ML20216D768

Rept to Duke Engineering & Services,Inc,On Allegations of Willfulness Related to Us NRC 971219 Demand for Info

ML20216D768
Person / Time
Site:	Maine Yankee
Issue date:	02/25/1998
From:	Mcgarry J, Wetterhahn M WINSTON & STRAWN
To:
Shared Package
ML20216C072	List: ML20216C069 ML20216D768
References
NUDOCS 9803170273
Download: ML20216D768 (34)

Text

__

1 O

REPORT TO DUKE ENGINEERING & SERVICES, INC.

ON ALLEGATIONS OF WILLFULNESS RELATED TO THE U.S. NUCLEAR REGULATORY COMMISSION DECEMBER 19,1997 O DEuANo rOR 1NrORuATION 1

Winston & Strawn J. Michael McGarry, III O' g31 73 980303 5 Mark J. Wetterhahn

• NISC PDR ,, February 25,1998 WPHUDSONLMH '!GS

._ _9

O. _N_

TABIE OF CONTENTS EQEL

1. EXEC! iTIVE S UMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D- I II. INTR ODUCTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-3 III. INVESTIGATIVE AND DELIBERATIVE PROCESS . . . . . . . . . . . . . . . . . . . . . . . . . . D-4 IV. LEGAL ANALYSIS OF WILLFULNESS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-5 V. FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . .................................D-7 VI. RESPONSE TO SECTION III OF THE DEMAND . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-10 A. Deliberateness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D- 10 ,

B. Response to Sections III.A and B . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-I I C. Response to Section III.C . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-20 D. Response to Section III.D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-23 VII. CONCLU S I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-31 w m eso w i ms D-ii j l

f l

l lO lV REPORT TO DUKE ENGINEERING & SERVICES, INC.

l ON ALLEGATIONS OF WILLFULNESS RELATED TO THE U.S. REGULATORY COMMISSION DECEMBER 19.1997 DEMAND FOR INFORMATION L EXECUTIVE

SUMMARY

l Winston & Strawn was requested to determine whether certain allegedly inadequate engmeermg I analyses and materially incomplete and inaccurate information relating to the SBLOCA analyses used by Maine Yankee Atomic Power Company to demonstrate compliance with 10CFR50.46 was the result of willfulness, either deliberateness or careless disregard, on the part of personnel of Yankee Atomic Electric Company (" Yankee Atomic"), many of whom recently became part of Duke Engineering and Services, Inc. ("DE&S"). To respond to the request, Winston &

Straw n reviewed the NRC's December 19,1997 Demand for Information (" Demand"),  ;

reviewed relevant documents and interviewed a number of witnesses. These actions were taken in conjunction with an expert Technical Review Team whose function was to determine the  ;

technical adequacy of the SBLOCA analyses in the areas raised by the NRC. l l

Willfulness, as used by the NRC, embraces a spectrum of actions ranging from deliberate intent *I to violate or falsify to and including careless disregard for NRC requirements. Deliberate l

,I_)

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 requirement or with disregard or utter unconcern of the consequences of whether there was i compliance. The existence of a reasonable justification for an action would defeat a charge of willfulness despite the fact that the action was ultimately found to violate NRC requirements. l We found no actions on the part of any individuals associated with the specific issues contained in the Demand that would involve deliberateness. We found all individuals m h open, honest and communicative to us. We found no specific intent to violate any NRC regulation Our evaluation, therefore, focused on whether there existed careless disregard for Commiss:on 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 Yankee Atomic for the Maine Yankee facility, the requirements of 10CFR50.46 were not met. The 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. Yankee Atomic had taken the position that the identification of the limiting break, combined with a sufficient i understanding of the physical phenomena which were occurring over the entire small break Q region, provided compliance with 10CFR50.46. The Technical Review Team recognized that V NRC expectations regarding 10CFR50.46 were not completely documented, but rather had been s m oso m nus D-1

I ,

I l

l(A),, communicated to the LOCA community through its interactions with the NRC. Utilizing the appropriate legal standad, and bearing in mind that the Yankee Atomic LOCA Group was

! isolated and not part of '.he LOCA community, it was determined that the applicable rcgulation could be read as it was by Yankee Atomic. Actions to implement the LOCA Group's j interpretation of the regulad~: did not evidence reckless indifference and, tle no careless disregard was found.  !

With regard to the second issue, the NRC asserted that Yankee Atomic caused a violation of 10CFR50.9 in that Yankee Atomic 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 wbn judged using the perspective ofits intended audience, it, one knowledgeable in the field, such as an NRC reviewer. Under the circumstances, we 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 l I

technical basis for one element of the SBLOCA analysis, the loss coefficient for the split downcomer 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 vith industry experience. The

(]

V Technical Review Team determined that a deficiency in the quay assurance of a confirmatory calculation existed, but there was a reasonable explanation as to why it occurred, and that the evaluation undertaken by Yankee Atomic was appropriate. No inadequate analysis existed and i the issue of deliberateness or careless disregard did not arise. l I

With regard to the fourth and last issue, the NRC asserted that Yankee Atomic memoranda used an unacceptable Best Estimate model rather than an approved Evaluation Modelin evaluating the effect of a decrease in steam generator pressure on the peak clad temperature in j 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 other 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 of the memoranda mischaracterized the Best Estimate model as the approved 10CFR50.46 Evaluation Model for the facility.

With regard to the use of the memoranda, we determined that it was understandable 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 l

design basis issues. However, it is also understandable that a Maine Yankee employee not expert in LOCA analyses would use these memoranda in performing th assigned 50.59 analyses. In evaluating whether careless disregard had occurred, we noted that the issue of whether decreased steam generator pressure impacted the design basis did not mature untillate 1992

!q g (ic, sor : months after receipt of the last memorandum), as evidenced by the Maine Yankee l

l m aoso m m s D-2

r h NRC Resident Inspector's suggestion that a 10CFR50.59 analysis should be performed. We found that the appreciation of the necessity for evaluating degoded conditions pursuant to 10CFR50.59 was in a state of transition at the time, with the NRC publishing Generic Letter 91-18 in late 1991. We 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 Yankee Atomic personnel did not meet the test for careless disregard of the regulations in that we 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, we determined that, ivhile in certain instances there may have been inadequate analysis associated with the SBLOCA analysis, there was neither deliberateness nor careless disregard resulting from the deficiencies discussed in the Demand. We found no willfulness on l the part of the two individuals or any other Yankee Atomic personnel. We therefore believe that j the Manager and Lead Engineer are capable of conducting their activities in the future in conformance. with NRC requirements. We conclude that there was nothing developed as a result of our investigation on the conduct of Yankee Atomic and/or DE&S personnel that would j prevent the improvements that we understand are being made from being successful and l resulting in DE&S' activities being conducted in full compliance with NRC requirements and (O expectations. l II. INTRODUCTION On December 19,1997, the U.S. Nuclear Regulatory Commission ("NRC") sent Duke Engineering and Services, Inc. ("DE&S") and Yankee Atomic Electric C< mpany ("YAEC" or

" Yankee Atomic") a Demand for Information (" Demand") conceming the provision of

" inadequate engineering analyses and materially incomplete and inaccurate information to an NRC licensee," namely Maine Yankee Atomic Power Company (" Maine Yankee" or "MYAPCo"). The subject matter of the Demand was a Loss-of-Coolant Accident ("LOCA")

analysis used by MYAPCo to demonstrate compliance with 10CFR50.46 and Appendix K to 10CFR50. Specifically:

YAEC prepared the small-break LOCA analysis which was utilized by MYAPCO during its operating Cycle 14 and to suppcrt its reload analyses for operating Cycles 14 and 15. See YAEC-1868. " Maine Yankee Small Break LOCA Analysis"(RELAP5YA SBLOCA analysis). YAEC also prepared the large-break LOCA analysis utilized by MYAPCO for Cycles 14 and 15. See YAEC-1160. " Application of Yankee WREM-BASED Generic PWR ECCS Evaluation Model to Maine Yankee" (WREM LBLOCA analysis).'

(q1

' Sec Letter of December 19,1997, transmitting Demand at 1.

muo nos D-3

l I i h

lU The Demand required DE&S to submit responses to the NRC within 30 days, which date was l extended pursuant to DE&S' request to February 27,1998. Among the matters raised in the l

Demand was an NRC request for "[a]n explanation why the NRC should not consider the

inadequate analyses, which apparently caused MYAPCo to be in vioh tion of NRC requirements, l to be the resuh of willfulness, either deliberateness or careless ohregard, on the part of YAEC .

and/or DE&S personnel."2 DE&S had purchased part of Yankee Atomic in late 1997, including )

the LOCA Group which was the focus of the Demand. l l

Winston & Strawn was requested by DE&S to address the willfulness aspect of the NRC's Demand. In response to this request, two attorneys from Winston & Strawn, J. Michael McGarry and Mark J. Wetterhahn (the " Willfulness Review Team"), both with a background in nuclear energy law, worked closely with the Technical Review Team chartered by DE&S to ,

report on the technical issues contained in the Demand. The Technical Review Team consisted of experts in the field of LOCA analysis and the other subjects contained in the Demand.

