effective programs and procedures that are in line with current industry practices and compliant with NRC requirements. Full transition and integration of the acquired YAEC organizations into DE&S will place the acquired YAEC work practices, products and services described within this report under the DE&S process improvement process. Transition of the acquired YAEC organizations into DE&S is currently scheduled for completion in July 1998.

DE&S found, based on its assessments, that the performance of, and the products and services provided by, the acquired YAEC organizations are acceptable. Meaningfulimprovements had been made during the past two years in the quality of YAEC procedures and the emphasis these procedures place on NRC requirements and NRC reporting. The technical quality of the work products currently being produced, as well as the professionalism and technical competence of the workforce, are consistently high. DE&S is systematically transitioning the acquired YAEC organizations to the DE&S project planning and work execution programs and procedures. The DE&S project planning process emphasizes clear definition and documentation of project specific requirements, both technical and organizational, and training of project personnel on these requirements. DE&S has strengthened the Bolton office leadership with a proven nuclear industry executive, William H. Rasin. DE&S is also currently engaged with each of the major nuclear clients formerly supported by YAEC to ensure that roles, responsibilities, and expectations are mutually understood and documented. These actions, coupled with the nuclear commitment and experience presented by DE&S should provide the NRC with a high degree of confidence that safety related analyses, products and services provided by DE&S to NRC power reactor licensees comply with the principles of: (1) adherence to NRC requiremer.s, (2) effective management control of safety-related activities, (3) accurate and complete communication with licensees and the NRC, and (4) conduct of work in accordance with NRC requirements.

A common finding or observation of the DE&S assessment teams svas the high degree of prefessionalism and openness of the employees within the acquired YAEC organizations. All individuals were found to be honest and communicative. While'in certain instances there may have been inadequate analysis, there was neither deliberateness nor careless disregard resulting from the deficiencies described in the Demand. DE&S found no willfulness on the part of the two individuals mentioned in the Demand and believes that these individuals are capable of conducting their activities in conformance with NRC requirements. DE&S also found no evidence that would question the sincerity and dedication of the acquired YAEC workforce, or l

would otherwise prevent DE&S activities to be conducted in full compliance with NRC requirements.

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New England is relatively unique in that most nuclear power stations are owned by several, relatively small utilities, may of whom have ownership in several stations. YAEC was the organizational entity formed to provide the organizational size and qualifications necessary to design and operate these nuclear stations. In fact, YAEC was the original NRC Licensee for the Maine Yankee, Vermont Yankee and Yankee Rowe nuclear stations. As a result of this multi-plant, multi-ownership situation, the organizational relationships between YAEC and the

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utilities was derived more from common nuclear history and experience rather than explicitly defined requirements. In particular, with respect to Maine Yankee, MYAPCo became the NRC Licensee in 1981. With this change, YAEC was no longer the primary interface with the NRC.

l Ahhough YAEC continued as the primary provider of engineering services to Maine Yankee, l many organizational relationships and responsibilities were not explicitly dermed at this point.

It was within this organizational framework that YAEC undertook the development, NRC approval, and application of RELAPSYA throughout the 1980's and 90's.

On December 1,1995, the Union Of Concemed Scientists (UCS) transmitted a letter to the State Nuclear Safety Advisor of the Maine State Planning Office statir.g that they had received anonymous documentation,".... purportedly from a longtime employee of the Yankee Atomic Electric Company...." alleging that the management at Maine Yankee deliberately falsified reports to the NRC in order to receive approval of an increase in the reactor's maximum allowable power level. The UCS letter further stated that the anonymous documentation alleged that management officials (at Yankee Atomic] manipulated computer calculations to avoid disclosing that the emergency core cooling systems at the Maine Yankee plant were inadequate to prevent overheating of the reactor fuel following a SBLOCA. The letter requested that Maine Yankee not be permitted to resume operation until a factual investigation of the anonymous allegations was completed. The NRC received these allegations on December 4,1995.

The NRC promptly dispatched an Assessment Team to YAEC headquarters in Bolton, Massachusetts on December 11 14,1995 to conduct a technical review of the allegations. The NRC Assessmant Team was accompanied by representatives of the State of Maine. On December 18,1995 the NRC held a public meetmg in Rockville, Maryland with MYAPCo and YAEC to discuss the findings of the technical review and to seek additional information. As a result of the technical review and subsequent evaluations, the NRC issued on January 3,1996 a

" Confirmatory Order Suspending Authority For and Limiting Power Operation and Containment Pressure, and Demand for Information" to MYAPCo. The order was based, in part, on the NRC's determination that Maine Yankee had not applied a computer code that was proposed to demomtrate compliance with the Emergency Core Cooling System (ECCS) requirements of 10CFR50.46 in a manner that conformed to the requirements of 10CFR50, Appendix K, nor to the conditions specified in the staff's safety evaluation report dated January 30,1989. Specifically, the confirmatory order and demand for information stated that MYAPCo allegedly could not demonstrate that the RELAP5YA code would reliably calculate the peak cladding temperature for all break sizes in the small break LOCA spectrum for Maine Yankee, nor had MYAPCo submitted the justification for the code options selected and other justifications and sensitivity studies to satisfy conditions in the NRC staff's safety evaluation.

On January 10,1996, MYAPCo provided an initial response to the Demand for Information.

On January 23,1996, MYAPCo responded to the Confirmatory Order. This letter also indicated that MYAPCo and YAEC had performed extensive technical evaluations of the issues rai3ed by the UCS allegations. MYAPCo provided reports to the NRC on January 22,1996 of two independent teams formed by MYAPCo and YAEC. On February 2,1996. MYAPCo provided a schedule for responding to the remaining items in the NRC's January 3,1996 Demand for Information. Included in this schedule was completion of mm- m 1-4

an independent SBLOCA ana:ysis to be performed by Siemens. The results of this analysis were submitted to the NRC on April 23,1996 by MYAPCo.

In response to the initial NRC investigations, an inquiry conducted by the NRC Office of Inspector General (010), and concerns expressed by the Governor of Maine, the NRC Chairman initiated a separate special investigation during the summer of 1996. This investigation was performed by an independent team of NRC experts and consultants with the objective of verifying YAEC engineering activities other than those related to SBLOCA analysis. The scope of this Independent Safety Assessment Team (IS AT) as stated in their report, dated October 7,1996 was:

"On May 31,1996, the staff was directed to perform an independent evaluation of Maine Yankee's safety performance. The overall goals of the independent safety assessment were to: (1) independently assess the conformance of Maine Yankee Atomic Power Station (MYAPS) to its design and licensing bases including appropriate reviews at the site and corporate offices; (2) independently assess operational safety performance giving risk perspectives where appropriate; (3) evaluate the effectiveness oflicensee self assessments, corrective actions, and improvement plans; (4) determine the root cause(s) of safety significant findings and draw conclusions on overall performance."

In general, the ISAT concluded: "The quality of engineering work was mixed but considered good overall. Strengths were noted in the capability and experience of the engineering staff, day-to-day engineering support of maintenance and operations, in the quality of most calculations, and in the routine use and application of analytic codes" On December 10,1996, MYAPCo responded to the ISAT report with a corrective action plan to address all of the issues identified in the ISAT report.

Continued review and investigation led the NRC to conclude by December,1997 that by YAEC's preparation and approval of the RELAPSYA SBLOCA analysis and the WREM Large Break LOCA (LBLOCA) analysis, and by YAEC's preparation and approval of the Core Performance Analysis Reports (CPARs) used to support Cycle 14 and Cycle 15 operation of Maine Yankee, YAEC caused MYAPCo to be in apparent violation of 10CFR50.46(a)(1). Specifically, the NRC stated that the RELAP5YA SBLOCA analysis and the WREM LBLOCA analysis, singly or combined, were not capable of acceptably calculating emergency core cooling system performance for the portion of the break spectrum between 2

0.35 ft and at least 0.6 ft 2Furthermore, the NRC stated that it was not possible to confirm that the limiting break had been identified and that the emergency core cooling system was capable of mitigating the most severe postulated accident. Moreover, the NRC concluded that YAEC provided MYAPCo with information that was not complete and accurate in all material respects regarding this noncompliance with 10CFR50.46(a)(1), and thus caused MYAPCo to apparently violate 10CFR50.9(a) by maintaining CPARs which contained information which was not complete and accurate in all material respects in connection with MYAPCo's Cycle 14 and Cycle 15 reload analyses.

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Additionally, the NRC concluded that as a result ofincorrect calculation of penetration factors, misapplication of the Alb-Chambre correlation, and inadequate QA review of YAEC 1868, fAEC caused hiYAPCo to rely on an unacceptable SBLOCA evaluation model which over

, predicted core cooling and overstated the conservatism of the evaluation model for Cycle 14 and Cycle 15 in apparent violation of 10CFR50.46(a)(1). Finally, the NRC concluded that by its use of an unacceptable "Best Estimate" SBLOCA analysis to determine the effects of a reduction in steam generator pressure on LOCA analyses, YAEC caused hiYAPCo to apparently violate 10CFR50.46(a)(1). Specifically, hiYAPCo relied upon this unacceptable "Best Estimate" SBLOCA evaluation model to calculate ECCS cooling performance in connection with a 10CFR50.59 analysis concerning the effects of a reduction in steam generator pressure.

As a result of these conclusions, the NRC issued the Demand on December 19,1997 to both DE&S and YAEC. In addition, concurrent with the issuance of the Demand, hiYAPCo was notified by the NRC of apparent violations associated with performing inadequate LOCA analyses and providing inaccurate information to the NRC.

1.5 DE&S Perspective DE&S, a wholly owned affiliate of Duke Energy Corporation (Duke), is committed to the highest standards ofintegrity and technical quality upon which its clients and the Nuclear Regulatory Commission (NRC) can rely without reservation. DE&S fully understands and serves its NRC Licensee clients in accordance with these basic principles:

a. adherence to NRC and industry requirements and standards,

b. effective management control of safety related activities,

c. accurate and complete communication with clients and the NRC, and

d. conduct of work in accordance with NRC and industry requirements.

Successful compliance with these principles has contributed, in large part, to the success of Duke's nuclear pro ~ gram. Just as importantly to DE&S, compliance with these principles in providing technical products and services to NRC Licensees is fundamental to the success of DE&S as a business. A detailed description of DE&S is provided in Appendix A, including:

(i) the company history, (ii) the current scope of nuclear services DE&S provides to NRC power reactor licensees, (iii) the DE&S December 1997 acquisition of YAEC assets, and (iv) the DE&S quality assurance, design control, and project planning programs.

Prior to the December 1997 acquisition of a portion ofits assets by DE&S, YAEC had conducted a number of assessments and implemented a number of corrective actions designed to address the concerns and issues raised by the NRC during, and subsequent to, its investigations of the December 1995 safety allegations regarding hiaine Yankee DE&S was aware of these allegations and the NRC's resulting investigations. DE&S conducted the necessary organizational assessments of YAEC during the "due diligence" phase of the acquisition. Aditionally, DE&S had planned to conduct a number of more detailed cmman 16

independent assessments of work processes, activities, and products after the acquisition as an element of the transition and integration of the acquired YAEC organizations into DE&S.

These DE&S assessments were accelerated and expanded to ensure that DE&S was fully responsive to the NRC's December 1997 Demand.

The focus of post acquisition actions by DE&S is straight forward - ensure that the new DE&S organizations are providing quality technical products and services that both comply with the principles listed above and meet the needs of DE&S clients and the NRC. This demonstration is essential for two reasons: (1) to ensure successful er ntinuity of ongoing work with quality products and services and (2) to allow a systematic transition from the acquired YAEC work processes and procedures to DE&S work process and procedures. Regarding continuity of quality products and services, the new DE&S organizations are currently working in accordance with pre existing YAEC work processes, programs, and procedures. This provides the most efficient means of assuring continuity of ongoing work at the time of the acquisition and continued compliance with governing regulatory commitments and requirements. Showing that the acquired processes and procedures reliably produce quality products and services ensures that needs of clients and the NRC are currently being satisfied.

Regarding transition of the acquired work processes and procedures, these will be systematically assimilated into existing DE&S programs and procedures. By design during the transition period, decreasing quantities of on going work will continue to be performed under the acquired YAEC procedures. This approach permits assimilation of these processes into DE&S programs and procedures in a systematic manner and reduces the potential disruption of ongoing work associated with an accelerated work process transition. Transition of the acquired YAEC organizations into DE&S is currently scheduled for completion in July 1998.

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2.0 DE&S EVALUATION AND ACTIONS 2.1 Daluption Approach DE&S review of the Demand indicates that the NRC has a number of concerns. In order to ensure a fully responsive reply to the Demand, DE&S concluded that a number of objectives had to be demonstratea. These objectives are to assure that:

a. The underlying cause(s) of events that led to the Demand have been identified and appropriate corrective actions have been successfully implemented,

b. Work products and practices exhibit an understanding of and adherence to governing NRC and industry requirements,

c. Werk products and practices exhibit management control over safety related activities,

d. Work product documentation is complete and accurate, and

e. Work products are technically correct and defensible.

Additionally, DE&S has to make a determination regarding the " willfulness" of actions and events summarized in the Demand and address the technicalissues stated in the Demand regarding safety analyses performed for the hiaine Yankee plant. These specific technical issues are:

a. It was not possible to confirm that the limiting break had bet.n identified and that the emergency core cooling system was capable of mitigating the most severe postulated accident for hiaine Yankee Cycles 14 and 15. [ Demand Section III.A)

b. Information provided to hiaine Yankee in support of Cycles 14 and 15 reload analyses was not complete and accurate in all material respects regarding compliance with 10CFR50.46(a)(1). (Demand Section Ill.B]

c. Incorrect calculation of penetration factors, misapplication of the Alb-Chamber correlation and inadequate QA review of YAEC-1868 produced a SBLOCA evaluation model that over predicted core cooling and overstated the conservatism of the evaluation model for hiaine Yankee Cycles 14 and 15. (Demand Section III.C)

d. A "Best Estimate" SBLOCA analysis, that was subsequently relied upon by hiaine Yankee in connection with a 10CFR50.59 analysis concerning the effects of a reduction in steam generator pressure, was inappropriately used to determine the effects of a reduction in steam generator pressure on LOCA analyses. [ Demand Section Ill.D]

All of these objectives were achieved by systematically subjecting the acquired YAEC crganizations to a series of horizontal and verticalindependent assessments. This process of systematic horizontal and vertical organizational assessments by DE&S was previously planned as an integral element of the transition and integration of the acquired YAEC organizations sumomm 2-1 J

into DE&S. The issuance of the Demand prompted DE&S to both accelerate and expand the scope of these assessments in order to be fully responsive to the Demand.

The independent assessments included an examination and evaluation of both the technical and process aspects of work performance. A significant focus was placed on LOCA and awociated safety analyses performed by the Noclear Engineering Department (NED), with detailed technical assessments conducted of these work products.[ Note: NED contains what was formerly the YAEC LOCA Group.) Additionally, assessments were conducted of:

(1) engineering and technical work practices and products other than those related to LOCA analyses,(2) the Quality Assurance assessment program, and (3) corrective actions initiated by YAEC in response to results of their root cause assessments. The specific DE&S assessment activities performed were:

a. Maine Yankee SBLOCA Analysis Assessment - This effort was an independent techn" '

review of the technicalissues summarized in the Demand.[ Appendix C)

b. Personnel Behavior Assessmera - This effort was an independent legal review of the

" willfulness" of the actions and events eummarized in the Demand. [ Appendix D)

c. Root Cause and Corrective Action Assessment - This effort was an independent review of the YAEC 1996 Root Cause Assessment and YAEC's 1996-1997 Correccive Actions.

[ Appendix E)

d. Engineering Process Assessment - This effort was an independent review of engineering / technical work practices used and work products produced during the past 3 years, other than those associated with LOCA analyses. [ Appendix F1

e. Quality Assurance Assessment This effort was an independent review of YAEC QA audits performed during the past 3 years. [ Appendix G)

f. Vermont Yankee Safety Analysis Assessment This effort was an independent technical review of: (i) Basis for Maintaining Operation (BMO) documents, (ii) Cycle 20 reload analyses, and (iii) containment analyses. (Appendix H)

g. Seabrook Safety Analysis Assessment This effort was an independent technical review of radiological consequence and RELAP5YA non LOCA transient analyses. [ Appendix I]

These assessments were conducted by individuals who are knowledgeable in, and independent of, the functional areas reviewed. Assessment team members were obtained from DE&S, Duke, subcontractors, and independent consultants. A listing of these individuals and their company affdiations is provided in Table 2-1. The assessment activities included (1) interviews with knowledgeable technical and management personnel, (2) reviews of governing work activity procedures, and (3) technical reviews of work products, typically calculations.

Throughout the course of these assessments, DE&S management met daily with the assessment teams to review progress and potential fmdings. When a specific weakness or potential deficiency was identified, it was entered into the acquired YAEC Condition Report

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System. Additionally, affected stations (e.g., Seabrook and Vermont Yankee) were notified of the weakness or potential deficiency for their own internal evaluation and follow-up.

The finding: and conclusions of each assessment are summarized in the Appendices, as l indicated previously in Section 2.1. DE&S management's evaluation of the overall findings and conclusions of the assessments is provided in below Section 2.2.

2.2 Evaluation Results Results of the independent assessment teams were integrated with DE&S management observations during and subsequent to the acquisition's "due diligence" assessments to produce an evaluation of the acquired YAEC organizations with respect to each evaluation objective listed in Section 2.1. This DE&S management evaluation found that the current performance of, and work products and services provided by, the acquired YAEC organizations are acceptable. Consistent with previous DE&S acquisition experience, the DE&S assessments did identify areas requiring management attention to ensure a successful organizational transition and integration. A dN:ussion of these areas and planned DE&S follow-up actions is provided in Section 2.J.

2.2.] Root Cause and Corrective Actions Validation Evaluation Objective:

Assure that the underlying cause(s) of events that led to the Demand have been identified and appropriate corrective actions have been successfsdly implemented.

DE&S Evaluation:

Three (3) underlying causes were identified by YAEC as a result of an assessment performed in early 1996: (1) dermition of organizational roles and responsibilities was less than adequate., (2) training of YAEC personnel was less than adequate, and (3) clarity and specificity of procedures was less than adequate. Each cause and its associated corrective actions is addressed separately below.

2.2.1.1 Organizational Resnonsibilities. Roles. and Communications The 1996 YAEC root cause assessment determined that the division of responsibilities / ownership and roles of organizations performing, controlling, administering, and managing activities for the hiaine Yankee plant were not always completely and clearly defined or understood by all parties involved in or impacted by the activity. DE&S concurs with this assessment. As described in Appendix E, many YAEC employees also believed that the relationship between YAEC and hiYAPCo was subject to varying dermitions, depending on what relationship was in the best commercialinterest of hiaine Yankee for a given situation. This less than adequate defmition of organizational roles and responsibilities clouded, and thus hindered, effective communications between YAEC and hiYAPCo and with the NRC.

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The DE&S assessments found that personnel currently have a good understanding of their individual and work group roles and responsibilities at the working level. This working level understanding of roles and l responsibilities was validated through the technical assessment teams and the l

process assessment teams. In particular, as described in Appendix E, the Root Cause and Corrective Action Assessment determined that corrective actions were initiated by YAEC to address definition of roles and responsibilities between YAEC and hiYAPCo. It should be noted that the subsequent 1997 decision by hiYAPCo to permanently shutdown hiaine Yankee rendered I many of these initiatives moot. Nonetheless, these corrective actions were focused primarily on hiYAPCo and did not fully address roles and responsibilities with other plants supported by YAEC. Additionally, the Engineering Process Assessment, which by design took a broader view of the organization than the relatively narrow focused technical assessment teams, determined that the employees within the acquired YAEC organizations have a limited overall understanding of the current DE&S organization.

Furthermore, on-going interactions wh DE&S clients formerly supported by YAEC indicate that additional clarifications regarding expectations and interface requirements with DE&S are warranted. (It should be noted that client representatives participated in an oversight role in these n: views, but were not interviewed as an element of the DE&S independent assessments.)

It is recognized by both employees within the acquired YAEC organizations and DE&S clients who were formerly supported by YAEC, that the acquisition of YAEC assets by DE&S has created an new organizational relationship. This new relationship is one of a client and a subcontractor, which removes the uncertainties caused by the co-ownership relationship that existed during the events described within the Demand. In fact, DE&S and its clients formerly supported by YAEC are in the process of negotiating contract modifications that more clearly define these respective roles, responsibilities, and expectations. As these contract modifications are enacted, DE&S workplace procedures and programs will be revised to conform to the specified contract requirements.

2.2.1.2 Comnany Training The 1996 YAEC root cause assessment determined that personnel at YAEC lacked formal company training and retraining in areas that are essential to performance of their day-to-day tasks. Informal company training and retraining referred, in part, to the read and sign training approach for day to-day work procedures. Although read and-sign is a useful training tool, it is not alone sufficient to ensure effective personnel training. Formal classroom instruction is at times necessary to ensure a complete understanding of codes, standards, procedures, regulations and management expectations.

DE&S concurs with this assessment.

smuomnes 2-4

The DE&S technical and process assessment teams found that the employees within the acquired YAEC organizations demonstrate a high level of competence and knowledge within their respective technical areas of expertise. The technical expertise of these employees was acknowledged ,

l consistently as a strength and will be discussed later. The Engineering Process Assessment and the Quality Assurance Assessments,in particular, reviewed training records for individuals originating, reviewing and approving documents that were reviewed as part of the assessment activities. These assessments found that personnel were appropriately qualified to perform their assigned functions. However, the Root Cause and Corrective Action Assessment determined that actions taken by YAEC to improve technical training were not complete and systematic. Although additional training had been introduced into the YAEC organization (specifically regarding NRC regulations and reporting requirements) and procedures governing training requirements were in place, evidence of a systematic commitment to improve

.he quality of personnel training was not found. While it was determined that more emphasis on personnel training is warranted, the cumulative findings of the assessment teams indicate that the employees within the acquired YAEC organizations are qualified to perform their work assignments.

2.2.1.3 Procedure Ouality The 1996 YAEC root cause assessment determined that procedures used for controlling the development of analyses were weak in defining important processes that appeared to be driven by personnel knowledge rather than by procedural guidance. These procedures did not require identification of the effects of analyses on licensing commitments or design basis documents.

YAEC recognized that governing analysis procedures needed to be sufficiently clear with respect to ensuring consideration of regulatory requirements, effective controls on the use of unverified data, compliance with QA requirements, updating of affected design basis and/or licensing documents. DE&S concurs with this assessment.

