ML20203L162

Rept of Root Cause Assessment Review

ML20203L162
Person / Time
Site:	Yankee Rowe
Issue date:	02/18/1998
From:	DUKE ENGINEERING & SERVICES
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ML20203L159	List: ML20203L155 ML20203L162 ML20203L193 ML20203L227 ML20203L245 ML20203L269
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NUDOCS 9803050381
Download: ML20203L162 (32)

Text

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Report of a ROOT CAUSE ASSESSMENT REVIEW Performed for DUKE ENGINEERING & SERVICES, INC.

Bolton, Massachusetts Performed by Powerdyne Corporation ,

t Reviewer / Preparer (b , - s Date bk\}\

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ROOT CAUSE ASSESSMENT REVIEW REPORT TABII OF CONTENTS 1.0 P U RPO S E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E - 1 2.0 SCOPE................................................................E-1 3.0 REVIEW PROCES S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E- 1 4.0 B AC KG R OUND . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E - 2 4.1 History of the Concerns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-2 4.2 Brief Company History - the Context for the Concerns . . . . . . . . . . . . . . . . . . . . . . . E-3 ,

5.0 GENERAL O B S ERVATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-5 5.1 Perspectives on the NRC's Concerns , . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-5 5.2 Organizational Responsibilities / Roles / Communications . . . . . . . . . . . . . . . . . . . . . . . E-7 5.3 Formal Company Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-8 5.5 . Licensing and Quality Assurance Involvement .............................E-13 5.6 Thinking Outside the Box, or Taking the Broader View . . . . . . . . . . . . . . . . . . . . . . E-14 6.0

SUMMARY

CONCLUSIONS AND RECOMMENDATIONS . . . . . . . . . . . . . . . . . . . . . E-16 6.1 Regarding Perspectives on the NRC's Concerns . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-16 6.2 Regarding Project Organizational Responsibilities / Roles / Communications . . . . . . . . E-16 6.3 Regarding Formal Company Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-17 6.4 Regarding Company Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E- 18 6.5 Regarding Licensing and Quality Assurance Involvement . . . . . . . . . . . . . . . . . . . . . E-19 6.6 Regarding Thinking Outside the Box, or Takirig the Broader View . . . . . . . . . . . . . . E-20 7.0 OVERALL CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..........E-21 LIST OF PERS ONS CONTACrED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-2 2 LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-23 QUALIFICATIONS OF REVIEWER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-25 APPENDIX A - LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . ............I-A APPENDIX B - LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I-B APPENDIX C - QUALIFICATIONS OF REVIEWER . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I-C

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b 1.0 PURPOSE On December 19,1997, the U.S. Nuclear Regulatory Commission (NRC) issued a Demand for Information Letter (Demand) to Yankee Atomic Electric Company (YAEC) and Duke Engineering & Services Company (DE&S) to obtain information the NRC considered necessary to determine whether the addressees should continue to provide engineering analyses, and in particular Loss of Coolant Accident (LOCA) analyses, to NRC power reactor licensees. This letter was issued as a result of an NRC inspection finding that LOCA analyses performed by l YAEC for Maine Yankee Atomic Power Co. (MYAPCo) were inadequate and tha inaccurate information concerning those analyses had been provided to the NRC.

l This report documents an independent Root Cause Assessment Review (RCAR or Review) performed by Powerdyne Corporation (Reviewer) of the discrepancies identined in the NRC's Demand. The purposes of this review were to:

(1) Validate that the root and contributing causes of activities described in the Demand letter had been appropriately identified; (2) Validate that the corrective actions taken to remedy the identified causes were appropriate.

2.0 S C 6 M.

The scope of this RCAR included review of relevant activities regarding a previously performed YAEC Self-Assessment (performed in April,1996) and corrective action identification (a December,1996 response to the Self-Assessment), reviews of documents associated with the discrepancies identified, review of relevant YAEC administrative and technical procedures, interviews with YAEC managers and other employees, and reviews of samples of analyses recently performed by the DE&S Bolton, Massachusetts Office. The Review was aimed at determining the validity of the identification of the root causes and corrective actions taken at the time of the Self-Assessment and their validity and relevance in the context of the current DE&S acquisition of the engineering organization of YAEC.

3.0 REVIEW PROCESS The Root Cause Assessment Review was conducted during the period January 19,1998 through January 23,1998 in the Bolton, Massachusetts office of DE&S.

The process began with reviews of the NRC's Demand letter and a December 19,1997 NRC letter to MYAPCo describing apparent violations stemming from discrepancies in their LOCA analyses which had be:n performed by YAEC in order to gain a clear understanding of the NRC's concerns, their understanding of the analyses discrepancies, and their position with regard to DE&S. Next, a report of the April,1996 YAEC Self-Assessmerawas reviewed along with the December,1996 Self-Assessment response by the engineering organization. This Self-Assessment hd been prompted by the NRC's initial identification of the MYAPCo. LOCA analyses discrepancies. Next, two parallel assessment paths were pursued, reviewing relevant administrative and technical procedures and interviewing management and production personnel

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l in the DE&S organization who were responsible for the specific LOCA analyses that had been

! questioned by the NRC as well as individuals responsible for various aspects of other engineering l

analyses. And finally, samples of recent analyses wera eviewed to determine if the types of concerns identified by the NRC or other concerns stik existed and if the corrective action; had been effective.

The RCAR was conducted by Donald C. Prevatte, an engineering and management consultant with Powerdyne Corporation, a company providing services to the USNRC, the U. S. Department of Energy, and the nuclear power industry since 1982 (Appendix C provides details of the reviewer's qualifications).

4.0 BACKGROUND

4.1 Historv of the Concerns In December 1995, the NRC conducted a technical review of the YAEC headquarters office in Bolton, Massachusetts and performed an investigation in response to allegations concerning the Maine Yankee LOCA analyses, which had been performed by YAEC. The NRC concluded from these activities that:

(1) By YAEC's preparation and approval cf the RELAP5YA small break LOCA analysis (SBLOCA)and the WREM large break LOCA analysis (LBLOCA), and by YAEC's preparation and approval of the Core Performance Analysis Reports (CPARs) used to support Cycle 14 and Cycle 15 operation at Maine Yankee, YAEC caused MYAPCo to be in apparent violation cf 10CFR s 50.46(a)(1).

(2) YAEC provided MYAPCo with information that was not complete and accurate in all material respects regarding this noncompliance with the above cited regulation, and thus caused MYAPCo to apparently violate 10CFR s 50.9(a) by maintaining CPARs which contained information which was not complete and accurate in all material resp: cts in connection with MYAPCo's Cycle 14 and Cycle 15 reload analyses.

(3) As a result ofits incorrect calculation of penetration factors, misapplication of the Alb-Chambre correlation, and inadequate review of YAEC-1868, YAEC caused MYAPCo to rely on an unacceptable evaluation model which overpredicted core cooling and overstated the conservatism of the evaluation model for Cycle 14 and Cycle 15 in apparent violation of 10CFR s 50.46(a)(1).

(4) 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 MYAPCo to apparently violate 10CFR ! 50.46(a)(1).

YAEC in April,1996 initiated a Self-Assessment conducted by its Quality Assurance Department and documented in RELAP5YA Self Assessment, Maine Yankee and Yankee Atomic,4/96, a report attached to YAEC Memorandum 632.TGS dated 4/30/96. This wamov ucs E-2  ;

effort identified the following three major common issues (root causes) in addition to numerous attendant and contributing causes for the NRC identified discrepancies:

(1) The divisions of responsibilities / ownership and roles of organizations performing, controlling, administering, and managNg activities for the Maine Yankee plant were not completely and clearly defined or understood by all parties involved in or impacted by the activities.

(2) YAEC personnel lacked formal company training end retraining in areas that were essential to performance of their day-to-day tasks.

(3) Procedures for controlling analyses were weak in defining important processes, and these processes were driven by personal knowledge rather than by procedural guidance. Most siecific to the NRC's concerns, the procedures did not require identification of the effects of analyses on licensing commitments or design basis documents.