The instant report describes the investigative process, the legal standards utilized, the determination as to willfulness and the bases for our findings for each of the four specific issues contained in Section III.A-D of the Demand, and the Willfulness Review Team's conclusion as to the ultimate issue contained in Section V.B of the Demand.

i III. INVESTIGATIVE AND DELIBERATIVE PROCESS is /

" At the outset, the Willfulness Review Team reviewed the Demand and associated documents to better understand the NRC's perspective, to identify the documents the NRC relied upon and to define the factual statements attributed to Yankee Atomic personnel. Additional relevant documents were also reviewed on a selective basis.

Because a determination of willfulness involves a mixed question of fact and law,it was necessary p rely upon determinations of fact made by the Technical Review Team. For example, the prem> INRC q-ntion V.B is that there was one or more " inadequate analyses" which caused a violation of NRC requirements which, in turn, resulted from willfulness on the part of YAEC and/or DE&S personnel. If the analyses discussed in Section III.A-D were determined in whole or in part to be technically adequate as judged by the appropiiate standards, then the need for a willfulness determination as to that element is obviated. However, even should a oetermination be made that any of the analyses discussed in Section III of the Demand was in one or more ways inadequate, the deliberations and recommendations of the expert panel are necessary to determine the nature and extent of that deficiency from a technical standpoint and in relation to the NRC regulations.

l As a result, the Willfulness Review Team closely coordinated its efforts with the Technical Review Team. In conjunction with the Technical Review Team, a number of questions were developed 5 order to assure that the Demand was completely analyzed and each of the issues, n

N 2

Demand at 18.

mmow.ms D-4 l

l

prenuses and facts contained therein was questioned. The Willfulness Review Team attended the pre-interview meeting of the Technical Review Team. We were also present at all the interviews conducted by the Technical Review Team, including those of the Manager and Lead Engineer of the LOCA Group. Generally, during those interviews, the Willfulness Review Team elicited

!_ background information and conducted some follow-up questioning directed to matters contained in this report; the bulk of the questioning was by the members of the Technical Review Team.

Interviews and ensuing discussions took place over the course of four days. During the L meetings, the Willfulness Review Team observed and participated in the deliberations of the l Technical Review Team. Utilizing the questions previously generated as a guide, we challenged the Technical Review Tm's conclusions to assure that they were understandable, defensible, and in a form that could be readily understood by an informed person and thus could be utilized as input to our deliberations.'

IV. LEGAL ANALYSIS OF WILLFULNESS In analyzing the question of willfulness, we used NRC guV.cace and precedents and judicial analyses in formulating a series of questions to guide our . e.tiberations. The term " willfulness,"

as used by the NRC, " embraces a spectrum of violations ranging from deliberate intent to violate or falsify to and including careless disregard for requirements."' Thus, at one end of the O " willfulness" spectrum are violations involving a " deliberate intent to violate or falsify."5 At the same end of the willful spectrum is 10CFR50.5 (1997), the NRC deliberate misconduct rule, which prohibits deliberate violations of NRC regulations and deliberate submittals ofinaccurate materialinformation to the NRC.' Fe poses of the deliberate misconduct rule, " deliberate miscenduct" is defined as "an intenm act or omission that the person hows . . . [w]ould cause a licensee to be in violation of aro s ule, regulation," or other NRC requirement.7 Thus, for We have reviewed the report of the Technical Review Team entitled " Yankee Atomic SBLOCA Technical Review Report (' Technical Review Report"). While certain portions of that report are cited herein, we have based our findings on the Technical Review Team's report in its entirety. A number of factual statements in this report are not contained in the Technical Review Report. These statements were based upon interviews conducted and information received.

NRC Enforcement Policy IV (C), 60 Fed. Reg. 34,381, 34,385 (1995).

s. g 8

10CFR50.5(a)(1)-(2). Section 50.5 was not cited in the Demand; however, reference to l the definitional language of Section 50.5 is instructive.

j

! 7 10CFR50.5(c)(1).

wmmoems D-5

i.

p a violation of an NRC regulation to be " deliberate," there must be both an intent to take a certain action and knowing intent to violate an applicable requirement.

l Examples of deliberate behavior, as previously cited by the NRC, include: a security supervisor reteaching exam materials after several guards answered a question incorrectly, allowing the guards to change their answers, then submitting the guards' revised answers for grading;8 making false statements on an application for access authorization;' and a contractor providing false information about the capability ofits product. These cases are examples of actions by individuals with specific knowledge that their actions violated NRC regulations. {

Deliberate behavior is, however, not the only standard applied to prove willfulness. At the other

~

end of the spectrum of" willful" violations are those that result from careless disregard.

" Careless disregard" is not defined in NRC regulations; however, a clear pronouncement by the -

NRC Staff regarding " careless disregard" was provided in the context of proposing the deliberate misconduct rule:

Careless disregard has been described as a showing of disregard fo.r a governing statute or an indifference to its requirements . . . . A finding of careless disregard indicates that the person acted with reckless indifference to the requirement, or with disregard (or utter unconcern) of the consequences or whether there was compliance. This recklessness involves, at a minimum, an unconcern as to whether a requirement was or will be violated, or a situation in which an individual blinds himself or herself to the realities of whether a violation has occurred or will occur."

8 Penncvivania Power & T .ight Co., EA 94-212 (May 9,1995) (issuing a Severity Level III violation to the licensee for violating 10CFR50.9(a)); Darryl R. Zdanavage, IA-95-011 -

(May 9,1995) (issuing a Severity LevelIII violation to the exam proctor for violating 10CFR50.5(a)(2)).

E;, Juan Guzman, IA 96-018 ( Apr.19,1996); Gerald O. Eckard, IA 94-016 (Aug. 3, )

'994). .

'* Thermal Science. Inc.. EA 95-009 (Oct.1,1996) (imposing a $900,000 fine on a l contractor for nine instances of" deliberately submitting information to the NRC that the contractor knew to be inaccurate or incomplete in some respect material to the NRC" regarding the performance capabilities ofits Thermo-lag fire barrier product).

55 Fed. Reg. 12,374,12,375 (Apr. 3,1990) (citations omitted). Sac alsa 52 Fed. Reg.

49,362,49,365 (Dec. 31,1987) ('The concept of ' careless disregard' goes beyond simple (p Q negligence . . . [it] connotes a reckless disregard or callous . . . indifference toward one's responsibilities or the consequences of one's actions.") (citations omitted). In sum," careless disregard" does not require a conscious decision to violate a known

! requirement; " careless disregard" requires that a violator act with indifference (beyond .

i wmosammms D-6

L  !

l r

l The NRC's description of " careless disregard" is similar to language that has been accepted by l the Supreme Court.i2 The Supreme Court has also held that if an individual makes a reasonable.

l- good faith effort to determine what constitutes a violation of the law, then he cannot be acting I with careless, or reckless, disregard." Accordingly, a willful violatic w..mnot esult if a licensee

{

has considered NRC requirements and reached the erroneous conclusion that its actions will not violate relevant statutory or regulatory provisions." The existence of a licensee's reasonable justification for its actions should, therefore, defeat a charge of willfulness, despite the fact that the licensee undertook an action that was ultimately found to violate NRC requirements.

V. FINDINGS The findings on the individualissues outlimd in Section III of the Demand and the actions of the individuals identified therein must be viewed in the context of the history of development of the SBLOCA code, RELAP5YA, by Yankee Atomic for Maine Yankee. In addition, key decisions l made by individuals other than the persons named in the Demand significantly affected the actions of the named individuals, the alternatives available to them, and their decision-nuking  ;

process. Perhaps equally important is an understanding of the internal reporting relationships, responsibilities, and philosophy of YAEC and its relationship with Maine Yankee, mere negligence) to the applicable requirement.

Although the federal courts have not defined " careless disregard," they have applied this standard to cases where an individual has shown " disregard for the governing statute and indifference to its requirements." Trans World Airlines. Inc. v. Thurston, l 469 U.S. I11,127 (1985)(citing United States v. Ill. Cent. R. Co.,303 U.S. 239,243 '

(1938)). See also United States v. Ill. Cent. R. Co.,303 U.S. 239,243 (1938)

(describing willfulness as conduct that "either intentionally disregards the statute or is plainly indifferent to its requirements") (quoting St. Louis & S.F. R. Co. v. U.S.,169 F. 69, 71 (8th Cir.1909)). Thus, the NRC and federal court definitions are generally.

consistent.

Tranc World Airlines. Inc. 469 U.S. at 129. l 1

Ses In re Wrnngler i nh.. LBP-89-39,30 NRC 746, 780.(1989), Icy'd on nihcr groundc.