The DE&S technical and process assessment teams found that current

workplace procedures are effectiec, are being properly implemented, and l reflect current industry standards and practices. Several assessments, in l particular the Engineering Process and the Root Cause and Corrective Action l Assessments, noted the presence of management initiatives to continuously l improve technical procedures. The Root Cause and Corrective Action

( Assessment did express concern regarding the relatively narrow focus of l

corrective actions on MYAPCo,instead of more globally addressing all YAEC clients. Nonetheless, the Assessment concluded that the root cause had been adequately addressed, and that the specific concerns of self assessment, analysis control, and NRC reporting have been addressed very well. These corrective actions and improvements were programmatic and applicable to all YAEC clients.

smmewines 2-5 1

The Engineering Process Assessment noted that the transition from YAEC to DE&S is creating some concern among former YAEC employees regarding the impact full integration into DE&S will have on their procedures. This concern is normal as staff face a transition from procedures with which they are familiar to a set of procedures that may be wry different in form and content. In fact, DE&S is aware of this and has implemented a systematic transition from YAEC to DE&S work practices, utilizing a process that ensures assimilating the best elements of YAEC procedures and work practices into the DE&S procedures and work practices. DE&S recognizes the importance of the transition from one procedure set to another, as well as the benefit of retaining the valuable lessons learned that have been incorporated into the YAEC procedures.

2.2.2 Adherence to NRC Reautrements Evaluation Objective:

Assure that current wrk products andpractices exhibit an understanding of and adherence to gover ting NRC and industry requirements.

DE&S Evalumaog:

Numerous work products and procedures were reviewed and interviews conducted by the assessment teams to address the stated evaluation objective with respect to the adequacy of the work products, governing procedures and personnel conducting the work. A common finding or observation from both the technical and process assessment teams was that the workforce consistently demonstrates a high level of competence and knowledge within their technical areas of expertise. Questions are answered accurately and correctly without hesitation. Typically, detailed explanations are provided without recourse to reference material. Available references and backup material are promptly provided upon request. Knowledge ofindustry practices and safety consciousness is evident.

Actions taken to improve the quality of technical procedures, especially with respect to treatment of NRC requirements are described in Subsection 2.2.1.3. Another general finding or observation of both the technical and process assessment teams is that work products and their goveming procedures are consistent with NRC requirements and

, generally accepted industry practices. In particular, a broad sampling of safety analyses found that analysis methods used follow guidance provided by the NRC's Regulatory Guides N.d/or Standard Review Plan. Technical analyses and documents are consistent with similar analyses and documents throughout the industry.

Assessment results show that the acquired YAEC organizations contain quality people producing quality products in accordance with NRC and industry guidelines. However, there are indicators that continued management attention is warranted. The circuustances surrounding the development, licensing, and utilization of RELAP5YA for the Maine Yankee SBLOCA analysis did not clearly display complete understanding

%-. 2-6

of goserning NRC regulations. (The DE&S evaluation of the RELAP5Y4 baves are discussed in Section 4.0 and Appendix C of this report.) Isolated simmions were idro identified by the technical assessment teams where a document's originator was either uncertain with respect to the licensing basis within which the product was to fu used or not sensitive to the necessary timeliness and documentation requirements to demonstrate compliance with a plant's licensing basis or an NRC requirement. Another indication mentioned by the Root Cause and Corrective Action Assessment refers to

[.

the limited nature of organied licensing support provided to the technical woniorce.

The timely involvement of licensing professionals to either determine or respond to questions regarding the current licensing basis requirements governing a particula:

analysis or project is of significant value to technical personnel.

3 As described above, procedures have been improved. training has been provided  %

regarding NRC regulations, and work products do display an awareness of and adherence to NRC technical requirements and industry standards. Therefore, the fundamental and necessary components are in place. Continued management attention will ensure maintenance of a work environment that is fully sensitized to compliance with both the letter and spirit of NRC requirements.

2.2.3 Manacenrent control Evaluation Objective:

Assure that current wrk products and practices exhibit management wntrol over safety related activities.

DE&S Evaluation:

Evidence of management control over sar ^y-related activities is generally sought by examination of several avenues: (1) clearly defined and communicated management expectations regarding organizational policies and work practices, (2) workforce understanding and compliance with management expectations and policies, (3) the formality of operations exhibited by the organization, (4) willingness of the workforce to identify problems to management, and (5) management visibility through out the workplace in a manner that reinforces management's priorities and expectations.

The Engineering Process, the Quality Assurance, and the Root Cause and Corrective Action Assessments each reviewed workplace procedures for conformance with applicable NRC requirements and industry standards. The collective findings of these assessments were that existing procedures properly incorporated applicable NRC and industry requirements, and were consistent with industry practices. Thus, one indication that management control is evident is the fact that management expectations regarding the manner in whien work is e ecoted is appropriately and effectively contained within workplace procedures.

Workforce understanding of management expectations is generally demonstrated through the quality of the work pioducts. As mentioned earlier and discussed later in muerma 2-7 l

Subsection 2.2.5, both the process and technical assessment teams generally found that work products were of a high technical quality and were performed in accordance with appropriate workplace procedures. The quality of technical work products provide ;

evidence of employee understanding of and compliance with manegement expectations '

as they are expressed within workplace procedures.

r As discussed previously, the early relationships that existed between YAEC and the NRC Licensees supported by YAEC were not always strictly defined or documented in detail. This approach supported a strong sense of comraderie and esprit de corp that in some ways served YAEC and its clients well, but also spawned the uncertainties regarding organizational roles, responsibilities, and communications that played such a significant role in the events described within the Demand. Corrective actions taken by YAEC inserted more discipline into the YAEC-MYAPCo relationship. However, the orgarDation is going through another transition with the DE&S acquisition.

Nonetheless, quality work products and services are being provided. DE&S management recognizes that the changed organizational relationship between the acquired YAEC organizations and the nuclear clients they suppon, combined with the DE&S commitment to full compliance with NRC requirements, warrants an increased management focus on discipline of operations, s

Appendix B, Section B.4, describes the recent development of and improvements in the YAEC reporting systems for deficiencies and employee concerns. The growth in usage of the deficiency reporting system is illustrated. A " healthy" number of deficiency reports is a positive sign which can be linked to the questioning attitude of the workforce and their ability and desire to identify potential problems without fear of retribution. Additionally, another common observation of both the process and technical assessment teams was the openness and honesty of the workforce throughout the assessment interviews. An impottant aspect of problemidentification pointed out within the Root Cause and Corrective Action Assessment is the concern about the degree of a questioning attitude present with respect to work assignments. A specific example was found where a calculation was requested by a client and performed.

However, the calculation was not sufficient to fulfill the needs of the client. Subsequent to completion and transmittal of the calculation, the discrepancy between the scope of the calculation and its intended use was internally identified and a deficiency report promptly initiated. It is a positive point that a questioning attitudt. within the organization discovered and reported the deficiency, albeit after completion and transmittal of the calculation.

Finally, management within the acquired YAEC organizations is visible throughout the work place, as evidenced by active participation in DE&S assessment teams' entrance and exit meetings (as well as the daily team status briefings), appropriate participation in technical meetings associated with ongoing client work, and appropriate participation in client interface meetings. Additionally, the Quality Assurance Assessment observed that active management participation in QA surveillance and audit oversight activities was evident throughout the past three years. Thus, management is visible in the workplace, particularly with respect to activities affecting quality.

muomu 2-3

2.2.4 Accurate and Comniere Documen?ation Evaluation Objective:

1 Assure that current uvrk product documentation is complete and accurate.

DE&S Evaluation:

A broad based approach was taken to assess the accuracy and completeness of documentation. Work product documentation reviewed included calculations, analyses, procedures, training records, technical documents, correspondence, and software documentation. The guideline used to judge the adequacy of analyses, calculations and other technical reports and documents was whether the documents were sufficiently detailed as to purpose, method, assumptions, inputs, references, and units that a technically qualified individual could understand the documents and verify their adequacy without recourse to the originator.

Overall, the findings and observations of the assessment teams were mixed regarding documentation quality. Some calculations were considered by reviewers to be excellent documents, in other cases, documents were difficult to follow, although in every situation where this occurred, the originator was able to satisfactorily respond to the reviewer's questions. Nevertheless, the above stated document quality guideline was not met in these cases. Additionally, situations were identified where training and/or qualification records for personnel participating either in the production of technical documents or surveillance of technical activities were either missing or incorrect. These situations were not widespread and were typically quickly remedied. One example is the Quality Assurance Assessment finding regarding missing or incorrect documentation supporting the training and qualifications of auditors and technical specialists in QA audits and surveillances. However, the assessment did conclude that the acquired YAEC QA program assessment process was of high quality and otherwise compliant with regulatory requirements. Another documentation concern identified by one of the technical assessment teams was the timeliness of completing formal documentation packages, particularly with respect to documentation of analyses performed to support operability determinations. Given the scope of the DE&S assessment teams, it is difficult to identify any clear trends based on the derived data, other than the assessments did observe a noticeable improvement in quality from older to more recent documentation packages.

In summary, documentation was found to be accurate, but in some cases discussions with a technical document's originator were necessary to provide clarifications or respond to review questions. Therefore, utilizing the guideline defined above, more focus is required to enhance work product documentation quality. DE&S is initiating actions with each client to ensure that their documentation needs and expectations are being met.

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2.2.3 Technical GuaHrv Evaluation Objective:

Assure that current nork products are tecimically correct and defensible.

DE&S Evaluation:

Numerous analyses, calculations, and technical documents were reviewed by the assessment teams to judge the technical adequacy of these work products. A common finding or observation from both the process and technical assessment teams was that no significant technical errors were identified in completed analyses. Minor problems were periodically identified, however, particularly in analyses that were "in process" and had not been reviewed and completed. It was also generally noted that an emphasis on the application of conservative approaches and/or assumptions is clearly evident.

Examples were identified of analyses that were described ar " state of the art" or "one of the best seen." Thus, it is clear from the assessment results that the technical work products produced by the acquired YAEC organizations are both technically correct and defensible.

2.3 Follow up Actions Identification of follow up actions by DE&S was accomplished through a DE&S senior management review of the evaluation results in the context of DE&S management expectations and standards, as well as those of DE&S clients, the commercial nuclear industry, and the NRC. These follow up actions are discussed in the Subsections below.

As previously stated, prior to receipt of the Demand, DE&S had planned to conduct assessments of the acquired YAEC organizations in the normal course of transitioning and integrating the acquired YAEC organizations into DE&S. These assessment results would not only assist DE&S management in constructing an effective transition and integration plan, but also would highlight functional and/or organizationalitems that warrant particular DE&S management attention. Previous DE&S acquisitions have shown that these " highlighted" resuhs from transition and integration assessments typically have been items that are either: (1) vulnerabilities requiring attention to ensure a successful transition or (2) areas where implementation of DE&S managem:nt expectations and work practices require attention to ensure successfulintegration. Although the Demand prompted DE&S to accelerate and expand the scope ofits planned assessments, the fundamental purpose and role of these assessments remained unchanged, namely a set of DE&S management tools to ensure successful transition and integration of the acquired YAEC organizations into DE&S.

DE&S management evaluation of assessment results (reference Section 2.2) found that the performance of, and work products and ser vices provided by, the acquired YAEC organizations are acceptable (e.g., meeting NRC requirements). Consistent with previous DE&S acquisition experience, the DE&S assessments did identify areas requiring management attention to ensure a successful organizational transition and integration. The follow-up actions described below address the four most important areas identified as requiring management mmomm 2-10

.m. ,,_ ,_ 4 a % _ _ a a __ e -_ _4% .,__ 2 _.m A. - _ __

attention to ensure successful transition and integration of the acquired YAEC organizations into DE&S. These four areas are: (1) organizational roles, responsibilities and communications, (2) personnel training, (3) nuclear regulation support, and (4) accurate and complete documentation.

Transition of the acquired YAEC organizations into DE&S is currently scheduled for completion in July 1998. It should also be noted that existing DE&S programs and procedures already address many areas identified by YAEC in 1996 as requiring improvement. One example of this is the Employee Concerns Program. However, one DE&S transition practice is  !

to incorporate the experience and lessons learned of acquired organizations into DE&S programs and procedures. This practice aUows DE&S to benefit from YAEC's experiences i and ensures that the process improvements previously implemented by YAEC are either l I

addressed by existing DE&S procedures or are incorporated into DE&S procedures.

The DE&S programs and procedures described within this report are revised on a routine basis. The:,e revisions are part of routine DE&S process improvements to maintain effective programs and procedures that are in line with current industry practices and compliant with NRC requirements. Full transition and integration of the acquired YAEC organizations into l

, DE&S will place the acquired YAEC work practices, products and services described within this report under the DE&S process improvement process.

2.3.1 Orcanirational Resnonsibilities. holes and Commmtications As part of the corrective actions in response to the YAEC root cause assessment, actions were taken by YAEC and MYAPCo to better define roles, responsibilities and communication requirements between YAEC and MYAPCo. However, a global effort to ensure clear definitions and understanding with all YACC clients was not specificaUy iindertaken. The DE&S acquisition of YAEC organizations introduces both the need and the opportunity to define organizational roles, responsibilities and communication requirements between DE&S and nuclear stations formerly supported by YAEC. To ensure a successful transition, DE&S actions are warranted to establish roles, responsibilities, relationships, and communication requirements that are properly defined and documented with each of the nuclear clients formerly supported by YAEC.

This action will also prevent recurrence of one of the identified YAEC root causes.

Discussions are underway with each of these nuclear clients to ensure a clear understanding of roles, responsibilities and communication requirements. The results of these discussions wiU be formally documented, with DE&S procedures revised as appropriate to reflect these un? rstandings. Additionally, these understandings will be formaUy communicated to employees and training conducted on any resulting procedure revisions.

A Management Directive has been issued to the acquired YAEC organizations formaUy communicating DE&S management expectations regarding interactions with DE&S clients. This Directive addresses three items: (1) feedback from the DE&S assessments regarding organizational roles and responsibilities, (2) DE&S management's view of the new organizational relationships created by the DE&S acquisition, and (3) DE&S management expectations regarding the role of DE&S as a service provider in maxw nms 2-11

~-m

communicating with and supporting commercial nuclear clients. A particular point to be noted as an element of DE&S management expectations on providing support to nuclear clients is the necessity of understanding the ultimate use of a requested analysis, calculation, or design. Routinely obtaining and documenting this understanding will ensure that the client's "real" problem or need is being addressed and that appropriate regulatory requirements are applied within the work product in a manner commensurate with its intended use.

In addition to addressing organizational relationships, an equally important relationship where ensuring clear definition is required at the individual project level. The DE&S project planning process is summarized in Appendix A, Section A.3.3. This process requires, in part, the identification of responsibilities and communication requirements for each project. The DE&S project planning process also requires that project personnel be trained on the requirements contained within the project plan. This training requirement includes project-specific roles, responsibilities, and communication requirements. Therefore, it is important to DE&S and its clients formerly supported by YAEC to promptly and systematically transition all on going work performed by the acquired YAEC organizations to the DE&S project planning system. Transition of the acquired YAEC organizations into DE&S is currently scheduled for completion in July 1998.

The complete transition of ongoing work and new work performed by the acquired YAEC organizations to the DE&S project planning process, in combination with the corporate level actions described above, will ensure that responsibilities, roles, and communication requirements are clearly defined and documented at both corporate and individual project levels.

2.3.2 Personnel Trainine Corrective actions were taken by YAEC in response to the 1996 root cause assessment to address immediate training deficiencies, particu!arly with respect to NRC reporting requirements. However, the need for a systematic training function and incomplete YAEC corrective actions are items that warrant DE&S management focus during transition and integration of the acquired YAEC organizations into DE&S. As a result, systematic integration of the acquired YAEC training functions into DE&S training processes is an important transition objective. Full integration will permit timely performance ofjob task analyses of the nuclear related work performed by the acquired YAEC organizations, allowing DE&S to systematically define each organization's training requirements. Overlaying the results of this analysis on the existing training program will permit the systematic identify training deficiencies and permit development personnel training matrices, from which individual training schedules will be developed.

2.3.3 Nuclear Reenlatorv Sunnort DE&S management must set and enforce the expectations ar.d standards regarding regulatory compliance. The DE&S evaluation found that existing procedures have been

.mte-nra 2-12

revised to appropriately address applicable NRC requirements and that YAEC had provided personnel training on NRC reportirg requirements. A strengthened training program will further er.sure workforce understanding of NRC requirements.

Noneth. 'ess, a continual message from DE&S management to the workforce reinforcing the importance of regulatory compliance is appropriate. DE&S has selected William H. Rasin to become DE&S Vice President of Nuclear, Fuel, and Quality 2

Assurance Services effective h1 arch 1,1998. hir. Rasin's organization consists of these former YAEC departments: Nuclear Engineering, Fuel Management, and Quality Assurance. hir. Rasin has an extensive background with Duke, and more recently with the Nuclear Energy Institute (NEI), in managing complex nuclear safety and regulatory issues. Appendix J provides a copy of hir. Rasin's resume. hir. Rasin's experience and presence in the DE&S Bolton office, will provide solid DE&S management emphasis to the Bolton office workforce in this area.

Another item requiring management focus during transition and integration is to ensure that appropriate nuclear licensing support is effectively integrated into ongoing work performed by the acquired organizations. The actions described in Subsection 2.3.1 provide means ofidentifying NRC regulatory support requirements at both the project and corporate level. Appendix A, Section A.I.5 provides a brief summary of DE&S licensing expertise.

2.3.4 Accurate and Comniere Dortmientation The variability in the quality of technical documents warrants increased DE&S management attention. The quality of recent technical documents indicates that work practices and procedures are producing work products and services supported by quality documentation. To further ensure the adequacy of supporting technical documentation, DE&S has initiated a number ofjoint client DE&S technical teams to review client specific documentation needs and expectations. The purpose of these teams is to determine whether existing technical documentation is sufficiently self contained to adequately meet their business and regulatory needs. DE&S procedures will be revised as appropriate to reflect these documentation needs and expectations. Employees will be trained on resulting procedure revisions.

A hianagement Directive has also been issued to the acquired YAEC organizations formally communicating two items: (1) the findings of the DE&S independent assessments regarding documentation quality and (2) DE&S management expectations regarding completeness of technical documentation. Additionally, the DE&S independent assessment process provides direct feedback to DE&S senior management regarding the effectiveness of the organization's implementation of management expectations and the maintenance of quality and compliant work products. Particular management emphasis will be placed on independent assessments throughout the transition from YAEC programs and procedures to DE&S programs and procedures to ensure that the acquired YAEC organizations continue to provide quality products and services throughout the transition and integration period.

ow-m 2-13

l TABLE 2.1 DE&S Indenendent Assessment Team Membershio Maine Yankee SBLOCA Analysis Assessment B. M. Dunn Framatome Technologies,Inc. Lynchburg, VA.

L E. Hochreiter Pennsylvania State University State College, PA G. B. Swindlehurst Duke Energy Corporation Charlotte, NC Z.L Taylor Duke Energy Corporation Charlotte, NC 1

PersonnelBehavior Assessenent J. M. McGarTy Winston & Strawn Washington, DC M.J. Wetterhahn Winston & Strawn Washington, DC Root Caus-end Corrective Action Validation D. C. Prevatte Powerdyne Corporation Fogelsville, PA Engineering Process Assessinent R. W. Bonisolli Duke Engineering & Services,Inc. Bolton, MA R. G. Eble Duke Engineering & Sersices, Inc. Vienna, VA R. C. Futrell Duke Engineering & Services, Inc. Charlotte, NC

.. B. Stringer Duke Engineering & Services,Inc. Charlotte, NC T. F. Wyke Independent Consultant to DE&S Charlotte, NC Quality Assurance Assessinent J. Gerson Duke Engineering & Services,Inc. Richland, WA P. R. Horsman Duke Engineering & Services, Inc. Charlotte, NC D. A. Walker Duke Engineering & Services, Inc. Norcross, GA i

Vermont Yankee Safety Analysis Assessment J. Atchison Scientech. Inc. Idaho Falls,ID B. Gitnick Scientech, Inc. Rockville, MD D. Prelewicz Scientech, Inc. Rockville, MD R.Tedesco Scientech, Inc. Rockville, MD

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1 TABLE 2.1 (Contined)

DE&S Independent Assessment Team hiembershin Seabrook Safety Analysis Assessment W. Arcieri Sekntech, Inc. Rockville, MD T. Hencey Scientech, Inc. Clearwater, FL D. Palmrose Scientech, Inc. Rockville, MD i

D. Vreeland Independent Consultant to Scientech, Inc.

4 I

c 4

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3.0 DE&S RESPONSE TO NRC CONCERNS The NRC identified concerns in Section IV of the Demand that relate to questions regarding the efficacy of permitting DE&S to continue providiag safety-related analyses to power reactor licensees.This Section provides the DE&S response to each of the stated NRC concerns.

3.1 Adherence to NRC Reoulrements Statement of NRC Concern:

"Theforegoing situation [as described within the Demand] raises serious questions conceming regardfor and adherence to NRC requirements ...... "

DE&S Resnonse:

DE&S is 100% committed to providing products and services to NRC power reactor licensees that are unquestionably compliant with all applicable NRC requirements and the licensing basis of the nuclear station for which the work is provided. This commitment is part of the history, the experience, and the reputation of both Duke and DE&S.

Independent assessments by DE&S of the acquired YAEC organizations show that meaningful improvements have been made in the treatment of NRC requirements within workplace procedures. Training has been provided to affected personnel on NRC requirements.

Additionally, current work products, primarily calculations and analyses are compliant with NRC methodology guidelines and exhibit knowledge of relevant NRC requirements. A strong bias to establish safety margins through use of conservative assumptions is present.

However, the DE&S evaluation also identified areas where improvements are warranted, primarily in strengthening executive leadership and strengthening licensing support of and interaction with ongoing work. Specific actions are underway to address these two areas.

Effective March 1,1998. Mr. William H. Rasin will become the DE&S Vice-President of Nuclear, Fuel, and Quality Assurance Sersices. This organization encompasses what were formerly the YAEC Nuclear Engineering, Fuel Management, and Quality Assurance Departments. (It should be noted that the LOCA Group resided within the YAEC Nuclear Engineering Department.] Appendix J provides a copy of Mr. Rasin's resume. The Nuclear Engineering Department has also been restructured and new management added to the department. Additionally, licensing functions associated with the acquired YAEC organizations will be fully integrated with existing DE&S licensing functions to strengthen the level of regulatory support provided to the workforce.