Subsequent to the Self-Assessment, corrective actions were taken in the YAEC's engineering organization, and in December,1996 a response to the Self-Assessment.

Yankee Atomic Memorandum NEDMY96-052,12/31/96, " Response to RELAP5YA Self-Assessment", was generated outlining those corrective actions.

On December 19,1997, the NRC issued its Demand for Information Letter to YAEC and DE&S, to be responded to within 30 days (later extended to 60 days), and in January, 1998, DE&S initiated this independent RCAR.

4.2 Brief Comnany History - the Context for the Concerns To clearly understand the concerns addressed by this Review and their root causes, one '

must understand their historical context, as was gleaned through interviews with several managers, including the YAEC Chairman and Chief Executive Officer who presided over the selling of YAEC's engineering services division to DE&S.

4 YAEC was one of the pioneers in commercial nuclear power in the United States, and was at the center of most of the commercial nuclear power endeavors in New England from the early years up until quite recently. New England was, and still is, somewhat unique, in that most of the nuclear power plants were owned by numerous, relative!y smalllocal utilities, many of whom held small interests in several of these plants. Few of these owners were large enough or sufficiently qualified to shoulder the complete responsibility of such a facility. YAEC, on tne other hand, had the organizational entity with the size and qualifications needed to design, operate, and maintain these plant, and in the cases of the Main Yankee, Vermont Yankee, Yankee Rowe plants, YAEC actually held the original operating licenses. As a result of this multi ownership, multi-plant situation, the relationships between the various utility owners and YAEC was, from the beginning, very familial, with the various roles and responsibilities understood by the parties mostly

, through their common historical perspectives rather than through explicit procedural ,

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requirements, although, as with other nuclear organizations, procedural controls tended to

increase with time.

In 1981, YAEC's relationship with Maine Yankee changed dramatically, with Maine Yankee taking over as the plant operating licensee holder. With this change, YAEC was no longer the primary point of contact for licensing related communications with the NRC for Maine Yankee. Although YAEC remained the primary engineering services provider, Maine Yankee could and did acquire services from other engineering firms. The precise organizational relationships and responsibilities between MY and YAEC, however, were not redefined or documented at this point, but facets of their former relationships appeared to have been assumed to remain by one or both parties, sometimes incorrectly. This was the environment in which the discrepancies discovered by the NRC were bred.

Another important historical context for these concerrc was the relationship between YAEC and the NRC with respect to the reporting of LOCA analyses and the unique regulatory requirements applicable to LOCA analyses. Traditionally, these analyses were pecformed by nuclear Steam supply system (NSSS) vendors, such as Westinghouse or General Electric, who worked in close concert with the NRC in the development and application of the codes used. YAEC had not participated in this early development work, and therefore, was not familiar with the informal precedents and expectations that had evolved, although it was aware of the CFR-mandated formal reporting requirements for LOCA analyses. It was not aware, for instance, of the NRC's expectations for regular dialorie during the development stages of codes that had become standard procedure with the NSSS vendors during the early days of the business. During the application of the NRC-approved RELAP5YA code for the Maine Yankee SBLOCA, YAEC had concluded that further interactions with the NRC were not required based on a closeout letter which Maine Yankee received from the NRC. This unfamiliarity with the NRC's expectations concerning dialogue on these code applications, though not mandated by the CFR's, appeared to have been a contributing cause to the NRC's displeasure with YAEC's performance.

During the twenty-one month period from the Self-Assessment until this Review, two other historical events transpired that significantly affected the organizational relationship between YAEC and the NRC, and YAEC's present and future relationship with the NRC:

(1) In August,1997, the owners of Maine Yankee made the decision to permanently shut down and decommission the plant and (2) on December 1997, YAEC was acquired by DE&S. These events rendered many of the specifics of the Self-Assessment observations and responses obsolete. However, all of the principles embodied in the Self-Assessment report were still valid and applicable to the current situation. Therefore, this review addressed the findings and responses to the Self-Assessment in the context of the situation as it existed at that time, as well as the embodied principles as they applied to the current context - an established organization with longstanding proven nuclear power industry capabilities, in the midst of redefining itself- its organization and culture, its new roles and responsibilities, its new participants, and its significantly different business environment.

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5.0 GENERAL OBSERVATIONS _

I This RCAR addressed each of the three major topics of concern identified in the Self-Assessment as well as other areas not addressed. The observations contained herein were based on a ,

combination of reviews of the assessment and respcase documents (see Appendix B fer a list of the documents reviewed), interviews with key managers in the former YAEC organization and with personnel with direct responsibilities for either production work performance or for support functions to the organization, such as training, quality assurance, and licensing (see Appendix A for a list of the personnel contacted), reviews of administrative and technical procedures (Appendix B), and reviews of a aampling of the production work (Appendix B). The following sections describe briefly the reviewers observations:

5.1 Persnectives on the NRC's Concerns Two YAEC engineers were identified in the NRC Demand letter as being primarily responsible for the concerns with the LOCA analyses for Maine Yankee - one of the individuals who produced the analyses and the responsible Group Manager. Both individuals were interviewed in this Review to assure that their perspectives of the NRC concerns were captured. Summarized, the NRC concerns were:

(1) YAEC performed no specific code analysis demonstrating compliance with 10CFR50.46 in the LOCA break range from 0.35 ft to 0.6 ft 2, the gap between the analyses capabilities of YAEC's small break and large break LOCA codes.

(2) YAEC did not adequately apprise MYAPCo of this analysis gap, thus causing them to violate 10CFR50.46 and 10CFR50.9, (3) YAEC's 10CFR50.46 analyses used incorrect or misapplied factors and correlations, and because of inadequate QA, failed to detect these errors, causing MYAPCo to violate 10CFR50.46, and (4) YAEC used an unacceptable "best estimate" small break analysis to resolve a concern with reduced Steam generator pressure, thereby causing MYAPCo to violate 10CFR50.46.

Both individuals contended that although, in hindsight, analysis errors and reporting errors were made, the NRC letter had incorrectly characterized these errors and had unjustly singled out YAEC and themselves as the cause of the concerns as follows:

(1) Although, as maintained by the NRC, there was no overlap in the range of break sizes for which specific code analyses were performed, the individuals felt that theirjudgement that the full range of the break spectrum was represented by the analyses that had been performed was reasonable, appropriate, justified, and valid at the time. This was based on several facts: (a) The trends of the results in the adjacent break ranges gave reasonable indication that peak cladding temperatures and the other critical parameters in the unanalyzed range would not peak or be higher than the results obtained in the analyzed regions. (b) They knew of no

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contrary indications for code analyses performed for other plants of similar size and design or in their previous experience. (c) Two subsequent code analyses that addressed the unanalyzed range - one a formal analysis performed by Siemens and the other an informal YAEC analysis using the small break LOCA code with adjustments to render it more stable in this region - both confirmed the validity of their judgements, and further, showed that the YAEC analyses were, indeed, very conservative. (It was also noted by the Reviewer that 10CFR50.46 does not specifically require that the code analyses be overlapping for the complete spectrum of break sizes as seemed to be implied by the NRC's Demand letter, although it does appear to imply that where gaps exist, these should be brought to the NRC's attention.)

(2) The YAEC small break LOCA analysis report to Maine Yankee specifically identified cede instability for break sizes above 0.35 ft 2, and these individuals stated that this code characteristic was discussed on several occasions with Maine Yankee's cognizant personnel. Although the NRC Demand letter acknowledged the report statements, it stated that the report's language would not signify to someone without LOCA code expertise that the code was not valid for breaks larger than 0.35 ft2 . However, the responsible YAEC individuals felt that the licensee audience for this report had been sufficiently qualified to understand the report's language, particularly in view of the attendant discussions that had taken place, and that therefore it was not reasonable to hold YAEC and themselves totally responsible for this miscommunication (3) The YAEC individuals acknowledged the analyses errors that were identified by the NRC and the failure to identify them in the checking process, but they maintained that the effects of these errors on the results were insignificant.

