ML20246E502

Intervenor Motion to Admit Contention,Or,In Alternative,To Reopen Record & Request for Hearing.* Urges That Further Low Power Testing Remain Suspended & Full Power OL Not Be Issued Pending Full Evidentiary Hearing Re 890622 Event

ML20246E502
Person / Time
Site:	Seabrook
Issue date:	07/21/1989
From:	Shannon J, Traficonte J MASSACHUSETTS, COMMONWEALTH OF
To:	Atomic Safety and Licensing Board Panel
Shared Package
ML20246E416	List: ML20246E411 ML20246E502 ML20246E525 ML20246E535
References
OL, NUDOCS 8908290154
Download: ML20246E502 (190)

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{{#Wiki_filter:_ p k L;. ( l . UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD' Before the Administrative Judges: Ivan W. Smith, Chairman Dr. Richard F. Cole Kenneth A. McCollom ) In the-Matter of ) Docket Nos. 50-443-OL ) 50-444-OL PUBLIC SERVICE COMPANY ) (Off-Site.EP) OF NEW HAMPSHIRE,.EI AL. ) ) .(Seabrook Station, Units;l and 2) ) July 21, 1989 ) INTERVENERS' MOTION TO ADMIT CONTENTION, OR, IN THE ALTERNATIVE, TO REOPEN THE' RECORD, AND REQUEST FOR' HEARING INTRODUCIlQH 'The Massachusetts Attorney General (Mass AG), Seacoast Anti Pollution League (SAPL), and New England Coalition on Nuclear Pollution (NECNP), (hereinafter " Interveners"), pursuant to S189(a)'of the Atomic Energy Act, move this Board to admit for litigation the Contention filed herewith as Exhibit 1 (" Contention"). In the alternative, Interveners request this Board to reopen the record, pursuant to 10 C.F.R. $2.734, and admit the Contention. 8908290154 890821 PDR ADDCK 05000443 0 PDR

~ Interveners' request a hearing on all issues raised by the Contention, prior to further low power operation 'or issuance of~as full power. license for Seabrook Station THE'CONTENTICE On' June 22, 1989, .during low power testing at.Seabrook

Station, the operating performance of plant personnel revealed serious deficiencies in training, management control, l~

supervision, communication, and procedures compliance plant operators deliberate'ly disregarded test procedures req i u ring prompt. shutdown of the reactor after a' steam dump val ve' failed open. The; failed valve caused changes in temperatureand pressure .which under start-up test procedures required manual tri p of .the reactor. Senior management personnel,' including the Vic President Nuclear Production, unit-shift su e

pervisor, assistant operations manager, and the operations manager, knew that continued operation violated test procedures Even when repeatedly advised of the violation by NRC inspectors these

, plant personnel continued to willfully violate test pr

ocedures, ignored NRC notifications, and initially refused to shutdown

~the reactor, Subsequently, senior management personnel provided " inaccurate and incomplete information to NRCon the shutdown, refused to acknowledge the seriousness of their proced ural i I } I

~ - _ - - _ i g the reactor and even suggested restart nresponse the NRC non-compliance, of these issues. Exhibit 2 In resolution operating license. without low power i s suspended Applicants' erating procedures underm ne ook has demonstrated Applicants' disregard of opLow power testing at Seabr adequately public safety. ent personnel are not that operators and managem necessary to assure procedures t trained or supervised. gulatory compliance are no operating proficiency and re 50 Appendix B (II). ing low power testing precludes part 10 C.F.R. adequate. Applicants' performance durthere is reasonable assurance that with Commission a finding at present that conformance will operate in ithout endangering the' Seebrook Stationlicensing requirements, or w $50.57(a). 10 C.F.R. regulations, blic. of health and safety of the pu willfully violated the terms ifications by Applicants' operators haveby ignoring repeated not 555.53(d). their operating license See 10 C.F.R. violations. d a low or NRC of test procedure t qualified to be grante presently, Applicants are no pending a full 50.57. 10 C.F.R. further full power license. olution of these matters, and a full power evidentiary hearing and res in suspended, low power testing should rema t issue. operating license should no 4 JURISDICTIOtiall issues raised in the over (5/25/89). This Board has jurisdiction at 5-9 slip op. ALAB 916, 29 NRC Content ion. 3- ~-_ ~~~~ _ ~ - - - - - - _ "-N

.) " .IEE CONTENTION IS TIMELY Pursuant to S189(a) of the Atomic Energy Act (Act), and as a consequence of the NRC's license suspension of Applicants' low powerLlicense, Exhibit 2, Interveners are entitled to a hearing /prioruto full power licensing.on their Contention and do not-have to meet any additional procedural requirements-such as successfully moving to reopen the record. 1. The Act provides filn anvioroceedino under this chaoter. for the orantino, suscendino, revokino, or amendino o,L_ goy liCRnig or construction permit, or application to transfer control, and in any proceeding for the issuance or modification of rules and regulations dealing with the' activities of licensees, and in any . proceeding for the payment of. compensation, an award or royalties'under Sections 2183, 2187, 2236(c) or 2238 of this title, the Commission shall crant a hearino upon the request of any persons whose interest may be affected by the proceeding, and shall admit any such person as a party to such proceeding. Act, S189(a); 42 U.S.C. S2239(a). (Emphasis supplied.) NRC's suspension of Applicants' low power operating license affords Interveners a statutory hearing right, since the suspension is expressly designated as a " proceeding" giving rise to a hearing under the Act.I' Cf. Commonwealth of 1/ The designated " proceedings" that trigger hearing rights under the Act have been strictly construed. Thus, for example, while an " amendment" to a license gives rise to a hearing, a decision by NRC to lift a license suspension, and reinstate the license, does not. [T]he extension of the term of a license constitutes a, L license " amendment" within the meaning of Section L' 189(a). Petitioners were therefore entitled to a l hearing. (footnote continued) l .1 4

} n. ..,y Massachusetts v. NRC, No. 88-2211 (First Cir. 6/29/89) slip op-p. 13 ("Because (license) reinstatement is not listed as a. specific action giving rise to a hearing, no hearing right is created by'S2239(a).") Exhibit 6. Having. commenced a $189(a) proceeding by suspending the . license, NRC must permit public participation in the proc on all issues relevant.and material to the suspension UCS_2. -HEC, 735_F.2d 1437, 1443 (1984) ("once a hearing on a licensing proceeding has begun, it must encompass all material factors . bearing on the licensing decision raised by the requester. such_ proceedings as are begun shall be formal, public hearings"). As defined by the Ccmmission, adequate training ~and -qualifications of plant operators and senior managementare a

fundamental and material precondition to licensing and continued operation.

(footnote continued) fall within one of the eight categories setIf a particu the section, no hearing need be granted by the forth in Commission..... Nor do we believe that Congress subsumed within the statutory category of licens suspensions. If it had, that license revocations are implicitly included inthen we should the category of license grants; but Congress found it Because none ofLthe actions specified in sectionnec 189(a) may be said to include the lifting of a rise to the rightsuspension, we conclude that such action does not license to a hearing. give EAD_1yis_Qhispo Mothers for Peace v. NRC, 751 F.2d 1287, 1312 and 1314 (D.C. Circuit, 1984). \\ I a

8 L Before it may issue an operating license for a nucl . power plant, the~ Commission is required-by its ear regulations to find that conformity with the application as amended,"the facility will oper in of the Commission"' "[t'jhere is reasonable a the that the activities authorized by the operati license can be conducted without endangering th ng health and safety of the public"; and "[t]he ap e is technically... activities authorized by the regulations." qualified to engage i plicant LO_thise findinos reactor coerators;have-beenis a_Aeterminati.Qn_that_the_ plant'sEnnec 11CADied.' tiQihers oroneriv trained and citing 10 C.F.R. for Peace, -(Emphasis supplied.)SS50.57(a)(2),EupEa, 751 F.2d at 1309, (3) and (4). Commission regulations. establishing adequate tr i a ning and qualifications for. plant. personnel as a precondition t o licensing, ERE e.g. 10 C.F.R. Appendix B (Quality Assurance Criteria), 10 C.F.R. Part 55 (operator licensing) 550.47 (technical qualification)), are in turn mandated by Congress. 42 U.S.C. $10226. (S306, Nuclear Waste policy Act, quoted Infra). As a matter of federal statute, and commission regulation, personnel training and qualifications are mate i r al to a license grant, continuing operation and, in this case, the license suspension flowing from serious deficiencies in the competence of Seabrook operators and management personn l e. Applicants' performance during low power testing ~~ demonstrated plant personnel are not trained or qualifi ed to meet minimum licensing requirements. NRC therefore suspended Applicants' low power license, indicating that the prior gtant of.a low power license was premature and not consistent with public safety. Interveners are entitled to a hearing on all _ _ _ _ _ - _ - - - _ _ - _ _ _

-issues raised by the Contention, including 1) the circumstances surrounding the suspension, 2) the lack of training and ' qualifications of plant operators and management personnel. 1). the: lack of adequate operating procedures, 4) whether Applicants meet minimum regulatory requirements to operate-the facility at low or full power in view of plant personnel incompetence,.and willful disregard of test procedures and NRC directives.A' In addition,-whether or not NRC elects to characterize the events of June 22 as a. license " suspension" is not dispositive of' Interveners'-rights to a hearing.1# Edison, by its voluntary shutdown and continued cessation of operations, made.it: unnecessary for the NRC'to revoke formally its license. The NRC nonetheless stated clearlv'and consistent 1v that it Equld not allow Pilcrim to restart until it was satisfied with Edison's improvements. IhE_fjL C theA thE_NRC did not call its decision to restart a " reinstatement" of the license is not controlling. Columbia Broadcasting System. Inc. v. United States, 316 U.S. 407, 416 (1942) ("the particular label placed upon [its action]~by-the Commission is not necessarily 1/ Unlike Commonwealth of Mass., sugIA,-Interveners here do not challenge a decision by NRC to lift a suspension,' and reinstate a-license, which would fall outside the " specific action.giving rise to a hearing" under the Act. Id. at p. 13. Indeed, NRC has made no such reinstatement decision'in this case. Pursuant to Interveners' statutory hearing rights, Interveners rather seek to participate with NRC, in a public hearing,-in assuring the public safety through a thorough exploration of the'above-referenced issues. l l 3/ Applicants themselves, however, have admitted that NRC ordered a " temporary suspension" following the' reactor shutdown. Exhibit 5. - 7

u L conclusive, for it is the substance of what the Commission _has purported to do and has done which decisive."). is Ibgis__ubstance of the_NRC's that Edison could not action _ygg ooerate license until_ th.tLEJLQ_a Uowed Pilorim oursuant t o. Lta it_to.d&__12 The decision allowing this was a reinstatement of the right to. operate pilgrim pursuant had been in effect prior to the shut-down.to the license that C.omc.2rlwe a 1 th._qL fiaLL Sup_f.A, S1ip op-pp. 11-12. o4 l -(Emphasis supplied.) Exhibit 6. It is undisputed that as a consequence of the events of June 22, 1989, Seabrook Station was shut down and operation cannot resume without appropriate corrective actions and .without prior-approval by the NRC Regional Administrator for Region 1. Exhibit 2. The " substance" of NRC's actions therefore constitutes a suspension of Applicants' low power license, and gives rise to hearing rights on all issues raised -by the' Contention. 2 '. Independent of these hearing rights arising from the suspension, Interveners are entitled to a hearing under S 189(a)'since Applicants' failures in training, management, and operator procedures during low power testing are relevant and material-to the orant of a full power license. Egg S 189(a); Mathers For Peace, 751 F.2d at 1309, quoted supra. Consistent with the Congressional mandate, 42 U.S.C. 10226, the Commission requires adequate operator training, management, procedures and performance as a precondition to full. power licensing. Recent events revealed by low power testing,

however, demonstrate that Applicants have not met these regulatory l 1

!^ t 5 requirements. Indeed, NRC has suspended further plant . operations at ans level of power until there'is full examination and corrective action concernin ' Bothlby' regulation, g the shutdown. therefore,.the Commission has determiand through s \\ ned that the training and . qualifications of plant personnel are material and rel the grant of a. full power. license for S evant to eabrook Station. Interveners are therefore entitled to raised by the Contention..

a. hearing on these issues UCS v.

NRC, (1984).