To provide additional assurance of quality work products that are in accordance with DE&S management expectations and in compliance with applicable NRC requirements throughout the transition and integration of the acquired organizations into DE&S, additional emphasis will be placed on the DE&S independent assessment process during the transition and integration period.

mumma 3-1

DE&S firmly believes that the actions described above, when cor.sidered in corrbination with other actions DE&S is undertaking and actions previously implemented by YAEC, assure that products and services provided to NRC power reactor licensees by DE&S comply with all applicable NRC requirements and the licensing basis of the nuclear station for which the work is provided.

3.2 Manacement Control Statement of NRC Concern:

"Theforegoing situation [as described within the Demand] raises serious questions ... ...

concerning management control and supervision over (NRC] licensed activities ........ "

DE&S Resnons :

DE&S is 100% committed to maintaining effective management control and supervision ofits safety related and NRC licensed activities. DE&S will not compromise on safety. A primary method of assuring compliance with this DE&S commitment is to conduct work, especially those related to nuclear safety, in a controlled and disciplined manner. This is also the culture Duke maintains to ensure successful execution of the responsibility to safely operate its three nuclear stations.

Independent assessments by DE&S of the acquired YAEC organizations show that a strong proceduralized work process that reflects NRC guidelines is in place. Significant improvements have been implemented in the deficiency reponing system and its active utilization is indicative of a workforce that openly identifies poteniial deficiencies. Additionally, the technical quality of work products demonstrates the effectiveness of the existing work processes in producing high technical quality. All are evidence of good management control.

However, the DE&S evaluation also identified areas where improvements are warranted, primarily in strengthening ti'e workforce understanding of organizational roles and responsibilities and independent oversight. %ecific actions underway to address these two areas include (1) clearly defining and forar y documenting regulatory and organizational interface requirements with nuclear clients formerly supported by YAEC and (2) the DE&S independent oversight function as described in Section 3.1. Recognizing that organizational relationships are changing both internally and with nuclear clients formerly supported by YAEC, the first action will prevent recurrence of the communication and organizational responsibility uncertainties that played such a large role in the events described within the Demand. The independent assessment function will provide DE&S management with direct feedback on the quality of work products, their conformance with DE&S management expectations, and their compliance with applicable NRC requirements throughout the transition and integration of the acquired YAEC organizations into DE&S.

Other actions being undertaken by DE&S provide further assurances of effective management control. Among these are the management actions described above in Section 3.1, the activities to fully integrate the acquired YAEC work processes and organizations into DE&S described

====== 3-2

throughout Section 2.3, and the management focus on strengthening the quality of workforce training as described in Subsection 2.3.2.

DE&S believes that actinns previously implemented by YAEC have produced a good infrastructure. Additional DE&S n magement actions to integrate this infrastructure into DE&S to clearly define organizational responsibilities, emphasize strong technical oversight in the workplace, and strengthen technicalleadership provide assurance that DE&S is exerting effective management control over all safety-related and potentially safety-related activities.

3 3 Accurate and Complete InformV.lan Statement of NRC Concern:

" Questions are raised as to whether YAEC and/or DE&S will in thefuture provide complete and accurate infonnation to licensees and to the NRC ....... "

DE&3 Response:

DE&S is 100% committed to providing accurate and complete information to its power reactor licensee clients and to the NRC, The commitment of DE&S to safety and the success of DE&S as a business mandate that all products and services provided to NRC power reactor licensees be accurate and complete. Through its affdiation with Duke, DE&S understards very clearly the importance of accurate and complete documentation in supporting nuclear station operations.

Independent assessments by DE&S of the acquired YAEC organizations show that the work products prodaed are consistently of a high technical quality, are consistent with NRC guidelines, tind meet requirements. However, the assessment results also showed a variability in the quality of the supporting documentation. The guideline used to judge the adequacy of analyses, calculations and other technical reports and documents was whether the documents were sufficiently detailed as to purpose, method, assumptions, inputs, references, and units that a technically qualir d individual could understand the documents and verify their adequacy without recourse to the originator. The assessments found some documentation which required a technical document's origin'ator to provide clarifications or respond to review questions.

Therefore, utilizing the guideline defined above indicated that more focus is required to enhance work product documentation quality. Within the scope of the DE&S assessment teams, no clear trends were identified other than a noticeable improvement in quality from older to more recent documents.

DE&S corrective actions are initially focused on conducting additional document reviews to determine whether any necessary programmatic actions are warranted. DE&S has initiated a number ofjoint client-DE&S technical teams to review client specific documentation for adequacy in meeting their business and iegulatory needs. The quality of recent technical documents does show that current work practices and procedures are producing quality documentation. The independent assessment function described previously will provide DE&S management with direct feedback regarding the quality of current work product documentation. Additionally, DE&S actions underway to document roles, responsibilities, and muc m 33

communication requirements with nuclear clients formerly supported by YAEC will strengthen the understanding of their documentation expectations and requirements. These client expectations and requirements will be communicated to the acquired organizations as part of DE&S transition training activities.

The DE&S project planning process requires clear dermition and documentation of client reqairements, methods used to ensure effective communication with the client, and training of project personnel on the requirements and provisions of the project plan. The DE&S project planning process is described in Appendix A, Section A.3.3. As described in Section 2.3, DE&S is transitioning ongoing work performed by the acquired YAEC organizations to this process. Completion of this transition will provide assurances of clear client communications at the individual project level.

DE&S believes that quality and cccuracy of recent technical documents shows that the acquired YAEC work practices and procedures are producing quality documentation.

However, older documentation may require recourse to the document's originator to provide clarifications or respond to review questions. DE&S believes that accurate and complete communication is currently being provided. Actions have been initiated with clients formerly supported by YAEC to ensure that each client's documentation needs and expectations are clearly understood and that existing documentation meets their business and regulatory needs.

These actions, combined with the transition of work to the DE&S planning process and the definition of organization responsibilita.s described earlier, ensure that accurate and complete documentation is provided to licensees and the NRC.

3.4 Conduct of Work Statement of NRC Concern:

" Questions are raised as to whether YAEC and/or DE&S are nilling and able to conduct their activities in accordance with the Commissions requirements ..... "

DE&S Resnonse:

DE&S is 100% committed to conducting its work activities in a manner that complies fully with all applicable NRC requirements and the licensing basis of the nuclear station for which the work is provided.

Independent assessments by DE&S of the acquired YAEC organizations show that: (1) a strong proceduralized work process that reflects NRC guidelines is in place, (2) the work products produced are consistently of a high technical quality, and (3) the work products produced are consistent with NRC guidelines and requirements. Additionally, the assessments noted that the workforce consistently demonstrates a high level of competence and knowledge within their technical areas of expertise, with an evident knowledge ofindustry practices and requirements.

However, the DE&S evaluation also identified areas where improvements were warranted.

Actions DE&S has undertaken to strengthen work practices and formality of operations have

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been described above. These actions include: (i) integration of the acquired YAEC work processes into the DE&S work processes, (ii) strengthening personnel training, (iii) formally documenting organizational roles, responsibilities, and communication requirements, and (iv) strengthening management leadership. Additionally, the independent assessment function described above provides DE&S management with direct feedback on the compliance of work process, practices, and products with DE&S management expectations and NRC requirements.

Other actions underway that support work process improvements are described throughout .

Section 2.3 The basic DE&S work processes are summarized in Appendix A, Section A.3. DE&S firmly believes that using its existing work processes as a foundation, in combination with the actions described above and the actions previously implemented by YAEC, provide assurance that DE&S work practices and conduct of work activities comply with all applicable NRC requiremeats.

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4.0 DE&S RFSPONSE TO NRC ISSUES The NRC has a specific set of technicalissues described in Section III of the Demand that relate to YAEC's development and utilization of RELAP5YA, primarily with respect to SBLOCA analyses.

This section provides a specific DE&S response to each of these issues.

4.1 LOCA Break Snectrum Statement of NRC Issue:

During Cycle 14 operations, and in the Cycle 14 and Cycle 15 CPAR analyses, YAEC caused biaine Yankee to use apparently unacceptable evaluation models which could not calculate or <

reliably calculate ECCSperfonnance. The models used were in apparent violation of 10CFR50.46(a)(1), because there sms a region of the small-break spectrum between break sizes of 0.35ft2 and at least 0.6ft2for which no acceptable evaluation model could either calculate or reliably calculate ECCS perfannance. The bianager and the Lead Engineer knew of this. In addition the oscillations and instability in the analysis became more severe at larger break sizes, increa' .g the risk that the limiting breaks had not been identified.

DE&S Resnonse:

The Maine Yankee SBLOCA analysis performed by YAEC with the RELAP5YA code consisted of a spectrum of break sizes ranging from 0.1 2ft to 0.35 ft2 . The 0.35 ft break size run was not completed prior to the PCT being reached. YAEC concluded that the limiting SBLOCA PCT was determined to be within this range of break sizes, and determined a 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.

YAEC maintained that this scope of SBLOCA analyses met the 10CFR50 Appendix K and 10CFR50.46 regulations. In response to questions from the RELAP5YA SBLOCA Technical Issues Assessment Team (Review Team), the LOCA staff provided additional verbal justification consisting of an explanation of SBLOCA phenomena which supported their conclusion.

The prior licensing basis SBLOCA analysis performed by Combustioa Engineering in 1977 established a trend ofincreasing PCT with break size, with the 0.5ft2 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 YAEC 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 draw a defmitive conclusion regarding the RELAP5YA PCTs for the unanalyzed portion of the Maine Yankee SBLOCA spectrum.

However, 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.

DE&S 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 size within the licensed break range. This does not mean that all break sizes need to be analyzed, but rather that the model must be capable of analyzing them. The RELAP5YA evaluation model has not sesomms 4-1

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 implementation.

DE&S also agrees with the Review Team's concurrence with the NRC's position that the YAEC 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 YAEC organization. The model is also considered to be unreliable in that an unexpected large change in results can occur with only a small change in the input to the code. YAEC considered the oscillations and instability to be non-physical, and that the PCTs predicted were sufficient. DE&S agrees with the Review Team position that the oscillations and instability are, in part, non-physical and due to fundamentallimi tations in the RELAPSYA code. Based on a broader knowledge of SBLOCA phenomena and results from other codes, YAEC was confident that SBLOCA was bounded by LBLOCA. Based on this expectation, YAEC accepted the results from RELAP5YA as adequate for showing compliance with the regulations. This may be a correct conclusion.

However, YAEC's conclusion is based, in part, on information beyond the demonstrated results of runs of the RELAP5YA code for Maine Yankee. DE&S concludes that this situation should have been communicated to the NRC prior to implementation.

The YAEC report YAEC-1868, which summarizes the results of the application of the FBLOCA evaluation model to Maine Yankee, discloses the results of the analyses, including the break spectrum analyzed, the termination of the 0.35 ft2 break case, and YAEC's explanation as to why the results meet the regulations. The illustrated results show the unstable and oscillatory behavior of the code. This report was reviewed by YAEC senior management and approved. YAEC-1868 was not submitted for NRC review based on YAEC's and Maine Yankee's understanding of a communication from the NRC Project Manager for Maine Yankee. Maine Yankee and YAEC expected that NRC review would be by a future inspection.

Not submitting YAEC-1868 for NRC revicw was an error. YAEC and Maine Yankee should have understood the NRC's expectations that such a 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, YAEC did provide Core Performance Analysis Reports (CPARs) to Maine Yankee with the deficiencies described above. This situation was not a result of any deliberate action to avoid compliance with the regulations, but rather a failure to understand NRC expectations. DE&S does not consider this 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 remained the same.

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4.2 Materially Accurate and Comolete Documentation Statement of NRC Concern:

As a result of YAEC's preparation and review of 1%EC-1868, IAEC provided bi1APCo with infonnation that was not complete and accurate in all material respects, and thus caused bi1APCO to be in apparent violation 10CFR50.9(a). CPARs maintainedfor infonnation and submitted to the NRC by hl1APCo, in support of Cycle 14 and Cycle 15 reload applications were apparently not complete and accurate in all material respects. hiY relied on 1AEC-1868 to prepare Cycle 14 and Cycle 15 CPARs in order to demonstrate compliance with 10CFR50.46.1AEC-1868, in its entirety, conceals the lack of an acceptable evaluation n'odel to calculate ECCS perfonnancefor a portion of the break spectn;m between 0.35ft' and at least 0.6ft'. As a result of the 01 investigation it appears that no Maine Yankee personnel realized that the RELAPS)A codefailed at 0.35ft2 or that there might be a portion of the break spectrumfor which there was no acceptable evaluation model to calculate ECCS performance, and that no one at YAEC infonned bi1APCo personnel that RELAPS1A had failed at 0.35ft'.

DE&S Response:

YAEC-1868 documents the results of the SBLOCA analyses performed by YAEC using the NRC-approved YAEC-1300P SBLOCA evaluation model. DE&S agrees with the Review Team that this document was sufficiently complete and accurate, and an appropriate summary of the SBLOCA calculations performed. However, the downcomer modeling changes that were not discussed in YAEC-1868 should tave been communicated to the NRC in YAEC-1868, as a revision to YAEC-1300P, or in some other communication. It is clear that YAEC did not appreciate the licensing significance of the downcomer modeling changes.

YAEC-1868 was understandable to its intended audience, and was suitable for a licensing submittal in support of Maine 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 regulatioas were consistent with the YAEC LOCA Group's understanding of the regulations.

DE&S concludes that YAEC'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 ir LOCA licensing processes to be complete and in compliance with the regulations.

Based on discussions with YAEC persorr.el, the problems with applying the RELAP5YA code to Maine Yankee were a regular 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, review documents, and make appropriate decisions on behalf of the utility. However, sumow-ms 4-3

utility staff generally rely on the LOCA organization to manage the technical details associated with compliance with Appendix K and 10CFR50.46. DE&S agrees with the Review Team that Maine Yankee staff, similar to typical utility staff, should have relied on the Yankee LOCA organization for LOCA licensing support. It was YAEC's understanding that Maine Yankee directed YAEC not to independently interface with the NRC staff. This was a root cause in the subsequent licensing process failures. Had YAEC executive management been more aggressive in maintaining the licensing function, YAEC personnel would have been better prepared to make decisions relative to NRC expectation and the interface with the NRC. This situation resulted in an absence of communication with the NRC LOCA technical staff for a number of years.

4.3 Evaluation Model Conservatism Statement of NRC Issue:

During Cycle 14 operations and in the Cycle 14 and Cycle 15 CPAR, l'AEC caused Ml'APCo to us an apparently unacceptable SBLOCA evaluation model which overpredicted core cooling. l'AEC-1868 apparently did not satisfy the requirements of10CFR50.46(a)(1),

because as a result ofincorrect calculations of the penetrationfactors, which arosefrom misapplication of the Alb Chambre penetration correlation, the analysis provides no basis to assumefull penetration of the emergency core cooling system injection and provides no basis to derive the loss coefficient of 600 usedfor the split downcomer nodali:ation. These deficiencies resulted in over prediction of core cooling and overstatement of the conservatism of the model. If the Alb Chambre correlation had been applied correctly, penetrationfactors would have been calculated in the range of 0.6657 to -0.7767, which is a meaningless result because the calculations uvuld have been less than :ero. Such calculations also indicate otherpossible errors in application of the Alb-Chambre correlation. Adequate QA review uvuld have revealed the errors and the unacceptability of the RELAP51'A SBLOCA analysis described in l'AEC-1868.

DE&S Resnonse:

Early applications of the RELAPSYA SBLOCA evaluation model to Maine Yankee 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 vessellower plenum more physical. Eventually an artificially large loss coefficient was introduced in the ' action connecting the two volumes representing a

$ split reactor vessellower downcomer. The due 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 based, in part, on the Alb-Chambre correlation. This correlation was applied to confirm that the amount of ECCS penetration predicted by RELAP5YA was conservative. DE&S agrees with the Review Team that this modeling approach was justified, and the use of a value of 600 obtained an amount of ECCS penetration that was consistent with industry experience. This modeling approach is not expected to result i m m w es 4-4 ,

in an overpredictio'n of core cooling, but since the calculations were not completed for all break sizes, it cannot be definitively confhTned.

An error was made in the application of the Alb-Chambre correlation. This arittunetic 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 calculation (value of 8), 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, which have the meaning of complete ECCS penetration. A person performing a quality assurance review is influenced by the result 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 YAEC 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 input to the correlation, and reasonable and valid results indicating significant ECCS penetration would have resulted. Therefore, although the application of the Alb-Chambre correlation was in error, the modeling which incorporated 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.

DE&S agrees with the Review Team that use of the Alb-Chambre correlation as the basis for the modeling approach which includes the junction loss coefficient of 600 in the reactor vessel downcomer is reasonable. It is recognized that the PCT results of the SBLOCA analyses for some of the break sizes are very sensitive to the value of the loss coefficient. This entire modeling approach was not presented to the NRC for review. YAEC 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 ofinputs. However, the Review Team concludes that due to the non-physical value used and the significance of this input parameter, it is in reality more of a model change, and in fact a very important aspect of the SBLOCA evaluation model. DE&S agrees with the Review Team's conclusion. This model should have been submitted to the NRC for review. It is also recognized that had there been a submittal of the Maine Yankee SBLOCA applications to the NRC, this modeling approach would have been discussed and would have been reviewed.

DE&S supports the Review Team's position that this model could have been approved by the NRC in this form or with some revision.

4.4 Best Estimate Analysis Utilization Statement of NRC Issue:

YAEC caused Maine Yankee to apparently violate 10CFR50.46(a)(1) by relying on an unacceptable SBLOCA evaluation model(Best Estimate RELAP5YA SBLOCA evaluation model) to calculate ECCS cooling performance in preparing a Section 50.59 analysis used to determine if a decrease in steam generatorpressure involved an unreviewed safety question.

Additionally, the proposed BE RELAPSYA evaluation model was not the approved SER computer code and did not comply with 10CFR50. Appendix K requirements. The NRC wmmowncs 4-5

indicates that is also reasonable to conclude that the bianager knew that the analysis which YAECperfonned regarding the effects of a reduction in steam generatorpressure on LOCA analyses as a safery analysis which would be used by biaine Yankee in a Section 50.59 analysis or other safety analysis. In view of the intended use of the YAEC analysis, the Afanager should have provided biaine Yankee with an analysis which met NRC requirements.

DE&S Resnonse:

Due to difficulties in applying the YAEC-1300P Appendix K RELAP5YA SBLOCA code to the Maine Yankee plant, the YAEC LOCA Group initiated a parallel effort to develop a "best-estimate"(BE) LOCA evaluation model, and so indicated in a memorandum. This modeling approach is an alternative approach to the traditional Appendix K approach. Since the YAEC-1300P topical report did not include this BE approach, a separate NRC approval of this method would be necessary prior to implementation. The Yankee LOCA Group Manager clearly understood that the BE approch 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. A report describing the proposed BE methodology was sent to the Maine Yankee staff. The YAEC LOCA Manager was under the impression that the report had been submitted to the NRC for review.

In 1990 Maine Yankee initiated a service request with YAEC 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 YAEC LOCA Group was responsible for addressing the LOCA aspects of this issue. In order to be responsive to the customer's request, YAEC 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 modelis erroneously stated to be the " licensing basis SBLOCA 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 analysis.

The 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, YAEC 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 YAEC would provide these results to Maine Yankee when completed. 'owever, the BE 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 mesomms 4-6 l

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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 YAEC that applied the BE SBLOCA model there were never any statements to the effect that the model was not NRC-approved.

The LOCA 91-04 memorandum actually stated the contrary, which was incorrect. This indicates a lack of administrative contml and communication within the Yankee LOCA Group and within the YAEC Maine Yankee . >jects Group, it also indicates a failure to appreciate the disclosures that should be made regarding application of unlicensed LOCA analysis technology in responding to plant support requests. There was also a failure to correct the situation when the results of unlicensed methods were found to have been released to the customer.

EE&S disagrees with the NRC position that unlicensed LOCA analyses cannot be used in performing some scoping safety evaluations, including input to 50.59 evaluations. Such applications can be appropriate provided that the application does not replace the analysis of record, and provided that the use of the unlicensed 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. YAEC's failure in this situation was in incorrectly characterizing the BE SBLOCA analysis as the licensing basis, and furthermore not stating that the analysis used unlicensed methods including restrictions on its use. In providing any licensee with a document, the supplier must consider possible uses of the document. If the use of the document must be restricted, such as one which uses unlicensed methods, the document must clearly state so.

It should be noted 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 be expected to not be significant. This is particularly true given that the analyses of record showed the SBLOCA PCTs to be significantly lower than the LBLOCA PCTs. An evaluation could have beenjustified without any SBLOCA analysis. An evaluation could also have been justified using BE analysis methodology provided that sufficient qualification was provided, and provided that the analysis of record was not replaced. Since such qualification was not provided with this evaluation which included references to the application of the BE SBLOCA methodology, this approach was not justified.

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5,0 DE&S RESPONSE TO NRC DENIAND FOR INFORh1ATION Section V of the Demand requested DEAS " nrovide information with respect to two topics.

Section 5.1 of this report provides the De ' sponse to the NRC question regarding continued performance of safety-related analyses for dkC Licensees. Section 5.2 of this report provides the DE&S response to the NRC question regarding the willfulness of DE&S personnel behavior with respect to the actions described within the Demand.

5.1- Continued Performance of Safetv Related Annivses Statement of NRC Information Need:

"[ Provide] an explanation why, in view of the matters setforth above [in the Demand], the NRC should pennit any NRC Licensee to use the services of YAEC LOCA Group and/or DE&S, to the extent that YAEC LOCA Group was transferred to DE&S, to perfonn LOCA analyses or any safety-related analysis to meet NRC requirements."

DE&S Response:

DE&S is committed to the highest standards ofintegrity and technical quality upon which its clients and the NRC can rely upon without reservation. As a wholly-owned affiliate of Duke, DE&S carries forward to the nuclear industry Duke's nuclear tradition built through four decades of nuclear design, construction, licensing, and operations experience. This tradition not only encompasses strict attention to and understanding of regulatory and industry requirements, but also emphasizes a technically strong workforce who are held accountable for their performance. The principles of: (1) adherence to NRC requirements, (2) effective management control of safety related activities, (3) accurate and complete communication with licensees and the NRC, and (4) conduct of work in accordance with NRC requirements are fundamental to the Duke /DE&S nuclear tradition. As a provider of management and technical services to the nuclear industry, nE&S cannot succeed as a commercial business without the total confidence ofits clients ant .ne NRC in this DE&S commitment to integrity and quality.