(4) The YAEC individuals acknowledged that, in retrospect, using the "best-estimate" small break LOCA analysis to resolve the concern with the lower Steam generator pressure was an error, but pointed out that it was used with the full knowledge, understanding, and concurrence of YAEC management and the licensee's cognizant engineers.

Since the NRC's inspection that had initially identified these concerns, a great deal of these individuals' efforts had been directed toward defending against the allegations surrounding them. This appeared to have taken its toll on their ability to contribute to normal YAEC working endeavors and in their personallives. Although the reviewer did not have the time to verify the accuracy of, or delve deeply into, the details of their individual stories, the perception derived from their interviews, other interviews, and other information gathered was that the personal damage these indivicaals had received and were continuing to receive was not commensurate with any technical or judgement mistakes they may have made.

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5.2 Orcanizational Resnonsibilities/ Roles / Communications The first major common issue identified by the RELAP5YA Self-Assessment was that "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 activities." Many of the interviewees contacted in the Self-Assessment also felt that the relationship with Maine Yankee was changeable, from contractor to integral part of the Maine Yankee organization, depending on who was contacted and/or which relationship was most advantageous to Maine Yankee. The lines of communications /

l reporting between the organizations and with external organizations, such as the NRC, were not clearly defined or understood. The Team recommended generating interface documents clearly defining the relationships between YAEC and Maine Yankee and clearly communicating these relationships to their respective staffs.

The Reviewer agreed with this assessmene it appeared that lack of formal def'mition of responsibilities, roles, lines of communications, etc. had been a major contributor to the analysis and reporting errors that were the focus of the NRC's criticism. The Reviewer believed further, based on interviews that revealed the common history of the two companies, that one of the root causes of this poor definition was this common history where such responsibilities, roles, and communications had been almost solely within the purview of YAEC as the license holder, and that this had been clearly understood, though not clearly documented, by both organizations during that early period. And that with the transfer of the license to Maine Yankee, YAEC had retained some of the previous responsibilities, but there had never been conscious, concerted decisions made in either organization concerning responsibility splits, with clear, written communications generated describing these responsibility splits. And further, for the vast majority of tasks undertaken by YAEC, the residual unwritten understandings between the companies after the license transfer had adequately provided quality work, and that the case in question was one of the few where these informal understandings had proved inadequate to provide the required assurance.

The response to the Self-Assessment on this issue found that Maine Yankee and YAEC managers were working to define the respective roles and responsibilities of their organizations working together and that an engineering responsibility matrix had been generated, but that further work was necessary by both organizations to refine the matrix and define the responsibilities. The response appeared to conclude that, at that point in time (December,1996), the corrective actions had not gone far enough to assuage the problem.

The Reviewer agreed with the response conclusion; however, subsequent to the response, these managers appeared to have made further progress in defining their organizational relationships, but with the advent of the Maine Yankee shutdown, this issue with respect to this licensee, became moot. However, the more global issue, assuring clear definition of responsibilities and lines of communication between YAEC and all clients, an issue that the Self Assessment Team had not been focused on, still remained.

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At the time of this Review, this larger issue had been recognized and discussed by all of tne managers interviewed, and a work control process outline procedure had been drafted in the Safety Assessment Group (part of the Nuclear Engineering Department) to address i

this issue. This procedure appeared to be very comprehensive, touching on all of the major points of concern, such as documentation of requested work from a client,

, definition of deliverables, tracking of work, scope dennition, plan and schedule, client l approval, work performance, and delivery and closecut. The reviewer could identify only

( a few additional significant poin:s that wece probably applicable to this procedure and should be incorporated as follows:

(1) Definition of the project organizational structure for all projects, identification by name the people in the organization positions, and identification by name of their counterparts or principal contact points in the clients' organizations.

(2) Detailed definition of the resouremhat will be required to accomplish a project, integrated with the project schedule, showing when and how these resources will be consumed.

(3) Routine tracking of scope accomplishment, meeting of milestones, and resources consumption tracking to assure that projects remain under control, i.e., delivering products within agreed upon schedules and budgets.

(4) Clearly defining the Quality Assurance and Licensing roles in all projects, or if they have no roles, clearly stating in project definition documents that they have no roles or that they have the minimal role of performing cursory reviews of project deliverables to assure that no new roles have been identified as a result of normal project evolution.

(5) Assuring that all these and the other points in the draft procedure are incorporated into any project contractual documents with DE&S clients to assure that scopes, roles, responsibilities, communications, etc. are clearly understood by all parties from the beginning.

5.3 Formal Comnanv Training The second major common issue identified by the RELAP5YA Self-Assessment was that

" Personal at Yankee Atomic lack formal training and retraining in areas that are essential to performance of their day-to-day tasks." The Self-Assessment also concluded that the training should include the relationship and hierarchy of pertinent codes, standards, and regulatory requirements, and company procedures. Although read-and-sign level was a useful and significant training tool, it was not adequate alone; classroom instruction was felt to be necessary also to assure uniform and complete understandirig of the codes, procedures, regulations, and management expectations.

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The Self Assessment Team recommended the following:

(1) That a formal training program, including classroom training, be established providing initial and periodic training on codes, regulations, and procedures pertinent to employees' areas of work responsibilities.

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l (2) That the Licensing Section Head be filled with someone specifically knowledgeable and dedicated to licensing coordination, or that a Licensing Engineer be specifically assigned to coordinate licensing activities.

(3) That a process be established to develop, review, document, and agree upon technical, regulatory, anc' licensing expectations of all matters in Project Engineering, NED, EED, PSD, etc.

Tite Reviewer agreed with the Self-Assessment Team's finding that a significant contributor to the discrepancies discovered by the NRC was lack of formal, job-specific training for YAEC employees and with the Team's recommendations.

The response memorandum to the Self-Assessment found that the only formal training performed to date had been on the reporting requirements of 10CFR50.9,10CFR50.59, 10CFR50.72, and 10CFR50.72. This included a 3-hour classroom overview of the NRC, 10CFR50.59, and other reporting rcquirements, and a 1-day training session by Winston and Strawn. The response memo also documented pending and planed upgrades of the Maine Yankee licensing staff, but no changes in the licensing area at YAEC were identified. The memo also documented that Self-Assessment Team Recommendation 3 above for establishing a project expectations process had been fulfilled by a procedure established in NED, although this Reviewer was unable to locate the described procedure.

The Reviewer concluded from the individual responses in the response memorandum and through interviews that although sporadic, uncoordinated activities in training had taken place, no credible, coordinated, formal training programs had been established or even mapped out at that time, that there had been no sigmficant activities in the YAEC organization related to generic coordination oflicensing activities and reviews (this will be addressed in more detail in a later observation), and that a coordinated expectations process had not been established. Therefore, the Reviewer concluded that YAEC's response to the training concern identified in the Self-Assessment had been inadequate, and that, at that time, training was stillinadequate to correct the concern.

The next question was, had subsequent activities in this area been adequate to correct the concern, i.e., (1) had a credible formal training program for employees had been established, and (2) if it had, was it effective?

The finc angs in this area were mixed and variable and were based on interviews with key managers, individuals performing work in the organization, an interview with the current company training coordinator, review of training related procedures, and inspection of sample training records.

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Training Procedures:

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Two procedures related to training were found and reviewed, WE-003, Rev 13, dated 6/27/97, Indoctrination of Per. ronnel, and WE-004, Rev 12, dated 6/27/97, Training.

These procedures were part of the Engineering Manual and appeared to adequately call out the training requirements for personnel performing engineering work, although it was not clear why two procedures were required to convey these requirements.

Procedure WE-003 contained a checklist form, Form WEMJ3-1, to document personnel training. Although this form covered many of the appropiate codes, standards, and regulations needed to effectively perform engineering work, it appeared to have the following weaknesses in the requirements it conveyed:

(1) It did not inct.x the internal company procedures designated the Nuclear Engineering Depanment Procedures or the Technical Administrative Guidelines.

(2) In its list of pertinent NRC rules and regulat;ons, it did not include 10CFR20, 10CFR50.9,10CFR50.46,10CFR50.71,10CFR50.72,10CFR50.73, and 10CFR50, Appendix K, two of which contain reporting requirements specific to the particular concerns identified by the NRC, and all of which contain reporting requirements pertinent to many day-to-day tasks engineering tasks in firms such as DE&S.