33 NRC 305 (1991) (A licensee's " serious albeit defective" efforts to comply with NRC j regulations were sufficient to defeat a conclusion of careless disregard of NRC requirements or willfulintent to violate NRC regulations.); In re Reich Geo-Physical. Inc.,

ALJ-85-1,22 NRC 941,957-58,962 (1985) (When a licensee has a reasonable basis for believing it is not violating NRC requirements, it is not guilty of carelens disregard or, concomitantly, a willful violation.). SCn also In re Georgia Power Co., DD-93-8,37 JRC 314,332 (1993), vacated on nihcr grounds,38 NRC 1 (1993) (holding violation was not O willful where licensee had employed a reasoned and deliberative, albeit incorrect, process regarding compliance).

L l wmesomnxs D-7 l-l c

l l~

l l

In response to a decision by YAEC management to broaden their analytical capabilities to l support their customers, a decision was made to initiate the RELAP5YA/ MODI development project in 1980 to analyze loss-of-coolant accidents in accordance with the requirements of

'10CFR50.46 and Appendix K to 10CFR50." The RELAP Code was originally developed by l INEL. In 1983, YAEC submitted a request to the NRC for approval of RELAP5YA as the code to meet the requirements ofII.K.3.30 of NUREG-0737 related to post-TMI issues.

l While the NRC review was in progress, Yankee Atomic applied the RELAP5YA code to the resolution of an issue for Maine Yankee raised by the NRC related to reactor coolant pump trips." During this time frame, it was recognized that there were some inherent difficulties in the RELAP5YA code which would require significant work to overcome. Certain individuals within the LOCA Group, including the Manager and Lead Engineer, although the former did not hold a management position at the time, requested that YAEC management permit them to develop a new code without the same inherent limitations as the code then under review by the NRC.

Management decided to proceed with the use of RELAP5YA. While the full details and the wisdom of such a management decision are well beyond this review, it indicates the mindset of the LOCA Group that they were to use RELAP5YA and overcome its difficulties in meeting the technical requirements of the plants for which they provided a SBLOCA Evaluation Model.

Also as determined by the Technical Review Team, the process by which approval of the RELAP5YA code was sought from the NRC in 1983, and which extended over a period of six years, was flawed in that in parallel with the approval of the Evaluation Model, plant-specific

applications were not performed. This resulted in the approval of a generic code which was not l significantly challenged and whose application to a specific reactor, g, Maine Yankee, was unproven, significantly lengthening the time necessary to obtain a final specific analysis which could be implemented for Maine Yankee.

In our view, these factors were an influence on the LOCA Group to make RELAP5YA work for Maine Yankee and increased the risk to both Maine Yankee and YAEC should that code not be demonstrated to comply with the requirements of 10CFR50.46. These critical decisions were made prior to the individual designated as Manager in the Demand being the incumbent in the  ;

position; he and the LOCA Group were saddled with these handicaps. i

" Technical Review Report a: III.A.  ;

i M. at III.B.

l M.

8' The Technical Review Team believes that Yankee Atomic should have realized that this step in the approval process was the industry standard. Technical Review Report at III.D. )

O We note, however, that a plant-specific application was not required by the NRC to be submitted in conjunction with the approval of the generic code.

i

.mmo ms D-8 L

It was also clear that the YAEC LOCA Group was working in relative isolation from the l 'LOCA community which was dominated by the nuclear steam supply system vendors and those companies selling core reloads. The Yankee Atomic LOCA Group was relatively small and, although all individuals interviewed appeared to be knowledgeable about small break LOCA computer models and phenomena and are technically qualified,2 they were challenged by resource limitations.2: Yankee Atomic management did not assure sufficient experience and depth in the staff of the RELAP5YA project team, in particular at the beginning of the project in the early 1980's.22 The Technical Review Team expressed its vi w that "a LOCA development program requires a very significant resource commitment which includes a sustained critical mass of expertise and continuity of key personnel."2s After the completion of the NRC's generic approval of RELAP5YA in 1989, a situation arose which led to some of the later deficiencies.2' During the generic RELAP5YA approval proce.;s, the LOCA Group was able to communicate directly with the NRC in furtherance of the licensing of that code; thereafter, application of the generic code to Maine Yankee became a Maine Yankee licensing matter which led Yankee Atomic personnel, including those in the LOCA Group, to believe that they could have no direct interface with the NRC.25 The structure preordained by Maine Yankee and Yankee Atomic management also appeared to inhibit the free flow of communications between the Yankee Atomic LOCA Group and their ultimate customers at Maine Yankee, ahhough Maine Yankee personnel appear to have been O> briefed periodically on the difficulties that the Yankee Atomic LOCA Group were encountering in implementing RELAP5YA for Maine Yankee.2' Formal communications from the LOCA Group were funneled to a Maine Yankee project group at Yankee Atomic and from :.' ere to a single point of contact at Maine Yankee to be disseminated to the appropriate Maine Yankee Tecimical Review Report at III.D.I. The Technical Review Team also noted that the culture at Yankee Atomic, which appeared to prefer self-sufficiency and independence in providing engineering services, contributed to the LOCA licensing implications. M. at III.D.2.

2

. M. at III.D.4.

2' M. at III.D.2.

22 g,

23 g,

24 The Technical Review Team found that communications during the general approval process appear to be adequate. Technical Review Report at III.D.6.

25 M. at III.D.6.

26 M. at IV.B.

m woso o m os D-9

V(D organization. Submittals intended for the NRC were made by a Maine Yankee licensing group which seemingly had little contact with the Yankee Atomic LOCA Group. While there may have been legitimate reasons which resulted in such lines of comrnunications (which reasons we did not explore and did not consider to be relevant to our inquiry), the fact is that communications were less than ideal and, in our view, contributed to the difnculty in resolving the issues discussed in the Demand.

Discussions with Yankee Atomic personnel revealed to us a rather uneven understanding of the NRC requirements related to design basis, the requirements of 10CFR50.59, and of the term

" safety-related." Inasmuch as these events occurred in the late 1980s and up to 1993, we attempted to judge this lack of appreciation and the resulting actions by YAEC employees by contemporaneous standards. Clearly, the emphasis with regard to meeting NRC standards was to assure adequate safety as opposed to today's focus on compliance. It was also clear that the employees of the LOCA Group considered themselves a technical resource and not part of the direct support of Maine Yankee operations. The LOCA Group was largely isolated and had minimal experience in operations: their entire emphasis was on technical development of information that could be used in the implementation of 10CFR50.46 and Appendix K to 10CFR50.

With this background, we turn to the issues posited in the Demand for Information,Section III.A-D.

G VL RESPONSE TO SECTION III OF THE DEMAND A. Deliberateness It is our conclusion, as bolstered by the opinion of the Technical Review Team, that l there was no deliberateness revealed in our investigation on the part of any YAEC or

{

DE&S employee or, for that matter, any other person within the scope of the issues covered by the Demand.27 We found allindividuals that we interviewed to be open, honest, and cooperative.2 If anything, they appeared to be overwhelmed by the situation before them, and they stilllacked an appreciation as to how matters that they perceived to be appropriate technical decisions -- or, at the most, technical differences among experts in their specialty -- had turned into an issue of willfulness in which their veracity was being challenged.

We did not encounter a situation where individuals tried to shift the blame to others; all were willing to defend what they believed to be valid technical decision-making within the scop: of the NRC regulations as they understood and interpreted them. Thus, we have determined that to the extent that there were inadequate analyses as discussed  !

p 27 Id. at III.D.4.

V 2' The Technical Review Team made a similar finding. Technical Review Report at III.D.4.

wmmomms D-10 l

j i

below and in the report of the Technical Review Team. which could have resulted in Maine Yankee being in violation of NRC requirements, they were not the result of deliberateness 'n the part of YAEC or DE&S personnel. We have no indication of any l ir lividuals proceeding despite knowledge that his or her actions were wrong. Thus we conclude, using the standards set forth above, that no deliberateness was involved.

B. haanse to Sections IIIA ==d B i

We turn now to a discussion of the four issues contained in Section III of the Demand to l determine whether there was careless disregard on the part of YAEC or DE&S -

l personnel which could have caused Maine Yankee to be in violation of NRC

. requirements. In doing so, we utilize the Technical Review Team's findings a's to whether there existed inadequate analyses in the specific areas delineated in Section III.A-D.

A. During Cycle 14 operations, and in the Cycle 14 and Cycle 15 reload analyse.c, Maine Yankee used apparently unacceptable Evaluation Models which could not calculate or reliably calculate ECCS performance.

B. MYAPCo maintained information and submitted to the NRC Core Performance

Analysis Reports, in support of Cycle 14 and Cycle 15 reload applications, which apparently were not complete and accurate in all material aspects.