DE&S has completed an extensive set ofindependent assessments of the acquired YAEC organizations. These assessments focused on ensuring that the acquired YAEC organizations are currently providing quality technical products and ser ices to their NRC licensee clients, in a manner that is compliant with NRC requirements. In fact, many of these same clients conducted licensee oversight activities during these assessments. The DE&S assessments addressed both process and technical aspects of the acquired YAEC organizations. Significant focus was placed on LOCA and associated safety analyses, as well as engineering work practices and products, quality assurance activities, and the effectiveness of corrective actions implemented by YAEC. The specific DE&S assessments conducted wete:

1. Maine Yankee SBLOCA Analysis Assessment - an independent technical review of the technical issues described in the Demand.

2. Personnel Behavior Assessment - an independent legal review of the " willfulness" of the actions and events summarized in the Demand.

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3. Root Cause and Corrective Action Assessment - an independent review of the YAEC root cause assessment and corrective actions.

4. Engineering Process Assessment - an independent review of engineering / technical work practices used and work products produced during the past 3 years, other than those associated with LOCA analyses.

! 5. Quality Assurance Assessment - an independent review of QA audits performed during the past 3 years.

6. Vermont Yankee Safety Analysis Assessment - an independent technical review of:

(i) Basis for Maintaining Operation (BMO) documents, (ii) Cycle 20 reload analyses, and (iii) containment analyses.

7. Seabrook Safety Analysis Assessment - an independent technical review cf radiological consequence, non LOCA transient analyses.

DE&S also assessed the underlying causes identified by YAEC that led to the events described in the Demand. These are:

1. Definition of organizational roles, responsibilities and communication requirements between YAEC and MYAPCo was less than adequate,

2. Training of YAEC personnel was less than adequate, and

3. Clarity and specificity of procedures was less than adequate.

Descriptions of the DE&S assessments, conclusions drawn from these assessments and follow-up actions identified by DE&S are provided in Sections 2.2. and 2.3. More detailed descriptions are provided in Appendices C - I.

DE&S management evaluated the assessment results to ensure that products and services being provided are high quality and meet the needs of DE&S clients and the NRC.

Additionally, areas for DE&S management attention were identified to ensure that products and services provided by the acquired organizations continue to meet the standards and expectations of DE&S management, the NRC and DE&S clients throughout the transition and integration of the acquired YAEC organizations into DE&S.

DE&S has concluded that the acquired YAEC organizations are providing quality products and services. In fact, DE&S did not identify any areas where the current performance of, or the products and services provided by, the acquired YAEC organizations is unacceptable. DE&S has also noted that meaningful improvements have been made in the quality of YAEC procedures and the emphasis these procedures place on NRC reouirements and NRC reporting. Additionally, DE&S has consistently found that the technical quality of the work products currently being produced, as well as the professionalism and technical competence of the workforce, are high. In summary, the DE&S assessments have shown that the acquired

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organizations fundamentally are technically sound, producing quality products and senices in accordance with NRC requirements and industry practices.

The DE&S assessments and management evaluation did identify areas that warrant DE&S management focus to ensure a successful transition and integration of the acquired YAEC organizations into DE&S. Although a number of work activities are required to effectively transition and integrate an acquired organization into DE&S, the transition assessments generally highlight a few items warranting management attention to ensure a successful transition and integration. In the case of YAEC, four items were identified. These four items are: (1) organizational roles, responsibilities, and communications, (2) personnel training, (3) nuclear regulation support, and (4) accurate and complete documentation. DE&S follow-up actions that address these fou items are described in Section 2.3. Each of these actions is either completed or underway. These actions are summarized as:

(1) Jointly defining and documenting with each nuclear client formerly served by YAEC the organizational roles, responsibilities, expectations, and communication requirements between DE&S and the client.

(2) Transitioning ongoing work performed by the acquired YAEC organizations into the DE&S project planning process. This process includes identifying requirements for clear definition and documentation of project level technical and organizational communication requirements, and training of project personnel to these requirements.

(3) Strengthaning the executive leadership in the Bolton office by naming, effective March 1, 1998, William H. R . sin as DE&S Vice-President of Nuclear, Fuel, and Quality Assurance Services. This organization includes the former YAEC Nuclear Engineering, Fuel Management, and Quality Assurance Departments.

(4) Communicating formally to the acquired YAEC organizations DE&S management expectations on: (1) the role and responsibilities of DE&S as a commercial provider on nuclear services and products to NRC Licensee clients, (2) the requirement to ensure that products and services meet the needs of the clients, and (3) the requirement to ensure that supporting documentation files are complete and accurate.

(5) Strengthening employee training through prompt integration of the acquired YAEC training functions into DE&S training functions, supporting the prompt identification of training needs and implementation of a systematic program for meeting these needs.

(6) Jointly reviewing with DE&S clients their client-specific documentation to ensure that supporting documentation meets each client's unique business and regulatory needs.

(7) Strengthening nuclear regulatory support to ongoing work through the prompt integration of DE&S licensing functions into the acquired YAEC organizations in such a manner that appropriate licensing support is provided.

In summary, DE&S found through its assessment process that the performance of, and the products and services provided by, the acquired YAEC organizations are acceptable.

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i Meaningful improvements have been made in the quality of YAEC procedures and the emphasis these procedures place on NRC requirements and NRC reporting. The technical quality of the work products being produced, as well as the professionalism and technical competence of the workforce, are high. DE&S is systematically transitioning the acquired YAEC organizations to the DE&S project plannmg and work execution programs and procedures. The DE&S project planning process emphasizes clear definition and documentation of project-specific requirements, both technical and organizational, and training of project personnel on these requirements. DE&S has strengthened the Bolton office leadership with a proven nuclear industry executive, William H. Rasin. DE&S is also currently engaged with each nuclear client formerly supported by YAEC to ensure that roles, responsibilities, and expectations are mutually understood and documented. For these reasons, combined with the nuclear commitment and experience represented by DE&S, the NRC should have a high degree of confidence that safety-related analyses, products, and services provided by DE&S to NRC power reactor licensees comply with the principles of: (1) adherence to NRC requiremen': (2) effective management control of safety-related activities, (3) accurate and complete communication with licensees and the NRC, and (4) conduct of work in accordance with NRC requirements 5.2 Willfulness of Personnel Behavior Statement of NRC Information Need:

"[ Provide) an explanation why the NRC should not consider the inadequate analyses, which apparently caused MYAPCO [ Maine Yankee Atomic Pour Company] to be in violation of NRC requirements, to be the result of willfid, either deliberateness or careless disregard, on the part of YAEC and/or DE&S personnel."

DE&S Resnonse:

A common finding or observation of the DE&S assessment teams was the high degree of professionalism and openness of the DE&S employees within the acquired YAEC organizations. DE&S has found no evidence that would question the sincerity and dedication of this workforce. The Demand focused on the actions of two individuals in relation to whether willfulness, either deliberateness or careless disregard, existed. DE&S found no willfulness on the part of the two individuals. DE&S, therefore, believes that the Manager and Lead Engineer are capable of conducting their activities in the future in conformance with NRC requirements.

Additionally, DE&S concludes that there was nothing found by the independent assessment of the conduct of YAEC (or DE&S) or its personnel that would prevent DE&S activities from being performed in full compliance with NRC requirements and erpectations. Appendix D provides detailed assessment of the willfulness associated with the events described within the Demand.

Willfulness, as used by the NRC, embraces a spectrum of actions ranging from deliberate intent to violate or falsify to and including careless disregard for NRC requirements. Deliberate misconduct is defined as "an intentional act or omission that the person knows would cause a licensee to be in violation of any regulation or other NRC requirement." On the other hand, a finding of careless disregard indicates that the person acted with reckless indifference to a

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i l v requirement or with disregard or utter unconcern of the consequences of whether there was compliance. The existence of a reasonablejustification for an action would defeat a charge of willfulness despite the fact that a person undertook an action that was ultimately found to violate NRC requirements.

DE&S found no actions on the part of any individuals associated with the specific issues contained in the Demand that would constitute deliberateness. Allindividuals where found to be gen, honest and communicative. No specific intent to violate any NRC regulation has been ideutified. The DE&S evaluation focused on whether there existed careless disregard for Commission requirements.

There were four specific allegations in the Demand, the first two of which had a common factual underpinning. The NRC alleged that because not all points of the SBLOCA spectrum could be reliably calculated by the code used by YAEC for the Maine Yankee facility, the requirements of 10CFR50.46 were not met. The RELAP5YA SBLOCA TechnicalIssues Assessment (Technical Review Team) concluded that the standard industry practice, as utilized by experts in the LOCA field, was that the code should have the capability of analyzing all points within the prescribed spectrum. YAEC had taken the position that the identification of the limiting break, combined with a sufficient understanding of the physical phenomena which were occurring over the entire small break region, provided compliance with 10CFR50.46. It is recognized that NRC expectations are not completely documented regarding 10CFR50.56, but rather had been communicated to the LOCA community through its interactions with NRC.

Utilizing the appropriate legal standard, and bearing in mind that the YAEC LOCA Group was ,

isolated and not part of the LOCA community, it was determined that the applicable regulation i could be read as it was by YAEC. Actions to implement the LOCA Group's interpretation of the regulations did not evidence reckless indifference and, thus, no careless disregard was found.

With regard to the second issue, the NRC asserted that YAEC caused a violation of 10CFR50.9 in that it provided inadequate information to Maine Yankee regarding the SBLOCA which did not reveal the code inadequacies discussed above. Many of the same considerations apply to this issue as to the previous one. In addition, the Technical Review Team found that on the whole the document submitted to Maine Yankee was sufficiently complete and accurate when judged using the perspective ofits intended audience,it, one knowledgeable in the field, such as an NRC reviewer. Under the circumstances, DE&S found that careless disregard of the regulations was not present.

The third issue relates to an assertion by the NRC that Maine Yankee had not provided a technical basis for one element of the SBLOCA analysis, the loss coefficient for the split down::omer nodalization, and, as a result, there was overprediction of core cooling and overstatement of the conservatism of the model. The Technical Review Team determined that the modeling approach utilized was reasonable and consistent with industry experience. The Technical Review Team determined that a deficiency in the quality assurance of a confirmatory calculation existed, but there was a reasonable explanation as to why it occurred, and that the evaluation undertaken by YAEC was appropriate. No inadequate analysis existed and the issue of deliberateness or careless disregard did not arise, mamomacs 5-5

With regard to the fourth and last issue, the NRC asserted that YAEC memoranda used an unacceptable Best Estimate model rather than an approved Evaluation blodel in evaluating the f effect of a decrease in steam generator pressure on the peak clad temperature in the small break region when it should have known that such analysis would form the basis of a 10CFR50.59 analysis. The Technical Review Team determined that methods o',her than the approved Evaluation Model could have been utilized for the work undertaken. However, the Technical Review Team felt that limitations on the use of such methods should be stated. Here certain memoranda mischaracterized the Best Estimate model as the approved 10CFR50.46 Evaluation Model for the facility.

With regard to the use of the memoranda, DE&S determined that it was at unreasonable to have failed to contemplate that such work products would be used in a 50.59 analysis inasmuch as they either represented scoping calculations and/or were not thought at the time to be addressing design basis issues. However, it is also understandable that a Maine Yankee employee not expert in LOCA analyses could use these memoranda in performing the assigned 50.59 analyses. In evaluating whether careless disregard had occurred, it wa.; noted that the issue of whether decreased steam generator pressure impacted the design basis did not mature untillate 1992 (i.e., some months after receipt of the last memorandum), as evidenced by the Maine Yankee NRC Resident Inspector's suggestion that a 10CFR50.59 analysis should be performed. It was also noted that the appreciation of the necessity for evaluating degraded conditions pursuant to 10CFR50.59 was in a state of transition a the time, with the NRC publishing Generic Letter 91-18 in late 1991. DE&S concluded that while a misstatement was made that the Best Estimate model was the licensing basis SBLOCA analysis and while it was used without prior discussion with the NRC, the actions of YAEC personnel did not meet the test for careless disregard of the regulations in that DE&S could not conclude that there existed a reckless disregard or careless indifference towards its responsibilities or the consequences of the actions taken under the circumstances assumed. Indeed, on the very date Maine Yankee's 10CFR50.59 analysis was internally approved, YAEC furnished Maine Yankee a draft 50.59 analysis on the precise subject based on its RELAP5YA Appendix K Evaluation Model.

In conclusion, DE&S determined that while in certain instances there may have been inadequate analysis associated with the SBLOCA analysis, there was neither deliberateness nor careless disregard resuking from the deficiencies discussed in the Demand. DE&S concludes that there was nothing identified by the independent assessment of the conduct of YAEC (or DE&S) or its personnel that would prevent DE&S activities from line conducted in full compliance with NRC requirements and expectations.

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A.1 COMPANY OVERVIEW A.I.1 Comnany Illstory Duke Energy Corporation (Duke) began offering management and technical services to electric utilities and related businesses in 1982. Five years laMr, Duke Engineering &

, Services (DE&S) was incorporated as a wholly owned affiliate of Duke to provide these l services to clients. (Figure A-1 shows the relationship of DE&S within the Duke corporate structure.) DE&S has grown continuously and expanded both the range and scope of services offered. Consistent with these developments, DE&S has gained considerable experience in working under a variety of organizational structures, teams and implementation formats. Because Duke is the only U.S. utility performing all major elements of a commercial nuclear program with its own resources, DE&S is uniquely qualified to bring the full spectrum of commercial nuclear experience to clients. Through DE&S, Duke's four decades of nuclear design, construction, licensing, and operations expertise is made available to other clients worldwide. With one of the strongest backgrounds in the nuclear industry, DE&S offers clients a depth of engineering experience and expertise available through only a few companies in the world..

In 1996 DE&S expanded its resource base and scope of nuclear services available to clients with the acquisition of VECTRA Technologies, Inc.'s (VECTRA) nuclear engineering, power services and government services businesses. DE&S continued its commitment to the nuclear industry in 1997 with its acquisition of the Nuclear Services, and related support divisions, of YAEC. DE&S currently provides engineering, environmental, hydroelectric, nuclear, renewable energy, power transmission, oil / gas production, and oil / gas processing services to clients world wide. Figure A-2 provides the current DE&S corporate structure.

A.I.2 History of Nuclear Involvement Nuclear power generation is a success story at Duke. From the experimental work of the Parr Station beginning in 1957, through the completion of Catawba Nuclear Station, Duke has performed and excelled in the complete process of licensing, procuring, designing, constructing, operating, and maintaining commercial nuclear pov.er pisnts. With more than 20 years oflarge capacity nuclear power generation experience, Duke's nuclear plants are continually recognized for their outstanding records in operational effectiveness. Oconee Nuclear Station holds the record for total electric generation in the U.S. McGuire Nuclear Station has received the top fuel efficiency ranking for multi-fuel stations. Catawba Nuclear Station has received the best thermal efficiency rating among pressurized water reactors in the U.S.

Duke's history of nuclear involvement began with the Oconee Nuclear Station. Its three units were completed in just seven years at a cost of less than half the industry average at that time. Today, Oconee remains one of the most efficient and least expensive generating stations in the country. In July 1991, Oconee Unit I became the first U.S. nuchar unit to produce 100 billion kilowatt-hours. McGuire was the next nuclear station and has consistently been one of the most efficient nuclear plants in the nation. When completed in 1986, Catawba, like the other stations, showed construction costs significantly below w m m o w as A-1

I industry averages. This station also enjoys a reputation foi efficiency in operations and I

maintenance.

In 1985, the Electric Power Research Institute (EPRI) began to develop plans for a standardized Advanced Light Water Reactor (ALWR) to meet the needs of utilities in the future. This program was led by the utilities in order to assure that the needs of the owner and operator were addressed in the next generation of nuclear plant designs. Based on ,

extensive nuclear experience. Duke played a central role in the development of the EPRI ALWR Requirements Document and subsequent developments such as the First-of a-Kind Engineering Program.

A.1.3 Industry Achievements Year after year, Duke's stations rank among the top in the nation for efficiency. The company's nuclear program is widely recognized and respected. According to Electric Light

& Pour, all three Duke nuclear plants were ranked in the top twenty nuclear plants in 1993 and 1994, based on lowest total O&M costs. Duke also aims for each station to be in the top 25 percent of the industry in capacity factor. The industry's top quartile capacity factor in 1994 was 90 percent.

Recently, Duke was honored by being named the first three-time recipient of the electric utility's highest honor - the Edison Award. The award, named for inventor Thomas Edison, was also presented to Duke in 1972 and 1984.

Consistent with the company's goal ofincreasing nuclear plant safety and reliability, Duke played a leadership role in establishing two important industry exchange and oversight groups - the Institute of Nuclear Power Operators (INPO) and the World Association of Nuclear Operators (WANO). At WANO's inaugural meeting in Moscow in May 1989, Duke's former chairman, William S. Lee, was named first president, serving until 1991. In addition, Lee served as Chairman of The Board for INPO fnom its founding in 1979, until 1982.

Duke was one of the first utilities in the nation to establish its own training program for nuclear plant operators. Duke offers comprehensive instruction for nuclear plant operators as well as for technicians in health physics, chemistry, maintenance, instrumentation and electronics. In August 1983, the Oconee Nuclear fa:ility became the first INPO accredited Plant Operator Training Program facility. All operator training programs are now accredited by INPO and have been approved by the NRC. Duke utilizes its on-sitc, full scope, control room simulators for initial license and requalification training of operators. Ali simulators are maintained with the highest degree of fidelity to the reference plant according to established standards for simulator configuration management.

A.I.4 Ennineerinn Exnertise DE&S brings a wealth of engineering expertise to clients' projects in civil / structural, electrical /I&C and mechanical / nuclear disciplines. Since 1904, when Duke Power Company's first hydroelectric station was constructed, employees have performed siting,

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quality assurance, design, licensing, construction, operations, maintenance and modifications for allits nuclear, fossil, hydroelectric .md pumped storage facilities. A complete mu'.ti discipline staffis maintained to provide ongoing engineering support to station opea ans and maintenance personnel. This stad assists with the resolution of daily problems involving all systerns and equ:pment, participates on evaluation teams, and provides engineering designs for the m'jor plant modifications performed each year. Many DE&S employees began their careers with Duke.

DE&S/ Duke employees represent all facets of engineering, including:

Engineers with civil / structural engineering expertise who are experienced in all facets of civil / structural engineering including foundation and structural design, piping analysis, support / restraint design and environmental engineering.

Engineers with mechanical engineering expertise and engineers with nuclear engineering expertise who are experienced in all facets of mechanical / nuclear engineering including mechanical systems and equipment design, HVAC, fire protection, piping, materials, nuclear fuels, safety analysis and radiation protection.

Engineers with electrical /I&C engineering expertise with experience in all facets of electrical /I&C engineering including electrical systems and equipment design and I&C design.

Approximately 75 percent of DE&S engineers are registered professional engineers.

The DE&S northeast regional office, located in Bolton, Massachusetts, is comprised of a 500-member technical staff. This team includes nuclear, systems, mechanical and electrical engineers, licensed asbestos abatement supervisors, maintenance, welding and materials, nondestructive examination, instrumentation control, radiological and civil engineers; safety analysts; nuclear physicists; environmental, materials and metallurgical scientists; computer systems analysts; quality assurance, training and administrative support personnel. This staffing represents a consolidation of the existing DE&S northeast region staff with the acquired staffing of YAEC.

A.1.5 Recul9torv/licensine Exnertise DE&S maintains a comprehensive licensing and regulatory compliance staff capab!e of dealing with the interwoven and complex regulations controlling nuclear and environmental work. Ranging from obtaining environmental permits at the local level, to supplementing the operational experience programs of a plant on the watch list, to licensing an ISFSI, DE&S licensing and regulatory compliance personnel and programs support nuclear clients every day. The YAEC acquisition compliments DE&S expertise with its direct experience in preparation for and participation in the NRC decommissioning hearing process, including interactions with stakeholders, based on the successful disposition of contentions raised in Yankee Rowe litigation.

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In addition, DE&S experience includes development of licensing strategies for spent fuel

storage under 10CFR Part 72 and Yankee Rowe's 10CFR Part 71 Certificate of Compliance application to the NRC and DOT for transportation of the Yankee Rowe steam generators as their own shipping casks.

DE&S assisted in the development of the Consumers Power Company's Big Rock Point Decommissioning Plan by providing both licensing and engineering services. DE&S advised Consumers Power on licensing strategies for addressing regulatory program requirements in a permanently shutdown mode and coordinated the development of the Decommissioning Plan and the accompanying Environmental Report. Additionally, DE&S supported licensing the recently completed steam generator replacement at St. Lucie 1.

DE&S also manages the effort of Louisiana Enrichment Services (LES) to license its Chiborne Enrichment Center, a proposed uranium enrichment facility to be located near Homer, Louisiana. This licensing effort is considered by many to be a bellwether for the future of commercial nuclear power and is being closely monitored by many inside and outside of the U.S. nuclear industry.

DE&S has, from the enactment of the National Environmental Policy Act (NEPA),

addressed the many radiological and non-radiological aspects oflicensing and operating nuclear generating facilities. This long history includes the preparation of Environmental Reports to assess the environmentalimpacts from proposed and operating facilities as well as facilities that may be allowed to operate beyond the original 40-year license term.

A.I.6 Nuclear Onerations and Maintenance Exnertise As an affiliate of Duke Energy, DE&S offers nearly three decades of nuclear operations and maintenance experience, along with a unique blend of technical excellence, cost consciousness, regulatory adherence, environmental sensitivity and quality assurance. In 1995, Duke's three nuclear plants operated at a capacity factor of 907c, compared to the national average of 787c. Nuclear operations is a primary responsibility of DE&S as a member of the Management and Integrating Contractor teams at the U.S. DOE's Hanford Site and Idaho National Engineering Laboratory.

DE&S works closely with clients to shorten outages, enhance efficiency and upgrade plant output, as well as assist with operational readiness reviews, integrated outage management, engineering support, risk assessments, in-service testing, instrument calibration and system modifications.

DE&S' shift advisor / mentor program provides real-time assessment, technical advice and operating philosophy coaching to on-shift operations management. By identifying areas for improvement and implementing solutions, DE&S shift advisors and mentors have improved operating personnel performance and overall plant performance at a number of U.S. nuclear stations including Comanche Peak Steam Electric Station, Indian Point 3, Watts Bar Nuclear Plant and Salem Nuclear Generating Station.