(3) 10CFR50, Appendix A was not identified as a mandatory regulation for review.

This seemed incongruous for an engineering firm in the nuclear business, since Appendix A contains the NRC's general design criteria (GDC's) for nuclear power plants. Additionally, this list should have also contained the 1967 proposed GDC's, since many older plants were licensed to these rules.

(4) The categories of training did not appear to include classroom training, or previous specific work-related experience.

Based on these reviews the Reviewer concluded that the procedures were adequate but needed improvement.

Training Records:

The Reviewer selected two individuals in the Thermal Hydraulics & Safety Group to verify their training records. Although line manac.:rs and supervisors had the prime responsibility for employee training, such records were required to be maintained as a QA record in RMS. The company Training Coordinator was requested to provide these individuals' records. Although the records showed that both individuals, during their long tenure with YAEC, had received extensive valuable training in technical and work-related subjects, the records showed very little training in codes, standards, regulations, or company procedures specifically relating to their work. Due to time constraints, the i

Reviewer was unable to determine if this observation was due to records not being adequately kept of other training received, such training not actually having been received, smuow,m E-10

or the records of such other training stored in a different location (acceptable practice as i long as RMS holds the official record copy).

Interviews:

Interviews revealed three different perspectives on training; from tiie eyes of the managers, the working level engineers, and the company Training Coordinator.

All managers and working level persons interviewed were not fully srisfied with the state of training as it existed. Although a comprehensive training program had been designed, the managers were frustrated because they felt that insufficient resources were available to support an effective training program. Additionally, since many of their people had been regularly required to work extended hours in recent years in order to meet client commitments, little time was left for training.

The company Training Coordinator was new to the job and appeared quite willing, dedicated, and conscientious. However, the Reviewer was surprised to discover that this individual had not been informed of the Self-Assessment Team's conclusion that inadequate training was one of the major causes for the discrepancies the NRC had identified, and that the Team had recommended corrective actions in the acea of training.

The scope of this position's responsibilities appeared to have been very narrowly defined to include primarily general and administrative training, e.g., employee benefits, EEO, interpersonal negotiation, time management, and similar areas, but no training in codes, regulations, or procedures, and no maintenance of employee records in these areas.

Putting all of this together, the Reviewer concluded that although a modicum of training applicable to the NRC's concems was being performed, overall, the training program was inadequate to provide the necessary assurance to the NRC that this cause for the weaknesses they had detected was being corrected.

5.4 Analvsis Control Procedures The third major common issue identified by the RELAP5YA Self-Assessment was that

" Procedures used for controlling the development of analyses are weak in defining important processes that currently appear to be driven by personnel [ sic] knowledge rather than by procedural guidance. The current procedures do not require identification of the effects of analyses on licensing commitments or design basis documents." The Self-Assessment went on to say that " Yankee Atomic needs to overhaul the procedure for controlling the preparation and issuance of analyses to assure that regulatory limitations and thresholds are considered upon completion of calculations; to assure that licensing design basis documents affected by analyses are changed when needed; to assure that the use of computer codes are within the SER bounds for which they were approved; to assure that measures are in place for the use, control and dissemination of unverified data; and to assure that procedures appropriately address the QA program requirements." The Self-Assessment also found that there was general confusion as to who was responsible for updating the FSAR and other Maine Yankee licensing documents.

mesmwxus E-11

m The Self Asse sment Team recommended that:

(1) Maine Yankee and YAEC managemem need to identify which organization has responsibility for FSAR update and assign a specific cognizant engineer this

, responsibility, l

(2) Maine Yankee and YAEC should revise procedures to assure that appropriate trigger mechanisms are in place to prompt FSAR updates when needed.

(3) Training in licensing and regulatory updating requirements should be instituted.

The Reviewer agreed with the Self-Assessment Team's determination that analyses control procedures were weak and with their recommendations in principle, although, as with other recomtrendations, the Reviewer felt that they were too narrowly focused on just the relationship between Maine Yankee and YAEC; they should have focused instead on the same type relationships with all of YAEC's clients.

The response memorandum to th Self-Assessment on this issue documented that YAEC procedures had bee.n revised to establish or lower thresholds and to ensure coordination, and six specific NED procedures were issued to address istues identified, including 10CFR50.46 reporting, review of computer software documentation, safety and relief valve modding (an EDR finding), FSAR and Technical Specification review and updating, 10CFR50.59 safety evaluations, distribution of information affecting licensing commitments, closeout of service / work requests, and calculation checklists.

The RCAR Reviewer performed a cursory sampling of procedures used by YAEC engineers in performing their work (insufficient time was available to perform a detailed review), with particular emphasis on those procedures addressing calculations, analyses, and licensing reporting (see documents reviewed list in Appendix B). The aims of this samplhg were to (1) determine if the particular concerns of the Self-Assessment, analyses control and licensing reporting, were adequately covered, (2) to verify that the programs and processes required by regulation were properly addressed, and (3) to ascertain if the procedures were clear, concise, understandable, coordinated with each other, and they made sense.

In general, the procedures sampled satisfied the Reviewer's aims; procedures were in place, had been recently updated, and appeared to provide very good control of analyses, and assured that reporting required by regulations was prcpirly performed; the programs and processes required by regulation were properly addressed, including the requirements for a 10CFR50, Appendix B Quality Assurance Program in procedure YOQAP-I-A, communication with federal regulatory agencies in TAG L.1,10CFR21 reporting in TAG No. 6 and WE-109,10CFR50.9 reporting in TAG No.18,10CFR50.46 reporting in NED Procedure No. 5, and reporting of changes in analytical methods in NED M. 8; and the procedures reviewed were generally clear, concise, understandable, and made sense.

Based on this sample review and comments from interviewees, the Reviewer conciti.ed that the weakness-in-procedures concern identified in the Self-Assessment had been s m mo m ra E-12

[

adequately addressed at the time of the response memo of December,1996; that the specine concerns of the Self Assessment, analyses control and NRC reporting had been l j_ well addressed in procedures; and that the company procedures had been further improved

in the time since that response, i

5.5 Licensing and Ounlity Assurance involvement One of the areas touched upon by the Self Assessment Team was the roles of Licensing and Quality Assurance in the production of analyses. Their observations and the resultant

responses were made in the context of the question, which oiganization, Maine Yankee or j YAEC, and which part of the organization was responsible for assuring that the regela'ory requirements and licensing commitments were satisfied and analyses work was performed in a quality manner? Although the specifics of their observations and the responses are

, obsolete today, the basic principles still apply; d work done on nuclear power facilii.es

! must be produced, reviewed, approved, applied, utilized, reported upon, etc., in the context of the regulatory requirements and licensing commitments applicable to that 4 facility and must be done in acco lance with a quality assurance program that conforms l with 10CFR50, Appendix B. DE&S, therefore, must, as a part ofits role as the consulting l 2xpert to whom clients turn for guidance, become more proactive in assuring that for d

work proposed and performed by the Bolton Office, the specific areas oflicensing and l quality assurance responsibility, e.g., reviewing the FS AR, reporting to the NRC, determining the licensing bases, performing audits of the work, are accurately and j completely identified, and the division of these responsibilities between the client and

DE&S is clearly delineated and documented.

!- In this vein then, the Reviewer interviewed the Manager, Regulatory and Industry Affairs.

The puipose was to ascertain if the nece:sity for routine Licensing involvement in future work, if no more than to determine that no involvement is required in certain tasks, was yet recognited in the new evolving mission / role of the DE&S Bolton Office. The interview revealed that, historically YAEC's licensing roles had been mostly limited to two areas: (1) Providing direct NRC interface for clients, who were essentially other members of the Yankee Atomic family, where YAEC possessed specific technical i expertise that required direct NRC contact, such as 10CFR50.46 LOCA analyses, and (2)

for newly emerging fields and issues, such as decommissioning, where YAEC acted as the  :

! licensing vanguard for clients. Involvement in other day-to day tasks for clients had not typicelly required Licensing's involvement. No specific plans had been made to address the more generic day to-day needs indicated by the Self Assessment findings and the changing role of the organization.