We treat subsections A and B of Section III of the Demand together since they are largely based upon common factual issues. Our inquiry was bounded by the small break analyses,iA, the small break spectrum up to and including a break size of approximately 0.6 ft.2. The NRC contends that in order for LOCA codes to be acceptable, they must not only be capable of calculating any point on the break spectrum but must be capable of producing reliable calculations. The NRC asserts that RELAP5YA, which is the code for evaluating the portion of the break spectrum up to 0.6 ft.2, was not capable of calculating break sizes of and greater than 0.35 ft.2. The NRC reasons that ifit was not possible to analyze any point on the break spectrum, it was not possible to confirm that the limiting break within the spectram had been identified. The NRC asserts that both the Manager and Lead Engineer were significant contributors to the preparation of the RELAP5YA code and should have recognized it did not meet the requirements of the regulations as interpreted by the NRC.

For their part, the Manager and Lead Engineer assert that they believed their implementation of the RELAP5YA code, as it applied to Maine Yankee, to be in compliance with the Commission's regulations and, in particular,10CFR50.46 and i- Appendix K.2' They reached this conclusion based upon the successful running of the RELAP5YA code at a number of break sizes smaller than 0.35 ft.2 and at 0.35 ft.2 until O

Technical Review Report at IV.A.

masomnus D-11 4

the code terminated." They asserted that they had identified the limiting break in the small break spectrum as being at 0.15 ft.2 and had done sufficient calculations and had a sufficient understanding of the physical phenomena which were occurring in the small break region, k, up to 0.6 ft.2, such that they could tell with reasonable assurance that a limiting break would not occur at 0.35 ft.2 or greater. They stated that the RELAP5YA code had not " failed" in the way alleged by the NRC, but it had terminated and had not been run further either at 0.35 ft.' or larger because it was not necessary.32 The Manager and Lead Engineer were of the view that the instabilities and oscillations which had been increasing as 0.35 ft.2 was approached, were conservatively accounted for in determining the Peak Clad Temperature (" PCT")" and would not have affected the ability of the code to reliably determine that the limiting break size for the small break spectrum had already been identified at smaller break sizes." They stated that 'the small break size did not contribute the limiting PCT which was in the large break q

( spectrum covered by another code."

Our evaluation starts with 50.46(a)(1)(I). In pertinent part, that section states: )

ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of .

postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated . . . . [T]he evaluation model must include sufficient supporting justification to show that the analytical technique realistically describes the behavior of the reactor system during a loss-of-coolant accident.

" 'M.

l' 33 M. .

32 l

YAEC-1868, Maine Yankee Small Break LOCA Analysis, dated June 1993 at 22, utilizes 1 the term " failure" to describe termination of a run due to numerical convergence errors even with very small time steps. The Manager and Lead Engineer state that the code stopped because an internal instruction related to minimum time steps caused it to do so.

They state the term " failure" has no pejorative connotation.

32 We use the term PCT to mean the highest temperature of the fuel cladding reached for a i particular break size and also to mean the highest fuel clad temperature for any assumed break size in the spectrum, usually modified by the word " limiting "

" Technical Review Report at IV.A.

" - M.

mom,.nm D-12 l_

)

I o

The Technical Review Team spent considerable time in discussing the proper interpretation of this portion of Section 50.46. Based upon its members' experience in l implementing the cited section, they viewed that the industry standard practice is that an Appendix K small break LOCA evaluation model be capable of analyzing any break size within the plant's small break LOCA licensing basis.36 That is not to say that the calculation model had to have been run for each point or that there could not be a point within the spectrum that the model code could not be capable of generating an analytical l

value, but, simply, that the capability should exist.37 An example discussed during the f

Technical Review Team's deliberations was that if a code could reliably calculate peak l clad temperatures at 0.3 ft.2 and 0.4 ft.2, the fact that the code was somehow incapable

! of performing a calculation at 0.35 ft.2 would not be a deficiency if the limiting break had

! been identified and it was understood why the code was incapable of calculating a value at that intermediate point. Based upon discussions with the individuals involved, it was the Technical Review Team's judgment that the RELAP5YA Evaluation Model has not demonstrated the capability to analyze the complete range of the historical Maine Yankee SBLOCA break spectrum.38 l

The Technical Review Team did state that Yankee Atomic's identification of the

! limiting PCT within the range of break sizes evaluated and conclusion that this was the limiting PCT for the small break spectrum based on the decreasing trend of PCT l _

for smaller and larger break sizes, was consistent with Yankee Atomic's l interpretation of 10CFR50.46 and Appendix K to 10CFR50." The Technical Review

'L Team noted that in response to its questions, the LOCA Group provided additional verbaljustification consisting of an explanation of SBLOCA phenomena which supported its conclusion." The Technical Review Team found that additional analysis results from other codes tend to support the Yankee Atomic position, although some of the trends are not consistent with that position.d' l

L M.

1

" M. See ahn Technical Review Report at III.D.I.

l 38 M.

M. at IV.A. We would note that a similar observation has been made in the Root Cause Assessment review performed by Powerdyne Corporation for DE&S in Section 5.1, l wherein it stated that 10CFR50.46 does not specifically require that the code analyses '

be overlapping for the complete spectrum of break sizes.

M.

di M.

mmosomuus D-13 l

Starting with the hypothesis that the course taken by the LOCA Group deviated from the industry standard practice to calculate PCTs over the entire small break spectrum, we examined the regulation to determine whether Yankee Atomic's interpretation,it, the focus of the regulation was the proper identification of the limiting break size and l limiting PCT, evidenced reckless indifference or utter unconcern of whether there was compliance with the regulation.'2 We believe that one reading of 10CFR50.46 could lead to the LOCA Group's interpretation. Section 50.46 requires that "ECCS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of noctulated. loss-of-coolant accidents of different sizcs, locations and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated." (Emphasis supplied.) The focus of the LOCA Group was in providing sufficient assurance that the limiting break had been identified; this is clearly the intent of the cited regulation. By its terms, the regulation requires that the PCT for "a number" of sizes be calculated. With the luniting PCT identified, the LOCA Group believed that running the remainder of the break. spectrum was unnecessary since they had confidence they had determined that the most severe break in the small break spectrum had been identified based upon an understanding of the physical phenomena involved.

The Technical Review Team noted that based on a broader knowledge of small break LOCA phenomen and results from other codes, Yankee Atomic was confident that small break LOCAs were bounded by large break LOCAs and based on this expectation, Yankee Atomic accepted the results from RELAP5YA as adequate for showing compliance with the regulations. The Technical Review Team stated its understanding that this may have been a correct conclusion, but because it was based, in part, on information beyond demonstrated results of runs of the RELAP5YA code for Maine Yankee, the :ituation should have been communicated to the NRC prior to implementation." The Technical Review Team concluded that a number ofissues related to RELAP5YA implementation (as well as a number of other issues related to 42 The Technical Review Team noted that the LOCA community's expectations "are not completely documented in specific references . . . ." Rather, "they have been clearly communicated to the industry's LOCA community through extensive and numerous interactions with the NRC (Reactor Systems Branch, the Advisory Committee on Reactor Safeguards (ACRS), and others) . . . ." Technical Review Report at III.D.I. While industry practice and NRC expectations are elements of our exploration of this issue, inasmuch as they are not regulatory requirements, our review must also focus on compliance with NRC regulations.

Id. at IV.A.

Id.

wmosooucs D-14

l l

the Demand) should have been communicated to the NRC prior to implementation.45

j. Clearly, in hindsight, that is the case.

l However, we understand the situation in which the LOCA Group found itself was not l entirely ofits own making and have taken this into account in our review. Maine l Yankee had been told that no submittal to the NRC to satisfy TMI Item II.K.3.31 was necessary.46 Communications with the NRC were understood by the LOCA Group to i be discouraged. Therefore, they did not believe they could informally communicate with the NRC. A corporate culture stressing independence, poor n,anagement oversight over the project, the LOCA Group's isolation from the remainder of the entities developing l LOCA codes, interactions with Maine Yankee, and actions by the NRC also contributed to difficulties in implementing the RELAP.5YA Code as applied to Maine Yankee.'7 I

With regard to our analysis of careless disregard, we considered the following elements:

< 1. The complete small break spectrum had not been analyzed.

l l 2. The Technical Review Team had questions as to the behavior of the PCT in the unanalyzed region based upon code behavior and previous results with other codes.

lg 3. The decision that all points on the comple'te small break spectrum did not have to lQ be run came after the LOCA Group was unsuccessful in completing the run at 0.35 ft.2 after significant effort over a number of years of working with RELAP5YA.

4. RELAP5YA was not run for a break size of 0.5 ft.2, the limiting small break size for l the Appendix K analysis of record.

l 5. It was a priority of the LOCA Group to complete the project.

l 6. If the LOCA Group had inquired of the NRC or determined its expectations from

others, it could have avoided the situation.

7. On its face,10CFR50.46 would not prohibit the combination of analysis and l i

technicaljustification based upon an understanding of physical phenomena. j i

) 45 Sce, , Technical Review Report at IV.A, IV.B and IV.D.

'6 Id. at IV.A.

'7 t Id. at III.D.I.