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DE&S has r.ssisted Salem in support of restart efforts. In May 1995, Public Service Electric

( & Gas Co. (PSE&G) began draft.;ng a starting lineup of DE&S system planners, system engineers and operations shift mentors. Teamed with existing PSE&G employees, DE&S is responsible for preparing Salem Units 1 and 2 for restart, obtaining all necessary regulatory approvals and assessing the plant's System Certification Process. DE&S has provided similar services, to Carolina Power & Light's Brunswick Nuclear Station, Houston Power &

Light's South Texas Project and New York Power Authority's Indian Point 3.

DE&S has developed state-of-the-art technologies and software to help streamline maintenance processes and meet the industry's continuous quest for safe, efficient and effective nuclear operations. DE&S provides full-scope preventive and predictive maintenance services, maintenance program optimization technologies and outage management services.

DE&S' Outage Maintenance Assessment (OMA) services help clients improve nuclear outage performance by reducing refueling outage duration and maintenance costs, while improving unit power availability. Using a computer software program, the OMA Analyzer, DE&S performs comprehensive analysis of nuclear outage maintenance work activities and tailors recommendations to meet plant-specific needs. DE&S has provided OMA services to Commonwealth Edison's LaSalle and Dresden nuclear stations, in addition to Duke Power plants.

DE&S is also helping clients design cost-effective preventive and predictive maintenance programs.

DE&S brings a high level of expertise in pumps, motors, breakers, wiring and valves to every predictive and preventive maintenance application.

DE&S has developed Maintenance Plusti, a comprehensive maintenance optimization process, which streamhnes nuclear stations' processes and provides consistent implementation between nuclear sites. Supported by a software application, this program features a proven strategy for improving equipment reliability, maximizing process availability and reducing maintenance costs.

A.2 YAEC ASSET ACOUISITION On December 1,1997 DE&S completed the acquisition of the assets of YAEC's Nuclear Services Division (NSD). The NSD assets acquired consist mainly of the engineering and environmental services business previously conducted by YAEC. DE&S acquired substantially all of the assets of the NSD, including NSD employees. The remainder of YAEC is still a separate corporate entity with no change in ownership as a result of the acquisition. A brief description of the major departments formerly within YAEC NSD and their disposition is provided below.

NORTHEAST UTILITIES PROJECT - The NU Project was established to support Northeast Utilities demands. The project is staffed with electrical, mechanical, systems and 1&C engineers mostly to perform modification design. A portion of these engineers have been permanently unuse=ms A-5  :

1 i

1 transferred to Seabrook. The remaining engineers have been assigned to support various DE&S i

projects.

l CONNECTICUT YANKEE PROJECT - The project consists of mechanical engineers, electrical engineers, I&C engineers, and systems engineers dedicated to the decommissioning of CY. The plans are to eventually move the staff to support the decommissioning of Connecticut Yankee ENGINEERING SERVICES (ESD) - This is a multi-disciplined de,.artment set up by YAEC to sell engineering services to non-affiliate clients. Personnel within this department have been assigned to support other DE&S projects.

ENVIRONMENTAL ENGINEERING - This department has extensive experience in the areas of Environmental Sciences, Emergency Planning (EP), Radiological Engineering (RE) and Radiation Protection (RP) Consulting & Health Physics (HP) Site Services. The EP group provides centralized Emergency Planning services to Vermont Yankee, Maine v ankee, Seabrook and Pilgrim. The RE, RP and HP groups are also supporting decommissioning of Yankee Rowe and Connecticut Yankee.

This department also provides Environmental Services for other New England non-nuclear utilities, other domestic and international nuclear utilities, the domestic commercial industries and the Federal Government.

ENVIRONMENTAL LABORATORY - This department works closely with Environmental Engineering and provides laboratory services in the following areas: (i) radiological sample analysis, (ii) in-vivo and in-vitro radio-bioassay, (iii) in-plant quality assurance program, (iv) personnel and environmental dosimetry monitoring, and (v) field and plant measurements FUEL MANAGEMENT - The department provides nuclear fuel procurement, fuel management and economic analysis, spent fuel storage, fuel and component fabrication engineering, core performance assessment and special nuclear material accountability services.

MAINE YANKEE PROJECT - The staff assigned to the Maine Project have been permanently transferred to Maine Yankee.

NUCLEAR ENGINEERING - This is a specialized engineering department created to resolve technical issues in support of continued plant operation. The services provided include: core reload analysis, plant safety / performance analysis, cost / risk-benefit and reliability analysis and specialized thermal hydraulic analysis.

PLANT SUPPORT (PSD) - The services provided by PSD are: in-service inspections (ISI)/non-destructive examinations (NDE), fire protection, materials evaluation, welding, maintenance, system engineering, in-service testing, thermography, computer aided design (CAD) and construction.

VERMONT YANKEE PROJECT - All of the staff assigned to the Vermont Project have been transferred permanently to Vermont Yankee.

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YANKEE ROWE PROJECT - This staff that supports the decommissioning of Yankee Rowe. The staff consists of systems engineers, licensing engineers, non-licensed operators, maintenance, security, HP, training p rsonnel, etc.

The other departments acquired by DE&S that are not discussed above include: Information  !

Resources, Treasury, Administrative / Security Services, and Health and Salty. All other aspects of ,

YAEC, including ownership of Yankee Rowe, remains unaffected by the DE&S acquisition.

A.3 DE&S PROGRAM DESCRIPTION This section summarizes the DE&S Quality Assurance Program, Design Control Program, and Project planning process A.3.1 OnnHty Assurance Pronram A.3.1.1 - Organi7ation The DE&S organizational structure for activities affecting quality is shown in Figure A-2. The functional responsibilities, levels of authority, and lines of communication for the various organizational entities are described below.

The President, DE&S is responsible for quality and is the highest level of management responsible for establishing DE&S quality policies, goals, and objectives. The President, DE&S has documented DE&S' commitment to quality in the " Quality Assurance Program Policy Statement." The Quality Assurance Program Policy Statement is contained in the DE&S QA Program Description (QAPD) and the DE&S QA Procedures Manual.

The Business Unit Managers are responsible for technical quality achievement for nuclear work assigned to their business units under the controls of the DE&S QA -

Program. These Business Unit Managers report to the Executive Vice President who reports directly to the President and have full technical authority over activities afDeting quality within their business unit.

The QA Manager reports to the President for quality assurance activities and is responsible for establishing and maintaining the DE&S Quality Assurance Program and verifying its effective implementation at all DE&S facility locations. The QA Manager is independent of managers responsible for performing quality affecting work and is sufficiently independent of cost and schedule considerations when in conflict with quality considerations. The QA Manager and President, DE&S jointly approve the DE&S QA Program Description and QA Procedures Manual. QA Program specific manuals (e.g., Standards Laboratory Operations Manual, Qualification and Testing Facility Reference Manual, Metallurgy Reference Manual, NDE Program Manual, Geotechnical Services Testing Manual, etc.) are jointly approved by the responsible Business Unit Manager and the QA Manager.

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L The QA Manager has direct access to the President and Business Unit Managers and is able to identify quality problems; initiate, recommend, or provide solutions; and verify hnplementation of solutions; and assure that further processing, delivery, installation, or use is controlled until proper disposition of a non-conformance, deficier.cy, or unsatisfactory condition has occurred.

The QA Manager has a quality assurance staff, condsting of auditors and support personnel. Tnese individuals report directly to the QA Manager. QA personnel may be located at she offices; however, they remain in contact with the QA Manager for direction.

All DE&S personnel have the authority to stop work. The ultimate authority to stop or resume work resides jointly with the responsible Business Unit Managers and the QA Manager. Matters that cannot be resolved at this level are escalated to the President for resolution.

A.3.1.2 Program Descrintion The DE&S QA Program meets 10CFR50, Appendix B,10CFR21, and ASME NQA-1-1989, through the NQA-lb-1991 Addenda. In addition, the DE&S QA Program can be tailored to meet other quality assurance requirements invoked by clients, such as ANSI N45.2, daughter standards of ANSI N45.2, and 10CFR72.

The DE&S QA Program Description (QAPD) Manual describes DE&S commitments to 10CFR50 Appendix B,10CFR21, and ASME NQA-1. The QAPD states policies, assigns responsibilities, and specifies requirements governing siting, design, construction, operation, and decommissioning of nuclear facilities. Snecific processes and controls on these processes, which implement these commitments, are specified in QA Procedures contained in the QA Procedures Manual and in QA Program specific manuals that have been approved by the QA Manager.

The QA Program provides for the planning and accomplishment of activities affecting quality under suitably controlled conditions. Controlled conditions include the use of appropriate equipment, suitable environmental conditions for accomplishing the activity, and assurance that prerequisites for the given activity have been satisfied. The QA Program provides for any special controls, processes, test equipment, tools, and skills to attain the required quality and for verification of quality. To the extent necessary, requirements contained in this QAPD are invoked on DE&S suppliers and the supplier's subcontractors.

The DE&S QA Program is mandatory for all activities affecting quality. When work cannot be acm'olished as specified in the implementing QA Procedure, or accomplishment of susa work would result in an adverse condition, work is stopped until proper corrective action is tak;n.

A graded QA program is implemented for Quality Affecting Activities. Categories are defined as:

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OA Condition 1 - Nuclear Safety Related OA Condition 2 - Radioactive Waste Related OA Condition 3 - Fire Protection Related (nuclear facilities)

OA Condition 4 - Seismic Category II Related This QAPD describes controls for QA Condition 1. Alternate controls for QA Conditions 2,3, and 4 are specified in implementing QA Procedures.

This quality assurance program may be used for non-nuclear projects as deeme necessary by the Business Unit Manager implementing the client contractual requirements.

A.3.1.3 Ouality Assurance Training Quality Assurance training is provided to all personnel performing activities affecting quality as determined by supervision. All DE&S permanent personnel, as well as loaned / pan-time personnel who perform quality affecting activities, receive Quality Assurance Indoctrination Training. QA Indoctrination Training includes general criteria, including applicable codes, standards, QA Procedures, QA Program elements, and job responsibilities and authorities. Detailed QA training is provided on the QA Program, applicable Project Plans and procedures prior to an employee beginning work affecting quality. Supervision is responsible for assuring personnel performing work under their supervision are trained.

The QA Training Coordinator (QATC) is appointed by the President, DE&S and is responsible for coordinating QA training activities for DEk3. The QATC serves as a centralized training support service for supervision in coordinating training and maintaining QA training records. This esponsibility is carried out as support for line management. Appropriate DE&S supervisory personnel are responsible for determining the type and extent of the training to be provided to an individual, and ensuring that the training is properly documented for personnel pe forming activities affecting quality.

A.3.1.4 Management Assessments The President conducts a Corporate Management Assessment every two years to determine if the Quality Assurance Program is effective on a corporate-wide basis.

The President appoints a team of DE&S managers or supervisors to conduct this assessment. Recommendations are provided to the President for action.

Business Unit Managers and the QA Manager conduct an Internal Management Assessment annually of QA activities under their control. The managers report results of the Internal Management Assessments to the President for review.

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A.3.1.5 Oualification/ Certification of Nondestructive Examination (NDE) Personnel DE&S maintains an NDE program described in the NDE Program Manual and may also use an approved outside agency to provide training and to administer examinations for certification by DE&S of qualified NDE personnel. In general, DE&S uses Duke as the approved outside agency. The ASME Recommended l Practice No. SNT-TC-1 A, Personnel Qualification and Certification in l Nondestructive Testing,1984 Edition applies as requirements for NDE personnel.

A.3.1.6 OA Program Renoning to Management Management is regularly informed on the quality of DE&S quality affecting work via results of audit reports, internal surveillance reports, Problem Investigation Reports, and management assessments. Corrective action is initiated and monitored until completion as necessary.

A.3.1.7 Oualification/ Certification ofInsnection and Test Personnel DE&S uses an approved outside agency to provide training and to administer examinations for certification by DE&S of qualified inspection and test personnel.

In general, DE&S uses Duke as the approved outside agency. Inspection and test personnel are qualified and certified in accardance with ANSI /ASME N45.2.6 -

1978, Qualifications ofInspection, Examination, and Test Personnel.

A.3.1.8 Oualification/ Certification of Personnel for ASME Code Desien Qualifications and certification of personnel (Registered Professional Engineers) engaged in certifying ASME Section III Code designs and documents, when required by ASME Section III, is conducted in accordance with ANSI /ASME N626.3, Qualifications and Duties of Specialized Professional Engineers.

A.3.2 Desien Control Procram Measures are established in QA Procedures to assure thu applicable requirements are correctly translated into design documents. Design inputs are specified. Controls are established for the selection and suittbility of application of materials, parts, equipment, and sucesses that are essential to the functions of structures, systems, and components.

Controle are established for the identification of design interfaces among participating organizations. QA Procedures specify controls for design input, design process, design verificatior,. design changes, and approval. These procedures include appropriate quantitative and/or qualitative acceptance criteria for determuung that important activities have been satisfactorily accomplished. Design documents are prepared, verified, and approved by qualified individuals. Design verification is performed by individuals who are independent of the preparer or approver. Design is verified by one or more of the following verification methods: design reviews, alternate calculations, or qualification tests. The method of design verification and results are documented. Design changes are governed by control measures commensurate with those applied to the original design. Computer

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OA Condition 1 - Nuclear Safety Related OA Condition 2 - Radioactive Waste Rclated QA. Condition 3 - Fire Protection Related (nuclear facilities)

OA Condition 4 - Seismic Category II Related This QAPD describes controls for QA Condition 1. Alternate controls for QA Conditions 2,3, and 4 are specified in implementing QA Procedures.

This quality assurance program may be used for non-nuclear projects as deemed necessary by the Business Unit Manager implementing the client contractual requirements.

A.3.1.3 Ouality Assurance Training Quality Assurance training is provided to all personnel performing activities affecting quality as determined by supervision. All DE&S permanent personnel, as well as loaned /part-time personnel who perform quality affecting activities, receive Quality Assurance Indoctrination Tnining. QA Indoctrination Training includes general criteria, including applicable codes, standards, QA Procedures, QA Program elements, and job responsibilities and authorities. Detailed QA training is provided on the QA Program, applicable Project Plans and procedures prior to an employee beginning work affecting quality. Supervision is responsible for assuring personnel performing work under the.r supervision are trained.

The QA Training Coordinator (QATC) is appointed by the President, DE&S and is responsible for coordinating QA training activities for DE&S. The QATC serves as a centralized training support service for supervision in coordinating training and maintaining QA training records. This responsibility is carried out as support for line management. Appropriate DE&S supervisory personnel are responsible for determining the type and extent of the training to be provided to an individual, and ensuring that the training is properly documented for personnel performing activities affecting quality.

A.3.1.4 Management Assessments The President conducts a Corporate Management Assessment every two years to determine if the Quality Assurance Program is effective on a corporate wide basis.

The President appoints a team of DE&S managers or supervisors :o conduct this assessment. Recommendations are provided to the President for action.

Business Unit Managers and the QA Manager conduct an Internal Management Assessment annually of QA activities under their control. The managers report results of the Internal Management Assessments to the President for review.

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A.3.1.5 Oustincation/ Certification of Nondestructive Examination (NDE) Personnel DE&S maintains an NDE program described in the NDE Program hianual and may also use an approved outside agency to provide training and to administer examinations for certification by DE&S of qualified NDE personnel. In general, DE&5 uses Duke as the approved outside agency. The ASME Recommended Practice No. SNT TC-1 A, Personnel Qualification and Certification in Nondestructive Testing,1984 Edi ion t applies as requirements for NDE personnel.

A.3.1.6 OA Program Reoorting to Management Management is regularly informed on the quality of DE&S quality affecting work via results of audit reports, internal surveillance reports, Problem Investigwon Reports, and management assessments. Corrective action is initiated and monitored until completion as necessary.

A.3.1.7 Oualification/ Certification ofInspection and Test Personnel DE&S uses an approved outside agency to provide training and to administer examinations for certification by DE&S of qualified inspection and test personnel.

In general, DE&S uses Duke as the approved outside agency. Inspection and test personnel are qualified and certified in accordance with ANSI /ASME N45.2.6 -

197f, Qualifications ofInspection, Examination, and Test Personnel.

A.? 1.8 Oualification/Certi6 cation of Personnel for ASME Code Design Qualifications and certification of personnel (Registered Professional Engineers) engaged ir c rtifying ASME Section III Code designs and documents, when s

required by ASME Section III, is conducted in accordance with ANSI /ASME N626.3, Qualificetions and Duties of Specialized Professional Engineers.

A.3.2 Desien Control Procram Measures are established in QA Procedures to assure that applicable requirements are correctly translated into design documents. Design inputs are specified. Controls are established for the selection and suitability of application of mataials, parts, equipment, and processes that are essential to the functions of structures, systems, and components.

Controls are established for the identification of design interfaces among participating organizations. QA Procedures specify controls for design input, design process, design verification, design changes, and approval. These procedures include appropriate quantitative and/or qualitative acceptance critcria for determining that important activities have been satisfactorily accomplished. Design documents are prepared, verified, and approved by qualified individuals. Design verification is performed by individuals who are independent of the preparer or approver. Design is verified by one or more of the following verification methods: design reviews, alternate calcuk.tions, or qualification tests. The method of design verification and results are documented. Design changes are governed by control measures commensurate with those applied to the original design. Computer maiowas A-10

software is verified and validated in accordance with the requirements of ASME NQA 2, Part 2.7 and ASME NQA 1, Suppleraent i15 2.

A.3.2.1 Desien Process Description Applicable design inputs (such as design bases, conceptual design reports, performance requirements, regulatory requirements, codes, and standards) are controlled in accordance with the following requirements:

• Design inputs are identified and documented, and their selectior. reviewed and approved by responsible management. ,

Design inputs are specified and approved to the level of detail necessary to permit the design work to be carried out in a correct manner that provides a consistent basis for making design decisions, accomplishing design verification, and evaluating design changes.

Changes from approved design inputs and reasons for the changes are identified, approved, documented, and controlled. Design inputs based on assumptions that require reverification are identified and controlled.

The design process is controlleu. Design work is prescribed and documented on a timely basis and to the level of detail necessary to permit the design process to be carried out in a correct manner. Design documents are adequate to support design, fabrication, construction, test, inspection, examination, and operation. Appropriate standards are identified and documented, and their selection reviewed and approved. Changes from specified standards, including the reasons for the change, are identified, approved, documented, and controlled. Controls for selecting and reviewing design methods, materials, parts, equipment, and processes that are essential to the function of an item for suitability of application are established.

Applicable information derived from experience, as set forth in reports or other documentation, is maae available to cognin.nt design personnel. Design documents are sufficiently detailed as to purpose, method, assumptions, design input, references, and units such that a person technically qualified in the subject can understand the documents and verify their adequacy without recourse to the originator. Controls for identifying r.ssemblies or components that are part of the item being designed are established. When such an assembly or component is a commercial grade item, that prior to its installation, is modified or selected by special inspection and/or testing to its requirements that are more restrictive than the supplier's published product description, the component part is represented as different from the commercial grade item in a manner traceable to a documented definition of the difference. Drawings, specifications, or other design output documents shall contain appropriate inspection, examination, and testing acceptance criteria.

Design analyses are planned, controlled, and documented. Design analysis documents are legible and in a form suitaole for reproduction, filing, and retrieval.

watsoeum A-11

Calculations are identifiable by subject (including structure, system or component to which the calculation applies), originator, reviewer, and date, or by other designators such that the calculations are traceable. Computer software used to perform design analyses is developed or qualified, and used according to the ret uirements of ASME NQA 2, Part 2.7 and ASME NQA 1 Supplement 115 2.

Documentation of design analyses includes definition of the objective of the analyses, definition of design inputs and their sources, results ofliterature searches ,

or other applicable background data, identification of assumptions and designation of those that must be verified as the design proceeds, identification of any computer calculation, including computer type, computer program (e.g.. name),

revision identification, inputs, outputs, and the bases (or reference thereto) supporting application of the computer program to the specific physical problem, review and approval, identification of analysis methods utilized, identification of the design / analysis results and demonstration that they meet the applicable acceptance criteria, and the conclusion of the design / analysis.

A.3.2.2 Design Verincation Nsign verification is performed using one or a combination of the following methods: design review; alternate calculations; or qualification testing. The particular design verification method is documented. The resuhs of design verification are documented, including the identification of the verifier. Design verification is performed by competent individuals or groups other than those who performed the original design but may be from the same organization.

Design verification is performed at appropriate times during the design process.

Verification is performed before release for procurement, manufacture, or construction or release to another organization for use in other design work. In some cases (such as when insufficient data exists) it may be necessary to release unverified designs to other organizations to support schedule requirements.

Unverified portions of the design are clearly identified and controlled. In all cases, design verification is completed before relying on the item to perform its function.

The extent of the design verification required is a function of the importance to safety, complexity of design, degree of standardization, state of the art, and similarity with previously proven designs. Use of previously proven designs is controlled. The applicability of standardized or previously proven designs are verified with respect to meeting pertinent design inputs for each application.

] Known problems affecting standaid or previously proven designs and their effects on other features is considered. The original design and associated verification measures are adequately documented and referenced in the files of subsequent application of the design. Changes in previously verified designs shall require re verification. Such verifications shallinclude the evaluation of the effects of those changes on the overall previously verified design and on any design analyses upon which the design is based.

Design reviews are controlled and are focused on assuring that: (i) design inputs were correctly selected and incorporated, (ii) assumptions necessary to perform smte m A 12

the design work were adequately described, reasonable, and, where necessary, reverified (iii) appropriate design methods were used , (iv) design outputs are reasonable compared to design inputs, and (v) necessary design inputs and verification requirements for interfacing organizations were specified in the design documents or in supporting implementing documents.

Alternate calculations focus on evaluating the appropriateness of assumptions, input data, and the computer program or other calculation method used, and the results verify the correctness of the original calculations or analyses.

If design adequacy is to be verified by qualification testing, the tests are identified, controlled, and documented. The test configuration is defined and documented.

Testing shall demonstrate the adequacy of performance under conditions that simulate the most adverse design conditions. Operating modes and environmental conditions in which the item must perform satisfactorily are considered in determining the most adverse design conditions. If the tests verify only specific design features, then the other features of the design are verified by other means.

Test results are documented and evaluated to ensure that test requirements have been met. If qualification testing indicates that a modification to an item is necessary to obtain acceptable performance, then the modification is documented and the item modified and retested cr otherwise verified to ensure satisfactory performance. Scaling laws are established, verified, and documented when tests are being performed on models or mockups. The results of model test work are subject to error analysis, where applicable, before using the results in final design work.