The only interview conducted in the Quality Assurance Department was with the Senior Quality Assurance Engineer that had directed the Self-Assessment. It appeared that early QA involvement with future work was a well recognized need at his level, probably as a result of his Self Assessment experience. He also indicated that QA activity in the organization had been quite high in recent years with numerous other assessments having been made. This should also be another positive factor in responding to the NRC's concerns.

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5.6 Thinkine Outside the Bot or Takine the Broader View As a part of the RCAR Review, the Reviewer, in consukation with the Manager of the Thermal Hydraulics & Safety Group, chose two recently generated calculations to review in order to ascertain the quality of the current production, review, and approval process.

Both were .e ociated with spent fuel pool cooling (see the documents reviewed list in Appendix L, The first was a thermal hydraulics analysis of Seabrook's new spent fuel storage racks, and the second was an analysis of passive cooling in the Maine Yankee spent fuel pool for the loss of normal cooling condition.

In both cases, the authors were well experience and qua'lified; in both cases, the objectives, inputs, and assumptions for the calculations were clearly and logically stated; and in both cases, .here appeared to be no computational error.

The Seabrook calculation also contained sensitivity runs which gave good indication of model and code validity. However, a possible cuhural weakness was revealed by this calculation. The model assumed no lateral communication between individual storage rack channels. The Reviewer knew from experience that some rack designs have longitudinal openings at the corners between channels that would allow lateral flow communication between channels. The Reviewer asked if the Seabrook racks had this design;if they did, the model used might have been invalid. The author did not know; she had not reviewed the rack hardware drawings. With subsequent review of these drawings the author verified that the channels were completely enclosed with no lateral flow paths, thereby confirming that model assumption. However, this indicated a possible process weakness; actual hardware documents ns always being consulted to verify that models truly represent the conditions intended to be modeled.

The Maine Yankee fuel pool cooling calculation's purpose was to determine the temperature in the spent fuel pool as a function of time with no active cooling. This task was part of an overall project to " support the desired operation of the SFP in the post-shutdown condition". The underlying apparent intent was to determine at what point in time was it no longer nece:sary to provide normal forced cooling without exceeding an acceptable fuel pool temperature in natural circulation conditions. The Reviewer found several errors in this calculation as follows:

(1) The calculation did not include radiant heat transfer from the pool surface, although it was considered by the author andjudged insignificant. At the higher pool temperatures addrersed (190*F,200'F and 210'F), this would appear, on initial inspection, to have been conservative. However, it was not, since this heat load added to the building would tend to raise building air temperature, thereby reducing the convective and evaporative heat transfer rates from the pool surface (the predominate heat transfer mechanisms), thus increasing the heat absorbed by the pool water, which would increase the rate of temperature rise, and thus reducing the time to reach the temperature limit.

(2) The evaporation rate, and therefore the heat transfer by that mechanism, was based on test relative hemidities taken with the building HVAC operating, removing both esooucs E-14

the water vapor and heat released inside the building from the fuel pool surface.

For the condition for whi h the analyses were required, the availability of the liVAC could not be r.e d. Without liVAC, the building temperature and relative humidity we ', rise substantially above the calculation values (the relative humidity ultimately to 1007c), thereby greatly reducing the effectiveness of the primary heat transfer mechanisms that were assumed.

(3) The evaporation rate was based on fuel pool level loss measurements taken over several days waaout makeup. However, no determination was made of any fuel poolleakage during that period. Such leakage is not uncommon and is normally detected by poolliner channel leakoff drains. This could have caused the results to be non-conservative.

In summary, the calculation set out to answer the wrong question, and in so doing developed an incorrect model(the control volume was too limited). The correct question should have been, at what point in time could the decay heat be removed from the stored spent fuel to the environment by way of the buildine by entirely passive mechanisms without exceeding the design limits on the fuel pool or any of the other associated equipment? An attendant question should have been, what would be the required water makeup rate to assure that under these conditions the water would not drop below an unacceptable level, considering that one of the primary heat transfer mechanisms would still be evaporation?

These discrepancies were discovered independently by the Reviewer; however, the author revealed during discussions that he had already discovered the overall concern of a too limited control volume, Condition Report 98 0003, dated 1/28/98 had just been written, and the licensee had been notified. The CR Committee and the author initiated a 10CFR21 evaluation proce.s regarding this error since these discoveries were also considered to have potential implications for similar analyses performed by or on behalf of other licensees on other plants.

The Reviewer considered that in the context of one of the NRC's primary concerns, quality and control of analyses, these finding merit increased management attention.

Discussions with the calculation authors and others led the Reviewer to conclude that a

! culture may exist in the Bolton Office, as in many organizations, of not taking a broad enough view of work assignments, not having a questioning attitude, not thinking outside the box;"the box" being the narrow boundaries thought to defiae one's working niche, in this last case,just performing an analysis but not defining the real questions, and in the

, first case, not consulting the design documents. In the last case, had the engineer taken l the broader view of asking more questions of the licensee, helping the licensee define the real questions, these errors might have been avoided.

Another element of thinking outside the box is realizing that in all cases, the reason a client comes to experts such as DE&S is because he or she has a problem, but in many if not most cases, he or she does not have a total, clear understanding of the problem. Part of the expert's role is to help the client define and package the problem, figure out what the realquestions are. Only by so doing can the consultant provide the true quality service omwmma E-15 l

l and products that the client, company management, the NRC, and the people performing the work all truly desire.

6.0 SUhihiARY CONCLUSIONS AND RECOMh1ENDATIONS i

6.1 Reemedine Persnectives on the NRC's Concerns conclusinns:

i (1) _ Although the DE&S individuals specifically identified by the NRC as having been primarily responsible for the analyses and reporting errors had indeed made some errors, the significance and mal intent ascribed to these errors by the NRC demand letter seemed disproportionate to the circumstances as they appeared to the Reviewer, Recommendatinns:

None. ,

6.2 Reenrtlino Project Ornanizational Responsibilities / Roles / Communications Conclosinns:

4 (1) The Self Assessment correctly concluded that a significant cause for the l discrepancies with the hiaine Yankee LOCA analyses and communications with 4 the NRC was poorly defined organizational responsibilities / roles / communications between YAEC and hiaine Yankee.

, (2) The Self Assessment was not focused on the more global underlying concern, that

the company had no written requirements that such responsibility / role /

communications definitions be clearly defined with all clients, not just hiaine Yankee.

(3) At the time of the Self Assessment response, although hiaine Yankee and YAEC ,

management discussions and plans had occurred concerning roles and responsibilities, and a responsibility matrix had been generated, little implementation of these plants had been effected.

(4) At the time of this Review,it was recognized by DE&S management that project definition was essential to assuring the quality of the company's work and the control of projects, and hence the company's continued success and profitability.

(5) The draft procedure, NED Work Control Process Outline, was a very good step in the right direction of establishing work control processes to assure the NRC and DE&S management that the quality of DE&S' work for nuclear clients and other clients will meet their high standards and expectations.

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Recommendations:

(1) That draft procedure NED Work Contro/ Process Outline be completed and approved at the earliest practicable date.

(2) That the additional points outlined in Section 5.2 of this report be incorporated into this procedure.

(3) That references in this procedure to specific clients, e.g., Vermont Yankee and North Atlantic, be removed and that it be made generic and applicable to all clients.

(4) That this procedure be made applicable to all DE&S work performed for clients by the Bolton, Massachusetts office, not just nuclear engineering work.

(5) That procedures governing the day to day work of the " Contracts Group"in the DE&S organization be generated or mvised to incorporate the applicable principles and provisions of this draft Trocedure.

6.3 Recardine Formal Company Trainine conclusions:

(1) Lack of formal employce training in the area of day to-day work recluirements, sucn as applicable codes, standards, regulations, and procedures was correctly identified by the Self Assessment Team as a major contributing cause for the discrepancies with the Maine Yankee LOCA analyses and communications with the NRC.