735.F.2d 1437, 1443 Interveners have-previously presented of their right argument in support to litigate' events, such as-those pre the Contention, sented by which arise out of low power testing 3. .That argument Exhibit is incorporated by reference in:the proposition that and is grounded low power testing is material 1to full power licensing and that the Commission h as_made successful completionfof low' power testing a preconditi on to the receipt of'a' full power license. Low power testing has'now specific defects in personnel training and revealed qualifications. UCS + c 9- }_ __ __ - _ = - - - - -

en v -

19 1

L therefore mandates that Interveners be afforded " formal, public hearings" on these. issues. UCS at 1443.S' E' For-this reason, Applicants'cannot avoid public scrutiny, through the hearing process, even if-they elect to forgo further testing pur.suant to the law-power license, where the same issues of concern raised in the Contention are material to full power' licensing.E' 4/ prior to the events involving the manual reactor trip, this Board declined to reach the issue of Interveners' hearing rights involving low power testing, apparently on grounds the issue was not ripe. See T. 28287. -5/ In response to MAG's. prior motion for. hearing on low power issues, Exhibit.3, Applicants proffered the argument that "{N]ever has the successful completion of low power testing been a" legal. prerequisite to issuance of a full power license." ' Applicants' Response to Motion of the Massachusetts Attorney General to Hold Open the Record pending Low power Testing and the Required Yearly Onsite Exercise and for Other Relief, (June 12, 1989) p. 5. Applicants' low power license is now suspended. prior to startup of the unit, to ADI level of power, adequate examination and corrective action involving personnel and procedural failures during low power testing must be. undertaken. Exhibit 2. Necessarily, NRC has determined in this case that the subject defects in low power testing are material to full power licensing and operation. This is consistent with well established NRC views. Egg Exhibit 3, pp. 3-6. Applicants' quibble over " legal prerequisite" should be rejected. E/ As UCS provides: "Thus, unlike in Bellotti, here the Commission has removed from the licensing hearing consideration of evidence that it considers relevant to a material issue in the section 189(a) proceeding as it has defined that issue. Not only is this different from the NRC position we condoned in Bellotti, in fact, it is in tension with Belloiti's conclusion, in the context of a license amendment, that "public participation is automatic with respect to all [section 189(a)) Commission actions that are potentially harmful to the public health and welfare." UCS 735 F.2d at 1443 citing Bellotti_y_._ (footnote continued) _ - - -

h p i REOPENING THE RECORD The. Contention provides the requisite basis and specificity, asserting that plant personnel are not adequately trained, or managed, and have willfully violated operating procedures, and-ignored NRC notifications of test procedute violatior.s. Applicants lack adequate procedures to reasonably assure that the facility will operate in accordance with regulatory requirements. These allegations principally arise ftom the events involving the Applicants' low power license suspension. Since the basis for Interveners' request for hearing flows 1from their statutory rights conferred under the Atomic Energy Act, those rights cannot be burdened with the higher standard imposed by NRC regulation for reopening the record. See'10 C.F.R. 52.734. we cannot conclude that the occortunity to seeA teopenino was an adecuate substitute f o r t he hear i rty_ cuaranteed petitioners as a matter of richt under section 189(a). In order to obtain reonenino, petitioners were recuired to show that they costesEgd_ new evidence whch was timelv; material in the sense that it would have resulted in a different outcome _ bad it been known earlier, and safety-significant. None (footnote continued) URC, 725 F.2d 1380 (D.C. Cir. 1983). Having determined that the management, training, and related issues are material Fn full power licensing, the Commission, unlike in Bellotti, cannot limit the scope of the " proceeding" to preclude hearing on these matters. Similarly, unlike Bellotti, where NRC imposed a license amendment to enhance safety, the present Juspension at Seabrook suggests the decision to grant the low power licer4se initially was unfounded and incorrect. Applicants presently do not meet even the minimum regulatoty requirements for operation and cannot reasonably assure the public safety. 10 CFR 50.57(a). The public therefore has a fundamental interest in participating in a hearing on the issues involving personnel and training, that are material to the original grant of the low power license, the suspension, and to any further operation of the facility at any level of . power.

w JV ~ of these'three criteri_a_jtpolies to reouests for a hearino under section 189(a). Under the latter -provision parties need only show that their~" interest may be affected" by a proceeding to bring about une of eight specified types of Commission. At most, parties must show that a particular issue is " material" in order to prevent its exclusion from a hearing under.4 section 189(a); this much our decision in Union of Concet.ned Scientists v. Nuclear Reculatory CgmmiSSi2D establishes. Mothers For Peace, 751 F.2d at 1316. (emphasis supplied). Accord, UCS v. NRC, 735 F.2d at 1443-1444.1# In the alternative, in the event this Board subjects Interveners' contention to the late filed contention criteria, 10 C.F.R. S2.714(a)(1), and/or the reopen the record standard, l'0 C.F.R. 52.734, Interveners' Contention meets these standards in any. event. LATE FILED CONTENTION STANDARD, 10 CFR 6 2.714(a)(1).

1. GOOD CAUSE Interveners' Contention principally arises out of the events of June 22, 1989, when Applicants disregarded operating 2/

The Appeal Board has recently cited the holding in the UCS decision. "[T]he Court also rejected the Commission's argument that-a party's hearing rights were protected because a patty could always seek to reopen the record if the exercise identified fundamental defects in the emergency plans." ALAB 918, (6/20/89) slip opinion p. 13, n.21. In the same opinion,, however, the Appeal Board nevertheless permitted imposition of the late filed criteria for contention admissibility, 10 C.F.R. S 2.~/14 ( a ) ( 1), as " reasonable procedural requirements". Ld. L - Interveners have appealed this decision, and continue to assert that imposition of late filed criteria is an impermissible infringement on Interveners' S189(a) hearing rights. Lt a m__ _m

s .and low power _ test procedures and NRC directives to promptly chut down the reactor. NRC indefinitely halted further operation at Seabrook Station and suspended Applicants' low power license. Important information on these events, including NRC's own report from its Augmented Inspection Team (AIT), has not yet been released. To minimize the anticipated charges by Applicants and the NRC Staff that the Contention filing is not timely, however, Interveners file the Contention based upon the limited information presently available, as d scribed below.E' 8/' This Board previously declined to grant Massachusetts Attorney General's request contentions on-low power testing.to establish

• schedule for filing Tr. 28287 Et 122 In eddition, the Appeal Board recently concluded that onsite exercise contention was not Intervenor's discovery of the information upon which it is based."" tendered promptly upon the 918.(6/20/89) slip op.

p. 16. Ege ALAB Interveners feel compelled to file this Contention atUnder these circumstances, time. -specificity under the regulations..It meets the minimum requirements for basis a this . Interveners anticipate, however, 10 CFR S 2.714(b). parhaps additional contentions willthat additional bases and later be filed based upon NRC's investigation, NRC production of documents responsive to Interveners' outstanding FOIA Request, and information obtained from other sources. response to CAL 89-11 indicates thatFor example, the Applicants' July 12, 1989 a Licensee Event Report will be submitted July 24, 1989. ovcnts of June 22 and 23 - for example,Other issues presented by the hardware issues, wall be raised by subsequent contentions based on information not yet available but which will be made available in the near future. Interveners note that regard.is very much a " Catch 22"the procedural context in this to decide what because they are called upon amount of information is sufficient d tailed information which would augment the contention anda to file an bases will be available shortly. It is for this very reasnn that low. power testing contentions. Interveners sought unsuccessfully a filing deadline f This fi reflects the l judgment that certain issues - training, ling ( are sufficiently supportable at this time while others - and management issues - hardware,. maintenance and licensee interface issues - are net without further information which is expected to be available in some cases in a few days. !a

l L l.O< On June 23,.following media' inquiries to this office, counsel for Mass AG called NRC and was advised that Sebbrook Station had been shut.downLas a result of problems during low power: testing. NRC stated-that the reactor wculd not be . restarted 1until these matters were explored, but declined to-elaborate.in detail. Following' June 23, through the public Document Room, Interveners obtained copies of Applicants' Preliminary Notifications of Event, Exhibit 4. These preliminary notifications provided only a bare outline of the circumstances surrounding,the shutdown and suspension. There is no clear reference to operator delay in tripping the reactor or 't. disregard by. senior plant personnel of NRC notifications of test. procedure violations prior to shut down. Similarly, Applicants' press releases on these events were inconsistent and, predictably, understated. Exhibit 5. The press release of June 22, alleges Applicants performed the shut down "in accordance with the strict technical criteria governing.tha current low power test program." On June 23, however, Applicants conceded that "NRC orders temporary suspension of low power testing. . control room operatots did not strictly follow a procedure in determining when to shut down the reactor." Again, however, this description of events-remained clipped, vague, and no reference given to repeated notifications by NRC of a procedural violation, or that plant l 14 1 1

Fk ye g . personnel knowingly. Operated the plant in contravention of test procedures.. .On.or about June 26, 1989,! Interveners were provided a copy .of the' confirmatory Action letter (CAL) which confirmed the license' suspension. Exhibit 2. The cover. letter to the CAL- [ advised of'an on going investigation by NRC of the June 22. H ' events, but no. substantia) details were provided on the-progress of.theiinvestigation or when it would be concluded. As of JulyL1, 1989, Thomas Murley of NRC suggested in the Boston Globa that NRC's own investigation was still in the preliminaryLstages: That is why we have an augmented inspect' ion team up there.and'are going to sit down with the licensee when they are done with their investigation to try and figure out what happened. As of the date-of this filing, NRC has not released a report on its investigation. On July; 13,.1989,E Applicants released to the press and l NRC its own report on the circumstances of June 22 involving the' shutdown.and suspension of the low power license.1E# See attached. Massachusetts Attorney General's office obtained a copy of the report on July 14, 1989, which provides Applicants' view of the June 22 events. I c E/ That same day, Massachusetts Attorney General requested a copy.of the report from Applicants' counsel and was advised that Ropes & Gray did not have the Report. 10/ See New Hampshire Yankee Response to Confirmatory Action p Letter 39-11 (7/12/89). . a_____-l___________.-_..

r t- .Upon receipt:of the report, Interveners immediately H. h ' consulted with their experts to prepare the Contention, L -Affidavit, and supporting materials. Although Applicants' L ' Report remains incomplete and uncorroborated, Interveners L p believe the report provides sufficient information to meet minimum requirements'for basis and specificity. Interveners have good cause'for filing the Contention on this date.11' 2) -Protection of Interveners' Interest There is no means other than by litigation of the Contention to protect Interveners' interest in ensuring that Seabrook Station operating procedures and management structure are' clearly defined, the operations and management personnel I adequately trained and qualified, and that facility operation at low.or full power will conform to regulatory requirements and reasonably assure the public safety. 10 c.F.R. 50.57(a). It is apparent that the Applicants, based on the NRC Staff's recommendation, were issued a low power license prematurely, and before Applicants were qualified to lawfully operate the plant at any power level. The Staff's error in judgment demonstrates that Interveners must protect their own interests 'll/ While the Contention is primarily grounded upon the above identified information, certain limited additional data, involving prior procedural non-compliance problems, are included as.part of the Contention basis. Sig Joint Affidavit of Gregory C. Minor and Steven C.

Sholly, p.

12, filed herewith..,.

k ( g. through' litigation of the Contention to reasonably assure public safety. 3. Develcoment of a Sound Record Interveners will contribute to the development of a sound i record through litigation of the issues-set forth in'the Contention, filed herewith. This adequately identifies the issues' Interveners seek to raise. ALAB 918 (6/2/89) s1. op. p. 20. In support of the Contention, Interveners will offer .the testimony of expert witnesses Gregory C. Minor and Steven C. 'S h o l l y'. 'The expert qualifications of Mr. Minor and Mr. Sholly are set forth in the attached affidavits and resumes, incorporated herein,.and are summarized as follows. Mr. Minor is Vice President of MHB Technical Associates, and has over 25 years of experience in the design, development, research, start-up testing, and management of nuclear reactor systems. For the past thirteen years, Mr. Minor has been a technical consultant and has participated in a variety of studies addressing nuclear facility management, and safety issues for various organizations, including the Department of Energy /Sandia National Laboratories, the Swedish Government, and the offices of several states' Attorneys General. Mr. Sholly is an Associate Consultant for MHB Technical Associates. For the last nine and a half years, Mr. Sholly has been engaged in analyzing technical nuclear safety, management,,. 17 -

O' design, construction, and regulatory issues and providing l technical. advice to a number of stateLand local governments. He has previously participated as an expert witness in proceedings before Atomic Safety and Licensing' Boards involving the Indian-point and Catawba facilities, and has presented testimony;before the United States Congress on nuclear safety issues. The testimony of Mr. Minor and Mr. Sholly, and the Applicants' admissions as set forth'in-Enclosure 4 to New Hampshire. Yankee. Response to Confirmatory Action Letter 89-11, and the Confirmatory Action Letter 89-11, filed herewith, may1 be summarized as follows: a. During low power testing at Seabrook Station, Applicants operated the facility in knowing violation of the test. procedures, the facility license application as amended, the-terms of their. operating license and Commission regulations, when operators failed to shut down the reactor after a steam dump valve failed open. b. Despite repeated notifications from NRC inspectors in the control room of the test procedure violation, plant personnel continued to operate the reactor and initially did not shut down.

fg, L b' p p c. Subsequently, senior. management personnel \\ n. provided inaccurate and incomplete information to NRC on the i shut down, did~not communicate to NRC their recognition of the seriousness'of the procedural non-compliance and even-suggested restarting.the1 reactor prior to resolution of these issues. E d. Commission quality assurance regulations require procedural compliance for activities affecting quality, which include operation of a reactor, as well as for indoctrination and training of; personnel, such as operators, as necessary to assure that suitable proficiency is achieved. 10 C.F.R. Part 50, Appendix B(II)(V). Applicants violated the Commission's quality e. assurance regulations mandating procedural compliance, it is apparent that the training program was not effective in this ' instance, and some improvement in the training program is essential if similar violations are to be prevented in the future. f. There is a strong safety assurance aspect o. the Commission's quality assurance regulations. Failure to follow procedures carries with it significant safety implications which cannot be ignored. . Prior procedural non-compliance problems mean the g. procedural non-compliance is not an isolated event, but is part of.a pattern of procedural non-compliance. -. - - - - - -

...,O, }, k" (: h. The following training and management procedures are notiadequate or in compliance with Commission regulations: . Operations Command and Control policy,' Operations. shift crew-procedures, management training and procedures for operation . oversight,' policy on procedural adherence, NHY post Trip Review ' process. 7 4 Whether Interest Represented By Existina parties In this. proceeding, no other party has raised or-is raising the issues set forth in the Contention. 5. ' Absence of Delav -presently, admission of the Contention will not delay the proceeding, in.that Applicants are barred from operating the facility at any power level until the issues raised by the shutdown and license suspension are resolved. Exhibit Admission will' expand the issues to be addressed through public hearing before this Board, but will not expand the issues which NRC has determined must be resolved prior to operation at any power level. REOPENING THE RECORD (2.734) 1. This motion arises out of the shutdown and license suspension on or after June 22, 1989. The motion is timely for the-reasons set forth above. In a4dition, the Contention presents ~ exceptionally grave issues which require hearing to assure.the public safety. l V _________-____D

L., b L ? 2. Through the Contention, this motion raises fundamental and significant safety and environmental issues on the training and qualifications of Seabrook operators and management personnel. To reasonably assure tae public safety, these E issues must be fully resolved prior to plant operation. 10-C.F.R. 50.57. The NRC~ Staff itself has recognized the significance of Interveners' concerns by suspending Applicants' . low power license. This suggests that even the Staff recognizes that Applicants' willful disregard of operating _ procedures, and NRC directives for shutdown presen's significant safety issues. Applicants' subsequent failure to L fully and fairly report the events to NRC underscores these ' legitimate concerns for public safety. As determined by NRC, the appropriate remedy for these safety related violations is suspension of the low power license. Applicants are barred from further operation at any powe level until complete resolution of these issues. This is consistent with the Commission's regulatory requirements where suspension should be ordered to protect the public safety.12# ll/ "A license. may be revoked, suspended, or modified, in whole or in_part, for any material false statement... in ..the supplemental or other statement of fact required by the Applicant' or because of conditions revealed by the application for license or statement of fact or any report, record, inspection, or other means which would warrant the Commission to refuse to grant a license on an original application or for failure to construct or operate a facility in accordance with the terms of the construction permit or license. or for violation of, or failure to observe, any of the terms and. provisions of the act, regulations, license permit, or order of the Commission." 10 C.F.R. 50.100. ___:_=-__