A.3.2.3 Design Changes Design changes are controlled. Changes to final designs, field changes, and nonconforming items dispositioned "use as-is" or " repair" are justified and are subject to design control measures commensurate with those applied to the original design. Design control measures for changes shall include provisions to ensure that the design analyses for the item are still valid. Changes are approved by the same affected groups or organizations that reviewed and approved the original design documents, unless otherwise approved by DE&S management. The design process and design verification methods and implementing documents are reviewed and modified, as necessary, when a significant design change is necessary because of an incorrect design. There design deficiencies are documented and resolved in accordance with approved procedures.

Field changes are incorporated into affected de 'gn documents when such incorporation is appropriate, and when a field uange is approved other than by revision to the affected design documents. Design changes that impact related implementing documents or training programs are communicated to affected organizations.

.mmsom m, A-13

\ . .

Design interfaces are identified and controlled. Design efforts are coordinated among interfacing organizations. Interface controls shallinclude the assignment of responsibility and the establishment ofimplementing documents among interfacing design organizations for the review, approval, release, distribution, and revision of documents involving design interfaces. Design information transmitted across interfaces are documented and controlled. The status of the design information or document provided is identified in transmittals. Where necessary, incomplete designs that require further evaluation, review, or approval are identified. When it is necessary to initially transmit the design information orally or by other informal means, design information is promptly confirmed through a controlled implementing document The QA Manager shall review design documents to assure inclusion of the applicable quality requirements as specified in implementing quality assurance procedures.

A.3.2.4 Software Utilization Computer Software is verified and validated. Software verification and validation activities shall ensure that the software adequately and correctly performs all intended functions; and ensure that the software does not perform any unintended function that either by itself or in combination with other functions can degrade the entire system. Software verification and validation activities are planned and performed for each system configuration that may impact the softwar . The results of software verification and validation activities are documented. Software verification and validation are performed by individuals other than those who designed the software. Software verification is performed during the software development to ensure that the products of a given cycle phase fulfill the requirements of the previous phase or phases. Applicable software life cycle phases are as defined by ASME NQA 2 Part 2.7.

Software validation is performed at the end of the implementation phase to ensure l that the code satisfies the requirements. Software validation activities, such as the development of test plans and test cases are integrated into each phase of the software life cycle. Testing is the primary method of software validation. Software testing is conducted in accordance with ASME NQA-1 Supplement 1IS-2. The validation of modifications is subject to selective regression testing to detect errors introduced during the modification of systems or system components, to verify that l

the modifications have not caused unintended adverse effects, or to verify that a modified system (s) or system component (s) still meets specified requirements.

i A configuration baseline is defined at the completion of each major phase of the l software development. Approved changes created subsequent to a baseline are added to the baseline. A baseline shall define the most recent approved software configuration.

umm- u A-14

l A labeling system for configuration items is implemented. The labeling system covers: uniquely identifies each configuration item; identifies changes to configuration items by revision; and provides the ability to uniquely identify each configuration of the revised software available for use.

Changes to software are formally documented. This documentation shall contain a description of the change, the rationale for the change, and the identification of affected baselines. The change is formally evaluated and approved by the organization responsible for the original design, unless an alternate organization has been given the authority to approve the changes. Only authorized changes are made to software baselines. Software verification activities are performed for the change as necessary to ensure the change is appropriately reflected in software documentation, and to ensure that document traceability is maintenance. Software validation is performed as necessary for the change.

The information that is needed to manage a configuration is documented. This information shall identify the approved configuration, the status of proposed changes to the configuration, the Status of approved changes, and information to support the functions of configuration identification, and configuration control.

A plan (s) for assuring software quality assurance is in existence for each new software project at the start of the software life cycle, or for procured software when it enters the purchaser's organization. This plan (s) may be prepared individually for each software project, or may exist as a generic document to be applied to software prepared within or procured by an organization, or may be incorporated into the overall quality assurance program. The plan for software quality assurance shall identify: the software products to which it applies; the organizations responsible for performing the work and achieving software quality and their tasks and responsibilities; required documentation; standards, conventions, techniques, or methodologies which shall guide the software development, as well as methods to assure compliance to the same; the required software reviews; and the methods for error reporting and corrective action.

Software requirements documentation outline the requirements that proposed software must satisfy. These requirements address, as applicable, the following:

(i) functionality - the functions the software is to perform, (ii) performance - the time related issues of software operation such as speed, recovery time, and response time (iii) design constraints imposed on implementation phase activities -

any elements that will restrict design options, (iv) attributes - non-time related issues of software operation such as portability, acceptance criteria, access control, and maintainability, and (v) external interfaces - interactions with people, hardware, and other software. An item can be defined as a software requirement only ifits achievement can be verified and validated. Software requirements are traceable throughout the remaining stages of the software development cycle.

Software design and implementation documentation includes a document or series of documents that shall contain: a description of the major components of the mm.wms A-15

software design as they relate to the software requirements; a technical description of the software with respect to the theoretical basis, mathematical model, control f flow, data flow, controllogic, and data structure; a description of the allowable or prescri'ad ranges for inputs and outputs: and the design described in a manner that can be translated into code.

A.3.2.5 Software Veofication Software verification and validation documentation shall describe the tasks, and criteria for accomplishing the verification of the software in each phase, and the validation of the software at the end of the development cycle. The documentation shall also specify the hardware and software configurations pertinent to the software verification and validation. The documentation is crganized in a manner that allows traceability to both the software requirements and the software design.

This documentation shall also contain the results of the execution of the software >

verification and validation activities, and shall include the results of reviews and tests, and a summary of the status of the software e.g., incomplete design performance and application requirements.

User documentation, as a minimum, shall include: user instructions that contain an introduction, a description of the user's interaction with the software, and a description of any required training necessary to use the software; input and output specifications; input and output formats; a description of system limitations; a description of anticipated errors and how the user can respond; and information for obtaining user and maintenance support.

Verification reviews shallidentify the participants and their specific responsibilities during the review and in the preparation and distribution of the review documentation. The reviewed documents are updated and placed under configuration control. Documentation of review comments and their disposition is retained until they are incorporated into the updated software. Comments and their disposition not incorporated are retained in accordance with established procedures. The review of software requirements is performed at the completion of the software requirements documentation. This review shall assure that the requirements are complete, verifiable, consistent and technically feasible. The review shall also assure that the requirements will result in feasible and usabic code. The software design review is held at the completion of the software design documentation. This review shall meet the design verification requirements of ASME NQA-1,3S-1. This review shall evaluate the technical adequacy of the design approach, and assure internal completeness, consistency, clarity, and correctness of the software design, and shall verify that the software design is traceable to the requirements. Upon completion of the testing phase (and the installation phase if necessary) the development cycle documentation is reviewed and approved to assure completion and acceptability.

A formal procedure of software problem and corrective action is established for software errors, and failures. This problem reporting system shall assure that m aso m m s A-16

problems are promptly reported to affected organizations to assure formal processing of problem resolutions. Problems in software may be classified by the organization responsible for the evaluation. Any classification system shall have defined criteria based on the impact of the software output. Corrective action by the responsible organization shall assure that: problems are identified, evaluated, documented, and, if required corrected; problems are assessed for impact on past and present applications of the software by the responsible organization; corrections or changes are controlled. and preventive actions and corrective actions results are provided to affected organizations.

A.3.2.6 Software control Controls are established to permit authorized and prevent unauthorized access to a computer system. Individuals or organizations developing and supplying software under contract are required to have policies and procedures that meet the applica-ble requirements of this Part as specified in procurement documents. The documentation that is required by this Part is delivered or made available by the supplier to the purchaser. The applicable requirements of this section are the responsibility of the purchaser upon receipt of software.. Typically this software enters the purchaser's organization at the start of the installation and checkout phase. The supplier shall report software errors, or failures, to the purchaser, and the purchaser shall report software errors to the supplier.

Software that iias been developed not using ASME NQA 2, Part 2.7 is placed under the configurction controls required by ASME NQA 2, Part 2.7 prior to use.

The user organization shall perform and document an evaluation to: determine its adequacy to support software operation and maintenance, and identify the activities to be performed and documents that are needed in order for the software to be placed under configuration control. This determination is documented and shall identify as a minimum: user application requirements; test plans and test cases required to validate the software for acceptability; user documentation. After the specified activities are performed, reviewed, and approved, the software is placed under configuration control.

As an alternate, the user organization shall obtain the above documentation from the supplier or perform a documented review of the documentation at the supplier facility to determine acceptability. This review shall meet the requirements as specified above.

The organization providing' software services shall have a plan (s) for software quality assurance that meets the requirements of Section 10. The user organization shall determine the adequacy of this plan.

A 3.2.7 Software Testing Computer Program Test requirements and acceptance criteria are provided or approved by the organization responsible for the design or use of the program to sumowm A-17

be tested unless otherwise designated. Required tests including (as appropriate) verification tests, hardware integration tests, and in use tests are controlled. Test requirements and acceptance criteria are based upon applicable design or other pertinent technical documents. Verification tests shall demonstrate the capability of the computer program to produce valid results for test problems encompassing the range of permitted usage dermed by the program documentation. Acceptable test problem solutions are as follows: hand calculations; data and information from technical literature. For programs used for operational control, tosting shall demonstrate required performance over the range of operation of the controlled function or process. Depending on the complexity of the computer program being tested, testing may range from a single test of the completed computer program to a series of tests performed at various stages of computer program development to verify correct translation between stages and proper working ofindividual mod-ules, followed by an overall computer program te:;t. Regardless of the number of stages of testing performed, verification testing is sufficient to establish that test requirements are satisfied and that the computer program produces a valid result for its intended function. Test problems are developed and documented to permit confumation of acceptable performance of the computer program in the operating system. Test problems are run whenever the computer program is installed on a different computer, or when significant hardware or operating system configura-tion changes are made. Periodic in-use manual or automatic self-check routines are prescribed and performed for those applications where computer failures or drift can affect required performance.

Test procedures or plans shall specify the following, as applicable: required tests and test sequence; required ranges ofinput parameters; identification of the stages at which testing is required; criteria for establishing test cases; requirements for testing logic branches; requirements for hardware integration; anticipated output values; acceptance criteria; and reports, records, standard formatting, and conventions.

Test results are documented. Verification test results are evaluated by a responsible authority to assure that test requirements have been satisfied.

Verification test records shall identify: computer program tested; computer hardware used; test equipment and calibrations, where applicable; date of test; tester or data recorder; simulation models used, where ap' icable; test problems; results and acceptability; action taken in connection with any deviations noted; and person evaluating test results In-use test results shallidentify: computer program tested; computer hardware used; test equipment and calibrations, where applicable; date of test; tester or data recorder; and acceptability smuomms A-18

A.3.3 Project Plannine Process Project planning is performed to specify the methods, controls, and procedures by which a scope of work, and the client specified technical and quality assurance requirements, are to De accomplished. Each client project performed under the DE&S QA Program requires preparation of a Project Plan. The Project Plan defines the applicable QA procedures and requirements, including the identification of specific QA Program manuals, client procedures, or other client specified project requirements to the project. Specific time limits are placed on the preparation and approval of a Project Plan with respect to authorization of work performance. The level of detail of the project plan is commensurate with the scope and complexity of the project. The Project Plan specifies, as a minimum:

• QA Condition (e.g., QA Condition 1. QA Condition 2, QA Condition 3 or QA l Condition 4);

• 10CFR21 Applicability (Do 10CFR21 requirements apply to the project?);

Project Description;

• QA Procedure identification (Identify specific QA procedures that apply to the project.);

• Interface Identification (identify internal / external project interfaces and controls utilized i to ensure that the proper project interface is achieved.);

• Software listing (Identify software and version to be used on the project);

• Deliverable listing (List deliverables, such as specifications and calculations, to be prepared as part of the project.);

• Design Input Definition (Identify source documents or list applicable design characteristics and/or functions to be used in developing design inputs, including applicable codes, standards and regulatory requirements.);

• Experience Information (Identify previous experience which is pertinent to the project, including lessons learned from similar projects);

• Scope of work (Provide detailed description of work to be performed);

• Special requirements (Identify special personnel certification requirements, client specified Codes / Standards, restrictions for Suppliers, special shipping / handling / storage requirements, etc.);

Documented training is required for all project persont.el on the requirements and procedures specified within the Project Plan. Preparation, control, revision, and close-out of Project Plans are governed by DE&S procedure. DE&S Project Plans are defined as QA records.

.mmowncs A-19

FIGURE A-1 Duke Energy Corporation Organization Rick Priory. Civirman & ChiefErecutive Oficer Paul Anderson,Prrsident & Chief 0perating Officer I i l l Jini IInckett Ilill Cooley Fred Fowler Dick Ransan Grrney President Grrnip President Gnnsp President Sr. Vice President EnergyServices Duke P<nver Energy Trtinsmission DusirstfledOpertriions Ruth Shaw Rich Osborne Richarif Illackburn Esecutne VicePresident Executire Vce Persident Executive Iice President Cinefitdarin. Officer ChiefFrrurtialOficer GenettiCounsci

/ N / T Duke Enginccting - Duke Energy

& Services Field Services

\ \ A  ;

/  %

N* I ^"*'

Duke Energy --

Trade & htarketing Dewtopimnt

\ A .

/

Duke Solutions / \

_ Duke Energy

\ Power Scrsices

/  %

Duke / Fluor Daniel T

-l/

Duke Energy

\ M International

, \ A Duke Energy / \

Tranvort & Trading _ Duke Energy Industrial

\ Auct Development

\ 4

.- - --.,ms A-20

i FIGURE A-2

~

Duke Engineering & Services, Inc. Organization J. F. Nonis. Jr.

President aml CEO W. J. Bowman -

Secretary & General Counsel J. M. Ilart W. O.'llenty S. IL Patrwa R. W. Benssil D.t Rehn >

Senior VP Ewecutiw VP Executive VP Executive VP Esecisive VP Corporate Group Federal G.oup Intemational & Petroleum Groir Energy & Envmmr . ntal Group Nuclear Group ,

1

- Strategic Programs - GeoEngineering Services - IHroleum Services - Pbwer Delivery - Mkfwestern Region i

- Treasurer and CFO - DE&S11anford -

Asia. Africa & Australia - Renewable Energy - Northeassern Regwn

- Corp Communications - DE&S Northwest - Central & South America - Generatbn Serv,ces - Fire Protection

- Qunlity Assurance - Las Vegas 05cc - Europe -

flystro & Water Sernces - Advances! Nuclear J

- Information Technoicgy Aiken Ollice - Enviewal - Westem Regbn

& Siting Services Administration - I AfiTCO -INEEL Field Sernees - Southeassem Region

& Pipelme initiatives

- Ifuman Resoun:es - LANL - Power Delivery Sernces - Technica: S.+ s

- VP- Stamng and - Dectwnmisming StafT Utiliration i

- Nuclear. Evel and Quahty Atsurance h t

m n wra m ms A-21 4

l l

APPENDIX B DESCRIPTION OF YAEC RESPONSE TO SAFETY ALLEGATIONS t

a 4

4

APPENDIX B DESCRIPTION OF YAEC RERPONSE TO SAFETY ALI EGATIONS i

TABLE OF CONTENTS EGEC B.1 TECHNICAL AS S ES S MENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 1 B.1.1 January 1996 YAEC Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B1 B.l.2 Siemens Analysis of Maine Yankee SBLOCA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3 B.I.3 NRC Independent Safety Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-4

! B.I.3.1 Assessment Description and Summary of Conclusions . . . . . . . . . . . . . . . . . . B 4 B. I .3.2 Approach to Assessment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B6 B. I.3.3 Technical Quality Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 6 B.I.3.4 Compliance with Safety Evaluation Reports Conditions . . . . . . . . . . . . . . . . B 6 B. I .3.5 Process Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-7 B.I.3.6 Analytic Code Validation Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 7 B.1.3.7 YAEC IS A Response Actions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-7 B.2 ROOT CAUS E AS S ES SMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 7 a B.2.1 Asse s sment Proce ss . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 8 B.2.2 Identified Cau se s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 8 B.2.3 Assessment Recommendations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 9 B.3 INITI AL CORRECTIVE ACTIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 10 4

B.4 PROCES S IMPR OVEMENN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 12 B.4.1 Corrective Action houss . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 12 B.4.1.1 History of YAEC Deficiency Reporting Systems . . . . . . . . . . . . . . . . . . . . . B 12 B.4.1.2 Condition Report System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12 B.4.1.3 Deficiency Trends . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 14

B.4.2 Employee Concems Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 15

! B.4.3 Engineering Instructions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 15 i- B.4.4 Joint Quality Audit Group . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 17 B.4.5 Functional Area Representatives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 17 B.4.6 Technical Specialist Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 18 B.4.7 Functional Area Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 18 i B.4.8 S elf Asse ssments . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . B 18 B.5 AUDITS AND ASSESSMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 18 B.S.1 Internal Audits / assessments of Engineering-Related Activities . . . . . . . . . . . . . . . . . . B-19 B.5.1,1 Ownership of Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B- 19 B.5.1.2 Adherence to Procedural / Code Requirements . . . . . . . . . . . . . . . . . . . . . . . B-19 B.5.1.3 Analysis Inconsistencies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-20 B.5.1.4 Control of Documentation / Records . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-20 B.5.1.5 Software Q A Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-20 smem-s B ii

I APPENDIX B 1

DESCRIPTION OF YAEC RERPONSE TO SAFETY AI LEG ATIONS J

s

TABLE OF CONTENTS l (Continued) g B.S.2 Joint Utility Management Audits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 21

B.S.2.1 1995 JUMA Audit of Vermont Yankee . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 21 4

B.S.2.2 1996 JUM A Audit of Maine Yankee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 22 B.S.2.3 1997 JUMA Audit of Vermont Yankee . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 23 i

i i

i e

l i

I d

t I

I v

spennsovaseia B iii l....-.. . - - - ., . . - .

APPENDIX B DESCRIPTION OF YAEC RESPONSE TO SAFETY ALLEG ATIONS Upon notification of the safety allegations transmitted by UCS in December 1995, YAEC promptly initiated a technical review of the allegations. Further YAEC review and interaction with the NRC, including the NRC Independent Safety Assessment for hiaine Yankee and the RELAPSYA (BWR)

LOCA applications audit performed in the summer of 1996, led to a number of technical / programmatic assessments and corrective actions. This appendix summarizes the more ugnificant of these assessments and actions taken by YAEC prior to the December 1997 DE&S acquisition. Details of actions summarized in this appendix are generally found in hiaine Yankee Atomic Power Station NRC licensing docket, Docket 50 309.

This appendix is nol to be construed as a response by YAEC to the Demand. This appendix only represents a DE&S summarization of actions taken by YAEC. This summary of YAEC actions is provided to serve as an aid to the reader in understanding the actions taken by and conclusions of DE&S in responding to the Demand.

B.1 TECHNICAL ASSESShfENT B.I.1 January 1996 YAEC Assessment The hiaine Yankee safety allegations transmitted in December 1995 immediately prompted hiYAPCo and YAEC executive management to form two corporate teams to evaluate the auegations: (1) a Response Te~ 3mposed of managers and technical specialists with responsibilities for the analyst oemg questioned and (2) an Independent Review Team compose of managers and technical specialists with no prior responsibilities for these analyses. The Response Team mission was designed to provide support for interaction with the NRC in reviews, as well as to evaluate the allegations, identify and resolve technical issues, and to ascertain whether any issues raised by the allegations or in subsequent revius adversely affected the conclusions of the safety analyses used to support past or future operation of the biaine Yankee plant. The Independent Review Team was also charged with assessing the impact of the allegations on the current licensing basis and the safety significance of the issues.

The Response Team supported the NRC inspection performed in response to the allegations during the week of December 11-14,1995 and provided additionalinformation to the NRC technicti staff. This team, along with executive management and other representatives from hiYAPCo and YAEC, met with NRR management and technical staff on December 18, 1995. The proceedings of these interactions with the NRC Staff demonstrated clearly that a lengthy breakdown in effective communications between YAEC and the Staff had occurred.

The review status of the Maine Yankee SBLOCA application did not meet NRC expectations. In retrospect, Yankee should have increased its direct communications with '

the NRC following the approval of RELAPSYA(PWR) for Maine Yankee application in 1989.

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Within the next four weeks the YAEC technical assessmerit of the suitability of using RELAP5YA(PWR) for demonstrating conformance with the criteria set forth in 10CFR50.46 was completed and provided to the NRC in Reference B.I. In that submittal YAEC concluded that the results obtained using RELAP5YA(PWR) were conservative and that SER conditions had been met.

in the January 1996 submittal, YAEC stated that the Maine Yankee ECCS performance during postulated SBLOCA conditions was acceptable and met all NRC regulations. YAEC

, also stated that during development of computer codes to analyze these conditions, some results were generated that,if valid, would not meet NRC requirements. liowever, these results were indicative of problems with the modeling used in development of the computer code and the Maine Yankee plant model. When a more appropriate plant model was developed, YAEC believed that the resuhs demonstrated that the Maine Yankee ECCS performance was adequate and all NRC requirements were met. Thus YAEC concluded that the operating limits for the plant were properly set on the basis of the limiting LBLOCA analysis using NRC approved methods. LBLOCA was consistently the historical LOCA licensing basis for Maine Yankee. Subsequent analyses that were performed by Siemens (Section B.I.2) reconfirmed this conclusion that the safety limits are conservative when established on the basis of the LBLOCA analysis.

The computer code RELAPSYA(PWR)*2 was approved by the NRC for use as a licensing method in evaluating the performance of the Maine Yankee ECCS under SBLOCA conditions.83 RELAPSYA(PWR) is a modified version of RELAPS/ MODI,8 d a commonly used computer code that was developed by the Idaho National Engineering Laboratory over the course of several years through the early 1980's. The RELAP5YA(PWR) code was submitted for NRC review in 1983 83 NRC employed the technical assistance of the Los Alamos National Laboratoy 8' and the Idaho National Engineering Laboratory8 ' in the review of the topical report. NRC requested responses to 197 questions on general application of the code in May 1984 and asked several additional questions on the SBLOCA application in September 1986. A series,of eight submittals responding to these questions was transmitted to the NRC between March 1984 and December 1986. A final submittal on SBLOCA application model improvements with appropriate comparisons to experimental test results was made in 1988. These submittals were supported by numerous meetings and discussions with the staff and their consultants and culminated in approval of the code for application to Maine Yankee in January 1989. In addition, the NRC staff and consultants had completed in 1986 the review and approval of RELAP5YA(PWR) analysis which was performed to provide guidance on Reactor Coolant Pump operation during SBLOCA transients for Maine Yankee.a te o The NRC approval of RELAP5YA(PWR) in 1989 included twelve specific conditions to be met in the use of this code for Maine Yankee. Several of these conditions specified that the code was required to be used in a certain manner. For instance, one of the conditions requires the code to be used assuming an emergency core cooling system water temperature of 200 'F. These conditions were met in the Maine Yankee application. ems n p;ye conditions relate to information to be provided in plant specific licensing applications that the NRC would review to examine how the code was used. The Response Team found that these submittals had not been made based on the understanding of a later NRC letter which

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indicated that this information should be retained for inspection." The Response Team verified the information was available for inspection by the NRC.