(2) At the time of the response memo to the Self Assessment Team's findings, very little had been done to develop and implement an effective company training program that would correct the problem.

(3) At the time of this Review, little additional work had been done to develop and implement an effective training program that would provide the desired assurance to company management and the NRC that this problem area had been corrected.

Recommendatiom:

(1) That a senior level management commitment of will and resources be made to develop and implement an in-depth company training program for all personnel providing services to nuclear utility clients with the goal of providing all such personnel with the specialized knowledge they require to perform their work in the quality manner expected by management and the NRC. (Although such a program would initially entail overhead costs, much of these costs could probably be recovered directly by offering elements of this training to clients, and indirectly by the enhanced credibility it would afford DE&S in tha nuclear services market.)

smuomms E-17

(2) That a well qualified individual be appointed as Training Manager, or similar title, and charged with the primary responsibility of developing and implementing the company's technical training program, and that this person be provided the authority and resources required to effectively carry out this responsibility.

6,4 Recarding Company Procedures

Conclusions:

(1) Weak procedures for controlling analyses and reporting of their effects on licensing commitments was correctly identified by the Self Assessment Team as a

significant contributing cause to the discrepancies found by the NRC in the YAEC LOCA analyses for Maine Yankee and for their concerns with reporting of LOCA analyses results and analyses changes.

(2) The weakness-in procedures concern had been adequately addressed by the time of the response memo of December,1996, and the specific concerns of the Self-Assessment, analyses control and NRC reporting, had been well addressed by company procedures.

(3) Company procedures re d:wed appeared to have been continuously improved in the time since the response to the Self Assessment memo of Dr. ember,1996.

(4) Some of the procedures still contain references to particular clients (vestiges of YAEC's previously close familial relationship with many of the New England plants), and virtually all of them refer to the company as YAEC. These appear inappropriate in the context YAEC's new evolving role as a technical services

, provider to more than just local clients, she acquisition by DE&S, and the acceleration of the evolution of the company's business as DE&S.

(5) Procedures do not appear to be optimally controlled, i.e., by a single company control procedure that spells out their format, authority, use, hierarchy, revision control, etc. As a result, formats are not consistent from one type to another, different types of procedures appear to address the same subject but contain different or complimentary or supplementary directions, and it is not always clear which procedural provisions are must-do requirements and which are simply recommendations. An example is several of the NED procedures which appear to address the same subject as the Engineering Manual procedures; these might be better controlled and present less potential for misunderstanding if they were combined into one procedure. Another example is the TAGS (Technical Administrative Guidelines) which contain directions concerning company policies and implementation of government regulations. The term " guidelines" appears to be misapplied for these procedures; th y should be mandatory, must-comply procedures and should be identified as soch in themselves and in a procedure-control proceoure.

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,, _ _ _ _ _ _ _ _ _ _ . 1

t i

, (6) The " Scope of Activities" section of the Engineering Manualintroduction describes its applicability to design changes and additions but does not denote I applicability to other engineering activities, such as performing engineering

calcu!ations, analyses, evaluations, etc., although it does contain specific procedures pertaining to these type activities.

(7) In NED Procedure No. 6, Safety Analysis Process, the term " safety analysis" is not defined, yet it appears to be describing calculations, for which there are two other procedures, NED No. 3 and WE 103. It is not clear from this procedure i how a safety analysis differs from a calculation.

4 Recommendations:

1 (1) A complete procedure review and revision process should be initiated with all the j company procedures to accomplish the following:

Define the philosophies, policies, management expectations, and business goals that will represent the direction of DE&S's Bolton Office, and integrate these with the specific requirements from both the YAEC and the DE&S procedures into new DE&S procedures for the Bolton Office.

Structure the new procedures to be generic with respect to clients, removing all specific references to YAEC's traditional clients, and the business practices and responsibility divisions that were unique to those

relationships.

Develop a procedure for procedures that spells out the format, authority,

! use, hierarchy, revision control, etc., for all DE&S Bohon Office procedures, j -

Assure coordination and consistency between the requirements of the various procedure types and procedures in different organizations.

l 6.5 Regardine Licensing and Oualliv Assurance Involvement i

Conclusions:

i I

, (1) As indicated by the Self Assessment, the licensing and quality assurance facets of f the YAEC organization had little involvement in the production, review, approval, d

delivery, and reporting of the Maine Yankee LOCA analyses that were the focus of the NRC's concerns, and that may have been a contributing factor to the discrepancies. Had these groups been involved as currently envisioned by the i

Reviewer, the potential for these discrepancies might have been somewhat reduced.

smt,soon s. E 19

(2) The relative roles of Maine Yankee's and YAEC's Licensing and QA Groups was not clearly defined or understood by either of the organizations, as was concluded by the Self Assessment.

(3) At the time of the response to the Self-Assessment,little had changed in YAEC to more clearly define those roles between these two companies or in the broader generic sense for YAEC.

- 1 (4) Even today it may not be fully recognized that Licensing and QA should be involved from the earliest stages of any work endeavor (the scope definition and proposal stages) to assure tnat, for d contracts, d of the licensing and QA tasks and responsibilities are identified and the assignment of these between the client and DE&S is clearly delineated and documented.

Recommendations:

(1) The definition of the roles and responsibilities of Licensing and Quality Assurance in the DE&S organization should be revised to include review, input, and client interface from the beginning of all contracts with nuclear clients, as well as other clients where licensing and QA functions may be required, in order to assure that d of the licensing and QA tasks and responsibilities are identified and the assignment of these between the client and DE&S is clearly delineated and documented.

(2) DE&S should establish and maintain a regular corporate presence in the NRC's regional and headquarters offices in order to keep them apprised of our capabilities and challenges in generic programs / analyses we are developing, to enable us to be more cognizant of generic industry concerns and NRC perspectives, and to provide assurance to the NRC that we are aggressively addressing their concern with poor communications between us and them.

6.6 Reenrdine Thinkine Outside the Box. or Takine the Broader View conclucions:

(1) Based on a sample of two recently performed calculations and analysis related conversations, the Reviewer concit.ded that the quality and control of analyses may not be at the level desired; however, the small L ple size was insufficient to base a final conclusion; further sampling would be required.

(2) A cuhure may exist in the Bohon Office of not taking the broader view, not having a questioning attitude,"not thinking outside the box"in the approach to work performance.

(3) A performance-based approach to reviews, i.e., sampling the prMuct, in this case analyses, provides the ultimate indication of whether or not corrective actions have

.mwmw u E-20

been effective. In this case, initial samples indicated that corrective actions may not have been effective.

RecommendMinns:

(1) That a broader sampling of recent analyses be performed to provide a more accurate picture of whether samples by the Reviewer were anomalies or were true indications that corrective actions had not been effective. Subsequent corrective actions should be based on that sampling.

(2) The Reviewer believes that taking the broader view, developing a questioning attitude, and thinking outside the box is a learned skill; therefore, it can be taught.

It is, therefore, recommended that a course ofinstruction be developed and provided to all Bolton Office employees on developing these skills.

(3) All analyses should be reviewed by a person or persons with very broad experience bases in the nuclear power industry. An essential element of this experience should include, but not be limited to, in-plant, hardware oriented operational or startup experience, in order to assure that analyses reflect or envelop the reality of actual hardware and actual plant conditions.

(4) Where possible, actual hardware design documents should be used in the development of anai, es models.

7.0 DVERALL CONCLUSIONS The Reviewer concluded that the Self-Assessment identified most, but not all, of the root and contributing causes for the discrepancies identified by the NRC in their December,1995 technical review. Most prominently missing from the Self Assessment was identification of the apparent culture in the Bolton Office of not taking the broad view in the approach to work and responsibilities.

The Reviewer concluded that the conective actions identified in the Self-Assessment were appropriate but that they had not all been effectively carried out. Most prominent was the inadequacy of company training to assure that employees have the knowledge of codes, standards, regulations, and procedures and other skills required to perform theirjob functions in quality manner.

The Reviewer concluded that problems may still exist in the area of quality of analyses; however, additional review should be performed in this area to determine if observations were anomalies or true indicators of a continuing problem area.