.w w As previously' cited in this motion, the Commission has 7 . promulgated extensive regulations to assure adequate. operator i and' plant personnel qualifications as.a precondition to licensinng. I' t. Every applicant for an operating license is requited to include, in its final safety analysis report,. information pertaining to the mangerial and administrative controls to be used to assure safe operation. The authority and duties of persons and organizations performing activities affecting the safety related functions of structures, systems, and components shall be clearly established and delineated in writing. l e a a The program shall provide for indoctrination and q training ~of personnel performing activites affecting quality as necessary to assure that suitable proficiency is achived and maintained. 10 C.F.R. part 50, App. B. (a) pursuant to $50.56, an oper: ting license may be issued by the Commission, up to the full term authorized by $50.51, upon finding that (4) The applicant is technically and financially qualified to engage in the activites authorized by the operating license in accordance with the regulations in this chapter. 10 C.F.R.-$50.57(4). (a) Requirements for the approval of an initial application. The Commission will approve an initial 3 application for a license pursuant to the regulations in this part, if it finds that -- 1 22 - =

a l ' ( 2 )' -Written examination and operating test. The applicant has passed the requisite written examination and operating testiin accordance with S555.41 and 55.45-or 55'.43 and 55.45. -These. examinations and tests determine whether.the applicant for an -A-operator's license has learned to operate a facility competently and safely, and additionally, in the case of a senior operator, whether the applicant has -learned to direct theLlicensed activities of licensed. operators competently and safely. 10.C-.F.R..555.33(a) (b) Any license may be revoked, suspended, or modified, in whole or in part: a a m (3) For willful violation of, or failure to observe any of the terms and conditions of the Act, or the license, or of.any rule, regulation, or order'of trhe Commission, or (4).For any conduct determined by the Commis.? ion to be a hazard to safe operation of the facility. 10 C.F.R. 555.61. LThese regulations are responsive to the Congressional mandate that the Commission assure the public safety by providing for adequate training and qualifications for nuclear plant personnel. The Nuclear Regulatory Commission is authorized and directed to promulgate regulations, or other appropriate Commission regulatory. guidance, for the training and~ qualifications of civilian nuclear powerplant. operators, supervisors, technicians and other appropriate' operating personnel. Such regulations or guidance shall establish simulator training requirements for applicants for civilian nuclear powerplant operator licenses and for operatur requalification programs; the requirements governing NRC administration of requalification examinations; requirements for operating rests at civilian nuclear- - -

\\ powerplant simulators, and' instructional requirements j for civilian nuclear powerplant licensee personnel training programs. Such. regulations or other regulatory guidance shall be promulgated by the y Commission within the 12-month period following j January 7, 1983, and the. Commission within the 12-month period following January 7, 1983, shall

submit a report to Congress setting forth the actions the Commission has taken with respect to fulfilling its obligations under this section.

42 USC $10226 (S306,. Nuclear Waste policy Act).11/ Bolb Congress and the Commission therefore have recognized that the training.and qualifications of plant operators and management were of such public safety significance as to warrant' specific legislation and rulemaking. This Board should similarly conclude that these issues, raised by this Contention, are significant safety issues. Both the license suspension and Commission regulations therefore plainly establish that the issues presented by the Contention are significant safety issues challenging the adequacy of the training, management, and procedures involving plant operators and senior management. -3. The issues raised by this Contention were not adjudicated before any Seabrook licensing board. Had the deficiencies in the Applicants' operator training and management review and control processes been known prior to licensing, it would have been likely that corrective actions would have been 11/ Following TMI, Congress recognized that adequate plant personnel training and qualifications were essential to public safety. "(L)ack of adequate power plant operator training had played a very significant role in the inability to control" the accident at TMI. Statements of Senator Weiker, Congressional Record, Vol. 128,

p. 515643 (Daily Edition, 12/20/82).

y:1 ,o l c required prior to that licensing just as the NRC has determined that such corrective actions are necessary prior to any further operation. 4 CONCLUSION

i. ;

For all the reasons set.forth above, the Interveners request that this Board admit the Contention for litigation and hold a hearing thereon. Respectfully submitted, JAMES M. SHANNON ATTORNEY ~ GENERAL COMMONWEALTH OF MASSACHUSETTS s J ^J By: John Traficonte Chief, Nuclear Safety Unit Matthew Brock Assistant Attorney General Nuclear Safety Unit .Public Protection Bureau One Ashburton Place Boston, Massachusetts 02108 (617) 727-2200 Dated.: July 21, 1989

EXHIBIT 1 1 9

7 L. l [. l-EXHIBIT I INTERVERORS' CONTIRTION FOLLOWING LICENSI_$MS.PIBSIM Events during Low Power Testing at Seabrbok Station on June 22, and 23, 1989 demonstrate that Applicants' plant operators, and management personnel, are not adequately trained or qualified, and lack adequate managerial and administrative procedures and controls, to operate the facility, at any level of power, in conformity with the facility application as amended, the provisions of the Atomic Energy'Act (AEA), and the rules and regulations of the Commission. There is no reasonable assurance that the activities authorized by an operating license can be conducted without endangering the health and safety of the public, or can be conducted in compliance with 10 C.F.R. Chapter 1, all as required by 10 C.F.R. $50.57. Low power testing also has demonstrated that Applicants have violated, and presently cannot meet the following preconditions for licensing to operate at any power level: 1) 10 C.F.R. Part 50, Appendix B which requires Every applicant for an operating license is required to include, information pertaining to the manaaerialin its final safety analy administrative and controls to be used to assure safe operation. e a a The authority and duties of persons and organizations performing activities affecting the safety related functions of structures,

systems, be clearly established and delineated in writing.and components sha a

a C_.___n__.._.___.

u Activities _affecting quality should be prescribed by ' documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, proceudres, or drawings shallcinclude appropriate quantitative or qualitative acceptance criteria for determining that imporant activities have been satisfactorily accomplished. T he_ p tog r a nt_ s h alL_proxisl e_.Lar _l ndac trina.tio n_a n d tralnino of oersonnel cerformino activities affecting ouality as necessary to assure that suitable proficiency is achieved and maintained. 2) 10 C.F.R.'S50.34(b)(6) which requires: (b) Final safety analysis report. Each application for a license to operate a. facility'shall include a final safety analysis report. The final safety analysis report shall include information that describes the facility, presents the design bases and' the limits on its operation, and presents a safety analysis of the structures, systems, and components and of the facility as-a whole, and shall' include the. following: (6) The following information concerning facility operation: (i) The applicant's organizational structure, allocations or responsibilities and authorities, and personnel qualifications requirements. (ii) Manaaerial and administrative' controls to be used to assure safe coeration. Appendix B, "Quali'v Assurance Criteria for Nuclear Power plants and Fuel Reprocessing plants," sets forth the requirements for such controls for nuclear power plants and fuel reprocessing plants. The information on the controls to be used for a' nuclear power plant or a fuel reprocessing plant shall include a discussion of how the applicable requirements of Appendix B will be satisfied.

3)

• 10 C.F.R. S55.53(d) which requires Each' license contains and is subject to the following conditions whether stated in the license or not:

L (d) The lic_ense ili_Eubj ect to, and the licenste shall observe, all aoolicable rules, regulations, and order _s of the Commission. BASIS A. On June 22, 1989, during low power testing at Seabrook .1 Station, plant personnel revealed serious deficiencies in their training and management procedures. Many plant personnel, including Operators, and New Hampshire Yankee Management, deliberately disregarded test procedures which required shut down of the reactor, when pressurizer levels fell below test criteria due to a failed valve. New Hampshire Yankee Response to Confirmatory Action Letter 89-11 (7/12/89) Enclosure 4, pp. 4-6 (hereafter identified as " Response", Attachment B.) Under start up test procedures, operators were required to shut down the reactor. Operators were aware of these procedures, but ignored them. Senior supervisors and management personnel, including the unit shift supervisor, operations manager, and the assistant operations manager knew that the reactor was operating in violation of test procedures, and'each, individually, had authority to order shut down. Each, however, failed to exercise that authority. Even when repeatedly notified of the violation by NRC inspectors in the control room, plant operators and management continued to operate the facility in violation of test procedures and initially refused to shut down the reactor. j 3-D

plant personnel's willful disregard of test procedures as well as NRC notifications of a violation, demonstrates serious deficiencies in training and management; and raises serious safety concerns due to the poor judgment of plant personnel, their unwillingness to admit error, and refusal to properly and timely respond to abnormal plant conditions. Subsequently, senior management perconnel provided inaccurate and incomplete information to NRC on the shut down, refused to acknowledge the seriousness of the procedural non-compliance, and even suggested restarting the reactor prior to resolution of these issues. Applicants' lack of candor in initially reporting on the shutdown to NRC raises additional, and fundamental, concerns as to Applicants' corporate character and management capability. Applicants' failure to comply with test procedures and the subsequent improper management actions, were of significant concern to NRC. Response, Attachment to Enclosure 4, p. 11. NRC therefore suspended Applicants' low power license. That suspension remains in effect and suggests significant safety concerns are presented by these events. Egg 10 C.F.R. S50.100. These events demonstrate that there is no reasonable 1 assurance that plant operation can be conducted in conformance i with Commission regulations and without endangering the health and safety of the public. 10 C.F.R. $50.57(a)(3). l l _4 l

U B. In their Response, Applicants have admitted;that the I- - procedures involving training, qualifications and management must be fundamentally reevaluated and revised prior to operation. Applicants thereby implicitly acknowledge that these primary safety-related procedures, that affect the quality of plant operation, presently are not adequate to reasonably assure public safety. These include: 1) The Operations Command and Control policy is not adequate, and did not function as intended. This was evidenced by the willful disregard by plant personnel, including operators'and management, of test procedures and NRC notifications to shutdown the reactor. Egg Response, Encl. 4, pp. 5-6. 2) The training and management procedures for the Operations Shift crew are not adequate. This was demonstrated when each member of the crew failed to respond to NRC notifications, or exercise his individual authority, to timely shutdown the reactor. Id. 3. Management procedures for oversight of plant operation are not adequate, and lines of authority and responsibility for management personnel are vague and confusing. This was demonstrated when senior personnel, including the Operations Manager, Assistant Operations Manager, and Unit Shift Supervisor, failed to override operators to _ _ _ _ -

i i promptly order plant shutdown in conformance with test procedures, id., or forcefully press their superiors for plant I ' shutdown. 4. The Station and Operations Management policy on- ? procedure adherence is unclear and confusing. . Clarification is ] 1 necessary to clearly delineate when deviation from procedures is acceptable, id. at p. 10, and when conformity to procedures is mandatory. This was demonstrated when certain plant personnel purportedly " misunderstood" the mandatory procedures ~ for plant shut down merely to be guidance, id., and disregarded l NRC notifications. In fact, since the same operators claim to have tripped the reactor in accordance with this test procedure. which.at least in that instance was not treated merely as " guidance", it appears that the operators simply chose to selectively follow operating procedures. l S. Present management training programs are not adequate. This was demonstrated when management failed to i exercise authority to order a reactor shutdown, or direct l operators to take this action. In addition, training must be provided with respect to all procedures identified for revision (see #1-4, incorporated by reference). ~6. The existing NHY post Trip Review Process is not adequate. Id. at 24. This was demonstrated when senior management provided wholly inaccurate and incomplete information on the circumstances surrounding the shutdown and - _ _ _ _ _ - _ _ _ _ -

k violated reporting requirements in 10 C.F.R. 50.72. Id. at 26. NHY improperly sought to' justify its violation of test procedures to NRC, and sought to restart the plant prior to resolution.of the problems revealed by the shutdown. Id. L C. In its application for an operating license, Applicantsfmade certain statements and representations as to the " managerial and administrative controls to be used to assure safe operation", 10 C.F.R. 550.34(b)(6), and as to the 1 -training program for " personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintaineJ". 10 C.F.R. part 50, Appendix B (II). Egg sections of FSAR, Attachment C hereto. Based, in part, upon these statements and representations in the application -which were incorporated into the low power license as preconditions to operation, 10 C.F.R. $50.57(a)(2), Applicants were issued an operating license. The events of June 22 demonstrate that Applicants have failed to comply with these statements and representations, and have violated these conditions of their license as follows: 1. FSAR S13.2.l(a) which requires licensed operator. training programs "to ensure that each individual can safely and ef fectively per 'orm various assignments to train a staff to operate and maintain the units safely, dependably, and economically". =___2----_

2.- FSAR $13.2.2 which requires a'" comprehensive J training program" for technical and management. staff "in vari'ous disciplines necessary to ensure that each can safely and effectively perform his' assignments".

3..