I emesaw ms D-15

l i

, f3

8. NRC expectations regarding 10CFR50.46 were not completely documented, but

! rather were communicated to the LOCA community through its interactions with the NRC.

9. The LOCA Group believed that the successful running of the RELAP5YA code at a number of break sizes smaller than 0.35 ft.2 and at 0.35 ft.2 until the code terminated, combined with a sufficient understanding of the physical phenomena which were occurring in the small break region,it, up to 0.6 ft.2, and an understanding of the differences between RELAP5YA and the code of record was sufficient to comply with 10CFR50.46.

10. Members of the LOCA Group were able to provide the Technical Review Team justification which supported their code. The Technical Review Team found that such results tend to support the Yankee Atomic position although some trends do not.

I

11. Although it was unable to draw a definitive conclusion regarding the PCTs for the i unanalyzed portion of the Maine Yankee SBLOCA spectrum, the Technical Review Team concluded that the SBLOCA PCTs for all the analyses reviewed met the 10CFR50.46 2200*F criterion and that SBLOCAs remain bounded by LBLOCAs.

( 12. A combination oflongstanding institutionalissues, isolation from the LOCA community, poor management, inadequate support within the Company, interactions with Maine Yankee, and actions by the NRC placed roadblocks in the way of the successful completion of this project.

I 13. The technical capabilities of the LOCA Group were acceptable for performing LOCA analysis, and the project included extensive assessment of RELAP5YA to scaled test facility SBLOCA data.

Weighing all these factors, we find that there was a basis for the LOCA Group's interpretation of 10CFR50.46, albeit one in contrast to NRC and industry expectations; we also find there was a basis for the LOCA Group to conclude they had identified the limiting break in the small break spectrum based upon their knowledge of the phenomena associated with small breaks. We therefore conclude that a reasonable

! justification existed for the LOCA Group's actions, and thus we cannot conclude that there was reckless indifference to the requirements of 10CFR50.46 or utter unconcern of the consequences in that the LOCA Group thought it was complying with a l permissible interpretation of 10CFR50.46. In these circumstances, we cannot say that an i individualin the LOCA Group blinded himself or herself to the realities of whether a violation would occur. We conclude that careless disregard of the Commission's requirements was not involved with respect to the issues in Section III.A of the Demand.

emmomms D-16

Turning now to subsection B, the NRC asserted that Maine Yankee maintained

' information in its files (YAEC-1868) and submitted to the NRC Core Performance Analysis Reports ("CPAR") in support of operating Cycle 14 and Cycle 15 reload l analyses which were not complete and accurate in all material respects in apparent L violation of 10CFR50.9(a). The NRC further asserted that the Manager never brought l the deficiencies to the attention of the cognizant Manager in charge of the Engineering l - Section of the Licensing and Engineering Group of Maine Yankee, who was the manager kept directly apprised by Yankee Atomic and by the Manager on the development of the specific RELAP5YA Evaluation Models.

For their part, the involved YAEC persor.nel maintain that they believed that they had

, complied with the requirements of 10CFR50.46(a)(1) and that the documents were l complete and accurate in all material respects and they adequately communicated with l Maine Yankee personnel.

i j It is clear that the Yankee Atomic LOCA Group expected that YAEC was subject to -

audit by the NRC with regard to the application of the Evaluation Model to Maine Yankee and its compliance with Commission regulations. We believe that this blief was reasonable, particularly in light of correspondence between Maine Yankee and the NRC ,

that such was contemplated.

l l The Technical Review Team concluded that YAEC-1868 documents the results of the -

! small break LOCA analysis performed by Yankee Atomic using the NRC-approved L YAEC-1300P small break LOCA Evaluation Model.5 It further concluded that the

- document was sufficiently complete and accurate as a summary of the small break LOCA calculations that were performed.5' It also found that it was understandable to its intended audience and was suitable for a licensing submittalin support of Maine Yankee.52 The Technical Review Report concluded that the amount of technical

l. information included was appropriate for any knowledgeable person to understand the l

d' l Our investigation did not reach the question whether individuals at Maine Yankee may l have been knowledgeable in the requirements of 10CFR50.46 and RELAP5YA. Su n.59, L infra. Sr.c alsa Technical Review Report at IV.B.

l l May.8,1989 letter to C.D. Frizzle, President, Maine Yankee Atomic Power Company .

l from Patrick M. Sears, Project Manager, NRC.

5 Technical Review Report at IV.B.

s' Id. As to the loss coefficient, sec n.78, infra.

52 Id. The Technical Review Report noted that the Abstract was potentially misleading in the use of the word " complete," but that the scope of the analysis as contained in the report is characterized correctly.

l muosomnm - D-17

O)

( results of the analysis." The Technical Review Team found there was no intended concealment ofinformation which would have identified any non-compliance with NRC j regulations." The Technical Review Team stated that YAEC-1868 could be understood '

by a knowledgeable engineer not trained in the LOCA licensing process to be complete and in compliance with the regulations." The Technical Review Team concluded that the compliance statement and the supporting analyses in YAEC-1868 would be understood by an NRC reviewer and would have led to interactions with Yankee ,

Atomic.56 l Our review of YAEC-1868 confirmed that there were instances which erroneously stated that the complete break spectrum was analyzed." However, we believe that the  !

document should be read as a whole. The arguments advanced by the Manager and Lead Engineer as to RELAP5YA's compliance with 10CFR50.46 are discussed in the body of the document itself." Inasmuch as this document was maintained at Maine Yankee and subject to audit by the NRC, we believe, as does the Technical Review Team, that any reviewer from the NRC " sent to audit compliance with 10CFR50.46 would have understood the compliance statements and supporting analyses and would

" M.

(N i V "

M. The Technical Review Team found that the statements in YAEC-1868 regarding I compliance with the NRC's regulations were consistent with the Yankee Atomic LOCA Group's understanding of the regulations.

M.

u g.

" Src, for example, the Abstract of YAEC-1868 which states that "[e] valuations were performed over a complete range of break sizes . . . ."

" Stc, for example, YAEC-1868 at 21-22. No runs of greater than 0.35 ft.2 are presented in the analysis. M. at 22.

" We believe that the standard advanced by the NRC to the effect that " language would not signify to an individual without expertise in LOCA code that RELAP5YA had failed and was not capable of calculating ECCS performance" (Demand at 11) is not the appropriate one for determining whether there had been careless disregard of the regulation. We believe the issue of concealment should be judged from the point of view of a cognizant NRC reviewer knowledgeable in computer codes used to satisfy Section 50.46. Src Virginia Electric Power Company (North Anna Power Station, Units 1 and 2), CLI-76-22, (g )

4 NRC 480,491 (1976). Such a reviewer or auditor would be able to appreciate the significance of the statements contained in YAEC-1868. The experts within the Technical Review Team were able to do so.

. m oso m m . D-18 l

l

l l

i lO have inquired further." Finally, we observe that the Technical Review Team felt that the 6

document was suitable for licensing ' which we understand to mean complete and accurate, subject to anticipated NRC questions and revision, as appropriate.

In addition to the considerations previously outlined in response to Section III.A of the Demand and in light of the above, we weighed the following considerations in

, determining whether careless disregard of NRC requirements existed related to Section III.B:

1. YAEC-1868 could be understood by a knowledgeable engineer not trained in the LOCA licensing process to be complete and in compliance with the regulations.

2. The Abstract of YAEC-1868 states that "[e] valuations were performed over a complete range of break sizes . . . ." Other sections also contain similar language.

3. YAEC-1868 was sufficiently complete and accurate as a summary of the SBLOCA calculations that were performed. The amount of technical information included was appropriate for any knowledgeable person to understand the results.

4. The scope of the analysis as contained within the report is characterized correctly

and there is no concealment regarding the " failure" of the run at 0.35 ft.2 and that no runs oflarger bre' k sizes within the small break spectrum were run.

5. But for the May 8,1989 NRC letter. YAEC-1868 would have been submitted to the NRC.

l

6. Yankee Atomic had a reasonable belief that the NRC would perform an audit on i YAEC-1868.  !

7. The statements in YAEC-1868 are consistent with the LOCA Group's  !

! understanding of the regulations.

8. The compliance statements and supponing analyses in YAEC-1868 would be j understood by an NRC reviewer and would have led to interactions with Yankee i o

Atomic, i i

\

l On this basis, we conclude that to the extent any deficiency existed, it did not represent l: reckless indifference to a requirement or utter unconcern whether there was compliance i

"- Moreover, the even more detailed calculations referenced in YAEC-1868 were available l

for review.

Technical Review Repon at IV.B.

m woso m na. D-19

l I

l^

with the requirement. We cannot say that any individual blinded himself or herself to w realities of whether a violation occurred. Thus, no careless disregard was involved.

C. . R*=an= to Sectlan HLC C. During Cycle 14 operations and in the Cycle 14 and Cycle 15 CPAR, MYAPCo used an apparently unacceptable SBLOCA Evaluation Model which over-predicted core cooling.