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4 1

l LIST OF PERSONS CONTACTED i

i 8Amt Title / Position / Function )

4 i Paul A.Bergeron Manager, Thermal Hydraulics & Safety Group j Kathleen E. Bocon - Training Supervisor i

James R. Chapman Director, Nuclear Engineering Department Don K. Davis Chairman & Chief Executive Officer, YAEC  :

) Cam DiNunzio Senior Quality Assurance Engineer

- Greg Hudson Project Director Dean A. Huggins Manager, Records Management Services Michael F. Kennedy Manager, Safety Assessment Group

! John M. Oddo Manager, Regulatory and Industry Affairs l Suzanne Palmer NED Engineer

! Liliane Schor Project Manager II

Stephen P. Schultz General Manager, Nuclear and Fuel Services i Michael W. Scott NED Engineer Ramu K. Sundaram NED Engineering Consultant i

I 1

i i

5 l

d 4

't

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LIST OF DOCUMENTS REVIEWED

1. USNRC Independent Safety Assessment of Maine Yankee Atomic Power Company, Executive Summary,10/7/96.

2. USNRC Letter to Mr. Charles D. Frizzle, President, Maine Yankee Atomic Power Company, 10/7/96.

3. RELAPSYA Self Assessment, Maine Yankee and Yankee Atomic,4/96.

4. Maine Yankee Letter to USNRC Chairman Shirley A. Jackson, CDF 96-192,12/10/96, Independent Safety Assessment

5. Yankee Atomic Memorandum NEDMY96 052,12/31/96, Response to RELAPSYA Self-Assessment.

6. USNRC Letter to Mr. Don K. Davis, President & Chief Executive Officer, Yankee Atomic Electric Company, and Mr. bhn F. Norris, President & Chief Executive Officer, Duke Engineering & Services Co.,12/19/97, 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 01 Report No.1-95 050).

7. Engineering Instruction WE 002, Rev 13,6/27/97, Design Document Control.

8. Engineering Instructiot WE-003, Rev 13,6/27/97, Indoctrination of Personnel.

9. Engineering Instruction WE-004, Rev 12,6/27/97, Training.

10. Engineering Instruction WE-103, Rev 17,9/19/97, Engineering Calculations and Analyses.

11. Engineerir.g Instruction WE-109, Rev 4,5/28/97, Engineering Deficiency Reports.

12. NRC letter to Mr. Michael B. Sellman, President, Maine Yankee Atomic Power Company, 12/19/97, Apparent Violations Stemming from NRC Office ofInvestigations Report Nos 1 050,196-025, and 1-96-043.

13. Technical Administrative Guideline No.1, Rev 12,8/1/97, Communication with Federal Regulvery Agencies on Behalf of Yankee Clients.

14. Technical Administrative Guideline No. 6, Rev 25,7/1/97,10CFR, Part 21 Reporting.

15. Technical Administrative Guideline No.18, Rev 1,1li24/97,10CFR50.9 Reporting Requirements.

16. Technical Administrative Guideline No. 22, Rev 2,8/19/97, Self-Assessment Program.

.mmmwm E-23

! 17. Technical Administrative Guideline No. 25. Rev 1,10/1/97, Condition Report System.

I

, 18. Nuclear Engineering Department Procedure No. 2. Rev 1,7/14/97, Microfilming and Tracking of Approved Calculations.

19. Nuclear Engineering Department Procedure No. 3 Rev 0,8/22/96 WE-103 Review Checklist.

20. Nuclear Engineering Department Procedure No. 5, Rev 0,12/11/96,10CFR50.46 Reposting.

21. Nuclear Engineering Department Procedure No. 6, Rev 2, S/7/97, Safety Analysis Process.

22. Nuclear Engineering Department Procedure No. 8. Rev 1,5/13/97. Changes to Analytical Methods.

23. Operational Quality Assurance Program, YOQAP 1 A, Rev 27,12/20/96.

24. Calculation Number MYC 2005, Rev 0,12/1/97, Spent Fuel Pool Passive Cooling.

! - 25. Calculation Number SBC 835. Rev 0,12/9/97, Spent Fuel Pool Thermal Hydraulics Analysis .

New Racks. ,

26. Service Request No. M 97 27A,9/4/97, Post Shutdown Safety Analysis.

27. - Condition Report No. 98 0003,1/28/98, Discrepancy with Maine Yankee fuel pool passive cooling calculation.

i 28. Draft NED Work Control Process Outline.

29. Execrpts from Maine Yankee Small Break LOCA Analysis,6/93,

, 30. Maine Yankee letter to Mr. William T. Russell, Director, Office of Nuclear Reactor Regulation, l USNRC, CDF 96 063,4/25/96, Submittal of Maine Yankee SBLOCA Licensing Analysis in -

4 Compliance with 10 CFR 50.46 and in Satisfaction of TM1 Action items II.K.3.30, II.K.3.31, II.K.3.5.

]

- 31. Maine Yankee letter to Mr. Frank J. Miraglia, Acting Director, Office of Nuclear Reactor Regulation, USNRC, MN 96-145, CDF-96-180,10/18/96, Re:Jonse to USNRC Request for i information (RFI)- Maine Yankee SBLOCA Analysis.

32. Yankee Atomie - Bolton Memorandum, W. J. Metevia and R. K. Sundaram to P. L. Anderson, j 1/2/90, Recommended Approach for SB LOCA II.K.3.31.

p 4

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OUALIFICATIONS OF REVIEWER Summan of Ouallnentionst The Root Cause Assessment Review relating to the (date) NRC Demand-for Information letter to Duke Engineering & Services, Inc. was performed by Mr. Donald C. Prevatte. Mr. Prevatte was well qualiDed to perform this review because of his extensive engineering and management experience in the nuclear power and other industries. This experienced spans more than 30 years in design, startup, testing, and inspection of nuclear and fossil power facilities, nuclear submarines, and jet engines, and management and direction of engineering organizations performing these activities. Approximately half of the Reviewer's experience is as an independent consultant providing engineering and management services to the power industry and government regulatory agencies, including the U.S. Nuclear Regulatory Commission and the U.S. Departinent of Energy, and practicing all of the skills necessary to operate a successful consulting business. The Reviewer has directed or participated in Afty seven performance based Team inspectionr of commercial and government nuclear facilities (see attached list). The Reviewer is a degreed mechanical engineer and a registered professional engineer. A more detailed description of the Reviewer's qualifications is provided in the attached resume.

= = = = = = E-25

- ~ ..- _ - .- __-_ - - -__- - .- -

1 I

POWERDYNE CO RPOR ATIO N 7924 WOODSBLUFF RUN. FOGFI c,VII I F PA 18051 (610) 398-9277. FAX (610) 398.Q222 Resume of  ;

DONALD C, PREVATTE l

SUMMARY

OF QUALIFICATIONS:

• More than 30 years of engineering and management experience in the design, startup, testing, and inspection of nuclear and fossil power facilities, submarines, and jet engines.

• Proven self starter in an entrepreneurial environ. ment, with a track record of running all aspects of a successful engineering consulting business.

• Highly analytical problem solving skills, with the ability to clearly define objectives, formulate logical, concise plans, maintain focus, and carry out plans.

• Outstanding organizational and time management skills. Ability to prioritize, manage multiple tasks, meet schedules / deadlines, Attentive to both " big picture" and details.

• Capable people manager, with the ability to build Teams, delegate, match skills to jobs, provide technical oversight, define goals, generate motivation, recognize contributions, and deal with non-contributors.

• Innovative, persistent, dedicated, versatile, enthusiastic, and conscientious.

• Excellent written, verbal, and interpersonal communication skiils.

EDUCATION:

Bachelor of Science Degree in Mechanical Engineering, North Carolina State University.

PROFESSIONAL EXPERIENCE:

1982 - present, Powerdyne Corporation President, Independent Consultant - Providing engineering and management consulting services to the power industry. Responsibilities include P&L, marketing, proposal preparation, contract negotiation and preparation, nuclear and fossil plant systems and equipment design, engineering analysis, technical document generation, plant inspections, project planning and management.