FSAR S13.5.1.3(25) which requires that testing during preoperational and initial operational' phases "is performed in accordance with written test procedures which incorporate or reference the requirements and acceptance limits contdined in applicable documents". The. failure of many operators and management personnel to comply with test procedures and NRC notifications for plant shutdown, or to accurately report events to NRC following shutdown, demonstrates the pervasive and fundamental defects in the cited programs and procedures. D. As further basis for this Contention, Interveners rely upon, and incorporate by reference, the affidavit of Gregory C. Minor and Steven C. Sholly, filed herewith. (Attachment A hereto). E. As further basis for'this Contention Interveners Reply upon, and incorporate by reference, Enclosure 4 (without Attachments) to Applicants' Response, filed herewith, Attachment B hereto. - _ _ _ - _ - _

(; i r a EXHIBIT B 4 %~..-_.-.:

t I i i 1 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD 3efore the Administrative Judges: Ivan W. Smith, Chairman Dr. Richard F. Cole Kenneth A. McCollom ) In the Matter of- ) Docket Nos. 50-443-OL ) 50-444-OL PUBLIC SERVICE COMPANY ) (Off-Site EP) OF NEW HAMPSHIRE, _E T_ _A _L. ) ) (Seabrook. Station, Units 1 and 2) ) May 31, 1989 ) MOTION OF THE MASSACHUSETTS ATTORNEY GENERAL TO HOLD OPEN THE RECORD PENDING LOW POWER TESTING AND THE REQUIRED YEARLY ONSITE EXERCISE AND FOR OTHER RELATED RELIEF e INTRODUCTION: GENERAL OVERVIEW OF THIS MOTION The Massachusetts' Attorney General (" Mass AG') moves that this Board hold open the record in the Seabrook full-power proceeding pending two events both of which are material and relevant to the issuance of a full-power license. These events i are: 1) the testing presently contemplated pursuant to a recently issued low-power operational license; and 2) the require'd yearly onsite exercise presently scheduled for the go )Y u + MvCv1~oey0~ t/fy. Y o

seek of September.25, 1989. In addition, the Mass AG reques s that this_ Board assert its jurisdiction over any litigation arising from these two material events, schedule a pre-hearing conference pursuant to 10 CFR 2.751a, 2.752 and 2.714 at which a' schedule for the filing of contentions may be set, and permit the Mass AG (and other Interveners who may be so inclined) the necessary access and observational opportunities to enaole the timely formulation and filing of contentions. -This motion addresses the following issues: A. ' Jurisdiction of this Board B. Materiality of Low-Power Testing to a Full-Power License C. Materiality of the September 1989 Onsite Exercise to a Full-Power License D. Extent and Nature of the Hearing Rights that Attach to all Issues Material to a Full-Power License 1. Basis for Request that Record on Full Power be Held Open 2. Need for a Schedule in Light of the Commission's Catawba holding l E. Access is Necessary to Provide Interveners Meaningful Right to a Hearing on these Material Issues 1. Access to Relevant Documents 2. Observational status F. Need for Expedited Disposition of this Motion and Request for Schedule for Briefing and Argument ARGUMENT 'A. JURISDICTION OF THIS BOARD In light of recent guidance from the Appeal Board there can be-no doubt that this Board has general jurisdiction "over all l LL__

Y. ' l. l matters pertaining now or in the future to the application for a license to operate Units 1 and 2 [ sic] of the Seabrook Station...." ALAS-916, May 24, 1989 Appeal Board Memorandum and Order at 6 of slip opinion (emphasis supplied). (Attached as Exhibit 1). The issues raised by this motion and the matters sought to be litigated by the Mass AG pertain directly to the issuance of a full-power license and as such are within the jurisdiction of this Board. B. MATERIALITY OF LOW-POWER TESTING FOR A FULL-POWER LICE In what follows the Mass AG sets forth in some detail the basis for his claim that low-power testing is an event (or series of events) that is material and relevant to the determin.ation by the Commission to issue a full-power license. 1. The testing, authorized in this case pursuant to a low-power license, is required by the Commission's regulations. An application for an operating license must set forth "[p]1ans for preoperational testing and initial operations." 10 CFR 50.34(b)(6)(iii). These plans are set forth in the FSAR and are reviewed by the NRC Staff pursuant to the Standard Review plan and several Regulatory Guidts. See Seabrook SER, NUREG-0896, Maren 1983, Chapter 14 (Attacned as Exnicit 2). This test program must be inclusive enough to " demonstrate that structures, systems, and components will perform satisfactorily in service' (SER at 14-4), and the results must be such that the ' successful completion of the program will demonstrate the functional adequacy of plant structures, systems and

components." Id. (Emphasis supplied.) See also 10 C.F.3 50.56, 50~.57 and Part 50, App. 3, XI and Part 50, App. A. GDC 1, 14, 18,.21, 30, 31, 32, 37, 40, 43, 46, 53 and 54. (All of which require testing of specific systems'and components.) It is on the basis of this testing program, inter alia, that the Commission will make the requisite finding that the Seabrook " facility will operate in conformity with the application as amended, the provisions of the'Act, and the' rules and regulations of the Commission." 10 C.F.R. 50.57(a)(2). Sjya also 50.57(a)(1) and (3). 2. Successful completion of an adequate low-power test is a precondition for the issuance of a full-power license-in i-those circumstances in which that testing is authorized under a separate' testing license, a. As the Appeal Board has noted: Low-power testing is a normal, necessary and . expected step in the life of every nuclear plant. This is true whether such testing is planned under the authorization of a separate fuel loading and low-power testing license. or scheduled as the first step toward operation under the authority of a full-power license. Pacific Gas & Electric Co. (Diablo Canyon Nuclear Power Plant, Units 1 and 2), ALAB-728, 17 NRC 777, 794-95 (1983). Because the Commission has determined that in this case low-power testing should precede the issuance of a full-power license, the successful completion of that testing itself becomes an issue material and relevant to that full power license. If this were not the case, then the Commission's statements

-n i L regarding'the relationship between a further delay in low-power testing and a concomitant delay in full-power licensing wouli be senseless.and unsupported.

See, e.g., CLI-39-08, May 13, L

1989, slip opinion at 27 b. Successful completion of an adequate low-power testing program is also a precondition to full-power licensing cecause of the purposes which low-power testing serves. Specifically, low-power testing should " allow the early discovery and correction of unforeseen but possible problems which may prevent or delay full-power operation. Long Island Lighting' Company, (Shoreham Nuclear Power Station), CLI-85-12,~21 NRC 1587, 1590 (1985)(emphasis supplied) (Attached as Exhibit 3). Thus, low-power testing may well reveal " undiscovered design and construction defects." 47 Ped. Reg. 30232, 30233 (July 13, 1982). 3. The Commission itself views preoperational and low-power testing as material to its full-power licensing decision. i See portions of NRC's Brief For Respondents, dated July 1983 filed with the D.C. Circuit in UCS v. NRC, No. 82-2053, 735 F.2d 1437 (1984) attached as Exhibit 4.

However, the NRC has asserted (and no doubt will again assert) that such "preoperational tests, low-power tests and power ascension tests are not the subject of Licensing Board hearings."

Respondent's Brief at 15 n.7. However, this position is no longer. tenable in light of the D.C. Circuit's decision in UCS

v. NRC.

The Court held that:

{ Once a hearing on a licensing proceeding is begun, it must encompass all naterial factors bearing.on the licensing decision raised by the requestor. 735 F.2d at 1443. (This case is attached as Exhibit 5.) 3ecause the successfal completion of an adequate low-power test progran is material and relevant to the issuance of a full-power license, the Mass AG has a right to litigate that testing.1 1/ As is clear from a review of UCS v. NRC and the relevant portions of the NRC's brief attached as Exhibit 4, the Commission defended its challenged 1982 rulemaking setting emergency exercises outside of the adjudicatory hearing process while maintaining such exercises as material to a licensing decision by comparing them to preoperational testing which the Commission asserted is similarly material but not litigable. In developing this comparison, the Commission expressly highlighted the extent to which both emergency exercises and preoperational testing involve ~ issues of judgment concerning complex matters. In fact, the Commission noted that: The full scale exercise is analogous to the evaluation of plant systems at operating power levels during power ascension testing. Where problems are discovered, they.are addressed before full commercial operation is achieved. On the other hand, the applicant's emergency plan is akin to the design of the facility.. Pre-operational testing is no less crucial to plant safety than the full-scale exercise is crucial to measuring emergency preparedness. 48 Fed. Reg. 16691, 16692,.16694 (April 19, 1983). The Commission no doubt believed diat since preoperational testing assertedly is not litigable and emergency exercises are analogous to-that testing, the Court would find that such exercises need not >be litigated. The Court, of course, ruled directly contra and held that the exercises are material to licensing and, therefore, litigable. Although unchallenged from that day to this, the NRC's assumption that it can by fiat exclude from litigation the low-power testing program which it acknowledges is material to a full-power license directly contradicts the decision of the Court in UCS v. NRC. For the same reasons the NRC argued to the D.C. Circuit that preoperational testing is analogous to emergency exercises and.should be treated similarly, the Commission cannor exclude such testing from the adjudicatory process without contravening UCS V. NRCo

l L t C. MATERIALITY OF THE SEPTEM3ER 1939 ONSITE EXERCISE A partial participation onsite emergency exercise is presently scheduled for the week of September 25, 1989 (Sea May 5, 1989 letter of William Lazarus to PSNH attached as Exhibit 6.) This emergency exercise is a requirement for the issuance of any full-power license to Seabrook after June 27 1989 (i.e., cne year after the last onsite exercise). This require.aent is unambiguously set forth at Part 50, App. E, IV, F.1: If the full participation exercise is conducted more than one year prior to issuance of an operating license for full power, an exercise which tests the licensee's onsite emergency plans shall be conducted within one year before issuance of an operating license for full power. As an emergency exercise material and relevant to the issuance of a full-power license, the Mass AG has a right under the Atomic Energy Act to litigate the results of this exercise. gCs v. sRC. The Commission as recently as May 1987 tacitly acknowledged the Mass AG's right to litigate an onsite exercise prior to the issuance of a full-power license. In promulgating its Final Rule setting two years as the timing requirement for a full participation exercise prior to full-power licensing, the Commission noted that a " pre-operational onsite exercise will continue to be required one year prior to full-power operation." 52 Fed. Reg. 16823, 16825 (May 6, 1987) (Attached as Exhibit 7). The Commission noted further that:

.Q it is clear enat the results of exercises are litigable in'the operating license proceeding, irrespective of when those exercises are held, long as the holding of an exercise is a so pre-license requirement. Id. at 16827 Finally, in response to a public comment concerning the retention.in.the new rule of the one-year requirement, the Commission stated as follows: Summary of Comment Edison disagreed with the last two sentences of the proposed rule which require the Applicant to conduct an exercise of its onsite plan if the offsite exercise is more than one year prior to issuance of the operating license. Edison argued that the additional test would be of marginal value and might tend to introduce additional issues into the operating license hearing. On this basis Edison recommended deletion of the last two sentences of the proposed rule. Commission Response The Commission disagrees that a pre-operational onsite exercise within one year before issuance of a full-power operating license is of marginal value.... And nince, unlike the situation with offsite exercises, no one has identified any existing response or timing difficulty with the onsite requirement, we find no reason to revise the requirement at this time. Id. at 16824-25. D. EXT 339 AND NATURE OF THE HEARING RIGHTS TRAT ATTACH TO ALL 'ATEET&L !$$VES 1. Request That The Record Be Held Open Pending Completion of Low-Power Testing and the September Onsite Exercise.

l l ) I i It is clear that once it is deter.91:ed thar low-power 1 testing and the Septemoer onsite exercise are material and relevant to the issuance of a full-power license, the Mass AG has a right to litigate these evrants. UCS v. NRC at 1443. Moreover,'this public hearing right is impermissioly burdened if the Mass AG must seek to reopen closed proceedings in order to secure these hearing rights. Id. at 1444. Put another way: to the extent that low-power testing and the next annual onsite exercise are themselves material to a future full-power license, the full-power record should not be closed before the Mass AG has had an opportunity, pursuant to the Commission's procedural regulations, to litigate these events. Just as the Interveners did not have to reopen the record to litigate the June 1988 exercise, so they should not have to do so to litigate low-power testing or the 1989 onsite exercise, the successful completion of which is a condition of full-power licensing. To this end, the Mass AG requests that this Board hold the record open pending the admission of any acceptable contentions filed in response to these two events. 2. Pre-hearing conference, late-filed contention standard and Catawba The Mass AG also requests that a pre-hearing conference be set at which a schedule can be determined for the filing of contentions arising out of low-power testing and the September 1989 onsite exercise. Further, the Mass AG seeks clarification as to whether any contention arising therefrom is subject to

the late-filed contention standard. Cf. Duke Power Comeanv (Catawba Nuclear Station, Units l_and 2) 17 NRC 1041-(1983). (Attached as Exhibit 8.) Obviously, if the Board holds a pre-nearing conference and sets a schedule for the filing of contentions, as it did for the June 1988 exercise, suen contentions should not be subject to the late-filed contention standard. In any event, the Mass AG requests that a deadline for.the submission of contentions be established so that in the event the late-filed contention standard is applicable, the Mass AG can have notice as to when, after these future events, contentions, although " late-filed,' are still filed with ' good cause... for failure to file on time." 10 C.F.R. 2.714(a)(1).2/ 4 E. ACCESS IS~NECESSARY TO PROVIDE MEANINGFUL RIGHT TO A H With regard to both low-power testing and the September onsite exercise, the Mass AG requests that this Board provide reasonable access and observational status to the Mass AG and his experts so that he might meaningfully observe both events. Such access should include, but not be limited to, all documents generated prior to, during and after low-power testing and the September onsite exercise which are relevant to an assecument of the performance of both the plant and all plant systems tested and the emergency onsite staff. i l \\ l L 2/ Obviously, intelligible contentions cannot be filed before the events occur. If every contention filed after the event is { nonetheless late-filed under Catawba, then the Mass AG seeks a { senedule so that he can establish " good cause" for the failure j L to file on time. j l.

l-F. REQUEST'FOR EXPEDITED DISPOSITION OF THIS MOTION The Mass AG seeks expedited disp cecause the ongoing low-power testi osition of this motion ng program is possibly of short duration and, in the event in whole or in part, the Board grants this motion the request status vis-a-vis this testin? programfor access and obse requires immediate relief. Respectfully submitted, COMMO!IWEALTH OF MASSAC JAMES M. SHANNON ATTORNEY GENERAL bN / n Traficonte ief, Nuclear Safety Unit ~~ One Ashburton Placeepartment of t Boston, MA 02108-1698 (617) 727-2200 DATED: May 31, 1989 1 ) ) l l

L.. 7;-i. I I 'f, 'k l ','t I-i.. ,e f. t1 I. I t 4 A L EXHIBIT C _________m-_____.m__.-_