During the inspection the week of December 11,1995, the NRC raised a number cf questions about the application of RELAP5YA(PWR) to the Maine Yankee SBLOCA analysis which could not be resolved satisfactorily during the four day review. These questions related to the fluctuations (oscillations)in peak clad temperature (PCT) calculated by the code, the time step sizes used in the analysis, the flow loss coefficient (K factor) developed and used in the analysis, and the range of break sizes analyzed. References B.1 and B.13 provided information addressing these questions with the intent of demonstrating that the RELAP5YA(PWR) application to Maine Yankee provided a conservative calculation for the PCT values for SBLOCA analysis.

Reference B.1 also summarized additional analyses that had been performed using an advanced version of RELAPS-MOD 3 that further supported the conclusions that had been reached in the RELAPSYA(PWR) analysis of Maine Yankee. However, efforts to justify the suitability of RELAPSYA use for Maine Yankee SBLOCA analyses were very limited subsequent to the submittal of Reference B.I. MYAPCo decided to retain Siemens to conduct an independent analysis of SBLOCA for Maine Yankee.

B.I.2 Siemens Analysis of Maine Yankee SBLOCA In Reference B.14 MYAPCo submitted analyses for SBLOCA analyses conducted by Siemens. The Siemens SBLOCA analyses were performed to support a nominal power of 2700 MWt. Features included a conservative local power assumption and variations in axial power profdes. RCP trip sensitivity studies were also performed. The analyses were performed to demonstrate that ECCS acceptance criteria, as stated in 10CFR50.46, were met.

Break spectrum calculations were performed to identify the limiting break size. Break sizes of 0.05,0.10,0.15,0.20,0.25, and 0.61 ft in2 the pump discharge side of a cold leg pipe were analyzed. An axial prorde peaked at a relative core height of 737c was used in the 2

break spectrum analysis. No core heatup was calculated for the 0.05 ft break size. The 0.10 ft 2break size resulted in the highest Peak Cladding Temperature (PCT) for the break spectrum cases.

Additional axial prorde calculations were performed at the limiting break size for axial profiles peaked at relative core heights of 52%,65%, and 85%. The 65% axial prorde case gave the highest PCT for the axial prorde sensitivities. Since the PCTs for the axial profdes peaked at relative core heights of 65%,73%, and 85% were relatively close, core cross flow loss coefficient sensitivity calculations, as required by the Siemens 3BLOCA methodology, were performed for these three axial prorde cases. The limiting PCT was for the axial profde peaked at a relative core height of 73% with the minimum core cross flow loss coefficient.

The PCT was 1781 *F.

All 10CFR50.46 criteria were demonstrated to be met for the limiting case:

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1. The limiting PCT was calculated to be 1781 *F, and was for IFBA fuel. The margin to l

the 10CFR50.46 PCT limit was shown to be in excess of 400*F.

1 l

2. The local cladding oxidation was calculation to be 1.5 percent.

3. The core wide metal water reaction was calculated to be less than 1 percent.

4. The core remains amenable to cooling by staying within the local oxidation criteria.

5. Flow rates from "..e Emergency Core Cooling System (ECCS) ensure that the core temperatures have been reduced to acceptably low values and long term cooling has been established.

The Reactor Coolant Pump (RCP) trip delay time evaluation consisted of both Evaluation Model(EM) analyses and best estimate analyses. The EM evaluation bounded RCP trip delay times of up to two minutes and demonstrated that 10CFR50.46 criteria are satisfied.

The limiting PCT for this analysis was calculated to be 1781 *F, and was for IFB A fuel. Best estimate calculations were performed to support RCP trip delay times from two minutes to 10 minutes. Break spectmm calculations were performed for break sizes of 0.10,0.15,0.25, 0.45, and 0.61 ft 2. Several RCP trip delay times were chosen for each break size calculation to atsure that a bounding case was analyzed for each break size. The limiting best estimate case was determined to be the 0.15 ft2 break,10-minute RCP trip delay time case, with a PCT of 1136*F.

The NRC had not completed its review of the Siemens analysis when MYAPCo decided to end plant operation. YAEC was unaware of any outstanding issues, however, that would have changed the overall conclusion reached by Siemens in this analysis.

The results from these Siemens analyses were similar to those determined with RELAP5YA with regard to important analysis trends and conclusions. The PCT calculated by Siemens was lower than determined with RELAPSYA, and thus supported YAEC conclusions that the RELAPSYA (PWR) results were conservative and well bounded by LBLOCA analyses.

B.1.3 NRC Independent Safety Assessment B.I.3.1 Assessment Descrintion and Summarv of Conclusions The NRC Independent Safety (ISA) review for Maine Yankee was conducted in response to findings developed by the NRC Office of the Inspector General (OlG) in a report dated May 8,1996. The ISA review team and approach was developed at the direction of the Chairman of the NRC to provide a thorough and substantial examination of all safety-related activities which supported Maine Yankee operation. One group from the 25 member ISA team was assigned for the duration of the effort to perform an assessment of the analytic code support provided for Maine Yankee YAEC. The approach and results of this review of YAEC safety-related analysis activities are described in this section.

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The IS A team was led by a senior NRC manager, and included representatives of the State of Maine. To ensure an independent perspective, the NRC members were selected from NRC offices other than the Orlice of Nuclear Reactor Regulation (NRR) and NRC Region I offices. The team management reported the results of this assessment directly to the Chairman of the NRC.

The ISA team conducted their reviews from June through October in 1996. The portion of the team that reviewed YAEC activities in the Bolton office included four full time NRC staff and expert industry consultants who performed on site review activities for four weeks over a calendar period of six weeks. In addition the team performed documentation reviews, benchmarking analysis, and assessment activities in NRC White Flint offices both prior and subsequent to these face to face inten:tionr. The scope of the examination included essentially all safety-related analyses, methods, computer codes, and procedures associated with

. work performed to support Maine Yankee activities, including that which was performed by the YAEC LOCA Group. Due to she on going NRC Office of Investigation (01) review, the only work that was exempt from the review was that which had been used specifically to develop the LOCA analysis limits for Maine Yankee.

The final report of the full NRC ISA team was transmitted to Maine Yankee Atomic Power Company on October 7,1996. The comments that follow were derived directly from an integration of all comments identified in the report that relate to the review of the Nuclear Engineering Department safety-related activities. As was stated in the ISA final report, this particular area of assessment (that is, review of the use and application of analytic codes) had not been typically reviewed as part of the NRC regulatory process. Consequently, a panel of acknowledged experts in the area of code development and phenomenology was assembled by the NRC to provide a critical review the findings and observations of the ISA team in this area. Findings regarding this aspect of the IS A team review as described in the IS A report were summarized as follows.

The use of analytic codes for safety analyses was found to be very good. Cycle specific core performanec analyses were found to be excellent. More complicated, less frequently performed safety analyses catained weaknesses, but the analyses were found to be acceptable based on compensating margin. YAEC did not have a written process to document how safety analyses conformed to SER conditions.

Some conditions were clearly known, considered, and used by YAEC. Other conditions could not be shown to be satisfied until additional analyses, i

assessments, and sensitivity studies were accomplished in response to ISA team l

requests. This new work demonstrated that all SER conditions had been satisfied, although the disposition of some issues required reliance on the known conservatism in specific accident analyses.

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B.1J.2 Anproach to Assessment The team evaluated the analytic code support provided by YAEC for MYAPCo to assure that Maine Yankee was operated within the bounds of the safety analyses.

l This assessment was performed by reviewing the YAEC process for conducting non LOCA safety analyses described in Chapter 14 of the Maine Yankee FSAR, and by performing an in depth review of two specific safety analyses: the Control Element Assembly (CEA) drop transient and the steam line rupture accident.

Selection of the dropped CEA transient for in-depth review provided a structured means to examine many of the computer codes used by YAEC. Selection of the steam line rupture analysis provided a forum for reviewing a dynamic accident analyzed with a complex systems code.

The overall review included:

(1) identification of the design basis analyses for postulated accidents and anticipated operational occurrences, (2) identification of codes, methods, and limitations, based on the team's review of topical reports and NRC Safety Evaluation Reports (SERs), and (3) an assessment of how limitations, restrictions, and boundary conditions are reflected in the safety analyses.

Certral to the assessment was the verification that conditions of approval contained in NRC SERs had been satisfied in the safety analyses. The IS A team also specifically examined code validation using guidance contained in Generic Letter 83-11, " Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions," February 8,1983.

B.1.3.3 Technical Ouallty Conclusions Cycle-specif'c core performance analyses, such as the CEA drop transient, which md many of the YAEC computer codes and methods, were found to be excellent.

An IS A team review of predicted and measured fuel bundle power distributions showed excellent agreement over several fuel cycles. Fuel performance calculations considered the multiple fuel types and multiple projected burmp histories, and were found to be excellcat overall. More complicated, le frequently performed systems safety analyses were found to contain weaknesses, such as those associated with the main steam line rupture (MSLR) ace: dent, but the analyses were found to be acceptable based on compensating margin. Overall, the use of analytic codes for safety analyses was evaluated as very good.

B.I.3.4 Compliance with Safety Evaluation Reports Conditions Compliance with conditions imposed on the use of analytic codes was verified for each of the 67 SER conditions affecting 13 codes. Although full compliance was

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confirmed, an audit trail with process documentation to assure compliance was not always available, thereby requiring in some cases additional analyses to verify compliance.

B,1.3.5 Process Conclusions YAEC did not have a written process to document how safety analyses conformed to code SER conditions. Also, YAEC did not have a documented process in place to identify and rank key phenomena for each of the transients and accidents in the safety analyses report and, in turn, to identify needed code validation and parametric study efforts. During the ISA, YAEC initiated preparation of a

Methods Overview Manual" which was designed to address these process issues.

B.I.3.6 Analvtle Code Validation Conclusions Some codes, such as the physics, fuels, and DNB codes, were found to have extensive validation to actual plant measurements and experimental data respectively. In contrast, the ISA team found that there was an over reliance on industry RETRAN code vatidation efforts, and that validation of RETRAN for the MSLR accident was weak. However, the ISA Team determined that regulatory requirements for this validation are not clear. Also, YAEC had submitted and received USNRC approval for using these codes. These submittals included verification of code applicability using validation to previously accepted methods.

No validation issaes affecting safety were identified.

B.I.3.7 YAEC ISA Response Actions The IS A review validated the overall technical quality of safety analysis work conducted by YAEC. Severalimprovement opportunities related to processes and documentation were identified. YAEC addressed these issues through improvements to processes used to validate, control, and document work products. Initiatives included improvements in procedural guidance and requirements, in program oversight (including use of technical specialists), and documentation development and review. T hese efforts are discussed in the following sections of this Appendix.

B.2 ROOT CAUSE ASSESSMENI In early 1996, MYAPCo and YAEC executive management also chartered an assessment team consisting of Maine Yankee and YAEC personnel to determine the underlying cause(s) of the allegations. This Assessment Team interviewed over 60 executive, management, supervisory and staff personnel from Maine Yankee and YEAC between February 1996 and April 1996. In April 1996, a report entitled "RELAPSYA Self Assessment" was issued by YAEC, identifying three prunaq causes for the allegation and nine areas requiring corrective action.

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B.2.1 Assessment Process The assessment team investigated the following programs, processes, and activities:

Commitment tracking Communications with the NRC Communications between MYAPCo and YAEC

• MYAPCo oversight of YAEC Definition of responsibilities between MYAPCo and YAEC

• 10CFR50.46 reporting procest.

Final Safety Analysis Report updating YAEC personnel knowledge of NRC regulatory fundamentals 10CFR50.59 evaluation process Employee Concerns process

• Document control Engineering procedures

• Personnel training Technical analyses The assessment team conducted extensive interviews of MYAPCo and YAEC personnel Interview results were summarized under each category, and evaluated by the assessment team to identify commonalities. The identified commonalities were then evaluated to determine recommendations for improvement or enhanceraent of apparent weaknesses.

B.2.2 Identified Causes The assessment team's evaluation identified the following underlying causes:

1. The division of responsibilities / ownership and roles of organizations performing, controlling, administering, and managing activities for the Maine Yankee plant are not always completely and clearly defined or understood by all parties involved in or impacted by the activity.

2. Personnel at YAEC require improved classroom. as opposed to training to the read and sign training approach.

3. Procedures used for controlling the development of analyses should be improved to define user actions rather than having processes driven by personnel knowledge. The current procedures do not require identification of the effects of analyses on licensing commitments or design basis documents.

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B.2.3 Aueument Recommendations In addition to the common causes noted above, the assessment team determined that near term management attention was necessary for the following areas:

1. Many of the interviewees in support organizations such as Nuclear Engineering, Fire  ;

Protection, Licensing, and Technical Support expressed that responsibilities for activities including safety enalysis report updates, commitment tracking, interpretation of NRC regulations, interface with the NRC, and distribution of MYAPCo documents are not clearly tmderstood. MYAPCo and YAEC management should define the responsibilities of each organization and clarify their relationship so that programs and processes can function more effic'ently.

2. YAEC should improve the controlling procedure for preparation and issuance of analyses to ensure thatt (1) regulatory limitations and thresholds are considered upon completion of calculations, (2) licensing design basis documents affected by analyses are changed when needed, (3) use of computer codes are within the NRC's Safety Evaluation Report (SER) bounds for which they were approved, (4) measures are in place for the use, control and dissemination of preliminary data and (5) procedures appropriately address QA program requirements.

3. The YAEC Employee Concerns Program should be evaluated to ensure that personnel are fully aware of the various mechanisms available to resolve an issue before it becomes an empicyee concein.

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4. Training programs need to be improved for technical personnel. Areas ofimportance include the relationship and hierarchy of pertinent codes, standards and regulnory requirements to licensing and design basis documentst and procedural training.

5. A commitment tracking system that interfaces with both Maine Yankee and YAEC needs to be developed to ensure that both formal and informal commitments are tracked and accounted for in both organizations. -

6. The licensing functions at MYAPCo and YAEC need to be strengthened using appropriate, sufficiently trained nersonnel who are knowledgeable in all aspects of licensing commitments.

7. YAEC should be included in communications between MYAPCo and the NRC when YAEC supported work is involved.

8. YAEC needs to strengthen the content ofits deliverables to i::clude key assumptions, values, and scope to support appropriate use by the receiver and user of the deliverable, m.moamn B9

9. Documentation ofimportant communications among the NRC, MYAPCo and YAEC (i.e., mutual understandings or agreements) should be strengthened to assure that commitments and agreements are captured and understood by all parties.

B.3 INITIAL CORRECTIVE ACTIOM Subsequent to the issuance of the assessment team's report, YAEC developed and implemented a plan to address each of the specific recommendations. These actions were:

1. Relationshin between Maine Ynnkee and YAEC was clarified.

Engineering for Maine Yankee was reorganized such that the engineering function was controlled directly by a Maine Yankee Engineering Manager and the communication for YAEC support was identified to be through a YAEC engineering coordinator. In the past engineering functions were controlled by a YAEC Project and Engineering Manager. The licensing function was totally at Maine Yankee with support from YAEC individuals who reported to Maine Yankee management.

2. Five (5) NED procedures were develoned in the following areas:

Analysis / calculation review checklist Safety and relief valve monitoring 10CFR50.46 reporting

• Safety analysis

• Changes to licensed methods These pocedures prosided additional guidance for NED personnel who were involved in the development and review of engineering analysis. The procedures described attributes that engineers used to assure that analyses met the administrative and technical requirements of the higher level procedures controlling design activities. They also clarified the steps for analyzing results to determine when results of analysis could potentially be reportable, especially under the requirements of 10CFR50.46.

3. Employee Concerns Program was revised.

YAEC formed a task force to review the method whereby reporting employee concerns were identified. As a result of this task force, the procedure was modified to make personnel aware of the various methods for bringing a concem to the attention of management without recourse to the individual. YAEC also established a twenty-four hour hotline that permits anonymous reporting of concems. The hotline also permits individuals to obtain the status of actions taken to resolve a concern

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4. Formal training on codes. standards. and regulatory reauirements was cravided to key nersonnel.

Individuals received training from professional organizations on the codes and standards applicable to the work in which they are involved. YAEC also committed to formal classroom training for all engineering personnel on the requirements of the Engineering Instruction Manual. The goal was to lower the incidence of" failure to follow procedures" and to increa e the quality of engineering output documents.

5. Commitment tracking systems were upgraded.

YAEC engineering functional area, developed tracking systems that allowed management to assign individuals to particular tasks and also to prioritize each task according to their importance and need.

6. YAEC suoported the increase and reorganization of the Maine Yankee licensing staff.

In 1996 the licensing ftmetion at Maine Yankee was reorganized and expanded so that it could better perform the licensing activities with the NRC. Maine Yankee licensing staff was increased to handle certain support work that was formerly done at YAEC. Although YAEC continued to provide some licensing support to Maine Yankee, it is clearly I reinforced that the fulllicensing function for Maine Yankee resided with MYAPCo. l l

Also identified were issues requiring further action by both MYAPCo and YAEC. A YAEC QA surveillance was conducted to verify the actions taken in response to the assessment team's report. The surveillance concluded that actions had been undertaken to resolve the identified issues and that a tracking system had been developed to identify actions, assign responsible individuals, and monitor the status of action completion.

In June 1997, YAEC QA cor.! acted a performance based audit of NED safety related engineering activities with the assistance of Scientech as independent technical specialists.

The focus of this audit was to perform a full scope review of the fuel reload analyses, as well as administrative issues including self assessment and corrective action. The Scientech team l l

report concluded that:

........the technical basis ..... appeared sound. Only minor deficiencies were identified in the calculational notebooks reviewed."

The audit also included an examination of sixteen YAEC self assessments and one functional area assessment conducted by NED. The audit concluded that NED was:

...... effective in identifying problems and areas for improvement and recommending corrective actions."

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- B.4 PROCESS IMPROVEMENTS YAEC recognized that improvements needed to be made in the me: hods for controlling work activities, training personnel, enhancing the design control procedures and increasing personnel awareness for reporting and resolving issues. During the latter part of 1996, a plan had been established for these unprovements. A brief summary of the key changes made at YAEC is provided in this section:

B.4.1 Corrective Action Process B.4.1.1 Historv of YAEC Deficiency Reportine Systems The YAEC Quality Assurance Program Manualincludes a commitment to ANSI N45.2.11 and NRC Regulatory Guide 1.64, both of which describe the Quality Assurance requirements for design control applied to nuclear power plants. ANSI N45.2.11 requires that procedures be established for reporting deficiencies [in design documents] and co:rective action to appropriate levels of supervision and management. Furthermore, the standard requires that a cause be determined for significant or recurring deficiencies or errors and that changes be made in the design process to prevent them for occurring again.

In order to comply with the standard's requirement, YAEC established an engineering instruction, " Engineering Deficiency Report (EDR)," which described the methods for reporting design deficiencies within all YAEC projects and engineering functions. The procedure was basic. It included a section to describe the deficiency, a section to describe the corrective and preventive action, and a section to indicate completion and verification of the corrective and preventive action. The procedure was in effect from 1989 to the end of 1997.

A second deficiency reporting system was the Status and Summary of Corrective Action (SSCA) report. This report was instituted in the 1970s to report findings identified during audits. This was in response to the requirements of ANSI N45.2.12.

B.4.1.2 Condition Report System Each of the systems described above provided different methods for reporting deficiencies. However, each system had common probims in regard to in the level of management attention afforded to monitor performance on a company wide basis and to prevent recurrence of problems. In addition, neither system had been designed to collectively analyze performance data on a periodic basis for determining adverse trends within the company.

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On May 31,1997 YAEC issued Technical Administrative Guideline No.25, c Mition Repod (CR) System as a replacement for the previous deficiency e wrting systems. The reason for issuance of the single reporting system was to assure a consistent method for rcporting, evaluating and dispositioning deficiencies. Tne system was a!so designed to provide a much lower threshold for reporting deficiencies. Significant improvements resulted from the CF process and are summarized as follows:

1. The system allows personnel to not only identify actual deficient problems in violation of a requirement but also potential problems. This created a system that could be an integral part of the self-assessm.nt process.

2. The system requires that all conditions reported under a CR be evaluated by th: CR Review Committee.

1

3. All CRs are evaluated by the individual and the individual's supervisor prior to corrective action evaluation to determine if the situatie could be potentially reportable to any local, state, federal or other regulatory agency. In addition, each of the CRs are evaluated te determine if they directly impact one or more plants and whether the adverse condition could potentially impact plant operations. If so, the CR system has provisions for reporting potential operability concerns directly and immediately to the plant.

4. The system established a Condition Report Review Committee (CRRC) which has representatives from the Executive Office, Quality Assurance, Environmental, Licensing, and Nuclear Engineering functional areas. The CRRC is responsible for the review of all CRs. The CRRC review requires evaluation of any specified corrective and preventive actions, assignment of corrective and preventive actions and action parties, determination of the appropriateness of reportability and operability, assignment of the significance of the CR, and evaluation of the need for a root cause analysis and follow up verification of corrective actions.

5. The CR system does not allow extensions for providing responses or for completing corrective actions. The system tracks the age of both responses (plans) and actions, and thus heightens the monitoring and awareness of overdue items by executive level and line management.

6. The CR system includes a uniform method of coding problems so that they can be trended periodically and analyzed for adverse process conditions which may not be apparent from examination ofisolated events.

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7. The Cu are controlled in a single tracking system and the status of open items is available to all personnel at all times through an integral, company-wide electronic network system.

YAEC personnel were formally trained on the use of the CR system and were given the opportunity to enhance the procedure prior to its formal issuance. In summary, the CR system provides a single problem reporting system with a lowered threshold for reporting problems and it receives the attention of executive and line management for evec, situation reported.