In this subsection, the NRC asserts that the LOCA Group misapplied the Alb-Chambre L

penetration correlation, which is an empirical calculation of the penetration factor of the :

!. injected ECCS water penetrating the downcomer annulus into the lower plenum. The-NRC concludes that because the analysis did not provide a basis to assuna full penetration of the emergency core cooling system injection, it provides no basis to derive the loss coefficient of 600 used for the split downcomer nodalization. The NRC further asserts that these deficiencies resulted in overprediction of core cooling and overstatement of the conservatism of the model. The NRC states that the Manager and lead Engineer should have realized during the work associated with the RELAP5YA analysis, that the Alb-Chambre correlation had been incorrectly applied. It states that adequate QA review would have revealed the errors and the unacceptability of the l RELAP5YA SBLOCA analysis described in YAEC-1868. The NRC ultimately states l ' that based upon these errors, YAEC caused Maine Yankee to rely on an unacceptable L SBLOCA Evaluation Model in violation of 10CFR50.46(a)(1).

l l The LOCA Group asserts that the NRC description does not reflect their reasoning concerning core penetration and the actions that they took. While admittedly there was l an arithmetic error in the implementation of the Alb-Chambre correlation, the results as presented in the calculation would not cause a reviewer providing QA oversight to have recognized that the Alb-Chambre correlation had been mistakenly calculated.

Furthermore, the LOCA Group states that even if the negative value of Alb-Chambre correlation had been discovered, a negative range,11, less than zero, would not be indicative of a meaningless result and merely would have caused them to further examine the values used in the correlation. The LOCA Group denies that such calculations also L mdicate other possible errors in application of the Alb-Chambre correlation.

l~

L The Technical Review Team concluded that the use of the Alb-Chambre correlation as a confirmation of the modeling approach which includes the junction loss coefficient of 600 in the reactor downcomer is reasonable given the available data and the deficiency in i.

62 It may be argued that Section III.C does not allege willfulness, but merely asserts that there were specific deficiencies in the LOCA analysis and related quality assurance errors.

However, we have analyzed this issue.

wmmam ms D-20 t-

l l

' the code. Early applications of the RELAP5YA small break LOCA 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. Eventually an artificially large loss coefficient was introduced in the junction connecting the two volumes representing a split reactor vessellower downcomer. The value of this loss coefficient was varied to l obtain a balance between the expected ECCS penetration and the effect on steam venting via the break." A value of 600 was selected as an appropriate value." The amount of ECCS penetration obtained with this modeling approach was justified, in l

part, by the Alb-Chambre correlation. This correlation was applied to confirrn that the '

amount of ECCS penetration predicted by RELAP5YA was conservative." The Review Team considered this approach to be an acceptable compensation for a RELAP5YA deficiency, and the use of a value of 600 obtained an amount of ECCS penetration that was consistent with industry experience." The Technical Review Team concluded that this modeling approach is not expected to result in an overprediction of core cooling, but since the calculations were not completed for all break sizes, it cannot be definitively l confirmed." l An arithmetic error was made in the application of the Alb-Chambre correlation, but not identified during the quality assurance process. The Technical Review Team found this to be a failure of the quality assurance process, but was able to conclude there was a n reasonable explanation as to why it occurred." The arithmetic error did not skew the

(^j result of the calculation, which was in the range of the expected result that complete penetration was predicted.7 The correlation can produce results in excess of the value of 1 (in this case, a value of 8) which have the meaning of complete ECCS penetration.7'  !

The Technical Review Team found that a person performing a quality assurance review is influenced by the result based on experience and expectations and the errors encountered here are the most difficult errors to identify. A correct application of the

" Technical Review Report at IV.C.

M.

65 g,

M.

" M. at 14, III.D.3, IV.C.

M. at IV.C.

l M.

7 M.

78 M.

wmmomms D-21

(

ty

i

'V Alb-Chambre correlation without the arithmetic error would have produced negative values indicating complete ECCS bypass. It was concluded that this result would have been immediately recognized by Yankee Atomic as non-physical for the SBLOCA conditions ofinterest.72 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.74 Tnerefore, the Technical Review Team concluded that although the application of the Alb-Chambre correlation contained an error, the modeling which incorporated the loss coefficient with a value of 600 remains valid and, thus, the results of the error in applying the Alb-Chambre correlation did not result in invalid input to the  ;

SBLOCA analyses."

l l

The Technical Review Team concluded that the use of the Alb-Chambre correlation as a l confirmation of the modeling approach which includes the junction loss coefficient of I 600 in the reactor vessel downcomer was reasonable given the available data and a need to address a deficiency in the code.76 The Technical Review Team believed that had there been a submittal of the Maine Yankee SBLOCA application to the NRC, this modeling approach would have been discussed and reviewed. The Technical Review Team belkvei that this model could have been approved by the NRC in this form or with some revision." i (o)

~'

It is therefore our conclusion that there is no inadequate analyses involved in subsection i III.C and, therefore, the issue of deliberateness or careless disregard < foes not arise. It is our view that the actions of the cited individuals were reasonable and there was no evidence of careless disregard."

M.

M.

M.

M.

76 M. The Technical Review Team recognizes that the PCT results of the SBLOCA analyses for some of the break sizes are very sensitive to the value of the loss coefficient.

M.

" While not raised by the NRC in its Demand, the Technical Review Team discussed whether the loss ccefficient changes required NRC approval inasmuch as it concluded that due to the non-physical value used and the significance of this input parameter, it is an (q) important model change. While the Technical Reviaw Team recognized that the loss coefficient was literally an input to the model, it concluded that in its opinion it was an mmsomm, D-22

F l

D. Remonse to Section IILD D. MYAPCo used an apparently unacceptable Best Estimate RELAPSYA SBLOCA l Evaluation Modelto calculate ECCSperformance.

l The NRC asserts that Maine Yankee performed a safety evaluation in order to determine if a decrease in steam generator pressure involved an unreviewed safety question pursuant to the requirements of 50.59 and, in doing so, used an unacceptable Best Estimate ("BE") RELAP5YA Evaluation Modelin evaluating small break ECCS LOCAs in apparent violation of 50.46(a)(1). The NRC states that Maine Yankee's March,1993 50.59 analysis relied on an analysis of the effect of reduction in steam generator pressure on ECCS performance prepared by Yankee Atomic vhich used, among other analyses, the Best Estimate RELAP5YA SBLOCA analysis. This document was entitled LOCA 91-04 which was approved by the Manager of the Yankee Atomic nuclear engineering department for the Manager of the LOCA Group. The Manager received a copy of a YAEC memorandum, Impact of Lower Steam Generator Pressure on the safety analysis, NED 91-18, which relied on LOCA 91-04 to evaluate the impact of reduced steam generator pressure on the small and LBLOCA analyses. A Yankee Atomic memorandum dated May 29,1992 entitled Steam Generator Pressure and Heat Transfer Coefficient Performance (TAG-MY-92-035) concluded that "the lower initial steam generator pressure did not affect the results of the licensing analysis,"

'Q

! D which conclusion was based in part on NED 91-18. The May 29,1992 memorandum states that it is " safety-related." The Manager approved the May 29,1992 memorandum.

The NRC asserts that a 50.59 analysis cannot confirm that the ECCS performance will I

be adequate unless the 50.59 analysis uses LOCA Evaluation Models acceptable to I demonstrate compliance with 50.46 and then reasons that the Best Estimate  !

RELAP5YA model could not be utilized because in January,1989, the NRC approval was for use of RELAP5YA as the Appendix K Evaluation Model for Maine Yankee.

l Additionally, the NRC states that there were some parts of the Best Estimate model which did not comply with Part 50, Appendix K. The NRC states that it is reasonable to conclude that the Manager of the Yankee Atomic LOCA Group was aware that the Best Estimate approach deviated from the approach approved by the NRC and that a Best

Estimate RELAP5YA model would not be acceptable for use in licensing matters, L without NRC approval. The NRC believes it is reasonable to conclude that the Manager i

important model change (Teclutical Review Report at III.D.1, IV.C). We note that the Technical Review Team recognized that had the Maine Yankee SBLOCA application been y directly submitted to the NRC, this model"could have been approved by the NRC in this form or with some revision"(Id. at IV.C). Because of these factors and the confusion that existed due to the NRC's May 8,1989 letter, we conclude that Yankee Atomic had a basis i for assuming that the NRC would have reviewed the matter in due course and, therefore, would have acted consistent with the Technical Review Team's expectations.

.mourm ms D-23 J

l knew that the analysis which Yankee Atomic had performed would be used by Maine l Yankee in a 50.59 analysis or other safety analyses. The NRC concludes that in view of the intended use of the YAEC analysis, the Manager should have provided Maine Yankee with an analysis which met NRC requirements and that by providing to Maine l Yankee an unacceptable analysis of the effects of reduction steam generator pressure, l Yankee Atomic caused Maine Yankee to apparently violate 10CFR50.46(a) by relying

, on an unacceptable SBLOCA Evaluation Model to c.alculate ECCS cooling performance in preparing a 50.'59 analysis.