28287 L 1 MR. DIGNAN: 'Are the originals being given to the 2 reporter? 3 JUDGE McCOLLOM: Yes, they are the signed ones, I 4 yes. 5 JUDGE SMITH: They are the only ones I had, in any 6 . event. 7 The two stipulations are listed separately in the 8 index as a separate category of item. 9 The Board is at this time denying the Attorney 10 General's motion to hold open the record for low power la testing and onsite exercise. 12 We don't reach the issue of whether, for example, 13 the onsite exercise might give rise to a right to a hearing. 14 In fact, our ruling would not turn on that at all even if we 15 were to assume that the onsite exercise -- and I'm just 16 using the onsite exercise as an example here, because that 17 is the one I think that is most likely to give a right to a 18 hearing. 19 But even if we were to assume that that onsite 20 exercise is likely to give rise to a right to a hearing and 21 that an issue may come out of it, we find that we do not 1 22 have the authority to hold open the record for that purpose. 23 The Board has general jurisdiction in accordance 24 with the Panel Chairman's reconstitution of the Licensing 25 Board Clarification to entertain the motion. But as we Beritage Reporting Corporation (202) 628-4888

28288 1 stated, no authority to grant the relief requested. 2 Essentially we agree very closely with Section A 3 of the Staff's response to the motion. 4-The Board, under the general authority granted by 5' the Commission in the Notice of Opportunity for a Hearing, 6 which is effected by the Constitution order by the Chairman 7 of the Panel, has two basic general areas of authority. 8 One, is to rule on the admissibility of 9 contentions. I might elaborate on that just a bit. 10 Also, we also have the responsibility and 11 authority to rule upon the standing of a petitioner to i 12-intervene, but that isn't relevant'here. 13 And then once having found that 'a contention 14 raises an issue for hearing, we have the authority and 15 jurisdiction to hear and decide those contentions. And, of 16' course, as everyone recognizes, we also have the additional 17 authority to raise an issue sui sponte, which again is not l ~18 involved in this consideration. { 19 We have no other authority, none whatever. 1 20 We have come to a natural ending to the 21 evident'iary hearing. There is no more testimony to be 22 presented to us. There is a astion which we understand will 23' be received to receive evidence. But properly before us ) 24 right now, there is just nothing more for us to do. l 25 We are required in those events to close the \\ l Heritage Reporting Corporation l (202) 628-4888 I

28289 1 -evidentiary record when we arrive at that normal milestone, (: 2 and we have come there today. 3 There is no place in any of the court cases cited 4 to us, any place in the Commission's regulations, no place 5 in any of the Commission's adjudicative decisions which 6 authorize the presiding officers to keep the evidentiary 7 record open solely to ease or to create higher threshold 8 burdens of a litigant. 9 It's a matter down the road, and it is quite 10 simply just beyond our authority. 11 Even if the NRC as a whole has the responsibility 12 to address the difference between reopening the record as 13 compared te a late-filed contention, or as compared to a 14 2.206 petition, even if that were the case, that authority 15 has not been granted to licensing boards, and that would be 16 true even if there is a v'>id in the Commission's system of 17 adjudicative fora. 18 And as stated, we found nothing in the cases -- 19 Union of Concerned Scientists v. NRC, or Mothers for Peace. 20 Everyone calls it " Mom's), the San Luis ObiRR2 case. The 21 Mothers for Peace -- that suggests to the contrary. ~ 22 Accordingly, this evidentiary record is now 23 closed. 24 MR. TRAFICONTE: Can I seek one -- 25 JUDGE SMITH: Indeed, we are not off the record, Heritage Reporting Corporation (202) 628-4888

28290 -1 -however. 2 MR. TRAFICONTE: Yes. 3 JUDGE SMITH: But the evidentiary record is 4 concluded -- 5. MR. TRAFICONTE: Yes. 6 JUDGE SMITH: -- as of this moment. 7 (The evidentiary record is herein closed.) 6 MR. TRAFICONTE: I understand that. 9 JUDGE SMITH: Zap. That's a magic moment. 10 MR. TRAFICONTE: The gun, I have a gun in my 11 pocket and I will fire -- 12 (Laughter) 13 MR. TRAFICONTE: That I understand. That I 14 understand. 15 I would like, in the form of a clarification, we 16 are intending, and I think I mentioned this once before, but 17 we are intending to file a contention arising out of the 18 events of last week'during low power testing. 19 It happens that, as the Board is aware, as part of 20 this motion, we were asserting a right to litigate the 21 events of low power testing, and we believe now there is 1 22 such an event loosely defined. 'We are not sure exactly what 23 occurred yet, but we have some sense of what occurred. 24 As a point of inquiry, based on my understanding i 25 of the jurisdictional lay of the land, I would prepare such l Beritage Reporting Corporation (202) 628-4888

r 1 EXHIBIT D \\ i i I 4 ) l ' l l - 1 l l l W b 1 4

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges: Ivan W. Smith, Chairman Dr. Richard F. Cole Kenneth A. McCollom In the Matter of PUBLIC SERVICE COMPANY OF Docket Nos.50-443 OL NEW HAMPSHIRE, ET AL. I.aw Power Operation /50 444-OL 21 July 1989 -(Seabrook Station, Units 1 & 2 JOINT AFF DAVIT OF GRE ORY C. MINOR A O STEVEN C. S4 011 Y I, Gregory C. Minor, do make oath and say: 1. My name is Gregory C. Minor. I am a Vice President of MHB Technical Associates. My bncinean address is 1723 Hamilton Avenue, Suite K, San Jose, California 95125. I received a B.S. in Electrical Engineering from the University of California, Berkeley, in 1960 and E a M.S. in Electrical P p-ering from Stanford University in 1966. 2.' -I have over twenty five years of experience in the design, development, research, start-up ' testing, and management of nuclear reactor systems. From 1960-1976, I worked for General Electric Company in the design, development, and testing of safety and control systems for nuclear plants. As part of these activities, I was responsible for system check out and /Go fgot $Whf f'

preparation of-pre-operational test procedures for a nuclear research reactor, including the direction of craft personnel performing the necessary system modi 5 cations. My responsibilities . Included equipment and systems design, as well as management of a large engineering group

responsible for new control room design. These new designs required the detailed evaluation of the' technical and ergonometric factors necessary to ensure that operations data presented to operators would provide them with the information necessary to make the proper decisions for safe operation.

. 3. . For the past thirteen years, I have been an independent technical consultant. In that capacity, I have participated in a variety of studies addressing nuclear facility' economic, management, and safety issues for various orgmal'a+ ions, including the Depanment of Energy /Sandia National laboratories, the Swedish Government, and the offices of several states' Attorneys General. I am currently a consultant on several nuclear plant cases in which design, management, and compliance with existing regulations are being investigated. 4 I am a member of the Nuclear Power Plant Standards Committee for the Instrument Society of America. Also, I participated in a Peer Review Group of the Nuclear Regulatory Commission's Three Mile Island Special Ir.quiry Group. Further details of my quali5cataons and professional saperience are summarized in my hatement of Prat===Innal Qualifications which is appended to this afSdavit as Attachament 1. I, Steven C Sboily, to make oath and say: 5. My name is Steven C. Sholly. Since September 1985, I have been employed as an Associate Consultant by MHB Technical Associates. My business address is 1723 Hamilton Avenue, Suite K, San Jose, California 95125. 2- ~ _ - -

1 6. I have been previously employed by the Union of Concerned Scientists as a Technical Research Associate and Risk Analyst from February 1981 to September 1985, and by the Three Mile Island Public Interest Resource Center as Research Coordinator and Project Director from January 1980 to January 1981. I also have non-nuclear experience in the wastewater treatment and science education fields from September 1975 to January 1980. I received a B.S. in Education, with a major in Eanh and Space Science and a minor in Environmental Education, from Shippensburg State College (now Shippensburg University), Shippensburg, Pennsylvania, in 1975. 7. For the last nine and a half years, I have been engaged in analyzing technical nuclear safety, management, design, construction, and regulatory issues and providing technical advice to state and local governments (including the States of California, New York, Ulinois, Pennsylvania, Maryland, Maine, Connecticut, and Massachusetts, and Suffolk County, New York) and Independent organizations on these issues. I have presented testimony concerning these I issues before the Connecticut Depanment of Public Utility Control on behalf of the Prosecutorial Division and Division of Consumer Counsel, before the California Public Utility Commission on behalf of the Division of Ratepayer Advocates, before the Pennsylvania Public Utility Commission on behalf of the Office of Consumer Advocate, and before the Massachusetts Department of Public Utilities on behalf of the Office of the Attorney General, Commonwealth of Massachusetts. I have also participated as an expert witness in proceedings before the Atomic Safety and Ucensing Board in the-Indian Point Special Investigation and the operating license review of the Catawba nuclear plant, and have presented testimony before the United States Congress on nuclear safety issues. I have previously served as principal investigator for monitoring startup I testing and low power operations of the Shoreham Nuclear Power Station for the New York Consumer Protection Board. Further details of my experience and qualifications are contained in my Statament of Profminnul Qualifications which is appended to this affidavit as Attachment 2. 1 3

l i I DISCUSSION i 8. MHB Technical Associates has been retained by the Office of the Attorcey General, Commonwealth of Massachusetts, to assist in a review of the results of the low power testing program for Seabrook Station, Unit 1. Seabrook Unit I received a license from the NRC permitting the conduct of such tests on May 26,1989 (License No. NPF 67). V In issuing that license, the NRC staff found that the facility would operate in conformance with the application, ] as amended, the provisions of the Atomic Energy Act, and the regulations of the Commission, and that there was reasonable assurance that the activities authorized by the operating license would be conducted in compliance with the Commission's regulations. y 'i i 9. The Commission's regulations require, in part, that the Final Safety ArJdysis Report (FSAR), submitted as part of the operating license application, include the managerial and administrative controls to be used to assure safe operation and the plans for preoperational tes and initial operations. # The regulations specifically reference 10 CFR 50, Appendix B, as setting forth the requirements for managerial and administrative controls for nuclear power plants. # Moreover, the Technical Specifications for Seabrook Unit 1 (Appendix A to License No. NPF-67) require that lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all 1/ Ixtter dated May 26,1989, from Steven A. Varga (Director, Division of Reactor Projects I/II, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commmion) to Edward A. Brown (President and Chief Executive Officer, New Hampshire Yankee i Division, Public Service Company of Nevf Hampshire),

Subject:

' Issuance of Facility Opemting Liceme NPF Seabrook Station, Unit No.1".

2/ These findings are required to be made by 10 CFR 50.57(a)(1) and (2). V 10 CFR 5034(b)(6)(ii) and 10 CFR 5034(b)(6)(iii). 10 CFR 5034(b)(6)(ii). .4.

perating organization positions.1/ Accordingly, as required by the Commission's regula o and by the Seabrook Station's Technical Specifications, the lines of authority and re e and communication must be established and defined down to the level of in operators. 9. On June 22, 1989, with the reactor at approximately 3% of full power, plant operators were performing a natural circulation test (1 ST-22). The reactor coolant pumps were tripped at 12:19 p.m. to initiate the test. At approximately 12:26 p.m., steam dump valve M 3011 went full open. Pressurizer level, which had been slowly decreasing as expected, conti to decrease even though letdown was at a minimum (approximately 10 gpm) and cha flow was almost at maximum (approximately 122 gpm). This further decrease was du of the primary side by steam released through the stuck-open steam dump valve MS PV-3011 approximately 12:29 p.m., pressurizer level decreased below 17%, which according to the requirement of the test procedure 1-ST-22 required a manual trip of the reactor. f/ - 10. The reactor was not tripped, however, until 12:36 p.m., some seven minutes later .when the Unit Shift Supervisor (SRO-licensed) ordered a manual trip in anticipation of a manual trip criterion of 2340 psig pressurizer pressure (the actual pressure was at 2310 p rising).1/ i i i ~ 5/ Office of Nuclear Reactor Regulation, T tehnie=1 SneciD' rions: semhrook station. Unit L Dadrat No. "H a Anrwn.i- "A" to Commission, N' REG-LT31, May 1989, Section 6.2.1.a.Jeanw No. b747, U.S. Nuclear Re 6/ New Hampshire Yankee, EVENT EVAT ITATION RFPORT: Vanual Remor Trin Dur na 1-S"-21 Namral Circulation Test at 1-2, Enclosure 2 to NYW-89086, letter dated ~ 12.u y 1919 from Edward A. Brown (President and Chief Executive Officer, New Hampshire Yankee) to William T. Russell (Regional Administrator, NRC Region I). 2/ Ibid. 5

• E"MN

New Ham = hire Yankee concedes that the Unit SMft Superintendent "should have 11. immedimely directed the reactorshut down when thepressumer levelreached the 17% test 12. New Ha='=h!re Yankee does not dispute that the mann=1 trip criteria listed in 1-ST-22 were exceeded for several minutes before a manual trip was initiated. 2/ New Hamp Yankee also does not dispute that the failure to initiate the manual trip as required by the Natura Circulation Test procedure (1-ST 22) is clearly a case of procedural noncompliance.10/11/ 13. New Hampshire Yankee also concedes that the procedural noncompliance violated the Seabrook Station Administrative Controls, Policy, and Guidance with respect to procedural adherence.12/ 14. Further, New Hmm,r= hire Yankee concedes that the Seabrook Station Operations Management Manual required that plant operation be conducted in accordance with applicable procedures.13/ 1/ New Hampshire Yankee, Manavernent Meetivenau Analvsis Reoort! Fatural Circulation ' Test. Mann=1 R== rear Trio sne Fa 'awun Actions at 20, Enclosure 4 to WN 89086, letter dated 12 July 1989 from Edward A. Brown (President and Chief Executive Officer, New Hampshire Yanker) to William T. Russell (Regional Administrator, NRC Region 1). PJ E at 4 ig/ New Hampshire Yankee, Ooerntional Tan-. evninadon of the Natural Circulation Test (1-ST-22) on J=== 21 : 989. at 4. Enclosure 3 to NYk89086, letter dated 12 July 1989 from Edward A. Brown (President and Chief Executive Officer, New Hampshire Yankee) to William T. Russell (Regional Admimstrator, NRC Region I). 11/ It should be noted that the Seabrook FSAR commits operations personnel to procedural adherence. Jeg, Seabrook FSAR, at 13.5-3. 12/ E at 5-6. 13/ New F==r= hire Yankee, Manneement Effectivenen Analysis Report: Natural Cirentation Test. Manh=1 R==ator Trb andTollowun Actiont at 7, Enclosure 4 to NYN-89086, letter Hmm,nchire Yankee) to William T. Russell (R(egional Administr -6 Jut 21 '99 :3:51