B.4.1.3 Deficiency Trends EDRs, SSCAs and CRs generated from 1995 to 1997 were evaluated to identify trends in deficiencies and to examine the effects of the recent charges to the deficiency report system.

In September 1997, the first trend report from the newly established Condition Report System was issued.

l Within these categories, it was difficult to recognize any type of trM i.r te t this point, since the system had been in place for four months. How ever, h was apparent that the threshold for initiating a deficiency report had been lowered significantly. Tite total number of various YAEC deficiency reports issued during the 1995 to 1997 time frame is shown below.

Year EDRs SSCAs ERs'" .CRs Total 1995 21 27 - -

98 1996 45 66 18 -

129 1997 37 320) -

199 268

") Event Reports were issued to Vermont Yenkee in lieu of SSCAs.

C) Includes 29 recommendations which were not issued as an SSCA.

Thirty-two percent (32%) more deficiencies were reported in 1996 than in 1995 and in 1997 there were one hundred-eight (108%) percent more deficiencies reported than 1996. This development was directly attributed to a lower threshold, more self-assessments being performed, and an increase in a questioning attitude by employees.

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B.4.2 Emplo ree Concerns Pronrnm The YAEC Employee Concerns Program was revised in 1997. The Employee Concerns Program now includes:

An employee hotline that is available twenty-four hours per day, seven days a week.

The hotline does not require identification of the individual and yet is capable of being queried by a concern's originator to determine the resolution sMtus of a concern.

Assurance of confidentiality for individuals reporting a concern.

Assurance that responses are provided to all concerns.

Provisions for ensuring personnel are aware of the various methods for reporting concerns, both internally and externally.

0.4.3 Ennineerine Instnictions The Engineering Instructions (WEs) were audited a number of times during the 1995 to 1997 period. The need for improvement was stressed during interviews conducted during the RELAPSYA Self Assessment. Although procedures met the requirements of ANSI N45.2.11, NQA-1, and Regulatory Guide 1.64, several major changes were made to the procedures to ensure that:

Calculations and analyses included a detailed review of uncertainties, Resuhs of calculatio md analyses are reviewed against allowables to determine if the results need to oe reported to any regulatory agencies, Preliminary reschs are so designated when transmitted and that they are followed up with approved calculations and compared to preliminary results, Records are established to provide configuration control so that potential revisions can be evaluated against previous results, Requirements are established to verify that contractors receive training on applicable YAEC procedures prior to starting work, Commercial software is verified on a case by case basis, c

Objectives of the analysis are clearly defined and determined to be met,

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• Analyses clearly specify the end uv: and restrictions on the use of analysis results, and

• Independent reviewer comments include a detailed description of the review and a detailed listing of all comments generated during the review of the analysis for documented dispositioning with the preparer.

YAEC Design Control Measures were evaluated and assessed. The purpose of this review was to compare and contrast YAEC's Procedures with respect to ANSI N45.2.11 (1974) (endorsed by Regulatory Guide 1.64, Revision 2) and ASME NQA 1-1989 through the NQA-lb-1991 Addenda).

Specifically, YAEC's Procedures were reviewed to establish that adequate and sufficient contrch exist and are documented for the following Design Control Criteria:

Program Requirements Design Input Requirements Design Process (Design Analysis, Drawings and Specifications)

Interface Control (Internal & External)

• Design Verification Document Control Design Change Controls Corrective Action Records Audits (Internal & External)

Software Quality Assurance Based on this assessment it was determined that YAEC's implementing procedures, as defined in various Engineering Instructions, Technical Administrative Guidelines and Quality Assurance Procedures, provide adequate and . sufficient controls in the area of Design Control. Further, these implementing procedures satisfy the requirements of:

ANSI N45.2.11 (1974)

• Regulatory Guide 1.64, Revision 2

• ASME NQA-1-1989 through the NQA-lb-1991 Addenda In addition, the most recent changes to these Engmeering Instruction, as documented in Engineering Modification Nos. 50 and 51, were reviewed. The Engineering Modifications provided a descriptive summary of the changes along with the reasons for the changes. In allinstances, changes to the Engineering Instructions were viewed as enhancements to the existing Design Control measures.

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B.4.4 .loint Ouality Audit Group In 1996, YAEC QA Department formed an organization named the Joint Quality Audit Group (JQAG) whose mission was to provide plant affiliate oversight of the YAEC engineering activities. The JQAG is comprised of QA representatives from, YAEC, Maine Yankee, Vermont Yankee, Seabrook, Northeast Utilities and Boston Edison. The JQAG meets quarterly and reviews oversight activities performed and planned. The basic mission of the JQAG has b en designed as follows:

1. Discuss YAEC audit and surveillance issues or concerns which are of common interest or which may have potentialimpact upon YAEC QA Department products and services;

2. Discuss YAEC QA Department Functional Area Assessment results including action plans and updates to the plans;

3. Facilitatt .iscussion of emerging industry issues and share available information from industry organizations, such as NRC, NEI, and EPRI;

4. Review and comment on the YAEC-QAS generated audit schedules for intemal and plant audits;

5. Facilitate participation on selected YAEC audits;

6. Report on the status of the YAEC audit program, upon request, to: (I) Nuclear Safety Review Committee (NS ARC) of plants receiving YAEC technical services, (ii) YAEC Board of Directors and (iii) other appropriate requestors.

7. Provide input to plans for an independent audit of YAEC.

8. Provide a forum for establishing a common audit program which can be used at each of the JQAG memb:r plants. This willinclude the development of procedures, schedules, reporting techniques and consistent follow-up of open items.

B.4.5 Functional Area Representatives In order to provide more frequent and consistent internal oversight of engineering activities, YAEC QAS established QA functional area representatives. The representatives are responsible for day to day contact with the applicable functional area, performing surveillances of fu .tional areas, reviewing procedures and assisting in resolving quality matters. In 1996 and 1997, the QA representatives t erformed a total of twenty-three (23) surveillances which supplemented the regularly scheduled audits.

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i i B.4.6 Technical Specialict Pronram

' An initiative was undertaken in 1996 to increase the use of technical specialists on the

! internal audits. During the period from 1995 to 1996, there were a total of thirty internal

. audits performed and there were technical specialists from Northeast Utilities, Scientech, l YAEC, Boston Edison, Seabrook, Vermont Yankee, South Carolina Electric & Gas, New York Power Authority, and also independent consultants to assist YAEC QA. The use of additional technical specialists allowed for more in-depth verification of technical matters and also for more objectiveness in reporting results.

. B.4.7 Functional Ama Acceccments Ir 1996 and 1997, YAEC departments were responsible for preparing a Functional Area Assessment (FAA). This process required the YAEC departments to evaluate their own j performance and grade themselves according to the results of audits, self assessments, l inspection reports, surveillances and management expectations. The reports included a l description of the tasks performed by the functional areas, a list of department strengths and weaknesses, and an improvement plan to address the weaknesses in the upcoming l year. In addition, Vermont Yankee had set client expectations to implement FAAs in i YAEC departments which perform a significant amount of work for them. The FAA was

used by the QA Department to determine the amount of oversight required in each functio ial area and also to determine generically the focus for every audit.

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B.4.8 Self Assessments I

i In 1995, few YAEC departments were performing self assessments. This was mainly due to the fact that management expectations were not communicated to alllevels of the

organization. This became widely known after an audit was performed in early 1996. As l a result of that audit, department management began performing self assessments and l staff personnel became more aware of the benefits of performing them. The process for

' documenting, tracking issues and corrective actions was inconsistent throughout the i company. In 1997, self assessments were being performed by all departments on a more routine basis. These assessments were value added, meaningful and very insightful as compared to the ones performed during 1995 and 1996. Departments were tracking issues and correcting them, they were being brought to the attention of management, and in some departments, benchmarking and performance goals were being used to evaluate their performance. A shortcoming of the process was that action items were not being tracked in a single system. This shortcoming is to be resolved in 1998.

B.5 AUDITS AND ASSESSMENTS

! Throughout 1995 to 1997, YAEC engineering activities were subjected to numerous audits by YAEC QA and independent assessments by outside organizations. Each YAEC organization m.momms B-18 l

1 involved in safetv-related activities was subjected to audits and/or assessments. The results of these activities arc summarized in the following subsections.

I B.5.1 Internal Audits / assessments of Encineering Related Activities Twenty-five (25) audits were performed by YAEC QA of engineering activities performed by YAEC project groups. Of these 25 audits, two (2) multi-department audits were conducted in 1997, rather than the individual department audits as performed in 1995 and 1996.These audits and assessments were both scheduled and unscheduled, addressing both routine QA audits and special audits in areas identified by YAEC management. Frequently, the audit teams included technical specialists. These specialists were specifically matched to the engineering disciplines which were the subject of the audits Specialists were obtained from: (I) within YAEC, (ii) plants inside and outside NRC Region I, and (iii) independent consulting firms. This use of technical specialists provides a valuable blending of technical end programmatic expertise.

Additionally, the knowledge and experience of technical specialists outside of YAEC allows the engineering activities being assessed to be benchmarked against non-YAEC organizations. Common areas requiring improvement were found in the areas of:

(I) program ownership, (ii) analyses inconsistencies, (iii) documentation control, (iv) department procedure enhancements, and (v) software control.

B.5.1.1 Ownershin of Procrams During various project audits, identified issues indicated that (1) interfaces between YAEC projects and their respective plants and (ii) own:rship of various technical programs (e.g., Appendix R, EQ, design control) were weak.

Actions were developed to address these weaknesses. As a result, clear and concise procedures were implemented and project meetings with responsible plant personnel were held to resolve and strengthen lines of communication and ownership.

B.S.I.2 Adherence to Procedural / Code Reauirements A number ofidentified weaknesses were " failure to follow" procedures / code requirements. The results indicated that engineering instructions governing implementation of code requirements were not cons:stently being followed.

Training sessions were established on the YAEC Engineering Instructions (WEs) and personnel accountability for procedural compliance was stressed.

NED checklists were implemented as part of calculation closure to ensure administrative compliance with the WEs.

Additionally, a task force was established to review the inconsistencies in the WEs and to revise them, as necessary. As a result of this effort, (I) procedures w m mso w acs B-19

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controlling the preparation and review of analyses are clear and concise for the user, and (ii) WEs interface more effectively with each other.

l B.5.1.3 Analysis Inconsistencies Issues identified during audits indicated inconsistencies in analyses. These I inconsistencies were brought to the attention of project management. Technical specialists utilized as members of the engineering audit teams noted that although analysis inconsistencies were identified, no circumstances were found where they affected the operability of the respective plants. However, a small percentage of them did result in the generation of either a Basis for Maintaining Operation (BMO) or a Justification for Continued Operation (JCO) document. Corrective actions focused on stronger design reviews, more descriptive assumptions and inputs within the analyses, and more thorough / detailed comment resolution ofindependent reviewer comments.

B.5.1.4 Control of Documentation / Records Audits identified documentation inconsistencies and record turnover issues.

Areas for controlling document turnover as a QA record to document control were reviewed in audits of the project groups at YAEC. Adherence to ANSI N45.2.9 in the past was confusing because of the general guidance provided for record storage. Corrective actions taken by the YAEC project groups included several task forces with Document Control personnel and the QA Manual revisions. Analysis procedures now contain detailed instructions for record control, addressing both temporary and permanent storage requirements.

B.5.1.5 Software OA Control Assessments indicated that software generated by engineers outside of the Software Control Library did not consistently meet engineering instruction requirements. Corrective actions generated by project groups for audit issues led to the revision of the engineering instructions for software control. The engineering instructions for calculation control was revised to provide better guidelines in the control of various computer codes generated. Revised software control procedures are more detailed and user friendly. Additionally, enhancements have been made to (I) the Software Control Library and its implementing procedure, (ii) the interface between the software control and calculation procedures, and (iii) the control of software used by project engineers.

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B.5.2 .Toint Utility Management Auditr 10CFR50 Appendix B, ANSI N45.2 and ANSI N18.7 require that all aspects of the QA Program be audited, and also that the adequacy and effectiveness of the QA program be regularly assessed. In addition, plant technical specifications require that the QA programs be audited at least every two years. Since YAEC QA supplied QA services to many Yankee plants, YAEC QA became a member of an organization called the Joint Utility Management Audit (JUMA) group in order to fulfill the3e NRC requirements and to remain independent for this activity. JUMA performs audits and assessments of member utilities with representatives independent of a specific utility. As an example, representatives from utilities A, B, and C would audit utility D, thus providing the independence and objective oversiew of QA program implementation required by federal regulations and industry standards. There were three JUMA audits performed at YAEC during the period of 1995-1996.

B.S.2.1 1995 JUMA Audit of Vermont Yankee The 1995 audit consisted of a team with representatives from Philadelphia Electric Company (PECO), South Carolina Gas & Electric ( SCE&G) and Commonwealth Edison ( Comed). The focus of the audit was YAEC functions for the Vermont Yankee plant QA program; the specific elements reviewed during the audit were as follows:

Corrective Action / Audits Peer Inspection / Audits / Surveillances Vendor Quality Assurance The audit of corrective action activities included a review of the adequacy of management support for the new Vermont Yankee corrective action system, QA involvement in the system, use of the system during audits, and threshold for reporting deficiencies. The auditors interviewed personnel and evaluated a sample of audit reports with findings, and a sample of corrective action documents. As a result of assessing this area, the audit team concluded that the Vermont Yankee corrective action system was effective and provided enhanced problem reporting, increased management awareners of problems, and simplified process control.

The second area assessed by the JUMA team, peer inspection, included an assessment of the adequacy of the QA audits and surveillances over peer inspection activities, adequacy of QA reviews of contractor QC, and effectiveness of attributes selected for peer inspection. The report concluded that audits and surveillances over peer inspection activities were adequate, but could be improved by ensuring aggressive corrective action was taken by audited organizations. In addition QA involvement in and oversight of wmmomms B-21

contractor QC activities could be increased. Attributes used to perform inspections were consistent with procedures and regulatory requirements.

The last area assessed during the 1995 JUMA audit was the Vendor QA activities. The specific elements reviewed under this area included:

Procurement Engineering and Vendor QA redundancy

• Communications between the Plant Maintenance, Procurement Engineering, and Vendor QA

=

Vendor QA Oversight of Receipt Inspection The audit concluded that the Procurement Engineering (PE) and Vendor QA (VQA) do not duplicate their activities. Communications among each of the involved procurement organizations is satisfactory, for the most part, although some improvements could be made between VQA and PE. Vendor QA oversight of receipt inspection was considered satisfactory.

The results of the audit did not identify any deficiencies in program implementation. There were ten (10) recommendations for improvements in the audited areas which were addressed and resolved by both Vermont Yankee and YAEC.

B.5.2.2 1936 JUMA Audit of Maine Yankee The 1996 JUMA audit was conducted of Maine Yankee QPD by a team consisting of Florida Power & Light, Entergy, and American Electric Power representatives. The scope of the audit included the in-plant audit program, the YAEC internal audit program and the plant QC inspection program.

The evaluation of the in plant audit program was accomplished by interviewing audit personnel and customers of the audit program, review of audit and surveillance documentation including reports, procedures and corrective action documentation. The report identified that the audit program was being effectiv:ly implemented and that the personnel conducting the audits had good knowledge of the areas being audited. The audit process made good use of technical specialists and the reports were well written. The audits that were reviewed were in compliance with the QA program and technical specifications.

The receipt and in-plant QC programs at MY were reviewed. ~1 he assessment included an examination of documentation, observation of activities and interviews with personnel. The audit team concluded that the QC program was effectively implemented and generally meets the needs of the plant.

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Opportunities for improvement included better oversight of receipt inspection, and more use of analytical equipment to verify material properties The last area assessed by the 1996 JUMA audit team was the YAEC internal audits of functional areas providing services to Maine Yankee. The evaluation included a review of audit schedules, plans, reports, corrective action documents, and interviews with selected personnel. Based on the reviews and interviews, the audit team concluded that the audits were very thorough and effective. Everyone on the audit team believed that the audits were more effective now than those performed five years earlier. They did, however, make several recommendations for improvement. First, they recommended that the audit group track the 10CFR50, Appendix B, elements and functional tasks audited to assure comprehensive coverage, and also to obtain input from audited organizations for activities to be considered during the audit. Lastly they recommended that audits increase the use of technical specialists from outside the region to facilitate bench marking. The audit team further identified a weakness with the oversight of YAEC audit activities by Maine Yankee.

Thev encouraged their participation in the newly formed Joint Quality Audit 0 (JQAG) which was established to provide the sponsor plants with the

, gnunity to input into YAEC audit plans, participate on audits, and to recommend corrective actions for future audits.

The 1996 JUM A audit did not identify any deficiencies; however, their were thirteen recommendations for improvement. Only three the recommendations noted above pertair.ed to YAEC. Subsequent to the audit, YAEC QA acted upon the JUMA recommendations by developing a comprehensive matrix identifying areas and elements audited, by requesting customer input to audit plans, and by utilizing technical specialists from outside the region to the maximum extent possible.

B.5.2.3 1997 JUMA Audit of Vermont Yankee The 1997 JUMA audit focused on the Vermont Yankee plant and assessed the corrective action programs, the audit program and vendor QA activities. The audit team consisted of representatives from ENTERGY, South Carolina Electric & Gas, PECO Energy and a peer evaluator from Vermont Yankee.

The evaluation of the corrective action system included a review of the newly instituted Condition Report (CR) system at YAEC. The audit team reviewed the procedure, a sample of reports, and interviews with personnel. The team concluded that the CR process has high level management involvement and that the QA personnelinvolved in the process displayed a lot of dedication and ownership. Their was some concern raised about the timeliness ofinstituting procedural changes for the system, and also for the backlog of overdue action mesown ms B-23

items. The QA department has since made improvements in the CR system based on the recommendations from JUMA. Less than a month after their audit, the CR procedure was revised to reflect the current practices for process control of the CR. Backlogged issues were brought to the attention of YAEC management and the overdue items dramatically reduced.

The audit team also evaluated the internal YAEC audit program by reviewing audit schedules, plans and reports. The team concluded that the audits were of sufficient depth and scope, and that they were performed by personnel who had enough technical experience to identify safety significant issues. The team also stated that the use of QA functional area representatives has improved communications between departments. Lastly, the team identified that the auditors utilized performance based information obtained from Functional Area Assessments (FAAs) to further explore weaknesses and strengtas during audits. There were four recommendations relative to improving the audit process. The first was to become more aggressive in assuring the timely resolution ofissues. The second recommended that management proyide better direction for the performance of self assessments. The third recommendation pertained to revising the surveillance and audit procedures so that they remained current with actual practices. Lastly, they recommended that the audit group review audit plans against the most recent FAA to assure adequate oversight. The recommendations were acted on by appropriate management and resolved prior to end of 1997.

The JUMA team also reviewed the vendor QA activities related to the Vermont Yankee Cycle 20 core design and reload analysis. The review specifically included an evaluation of the organizational interfaces, oversight by QA and Fuels departments, and exchange of key design documents. The JUMA auditors concluded that vendor QA activities were effective. There were four recommendations for improvement in the areas of developing a reload interface procedure, establishment of a single point interface contact, development of an emerging issue action plan, and changing the time requirement for issuance of vendor audit and surveillance reports. Each of the recommendations in the vendor QA area have been addressed by QA management and resolved.

The conclusion reached by the 1997 JUMA audit team was that areas evaluated were adequate and effective. There were no deficiencies identified, however there were fourteen recommendations for improvement, all of which were addressed by QA management.

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s APPENDIX B REFERENCES B.1 Letter, G. D. Whittier (MYAPCo) to USNRC, MN-96-08, " Docket No. 50-309, " Maine Yankee Atomic Power Station Response Team and Independent Review Team Reports,"

dated January 22,1996.

B.2 R. T. Fernandez, et al., "RELAP5YA - A Computer Program for LWR Thermal-Hydraulic Analysis," Report YAEC-1300P, Volumes 1,2, and 3, October 1982.

B.3 Letter, USNRC to Maine Yankee, " Acceptance for Referencing Topical Report YAEC-1300P, Volumes 1,2,3, RELAP5YA: A Computer Program for Light Water Reactor System Thermal-Hydraulic Analysis," dated January 30,1989, with enclosures. ,

B.4 Ransom, V. H., et al., "RELAP5/ MODI Code Manual, Volumes I and 2, NUREG/CR-1826, March 1982.

B.5 Letter, L. H. Heider (YAEC) to D. G. Eisenhut (USNRC), FYR 83-9, FVY 83 4 and MN 83-12, dated January 14,1983.

B.6 Willeutt, G. J. E., Jr., Letter Q-7-83-559, " Review of Yankee Atomic RELAP5YA Small Break LOCA Model," Los Alamos National Laboratory, dated November 17,1953 (from Reference NMY 86-51).

B.7 Fineman, C. P., " Technical Evaluation Report: RELAP5YA Computer Program for Use in PWR Small Break Analysis", Idaho National Engineering Laboratory, EGG-TFM-7933, dated May 1988.

B.8 Letter, G. D. Whittier (MYAPC) to D. G. Eisenhut (USNRC), MN-84-56, " Resolution of TMl Action item II.K.3.5, ' Automatic Trip of Reactor Coolant Pumps'," dated April 9,1984.

B.9 Letter, USNRC to Maine Yankee, " Maine Yankee Reactor Coolant Pump Trip," dated April 15,1986, NMY 86-51.

B.10 YAEC-1868, " Maine Yankee Small Break LOCA Analyses," 1993.

B.11 Letter, James R. Hebert (MYAPCo) to W. T. Russell (USNRC), MN-96-018, dated February 21,1996.

B.12 Letter, Patrii M. Sears (USNRC) to C. D. Frizzle (MYAPC), " Maine Yankee:

Implementation of Small Break LOCA Analysis, NUREG-0737 II.K.3.30 and II.K.3.31 (TAC 48176), dated May 8,1989.

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B.13 12tter, S. P. Schultz (YAEC) to E. H. Tr'ottier (USNRC), "Information Regarding the Small Break Loss of Coolant Accident (SBLOCA) Analysis for Cycle 15 Operation at Maine I

l Yankee," dated December 14,1995.'

B.14 Letter, C. D. Frizzle (MYAPCo) to W. T. Russell (USNRC), MN-96-056, " Submittal of Maine Yankee SBLOCA Licensing Analysis in Compliance with 10CFR50.46 and in Satisfaction of TMl Action items II.K.3.30, II.K.3.31, and II.K.3.5," dated April 25,1996 L

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APPENDIX C RELAP5YA SBLOCA TECHNICAL ISSUES ASSESSMENT summw ncs

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