There are two separate elements of our evaluation of this issue. The first relates to the Maine Yankee service requests which were issued in 1990. These asked Yankee Atomic i to perform what the Technical Review Team fcund to be a " scoping analysis" of the effect of reduced steam generator pressure on the licensing basis transients and accidents." Steam generator pressure was decreasing steadily due to fouling of the steam generator tubes. The Yankee Atomic LOCA Group was responsible for

~

addressing the LOCA aspects of this situation."

We examined the use of the Best Estimate evaluation model to respond to those requests and the appropriateness of the response. In early 1990, due to difficulties in applying the generic Appendix K RELAP5YA small break LOCA code to the Maine m Yankee plant within the prescribed time, the Manager had recommended a parallel alternative Best Estimate Evaluation Model" Since the previously submitted topical report did not include the Best Estimate approach, separate NRC approval would be necessary prior to implementation.32 The Manager rr anded submittal of the Best Estimre:e model to the NRC, presumably for NRC t..mw and approval. The Best Estimate approach was memorialized in a memorandum dated August 1,1990." No NRC approval for the Best Estimate approach was sought by Maine Yankee." For some time, the Manager was under the impression that the Best Estimate approach methodology report had been submitted to the NRC by Maine Yankee for review."

" Technical Rwiew Report at IV.D.

Id.

" January 2,1990 Memorandum to P.L. Anderson.

L

" Technical Review Report at IV.D.

" August 1,1990 Memorandum to P.L. Anderson (LOCA 90-110).

" The reasons for this are unclear, but not germane to our findings.

I " Technical Review Report at IV.D.

wmmo . D-24

]

(-

LOCA-91-04, dated' January 25,1991, and NED 91-18, dated January 28,1991, which -

- incorporates LOCA-91-04, are the memoranda responses to the 1990 service requests.

In order to be responsive to the customer's request, Yankee Atomic used the only available LOCA analysis tool at that time, the Best Estimate model, to assess the effect of reduced steam generator pressure on the analysis of record." The author's i- characterization in LOCA-91-04 that the Best Estimate model was the licensing basis for i

small break LOCA analysis at the facility was incorrect and never corrected until the issuance of TAG-MY-93-012." The analysis of record at this time was the 1977 Combustion Engineering analysis. As noted by the NRC, the Manager did not sign the 1991 memoranda."

l The Technical Review Team noted that the effect of reduced steam generator pressure on the SBLOCA results for the magnitude of steam generator tube fouling and plugging j that was being evaluated would not be expected to be significant. This is particularly true given that the analyses of record showed the SBLOCA PCTs to be lower than the l LBLOCA PCTs. The Technical Review Team found that an evaluation could have been justified without any SBLOCA analysis. An evaluation could also have beenjustified using the Best Estimate analysis methodology provided that sufficient qualification was included, and provided that the analysis of record was not replaced. Given this finding, it would be possible to conclude that no tecimical inadequacy existed,it, that a non-approved LOCA evaluation model could be appropriately utilized to support a

, 10CFR50.59 analysis. However, we complete our analysis based on the assumption that I use of the Best Estimate model was not appropriate.

l The Technical Review Team disagreed with the NRC's premise that only approved Appendix K evaluation models can be used in performing some scoping safety evaluations, including input to 50.59 evaluations." The Technical Review Teun found that models such as Best Estimate models can be appropriately used provided that the

, application does not replace the analysis of record, and provided that the use of the

( analysis method is clearly stated and justified." It noted that if there is any doubt ,

l regarding the appropriateness of such an application, then the NRC should be consulted

( 1 l

Id.

" We have not interviewed the author of this document who we understand left the l-employ of Yankee Atomic in the early 1990s.

" . The Manager informed us that he likely would have recognized that the characterization of

' the Best Estimate modelin the memorandum was incorrect.

" Technical Review Report at IV.D.

Id.

mmomms D-25

l i

prior to implementing the results of the analysis." It found that Yankee Atomic's failure in this situation was in incorrectly characterizing the Best Estimate LOCA analysis as the licensing basis analysis, and furthermore not stating that the analysis used non-NRC

]

approved methods and had restrictions on its use." l The Technical Review Team noted that the effect of reduced steam generator pressure on the SBLOCA results for the magnitude of steam generator tube fouling and plugging j that was being evaluated would not be expected to be significant." This is particularly true given that the analyses of record showed the small break LOCA PCTs to be lower than the Large Break LOCA PC'. s. The Technical Review Team found that an evaluation could have been justified without any SBLOCA analysis." An evaluation could also have been justified using the Best Estimate analysis methodology provided that sufficient qualification was included, and provided that the analysis of record was not replaced." Given this finding, it would be possible to conclude that no technical inadequacy existed,it, that a non-approved LOCA evaluation model could be appropriately utilized to support a 10CFR50.59 analysis. However, we complete our 1 analysis based on the assumption that use of the Best Estimate model was not I appropriate.

We understand that the author of LOCA-91-04 may have believed he was preparing a scoping document to answer a hypothetical question from Maine Yankee for which the Best Estimate approach was appropriate. We reviewed the Service Requests from ,

Maine Yankee which resulted in LOCA 91-04 and NED 91-18." We believe that l based upon the amount of time permitted to comple t the evaluation, it was questionable whether anyone contemplated that othei than a scoping analysis was requested. The Technical Review Team agreed that a scoping analysis was requested." It is recognized that the Service Requests speak to bounding of the safety analysis; however, at the time, there was not the present sensitivity to the term M.

" M.

M.

M.  ;

" M.

" Service Request M-90-155 (November 1,1990): Service Request M-90-158A p)

(November 23,1990).

Technical Review Report at IV.D.

wmmomms D-26 l

.)

i n

U " safety analysis" or to the need to analyze design basis deviations." Because the Best Estimate model wr yielding significantly higher peak clad temperatures than the Evaluation Model of record (while still below the acceptance criterion of 2200 F set forth in 10CFR50.46(b)(1)), it may have been felt that it was more conservative to use this approach to determine whether the change in steam generator pressure caused the PCT to increase. In the circumstances, we conclude that there was a reasonable basis for the use of a Best Estimate anodel at the time despite the error la its representation as the liv m hasis analysis.

The second element invohn . iesponse to the April 9,1992 Service Request No.

M-92-42, which asked for a determination of the minimum steam generator pressure that can be supported by analysis. A target value of no greater than 743 psig actual pressure was indicated as desirable. A request for the provisions of the " uncertainties that should be used with the computer and Main Control Board indications of steam generator pressure" was also made. In response to the request to determine the minimum steam generator pressure that can be supported by the analysis, on May 4, 1992 Yankee Atomic stated that this portion of the request "will be completed aft er the revised SBLOCA model is completed."" The response indicated that the second half of the request had already been completed, referencing a memorandum, TAG-MY 91030.'"

A y") The May 29,1992 memorandum, TAG-MY-92-035, referenced NED 91-18. 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 Best Estimate model. TAG-MY-92-035 was referenced by Maine Yankee in a 50.59 l

" This is particularly true from the perspective of an analyst (or even the Manager) in the LOCA Group who was far removed from day-to-day issues associated with keeping the design basis and safety analysis current. The Technical Review Team stated that the failure to include statements that the model was not NRC-approved also indicates a lack l of administrative control and communication with the LOCA Group and the Yankee  ;

Atomic Maine Yaakee Project Group and a failure to appreciate the disclosures that I I

should be made regarding application of non-NRC approved LOCA analysis methodology in responding to plant support requests. Technical Review Report at IV.D.

" This response was contained in the YNSD Response section of Service Request No.

M-92-42. We understand that it was the Manager's intent from April 1992 forward that i the Appendix K small break LOCA analysis be used in conjunction with the analysis of the i lowered steam generator pressure issue. However, this intent does not appear to have been communicated to the individual at Maine Yankee performing the 50.59 review.

'" We understand that as to the small break LOCn, sae technical bases of the two g] memoranda, TAG-MY-92-030 and TAG-MY-92-035 are similar; both reference (J

NED 91-18.

mmmm D-27

q V evaluation dated April 12,1993 as part of the justification for operation with reduced steam generator pressure. The analysis based on the Appendix K SBLOCA results, i.A, RELAP5YA, were forwarded to Maine Yankee also on April 12,1993, along with a draft 50.59 evaluation for Maine Yankee's use. These results were not referenced by Maine Yankee until January 13,1994, when the original Maine Yankee 50.59 evaluation was revised.