---

A09 266--T1 Mi I C C

15. New Ham:=hi e Yankee, however, fails to acknowledge that the " pro f noncompliance" (which it concedes occurred) represents a violation of 10 C Criterion V. L4/ Specifically, Csiterion V, states, " Activities affecting documentedinstructions, procedures, or drawings, of a type appropriate to be accomplished in accordance with these instructions, procedures, or draw Instructions, procedures, or drawirgs shall include Appropriate quantitative or qualitat determining that important activities have been satisfactorily accomp that the conduct of tests while a reactor is at power is an " activity affecting meaning of Criterion V of 10 CFR 50, Appendix B. (Indeed, in order to con first required an operating license issued by the NRC, without which even b critical would have violated the Atomic Energy Act.) There also can be no d 1-ST-22 was the " procedure", within the meaning of Criterion V of 10 CFR 50 governed the conduct of the Natural Circulation Test. Indeed, it should be noted 1-ST-22 was prepared by the Startup Test organization, and reviewed and a Operation Review Committee.11/ The procedure was also reviewed by V L4/ Commission,' sin addition since the Seabrook Station Unit 1 Ucense No. N tions (Para; , Criterion V, also means tha),the license comply with 10 50 Appencix t terms of the operatinglicense. 11/ The Station Operation Review Committee Manager ca all matters related to nuclear saf(ety. The SORC is cha Manager Manager,, and includes members ' as follows: Assistant Station Manager, Operations Technical Supert r, Maintenance Instrumentation and Contro Department Supervisor, ReactorDe artmentSupervisor, L Supervisor, Health Physics Department Supervisor, Technical Support Supe Chemistry De nt Supervisor. The SORC is specifically charged to review all procedures Mg, Offke of Nuc; nuclear safe, and all tests and nments that affect nuclear safety. ear Reactor gulation, Techai (wifica+ianc (**braak Station. ? Commusion. NUMG-1331, May 1989, Section 6.4.1. nit 1. Docke 7

G [ Westinghouse, the Seabrook NSSS vendor. M/ In addition, the procedure was reviewed by New Hampshire Yankee Quality Assurance.12/ 16. There is also no dispute, therefore, that compliance with the provisions of that procedure, including manual trip criteria, was mandatory. Indeed, prior to the test, the operating crew read the entire test procedure, and the Test Director briefed these personnel and made individual copies of the manual reactor trip criteria for these personnel. Accordingly, inasmuch as an activity affecting quality was not performed in accordance with the governing procedure, this constitutes a violation of Criterion V of Appendix B to 10 CFR Part 50. 17. Moreover,10 CFR 50, Appendix B, Criterion II, provides that the quality assurance program "shall provide for indoctrination and training of personnel performing activities affecdng quality as necersary to assure that suitable proficiency & achieved and maintained".18] It is beyond dispute that being an operator at the controls of a nuclear power reactor is an " activity affecting qualsy" within the meaning of 10 CFR 50, Appendix B. Accordingly, indoctrination and training oflicensed operators of the requirements of the NRCis required. 18. New Hampshire Yankee states that under normal conditions operators function in a crew consisting of three members. During the conduct of the Natural Circulation Test, there were Sve crew members. New Hampshire Yankee cimims that the additional two members in the shift M/ New H== W Yankee, Manavement Meetiveness Analysis Renort: Natural Circulation Test Man,=al D=Mor Tri, and Fol owun Actiont at 3. Enclosure 4 to NYN-89086, letter u dated 12 July 1989 from & ward A. Brbwn (President and Chief Executive 05.cer, New Hampshire Yankee) to William T. Russell (Regional Administrator, NRC Region I). 17/ Id., at 8. 3/ 10 CFR 50, Appendix B, Criterion 11, "Quakty Assurance Program". 8-

crew "made the reporting relationships, lines of authority, and communications more compier". 3R/ This is misleading. All five members of the shift crew at the time of the Natural Circulation Test had the authority to order the reactor to be tripped or the authority to manually trip the reactor (all were licensed operators, and at least the Unit Shift Superintendent was SRO-licensed). Whatever " complications" might have been Introduced by the addition of two members 1 to the normal shift crew of three members, none of these complications change the fact that cash of the shift crew members possessed the authority to promptly terminate the test by either ordering the manual trip of the reactor or by manually tripping the reactor himself once the test conditions were exceeded, and, indeed, the terms of the test procedure (1 ST-22) required that t reactor be manually tripped when the test conditions were exceeded.11/ 19. Nonetheless, the shift crew permitted power operation to continue for approximately seven minutes until a second test condition was nearly exceeded before manuall tripping the reactor. Indeed, the Unit Shift Supervisor was cognizant of the requirements of procedure 1-ST-22, and was cognizant of the pressurizer level throughout the test, even to the point of conferring several times with the primary side control board operator and informing the 12/ New Hampshire Yankee, Operational r=="** evaluatinn of the Natural Circulation Test (1-ST-22) on June M Edward A. Brown ((President and Chief Executive Officer, New H William T. Russell (Regional Administrator, NRC Region I). I at/ Note that the Seabrook Unit 1 Technical Specifications,(Appendix A to License No. NPF.

67) require that the minimum shift crew composition while m Mode 1,2,3, or 4 consist of a total of $g licensed oxrators (two SROs and two ROs), and not gi caf tsc as asserted by NHY. Og,, Of5cc of Nuclear Reactor Regulation, "echnici Snec onc Seabrook State n

_ mt 1. nacket No. 50 441 Anoendix "A" to Jeanne No. F-67. U.S. Nuclear Regu story hission, NUREG-1331, May 1989, Section 6.2.2.a and Table 6.2-1. Power operstion at less than or equal to 5% of rated thermal power is Mode 2. Id., at Table 1.2. Similarly, the Seabrook FSAR reflects a four-member shift crew (Shift Superintendent, Unit Shift Supervisor, a Su msory Control Room Operator, and a Control Room Operator). Sec, Seabrook FS at 13.1 10. 11/ New Hampshire Yankee, Manmoement Effectiveness Annivsis Renort Natural Circulation Teat Manual Reactor Tri) and Followun Actions. at 2, Enclosure 4 to NYN 89086, letter Hampshire Yankee) to William T. Russell (R(egional Admmistra _ _ _ _ _ _ _ _ _ _ - _ _ _

1 Test Director that the pressurizer level was approaching and later was b three separate r:sti5 cations were made to New Hampshire Yankee pe who were present in the control room at the time of the test that the 17% pre trip criterion had been exceeded. NRC personnel informed the Startup Manager, the Test h Director.(who informed the Unit ShQ 9 pervisor, who acknowledged this Assistant Operations Manager of this fact. 23/ The Assistant Operatics Ma test trip criteria with the Test Director and confirmed that the pressurizer leve conveyed to the Unit Shift Supervisor. Thus, despite a number of managemen personnel knowing of the low pressurizer level condition, none of them acted to sh recommend shut.down of the reactor. As the Assistant Operations Manage

plant status to determine whether to personally intercede with the Unit Shift regard to actions being taken, the Unit Shift Supervisor (responding to ano pressurizer pressure increase) directed the second of the two primary side board

. manually trip the reactor. 33/ Clearly, the aimataneaum failure of five licensed op comply with the explicit written requirements of a procedure governing power ope l they were engaged is a serious matter, and one which is not easily explai . judgment. Moreover, none of the other licensed personnel and managers w apparently not "at the controls", acted to recommend shut down. 3d/ 21/ Note that the Seabrook Unit 1 Technical Specifications (Appendix A to Ucense lice)nse. fee, OfBee of Nuclear Reactor Regulation, Tsch Re= tic a. L nTt i h=hrook I Regu atory C+---- 3==4on, NUR3G 1331, May 1989, Section 6.2.2.f k443. A:w..t 'A" to Jamaa* No. NPF47. l..S. Nuclear 23/ New Hampshire r w=u m-Yankee, Mae=== Ehaiv==== Analva s Renort: Natural Ci culation

• ae Tri s md Foi ar. Act ans Hamaahim Yankee) to William T. Russell (R(egio osure 4 to NYN-89016, letter 3d/

According to NHY, there were 6 management personn21,15 operators on shift resumably includt'ag the five operators at the controls), and 16 additional opera o as part of their training. Also present in the coutrol room at the time of the incident were 8 test wrsonnel,4 training department persoanel,3 people fro representative from "SEG, and one representative from SAT. Sag,Id., at 5. _x---- - ^

r1 20. In addition, New Hampshire Yankee concedes that the Startup Test Director should hnve *strongly and immediate& recommended shutting the reactor down, in a clearprecise man the Unit Shift Supervisor when the test trip criteria were reached, but that the Startup Director failed to do so. New Hampshire Yankee also concedes that the Starmp Manage Test Director should have been "immediately respomive to the information provided by the impector". 36/ 21. New Hampshire Yankee states that, "The @erations personnel believed that the Startup Test Procedures, by their nature, were inherently flexible and did not require verbatim compliance." 26/ This is incorrect and misleading, since these procedures were being used to control operation of a power reactor, and were approved and issued for use for that purpose in accordance with 10 CFR 50, Appendix B, Criterion V of which requires compliance with written procedures governing activities affecting quality. 22. Accordingly, based on the above, it is our conclusion that a violation of 10 CFR 50, Appendix B, Criterion V, occurred in that licensed operators at the controls of Seabrook Unit 1 on June 22,1989, failed to manually trip the reactor when the procedure 1 ST-22 manual criterion of 17% pressurizer level was reached. Moreover,it is apparent that the training pr l was not effective in this instance. Therefore, some improvement in the training program is essential if similar violations are to be prevented in the future. 23. New Hampshire Yankee states, *At no time during the tnnsient were any technical specepcation parameters or design limits acceded nor was there any danger to public health and 25/ 11, at 6. 26/ Id. at 10. . c___ _a a x __u a ,as

l safety, to thesafety ofplantpersonnelor toplant equipment." 21/ Even if this is tecbolcally c it is irrelevant to the violation of 10 CFR 50, Appendix B described above. Dere is a s safety assurance aspect to the Commission's quality assurance regulations, and these re cannot be waived by licensee personnel at their convenience in favor of a less limiting set of criteria.~ Failure to follow procedures carries with it significant safety implications which cannot b ignored.- 24 Moreover, it should be noted that the procedural noncompliance which occurred 'during the Natural Circulation Test on 22 June 1989 may not be an isolated event. There it at least some evidence of other possibly related procedural noncompliance problems. Intpection Report 89-3 (2/28/89 4/24/89) states: 38/ Four incidents ted a potential reduction in addition to detailin the condxt of routmeplant Two involvedfaihuas to pmperty position and lock a valve to ( preventpot boron dilution of the reactor coolant system. Another concemed the generanon of a reactor trip ' when steam genemtor levels were allowed to drop to the trip whdein wet A event involved the inadvertent opemnq of a s&poweropemtedreliefsafety impact, they may e apotential weakness m what has been l&C testing. While none of these inculents to date, an ewellent operating recont M. More recently, in a memorandum to the Director of the Office of Nuclear Reactor Regulation, the NRC Regional Administrator (Mr. Russell) stated: 32/ 22/ New Hampshire Yankee letter NYN-89086, dated 12 July 1989, from Edward A. Brown (President and Chief Executive Officer, New Hampslure Yankee) to William T. Russell (Regional Admimstrator, NRC Region I), at 1. 24/ NRC Inspection Report 50443 dilutionincident described abbve/89-3, Inspection Report Summary, at 2.The boron was the subject of a Notice of Violation (89-03-02). 31/ NRC memorandum dated 5/3/89 from William T. Russell to Thomas E. Murley, "Scabrook Station, Unit 1 Low Power License Recommendation, Supplement 1". _

Recently, sevemi minor events and problems have occuned at the site which raise NRC concem over the attention to detailin ope ations displayed by theplant staff The most signtpcant of these involved two violations of the special zero-power license conditions anociated wnth locked valves. While none of the mdividual events had an impact on safety, taken together they indicated apossible declining trend infacilityperformance. 26. Based on the above circumstances, we conclude that the procedural noncomplianc of 22 June 1989 is not an isolated event, but is rather part of a pattern of procedural noncompliance at the Seabrook Station. Moreover, the 22 June 1989 cvent is much more rerious since it represents a pervasive noncompliance, involving the entire shift crew on duty at the ti . of the incident. Signed ur:dcr the pains and penalties of perjury this 21st day of July 1989. M n=5 Gregory C.ho/ Steven C. Sholly . ___lut z t _ _ : Ra n : ss _ _.-m.

ATTACHMENT 1 Statement of Professional Qualifications Gregory C. Minor j ) l l l l JuL 21 59 33:5g -ara ann _n-

.l I i PROPPetf0NAL QUA1 YPicATIONS OF GREGORY C. MINOR GREGORY C.nGNOR MMB Toshnical Associates i< 1723 Hassilton Avesse i Seite K San Jose, California 95125 (dos)266 2716 EXEEEEN2r uns to Passawr j - m.E"- . MMB T, a' t A=='.e. h w c M _.a 8t*- 4 and energy consultant to state, federal, and private orgammations and individuals. Maior actMtles inclade studies of safety and risk involved in energy generation, providag technical conomitig to legislative, regulatory, public and penate groups and espert witness in behalf of state orpsminations and citissas' groups, Was co editor of a angue of the Reactor Safety Study (WAgH 1400) for the Union of Concerned Scientists and co aather of a risk analysis of Swedish reactors for the Swedish Energy e '"= Served on the Peer Review Group of the NRC/rMI Syncial Ingeiry Groep (Rogovia ca==Im). Actively inwohed in the Nuclear Power Plant Standards Comunittee work for the lastrument Socwty of America (ISA). 1972 1976 as -- -- p_' a r. __ J a t _. -- - -. - v.4 - - a r_- _ q vt.4 c,_y_-- .Nacinar Enarav Divisinnt San Jaan California Managed a design and i.- *, ' group of thirty four engineers and support persomaci designias systems for ese in the measurement, metrol and operation of anclear reactors, lavehed coordination with other. reador design organizations, the Nuclear Regulatory c_ and costomers, both overseas and domestic.