1987 - Present: Participated in fifty-seven performance based Team inspections of nuclear power facilities for the Nv: lear Regulatory Commission, the Department of Energy, and nuclear utilities.

Performed reviews of design, maintenance, testing, operations, and human performance, and served as Team Leader.

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l 1982 - 1992: Provided engineering and management consuhing services to Pennsylvania Power & Light l Company for the Susquehanna Steam Electric Station. Supervised the Engineering Planning and j Scheduling Group, gaining experience in Critical Path Method project planning and controls using i Project /2 computer code. Developed cost analyses and position papers to prepare client for a Public

Utilities Commission management audit and to support a rate increase request.11. the Power Uprate Project, performed system design evaluations, engineering analyses, and systems redesign to support a 5% power uprate. Evaluated the reactor building HVAC and chilled water systems, portions of the emergency service water system, the standby gas treatment system, the main condenser, and motor operated valves with regard to NRC Generic Letter 8910 requirements. Performed analyses of reactor building heat loads and building coolers. In the Design Bases Document Project, performed licensing l requirements / commitments research. During plant startup and early operations, generated analyses of

high energy pipe breaks, two phase jet impingement, and suppression pool heatup, and directed leak before break analyses. Developed plant modifications and procedure changes to mitigate MSIV 4

leakage, and generated modifications for the diesel generator stating air system and the auxiliary boiler

feed pump seal cooling system.

i 1987: Updated / redesigned the 10CFR50.59 Safety Evaluation Program for Point Beach Nuclear Plant, j 1981 - 1982, General Physics Corporation i Director, Power Engineering - Directed twenty-eight engineers in design, analysis, and field engineering l activities. Responsibilities included technical supervision, marketing, recruiting, planning and budget development. Directed Appendix J containment retests at the Brunswick Steam Electric Plant.

4 1980 - 1981, Quadrex Corporation Site Manager - Supervised nineteen Quadrex engineers assigned to PP&L's Susquehanna Steam Electric Station, Nuclear Plant Engineering Department. Developed the conceptual design of the On Site Low Level Radwaste Holding Facility and provided project management consultation.

1977 - 1980, Brown & Root, Inc.

Discipline Supervisor Electrical and Instrumentation and Controls Design Groups on the South Texas I Project, Responsible for the technical direction and production output of these disciplines. Developed a 2

multi-discipline design review program. Supervised the rewrite of the System Descriptions.

Lead Mechanical Startup Engineer - Developed startup schedule, wrote test procedures, and i performed " hands-on" testing at Parish Generating Station (fossil). Supervised boiler chemical i cleaning / steam blows.

i 1972 1977, United Engineers & Constructors, Inc.

~

- Lead Startup Engineer - Successfully marketed the preoperational testing program contract for the Salem Generating Station, Unit 2. Developed the program and supervised sixteen startt.p engineers.

Senior Startup Engineer - At the Brunswick STEAM Electric Plant, responsible for the reactor vessel hydrostatic tests, all contsment leakage rate and structural integrity tests, integrated system flushes, ,

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diesel generator acceptance tests, various systems preoperational tests, and the main condenser tests.

Coordinated systems completion for testing. Directed the first plant outage after full power operation.

1971 -1972, Newport News Shipbuilding & Dry Dock Company Mechanical Test Engineer Wrote procedures and performed post overhaul testing of submarine reactor plant systems. Attended Shift Test Engineer School.

l 1969 - 1971, Pratt & Whitney Aircraft Corporation Design Engineer - Designed jet engine hardware for F 15/F-16 fighter aircraft. Performed finite element fatigue analyses using state of the art computer codes and hardware.

REGISTRATION: Registered Professional Engineer, Texas.

MILITARY SERVICE: US Navy, Lieutenant, Navigator, Main Propulsion Assistant, Legal Officer.

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TEAM INSPECTIONS PARTICIPATION BY D.C. PREVATTE Bad fJjgg S3 stems Imnection Tvne/ Area Turkey Point NRC Intale Cooling Water SWOP 1 Oserview/ Design Sequoyah NRC Emergency Raw Cooling Water SWOPl/ Design SALP Followup (2 inspections) Various Inspection / Design Hatch NRC Plant /RHR Service Water SWOP!/ Design Farley NRC Emergency Service Water SWOP!/ Design Arkansas Nuclear i NRC Emergency Service Water Diagnostic Evaluation / Design Zion NRC Emergency Service Water Diagnostic Evaluation / Design (3 inspections) NRC Service Water /CCW hinterial Condition Evaluation NRC Service Water SWOP 1 Followup / Design Watts Bar NRC Essential Raw Cooling Water Integrated Design inspection (2 inspections) NRC Component Cooling Water Vertical Slice Review Evaluation San Onofre NRC Saltwater / Component Cooling Integrated Testing / Design Oconee NRC Emergency Service Water SWOP!/ Design /hf aintenancerresting _

Surry NRC Emergency Service Water SWOPl/ Design / Maintenance / Testing Palo Verde NRC Essential Service Water Systematic Test Performance (2 inspections) NRC Emergency Power Diagnostic Evaluation / Design Catawba NRC Nuclear Service Water SWOPl/ Design 41aintenance/ Testing Crptal River NRC Nuclear Service Water SWOPIOverview/ Followup / Design hiaine Yankee NRC Component Cooling Water SSFI/ Design WNP 2 NRC Emerfency Power SSFI/ Design (2 inspections) NRC Various SSOhtt/ Design Cooper NRC Emergency Power SSFl/ Design (2 inspections) NRC SLC/CRD/RPS SSFI/DesiFn Robinson 2 NRC Emergency Power SSFI/ Design Fitzpatrick NRC Emergency Power SSFI/ Design 4taintenance Rancho Seco NRC Emergency Power SSFVDesign (2 inspections) NRC Auxiliary Feedwater SSFl/ Design Point Beach Utility Emergency Power SSF1/ Design Diable Canpn Utility - Auxiliary Feedwater SSFI/ Design hicGuire NRC Auxiliary Feedwater Diagpostic Evaluation / Design hionticello -- NRC- Core Spray SSFUDesign immmun E 29 i

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Dant gligat Sistems inanection Tyne/Atta 9 Mile Point i NRC Core Spray,HPCI SSFI/ Design Trojan Utility Control Room Ventilation SSFI/ Maintenance Grand Gulf Utility Standby Liquid Control SSF1/ Design J (2 inspections) NRC Standby Service Water Engineering Teaminspection j

?

l Duane Arnold Utility HPCI SSFI/ Maintenance Shoreham Uti ity Evaluation of SSFI/SSOMI - SSFI/ Design BrownsIbrry Utility RHR SSFI/ Maintenance

(2 inspections) NRC Various Modifications / Design WolfCreek NRC Various SSOMI/ Design (2 inspections) NRC ESW, AIM'and DieselGen. Engineering & Technical Support 4

1

, Dresden NRC Various SSOMI FollowupInspection/ Design (2 inspections) NRC Various Engineering & Tect nical Support insp.

North Anna NRC Various SSOMI/ Design 1 Hope Creek Utility Liquid and Solid Radwaste System Functional Evaluation Salem . Utility Various Outage Modification inspection  ;

(4 inspections) Utility Radwaste Systems SSFI/ Design Utility Fuel Handling Building HVAC SSFI/ Design Response Team Utility Control Room HVAC Design Review Inspection Fermi 11 NRC HPCI, Inst Air SSF1/ Design Savannah River DOE Various Human Performance Inspection D. C. Cook NRC CentrifugalCharging SOPl/ Design Kewaunee NRC AITV. AuuliaryCoolant SOP!/ Design Ft.Calhoun NRC Al%', CCW, Raw Water Engineering & Technical Support / Design l

Millstone 3 - NRC Charging Sptem SSF1/ Design i

Paducah Gaseous USEC Autoclave System SSFlfTeamleader/ Design Diffusion Plant

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9 I

APPENDIX y ENGINEERING PROCESS ASSESSMENT I

MMD

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