We do not believe that it was appropriate to utilize NED 91-18 which relied on a Best Estimate approach without identifying it as being a non-approved LOCA model inasmuch as at that time, iA, by April 1992, the LOCA Group knew that the Best Estimate model had not been submitted to the NRC. However, the Technical Review Team did conclude that its use would have been permissible provided that the application does not replace the analysis of record and provided that the use of the analysis method is clearly stated and justified. This was not done.2 ' Furthermore, in 1992, Yankee Atomic should have recognized that the Maine Yankee request was for more than a scoping analysis. We believe that Yankee Atomic should have recognized that its memorandum, TAG-MY-92-035, could possibly be utilized by Maine Yankee for safety related analyses.

Our next task was to determine whether the inadequacies described above were the result of careless disregard. In the 1992 time frame, utilities were only beginning to recognize that changes in plant conditions could constitute a change, test or experiment which would bring into play the requirements of 10CFR50.59. Whereas the steam generator pressure was an input to various station analyses, it was not a controlled variable nor do we understand that it could be found in the Technical Specifications for the facility. At the time, it was known that the LOCA small break analysis was not sensitive to the value of the steam generator pressure. 2 This understanding was confirmed by later analysis. Generic I2tter 91-18, "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability," which was issued on November 7,1991, increased nuclear utilities' sensitivity of the need to prepare a 50.59 analyses for "as-found" degraded conditions, but such appreciation does not Technical Review Report at IV.D. Technical Review Report states that there was a period of time available to correct the situation when the results of non-NRC approved methods were found to have been released to the customer. M. We are not aware that during the time in question, anyone had an affumative appreciation that NED-91-18 was incorrect. Clearly it should have been reviewed again when TAG-MY-92-035 was issued, O but was not. This is a deficiency.

G - u.

mmowms D-28 ,

l

f l

f IO appear to have worked its way to the LOCA Group until much later.2 3 We do not believe that the period of time between the issuance of the generic letter and the issuance of TAG-MY-92-035 was such that it was likely that the isolated Yankee Atomic LOCA Group would have appreciated the effect of Generic Letter 91-18 upon the Group's activities. j The most compelling piece of evidence that we obtained on this issoe was an indication that the impetus for Maine Yankee preparing a 50.59 evaluation rela;cd to !

the lowered steam generator pressure issue, iA, based upon the May 29,1992 ]

memorandum, came from its NRC Resident inspector some months after Maine Yankee received the May 29,1992 memorandum. This acts to confirm that Yankee Atomic may not have been unreasonable in not anticipating that its memo would be used in a 50.59 evaluation. If the requirement to prepare a 10CFR50.59 evaluation was not within the contemplation of the utility which was in a better position to understand its design basis, the expectations of the NRC and its own procedures to implement 10CFR50.59, this is a strong indication to us that even if Yankee Atomic )

should have anticipated the design basis implications of its May 29,1992 l

memorandum, its failure did act amount to careless disregard. l Also instructive are Yankee Atomic's actions and statements in an April 12,1993, memor mdum, TAG-MY-93-012, on the same subject ratter. This memorandum

()

l incorporated the lower steam generator pressure of 735 psig in the safety analysis, I

'~

utilizing RELAP5YA in the Appendix K mode as promised in the Service Request. With regaro to compliance with 10CFR50.59, that memorandum stated in a section entitled

" Safety Evaluation":

This memo is safety related. It provides Maine Yankee with a single i method for monitoring the full power SG pressure anJ overall heat I transfer coefficient (UA) to verify that these parameters are within the safety analysis envelope. Since this memo theoretically expands the operating space of the plant from what is actually reported in Appendix D of the FSAR, it is our opinion that incorporation of the Generic Letter 91-18 states:

The additions to the NRC Inspection Manual are based upon previously issued guidance. However, because of the complexity involved in operability determinations and the resolution of degraded and nonconforming conditions, there have been differences in application by NRC staff during past inspection activities. Thus, the purpose of publishing this guidance is to ensure consistency in application of this q guidance by the NRC. Regional inspection personnel have been briefed on Q this guidance. The NRC will conduct further training on these topics to ensure uniform staff understanding.

[

l mnwom u D-29 L

l;

. (3 V information _into Cycle 13 operation may require a 10CFR50.59 evaluation. Attachment C contains 50.59 eve.luation information (per Maine Yankee Procedure Number 0-06-4) which can be useo by Maine

j. Yankee to complete their formal review. It is not intended to be a j c.omnlete or formni 50.59 evninntion. (emphasis in original)'"

We believe that this language reflects Yankee Atomic's new appreciation of the changing attitude in the nuclear industry as to the circumstances for which a 50.59

! analysis need be generated.8 5 Our evaluation of reckless disregard considered the following factors:

i

1. The Best Estimate model was the only small break evaluation model available at the l time.

l

2. The Best Estimate approach was yielding PCTs higher than the Evaluation Model ofrecord. 1

3. The Best Estimate model was incorrectly indicated as being the licensing basis evaluation model.

/ 4. In 1991, there was considerable confusion for reasons beyond the control of the ,

LOCA Group as to the status of the Best Estimate model.

5. LOCA-91-04 could have been considered a " scoping" evaluation.  !

i

6. The Manager should have known that affirmative NRC approval of the Best j Estimate model was needed and in 1992 that it had not yet been received. l l

I

7. There is no evidence that the Manager was familiar with the contents of l NED-91-18 cnd LOCA 91-04.

8. The Manager apparently approved the May 29,1992 rnemorandum without I

reviewing the references, including NED 91-18 and LOCA 91-04.

'" ' It is ironic that this memorandum was dated the same day as Maine Yankee's approval

date for its 50.59 analysis related to the May 29,1992 memorandum. We understand that l Yankee Atomic was not aware of Maine Yankee's 50.59 analysis at the time it issued the i memorandum.

85 Some time before this memorandum was prepared, Maine Yankee requested that Yankee l Atomic prepare draft 50.59 input for any analyses it performed which could affect the facility's safety evaluation. We understand that the April 12,1993 memorandum, TAG-MY-93-012, was among the first to incorporate such an evaluation.

t smusmemmi D-30 l

i

9. Failing to review all references was an error, but understandable. 'i l

10. Yankee Atomic personnel should have recognized that the May 29,1992 memorandum had the potential to be utilized in a manner that affected the safety analysis. 1 i

11. From April 1992 forward, the Manager intended.that the Appendix K SBLOCA analysis would be used to reanalyze a lower steam generator pressure limit than had been previously assumed, but such intent was not communicated to the individual

. performing the 50.59 review. -i

12. Maine Yankee personnel also did not contemplate the need for a 50.59 analysis relating to reduced s'eam generator pressure until suggested by the NRC Resident inspector some months subsequently.

13. The regulatory position that degraded conditions would have to be reflected in the safety analysis via a 50 59 esaluation in the sane way as affirmative changes in the a design had only recently been addressed by the NRC. Apparcatly this guidance had not had an opportunity to have affected the thinking and actions of LOCA Group personnel with regard to steam generator pressure degradation.

14. Yankee Atomic's Aptil 12,1993 memorandum, TAG-MY-93-012, does contain a draft 10CFR50.59 analysis.

l

15. The Technical Review Team concluded that analytical techniques other than tlie approved Evaluation Model can be used to prepare 10CFR50.59 evaluations if -

sufficient disclosure is made.

l We cannct conclude that thern existed a reckless disregard or callous indifference I toward their responsibilities or the consequence of their cctions on the part of Yankee Atomic personnel. In 'the 1991-92 time frame, Yankee Atomic did not appreciate that  ;

steam generator pressure degradation affected the design basis; however, as this issue matured, so did Yankee Atomic's thinking as evidenced in its April 12,1993 )

memorandum, TAG MY-93-012. Yankee Atomic did not reasonably have kmowledge  ;

that its behavior was likely to result in a violation of NRC requirements as it reasonably j could have been expected to understand them at the time. Therefore, we conclude that I there has net been careless disregard of the regulations.

VII. CONCLUSION Section V.B of the Demand states as follows:

An explanation why the NRC should not consider the inadequate analyses, j O- which apparently caused MYAPCo to be in violation of NRC j I

wmmo-nos D-31 j l

3

.()

V requirements, to be the result of willfulness, either deliberateness or careless disregard, on the part of YAEC and/or DE&S personnel.

Lo. 1 The previous sections state the conclusions of the Willfulnesr, Review Team as to its findings as to the specific matters within the scope of the Demand. The Demand focused on the actions of two individuals in relation to whether willfulnesr., either deliberateness or careless disregard, existed. We found no willfulness on the part of the two individuals or any other Yankee Atomic insonnel. We therefore believ9 that the Manager and Lead Engineer are capable of I conducting their activities in the future in conformance with NRC requirements.

l We conclude that there was rotMng developed as a result of out investigaGon ou the conduct of Yankee Atomic and/or DE&S personnel that would prevent the improvements that we understand are being made from being successful and resulting in DE&S' activities being conducted in full compliimce with NRC requirements and expectations.

i

\

.f I

U l

}

+

\ s F

(  ;

I

> i 4

O b

m tem e nna: D-32

't h 3

'ai 2  ;