• 7 ""ths incladed coordinating and managing and design and development of control systems, safety systems, and new control concepts for use on the mest gaaeration of reactors. The position included

. ::f "

• y for standards applicable to control and instrumentation, as well as the design of abort-term solutions to Sold probicass. The disciplines involved incloded electrical and me-chanical engineerias, seismuc design and process computer' control /programmag, and equipescat qvahbaria=

1970 - 1972 M -- - - m. ar e;_.: J s o.. n--t_% G,.,,m1 pt.,*ric Comnanv. Nneta=e Enerav Divisina. San Jose. Cahfornia Managed a group of seven engineers and two support personnelin the design and preparation of the detailed system drawings and control documents relating to safety and emergency systems for nuclear reactors. Responsibility required coordination with other desiga 1 J ut. 21 *59 14: 01 20 -266- ~' l 4 9 - - f ESI-W

1 orgnaianticas and interaction with the customer's engiacering personacl, as web as r Pessonnel. 1963-19m i r'- ' - w- ' n (=- - 3 ye_- - 4 c,- ; g, my g._j. y g g,q BM for the design of specific control and instrumentation systems for nuclcar 14 ad dosism ;::,- Nh for various subsystems of instrumentation used to reactors. measure neutron flus in the reamor during startup and intermediate power operation. Perforsned lead syntam design faar+ian is the design of a tudor system for measuring the power generated in ancinar reamars. Other responsibilities included on site chadout and testing of a complete reactor control system at an experimental reactor is the Southwest. Recoind patent for Nuclear Power Monitoring System. 1960 - 1963 _i'- > u- --e prw, c m1 pt,,. ie c c g.,,,,,,,, ;, w,,hio,o, ("'"'=' - ' Ari - - = Rotating assignments in a variety of disciplines: Engineer, reactor malatenance and lastrumcat design, KE and D reactors, Hanford, Washington, circuit design and equipament maintenance cooth% Design caginear, Microwave Dpcat, Palo Alto, California. Work on design of cavity coupiers for Miaownw Travsling Wave Tubes (TWT). Design engineer, Camputer D_ ;1^--- ^ Phoenis, Arumna. Design of core driving circuitry. Design engineer, Atomic Power Equipment Department, San Jose, California. . Ciremit design and smalysis. Design engineer, Space Systems Department, Santa Barbara, California. Prepared control portion of satellite proposal Technical Staff - Taehale=1 Military Planning Operation. (TEMPO), Santa Barbara, California. Picpare analysea of missile exchanges. During this period, completed three year Ocaeral Electric program of cuensive education in advanced engineering principles of higher mathematics, probability and analysis. Also conopleted courses in Kepner Tregoe, Effecthe Presentation, Management Traming Program, and various technicalseminars. 2 g e 'a 1 ~ ' ' ~ ~ ^

EDUCATION Umsversity of California at Berkeley, BSEE,1960. Advassed Coor= is eing three yar curric.ium, General El.ctric Company,1963. v Stanford Unhorairy, MSEE,1966. HONORS AND ASSOCIATIONS Tau Beta Pi'-f x? g Honorary Society Co-holder of U.S. Patent No. 3,565-7G0,' Nuclear Reactor Power Monitoring System,' February,1971. Member: American Associarian for the Advancement of Science. Member: Nuclear Power Plant Standards cammittee, Instrume.at Society of America. PUBLICATIONS AND TESTTMONY 1. G. C. Minor, S. E. Moore, ' Control Rod Signal Multiplexing,' IEEE Transactions on Nuck,ar Science, Vol. NS.19, February 1972. 2. -* G. C. Minor, W. G. Milam, 'As Integrated ControlRoom System for a Nuclear Power Plant," ' NEDO 10658, presented at International Nuclear Industries Fair and Terhaaral Meetings, October,1972, Basic, Swkzerland. 3. The above article was also published in the German Technical Magazine, NT, March,1973. 4 Testimony of G. C. Minor, D. G. Bridenbaugh, and R. B. Hubbard before the Joint Comuni* tee on Atonnic Energy, Hearing held February 18,1976, and published by the Union of Concerned Sdestista, Cambridge, Manachusetts. 5. Testimor.y of G. C. Minor, D. G. Bridenbaugh, and R. B. Hubbard before the California State Assembly Comunittee on Resources, Land Use, and Energy, March 8,1976. Testimony of G. C. Minor and R. B. Hubbed before the California State Senate Committee on 6. Public Utilaies, Transit, and Energy, March 23,1976. 7. Tr.stimony of G.*C. Minor regarding the Grafentheinfeld Nuclear Plant, March 16 17, 1977, Wurzbeers, Germany. l 8. Testimony of G. C. Minor before the Cluff Lake Board of Inquiry, Regina, Saskatchewan, Canada, September 21,1977. 1 l l l Lnwn ua ~~--

9. na m' ' - d M ' n--a r A R:-f:;; d the NRC m-a * ^" ^ kd WASM. Id1B fNUREG 75/014) H. Kendall, et al, edstad by G. C. Minor and R. B. Hubbard for the Unios of Concerned cc.elara August,1977. i 10. . $- ^.. - a u*a f -- -^ MMB TM Mg January,1m. (Pubbshed by Swedah Departament of Industry as Docessent Dat im:1) Testimsosy by G. C. Minor before the Wisconsin Public Service Commission. February 13,1975, 11. i - ar e - ' - e - w r,.. e w -*e - -- -. 12. Testheosy by G. c. Minor before the CaMarnia taginiature Assembly c== ire

• an Resourees, Land Use, and Energy, AB 3108, April 26,1978, Sacramento, California.

13. Presentation by G. C. Minor before the Federal Ministry for Research and Technology (BMPT), Meeting on Reactor Safety Research, "-+ !==h in " ' -- * =~w August 21, and September 1,1978, Bonn, Germany.. i Testimony of G. C. Minor, D. G. Bridenbaugh, and R. B. Hubbard, before the Atontic Safety 14. and IJcansing Board, September 25,1978, in the matter of Black Pos Nuclear Power Station Casatruedes Parasit Hearings, Tulsa, Oldahoma, 15. Testimony of G. C. Maar, ASLB Hearings Related to TMI 2 Accident, Rancho Seco Power Plant, on behalf of Friends of the Earth, P- - T - - 13,1979. 16. Taari===y of G. C. Minor before the Michigan State Igislature, Special Joint ca-i :a* on Nuclear Energy, I==A =*ie== d %eme * - ' k - ^ for Nii 1= Pca Pl==*= la hushigna, odober is,1979. 17. A cri+6.i View of M---*

• " ^ - by G. C. Minor, paper prescated to the Annenean Amaarimeiam for the Advancessent of Science, Symposium on Nuclear Reacsor Safety, Jamaary 7, 1900, San Frandsco, California.

{ {

8.

He 5" - af

• an *- f :- of N='= P-P1-paper presented at Foruns on Swedah Nacinar Referendum, WW Swedes, Masch 1,1900.

39, v - -- e n- < _ _ pu. c__._,,_,__. N CA MHB Th Amociat% September 1980, prepared for the Minnesota Pollution Control Agency, Roseville, MN. 20. Testimony of G. C. Minor and D. G. "- ": ' -j-before the New York State Public Service Ca==*=alaa, Ch='- N~t-P6* Pc- -

• F '~ ' in the matter of long Island Lighting Company Temporary Rate Case, case IP 2Tnd September 22,1990.

21. Direct testimony of Dale G. Brideabeugh and Gregory C. Minor before the New York State } Public Service Ca==lamir=, Y=:-- 5 *-- s P.n. Caa.ida Rrig af Cl.a k-- Net *=- Pasar Station C;qpts pad Schadnis. in the matter of Long Island Lighting Corapany Temporary Rate Case, Case Number 27774, September 29,1980. 22. Sw=-- f== *= - ' c61, E6 Crird-- MHB Techsucal Associates, January,1981, prepared for and available from the Swedish Nuclear Power laspectorate, Stockholm, Sweden. -a. i l l .fut It 'ss u:u

'O 23. Testimony of G. C Moor and D. G. Brideabeugh before the New Jersey Board of Pubbe Utilities, O Creek IM) Ed -'- Oue== Ing *m the matter of the Petition of Jerary Central Power and Ught Company for approval of as increase in the rates for niectric servise and adjestment clause and factor for sud service, OAL Docket No. PUC 3518-80, Dochet Nos. 306 2g5,807 48B, Febramry19,198L 24. Testimony of G. C Minor and D.G. WM j osPORV's W ". _ **- ~ Diablo Canyos Operating 13ccese hearing before ASLB, la the matter of Paede Gas and Eledric Conspany (Diablo Canyon Nuclear Power Finat Unks 1 and 2), Docket Nos. $0 275-01, S 32301 January 11,1982. 25. Testimony of G. C Moor and R. B. Hubhard on 8'-- = = = = - - P'- ~7 Diablo Canyon Operating License hearing before ASIA, Docket Nos. 50 275 01,50323 OL, J '11,1982, 26. s e 1=> ~' ===de' " F na*e Crhs'- P'- If # = "1. MHB Technical Assodam., Febr.eri 1982, prepared for and available inaa the Sweesh Nuclear rower laspectorate, 3;c.c.cha, Sweden. 27. Taariana=y of G. C. Minor, R. B. Hubbard, M. W. Goldsmith, S. J. Hamood os bahalf of SuBolk County, before the Atomic Safety and licensing Board, in the matter of Long Island Ughting Coss me _a_ pany, Shoreham Nuclear Power Stados, Unit 1, regarding Contention 7B,188881 . A *- - Tae- =r'= Docket No.30 322 0L, April 13,1982. 28. Testimoey of G. C. Minor and D. G. E'i 4 as behalf of Suffolk County, before the Atomic Safety and IJacesing Board,in the matter of Long Island IJghting Company, shortham Nuclear Power Station, Unit 1, regardi's S ~ " "= ~ 0- ~~ " It ?~ S'* ~~"' Vahn Fallara. Docket no.30-322 01, April 13,1)S2. 29. Testimony of G. C. Minor and R. B. Hubbard om behalf of SuNolk Couary, before the Atosaic Safety and thansing Board, in the maner of long Island ughting Company, Shoreham Nuclear Power Station, Unit 1, regarding h8u Commtv re % M w 50c e--~i= 1 Post. AccadsatMaastoring. Docket No. 50 32241, May 25,1982. 30. Testimony of G. C. Mmor and D. G. Brid*=hangh on behalf of Suffolk Casety, before the Atonde Safety and IJcensing Board, in the matter oflaag Island Lightas Company, Shoreham Nuclear Power Station, Unit 1, regardag t=M county C=*-;~ 21 ERV Test ra_,.sa, Docket No. 50 322 OL, May 25,1982. 31. Testimony of G. C. Minor and D, G. Brideabaugh on behalf of Suffolk County, before the Atomic Safety and 1 hassar Board, in the matter of Long Island Lighting Company, Shoreham Nuclear Power Station, Unit 1, regarding n-8-An of SRV Ci at' ---- Docket No. 50 322 OL, June 14,1982, 32. Testimony of f J. C. Minor on behalf of Suffolk County, before the Atosnic Safety and 1Jcenatog Board, in the matter oflong Island Lightmg Company, Shoreham Nuclear Power Station Unit 1, regarding Envir==-- =1 On M"-=N. Docket No. 50-322 0L, January 18,1983. 33. Teethnony of G. C. Minor and D. G. Bridenbaugh before the Pennsylvania Public Utility Commission, on behalf of the Office of Consumer Advocare, R**=edine the Cast of 5-

i rw -.:-- the C--- =rk.... cr== pharie saada-Unit 1. Re: Penar>ivania Power and ught, Docket No. R-822189, March 18,1983. 34. Sw'-ral testimony of G. C. Minor, R, B. Hubbard, and M. W. Goldanuth on behalf of Su5cEt County, before the Atonne Safety and Lie-h: Board, in the matter of Long Island Ughting Corapany, Shoreham Nuclear Power $wion, Unit 1, regardag Safety classification " 9..e-- 1 ata. (en r ,taa 7Bt Docket !:o.50 322, March 23,1983. 35. Verbal testimony before the District Court Judge in the case of Sierra Club et. at vs. DOE regarding the Ocan ur of Uranium MillTailb;lp June 20,1983. 36. s== e a - '-*---"ta===d C = F :' e Cu * - F 'l n=-1 MHB Technical Aunciates, Jame,1983, prepared for and available from the Smiish Nuclear Power Inspectorate, StWal=. Sweden. 37. s.- -*i Ev bariaa N- -m Reat== R =n - A fald=1 Ev=6= dan MHB Technical Amaarnates, June,1983, prepared for and available frees the Swedah Nuclear Power Inspedorats, Stockholm, Sweden. 38. Testimony of G. C. Minor, F. C, Finlayson, and E. P. Radford before the Atomic Safety and Uceasing Board, in the Matter of Ices Island Lighting Company, Shoreham Nudcar Power Station, Unit 1, regarding F--rv Pf==6 - Ev--*4aa Ti=>< =A Danes fred-65. 23.D and 23_ML Docket No. 50 322-OL 3, November 18,1983. 19. Testimony of G. C. Mioco, Siaewe!! 'B' Power Station Public Inquiry, Proof of Evidence Ramardmqr Sainty Issues. December,1983. 40. Testimony of D. O. Brideabaugh, L. M. D-Ll aa, R. B. Habbard and G. C. Minor before the State of New York Public Service Ccamission, PSC Case No. 27563, in the matter of long Islandl Ughting Company Prana *A:-- to I d' = the cmt of the harch*= Nucl =e Generating Facilqv -- Phase IL on behalf of County of Suffolk, February 10,1964. 41. Testissouy of Fred C. Finlaysos, Gregory C. Minor and Edward P. Radford before the Atonde Safety and f W : Board, in the Matter of Long Island Lighting Company, Shoreham Nuclear Power Station, Unit 1, on behalf of Suffolk County Regarding F-==w Pf=M= Nita ina (CQatandos.fd), Docket No. 50 322 OL, March 21,1964. 42. Testimony of G. Dennis Eley, C. Joha Amith, Gregory C. Minor and Dale G. Bridenbsogh before the Atonde Safety and L1===:= Board, in the matter of long Island Lighting company, Shoreham Nuclear Pcwcr Station Unit 1, regardag mn Dwn G,.a.atars =A 20 MW Gas InghiaG, Docket No. 50 322-OL, March 21,1984. 43. Reviaed Taatima=y of Gregory C. Minor before the Atomic Safety and Ucensing Board, in the matter of Long Island Ugidag Company, Shoreham Nuie Faer Station Unit 1, on behalf of Suffolk County regarding F=--w Pi=Ma

1. =m w.1 m =:w (c^attadena fliand1&L Ncket No. 50 32241, July 30,1964.

Teatimony of Dr. Christian Meyer, Dr. Jose Roesset, and Gregory C. Minor before the Atomic 44 Safety and Licensing Board, in the matter of Long 1sland Ughting Company, Shoreham Nuclear Power Station Unit 1, on behalf of Suffolk County, regarding law Power Henria== Sainmic raahitittee of AC Power Sources Docket No. 50-322 OL, July 1984._ 6- ,'un 3: 33 :a:ga 405 2EE 'ian cage 207

4 45. Surrebuttal Testianomy of Dale G. Brideatmugh, Lynn M. Danielson, Ridard B. Hubbard, and i Gregosy C Minor Before the New York State Pubbe sernoe r%====la=, FSC Case No. 27563, Shoreham Naciser Station, Lees Island I.lghting Company, on behalf of sufolk County and New York Stats Cossamer Procession Buard, regardlag Innatigadan af the cost of the Shanahasa Nuclear Gamerating Fadliev. Omober 4,1984. 46. Direst Tesdmony of Dale O. L" - f. Lynn M. Danieleos and Gregory C. Minor on behalf of Maammeh==erk Attorney Gameral, DPU 84145, before the Massachassets Deparnacet of PubEc Utihiiss, regarding 6' -- of T- '% by Fiedl-- c.a. - A ru,i, r w Cassammy for Saabrook Unit i November 23,1964,84 pas. 47. Direct Tamrianany of Dale G. Brideabaugh, Lynn M. D-=I 6= and Gregory C Minor ce behalf of Maine PubEc UtiEties Co==l==laa Star regarding Pr d-af ra== of '"13 Unit i Docket No. 84113, Deceanbar 21,1984. 48. Direct Testimony of Dale G. E'

• 1 O and Gespory C. Minor on behalf of SaSelk County

_4_ ym_ - _, m -i c--- 2., r -t Docket No. 50-322 0L, Ja==ary 25, 1985. 49. Direct Testianomy of Dale G. Bridenbeugh, Lyme M. r =. and Oregory C. Minor on behalf of the vermost Dv of Public Servios, PsB Doder No. 5030, regarding ZDuisseus e _ _i v=w " q_..s c_ __d c-. for C-4 i November u.1985. 50. Surrebettal testimony of Gregory C. Minor on behalf of the Vermont rnw of PabBe Service, PSB Docket No. 5030, Prudmann of Central Vac=a= 1h W: sernae c-da-ce= kgjaalungu, December 13,1985. 51. Direct Teatimony of Dale O. Bridambaugh, Oregory C. Minor, Lyma K. Price, and Steven C. Shally on behalf of State of ra=aamieur Deparansat of Pubbe Utility Control Prosecutorial Dwison and DMalon of Consumer Counsel regarding the Predamme of Espandstarna on Mikant.Mau, Docket No. 83 0703, February la,1986. 52. Direct Temi-a y of Dale G. Bridsabaugh and Gregory C. Minor ce behalf of Massachusetts Anorney oemaral regarding the F- - of T __ hv New F- S_= ca for ^ Saahrook Unit 1 Dookat Nos. ER45 646 000, ER45 647 000, February 21,1986. 53. Direct Testimony of Oropory C. Minor on behalf of the Prosecutorial Division of CDPUC regarding rup e-.:a-Pr' for "'" w Unit 1 Docket No. ER 85 72M01 Mareh 19,1986. Direst Testiadsy of Dale G. Bridenbaugh and Gregory C. Minor ce behalf of Massachusetts 54 Attorney Oeacral regarding wupra c.g _nta. & -- rar wr6-Unit i Docket No. 85 210, Meek 19,1996. 55. Drect Temissemy of Dale G. Bridenbaugh and Gregory C. Minor on behalf of Masseelmsetts Attorney General regarding WMWea's Pe=--ein! Om=3a. naru ..A n. =. a c..ie.i &8d;*ia-an utwa=, Unit i Docket No. 85 270, March 19,1986. 7

P i \\ 56. Rebuttal Testbnomy of Dale G. Bridenbeugh and Gregory C. Minor on behalf of Massachusetts Attorney Genersi regarding Rab **=8 in New #f ' Paner er - - /t -" c--i i Docket Nos. ER45 646 001, ER45 647 001, April 2,1936. 57. Direst Tealmsomy of Dale G. W'-:M and Gregory C. Minor on behar of State of Maine Star of Pubes UtEldes % =l==ia= regarding F-4= '. af " = Unir i la r the naamer of Maine Power Compesy Proposed Increase in Rases, Decket No. g5 212, Apru 21, ) 1 19e6. Sg. ' " '= d the t' - _d. ; 4 A.e =t for N-6-- h- -w P-- ' - far the rass of New - c YJuk, prepared for the State of New York Constaaer Protection Board, by MHB Tarh= cal Associates, Jane 1936. 59. Direst Tamei==y of Dale O. Bridenbeugh and Gregory C. Maar on behalf of the Vermont Departssent of Pubuc Service, regarding Pr= t-of twea hv eern.1 Vrt==t N& t_t Caranratian for Minetane i Docket No. 5132, Asqgest 25,1936, 60. Serrebuttal Tami==y of Gregory C. Manor la the matter of hrsey Central Power and Light Company, regardnes TMr n-- e a r_'.- -- - t =h (Oral tanima=y), OAL Docket No. PUC 793D45, BPU Docket No. ER851116, September 11,1936, Serrebunal Testhnomy of Gregory C. Minor on behalf of State of Varmont Departusant of 61. Pubue Service, regarding CYP 1/NU r= =' -- S'--- =. ' ' to 'Minstane Unit 1 Docket No. 5132, Nouseber 6,1936. 62. Direct Testimony of Gregory C. Minor and Lyma K. Price on behalf of State of Vermont D7 est of Publie Serwce, regarding Pradence of 8'== "=ca for ""-- '- t Docket No. 3132, Desember 31,1936. 63. Dired Testimony of Gregory C. Minor on behalf of Suffolk Couary, befors the Atossic Safety and Licensing Board, concerning " ?. % A d= a = -- 'm__ (e-, t. M in the matter ofI4eg Island Lighting Company, Shoreham Nucieer Power Station, Unit 1, Dochet No. 50-322 0L.5, February 27,1937. 64. Direct Tamima=y of Gregory C. Abr et. al on behalf of the State of New York and Suffolk County,before the Atomic Safay and uneasing Board, regarding ne sense of the Emarunacy ___ _. y c _.-._ m, %.g,,,,,, % m;,, ,,,y, Shoreham Nuclear Power Station, Unit 1 Docket No. 50 322 01 3, April 6,1937. 61 Direct Testimony of Gregory C. Minor regarding h- - -; F8- - R=='= PM*a. Vanitoring and Demontaanimation. Shoreham Docket 50 322 OL 3 (Emerp:Scy Plassag), April 13, 1987, Testimony of Jessory C. Minor Steam C. sbolly et. at on behalf of Suffolk County, regarding 66. inm = --- e - _ Pr 7 =" before the Atomic Safery and Licensing Board, la j' the matter af Lang Island Lighting Conspany, Shorshem Nuclear Power Station Unit 1, Docket No. 30 333 OL.3, AprE 13,1937. 67. Robottal Testimony of Gregory C. Mmor and Steves C. ShoDy on behalf of Suffolk County regarding IR PO's R=are re=s, n (Rahurral o T=L aav of I.mwis G. Hiil===t in the t 8

maner of long Island IJghting Compsey, Shorchen Nuclear Power Station, Unit 1. Docka No. 532ML3, May 27,1987, 68. Direst Tesdmony of Dale G. Bridenbaugh and Gregory C Minor on behalf of Massachusetts Amarapy General, before the Federal Emergy Regulatory Commission, regarding canni Elmark e- -t. m.w.a to eaa vak 2 raa m_4.= y-706 001,Juh31,1937. - 2 Docket No. N-69. Direst Testimony of Dale G. Bd'".' and Gregory C. Minor before the Pennsylvania P:fdie Utility e '=2 Regarding '- 790">R314, OCA Statement No. 2, Angust 31,1937.V+e Unh 1 1m o=== Docket No.1-10. Orel teethnomy of Gregory C Minor Before the IEinois PoOution Control Board on behar of ReedCustar Casamunity Unic School District No. 255 U, re: M- - ' c= ?=' i Scyteanber 8,1908, Cme PCB 87 209. 71. Teenseoay of Oregory C. Minor in the U. S. District Court, Brooidyn, New York, September 3 190s, m = - af a^5 a f n h at Cane CV87 646. c 72, cm aa m-_ e * :. r-Rn' lasses 5,10, and 24, prepared by MHB Technical Aa=3a8= for The Ohio State Umberairy Nuclear Engineering Progress Esport Review Panel, I Public Utitry f= -- of Ohlo, Omober WOS. 73. Dired T xy and Exhibits of Dale G. 2f' ".'. Gengary C. Minor and Steven C. Sholly on Belif of Massachusetts Department of the Attorney Ocaeral, Re: Pilgrha Nuclear Power Station, Imusanannan af Pilarim Omaar. DFU 86 28, November 30,1988. 74. Supplemental Testimony of Dale G. O,- ' ;', Oregory C Minor and Simes C. Shop on Behalf of 58-y " m Department of the Astorney Oeseral, Re: Pugrim Nuclear Power Station, 'r = d E

• C=== DFU 88 28, January 20,1939 Enidhit AG 2, 75.

Testissany of Oregory C. Minor, U. S. District Court, Brooklyn, New York, February 3,1909, re: ca -av af * " + m LR en ar mL Case 37 QV. 646 (JBW). i 76. Surrobuttal Testlanomy of Dalt O. Brid== haugh, Gregory C Minor and Steven C shoDy on BahaK of uma Department of the Attorney General, Re: P9sries Nuclear Power Statica, L ^- ~'- af % rim Ow-DFU 86 38, February 13,1989, Exhibit AG 74. 77. Surrobustal Testissooy of Dale G. Bridenbesch, Gregory C Minor and Steven C Shouy oc Behnif of Massachusetts Departiment of the Attorney Ocaeral, Re: Pilgrha Nuclear Power Station, InAla= af Pilah Os-- DFU 88 28, February 17,1939, Eskibit AG-93. i i 9

l ATTACHMENT 2 Statement of Professional QuaH5 cation Steven C. Sholly i i I l j

ran85mONAL OlW MCAMONS OF s n e VEN c. EMnt3.y STRVEN C.EHOLLY MHB Tedmini Amoeistes 1725 HamCaos Avenue 1 Saite E SeeJose, California p5125 (40s)2s6 2n6 arreuBNG: t;' i 1985 PRESENT p__.^' - M T * '-1 AM - E-== h FM--J i Anodes. in energy consakiss rm that specianau in -a -u i and --,-ts of a 1 energy prodnetion facitieima, espedaDy suelear, for local, state, and federal somramaats and privets orgnaisations. MHB is amensively involved in., bsri proceediep and the properatina of st.dia ned reports. Co.d.a rasard, write = ports, partidesse in decovery proces is nemissary ; - " ;. develop wi-a y and aber for resstatory p -- " ;. and respond to cliset inquiries. Clients how included: State of California, State of New York,Staes ofInimais. February 19st-sepeseber 19a5 .,. <. e m. . y.. = -s m 1".=. u or er >~ - w* D& maamarch assedate and thk amanyet sor pubne imeerest youp bened in cambrides, F --

• m. that speciahms in --iAt the impen of advanced technoiosas on acaery, principany la the aroes of aram control and ry. Toshmicei ark W on ancisar power plant ans.cy, with emphasis on probabuisde rink -, radioiseises amersneer planning and pequ 'n. and generic safety issues. Conducted resserch, prepared reports and stadina partielpeted is adsdalstrative proceedags befort the U.S. Nucieer Regulatory F-

m demeloped testhmony, analysed NRC rais-making proposals and drat reports and prepare conseests thereon, and responded to ingairies fremt sponsors, the pseerst pobEs, and the media. Participated as a messbar of the Panel on ACRS E8emivenear. (1985), tbs Panel on Regulatory Uses of Probabinstic Rink Ammanament (Pear Review of NUREG-1050; 1994), Imited Observer to NRC Peer Review snestage on the soaree term r---=* (RhG 2104; 19851996), and the ' ', _ ' _ Advisory Committee on Noelcar Risk for the Nuclear Risk Task Peres d the Nadonal Anandath oflasureDeg rasnminain-*rs (1984). 1

6 January 1930-Jamanry1931 r..' - e - u y-- --a tz_= r=.,e,. The.e Mile r '-a Nm tar =s n=_.% Cantar.% Pammavhmaia Prodded adualmistrative direction and taordinated research prtincts for a potec interest groep based in Harrisbars, PennsylvaA centered around hsens related to the nros Mile Island Neelear Poser Plant. Proper d inmedraming proposals, tracked proyees of UJ. Nudoar Regalatory N ' 'a U.S. C., r r of Easegy, and General Putec Utihtaas actwitnes 1 commerning cleanup of Three Mile laiend Unit 2 and preparation for restart of Three Mile Island Unit 1, and magitared developments related to emergency piamming, the Saandal health of General Pubhc Utilldes, and NRC rata-aW anions related to nres Mile Island. 5 July 197s-Jaaaary1930 g. m..-.__. y _ n_ _ _ _,,,,, -_ _.-- T,=e -e t'._= L Tm. +'aM =1 Apharer. Marshaw. Penaswivansa Chief Biological Process Operator at a 2.5 alWas gallom per day tertiary, acdvared stadge, wassawater treatment plant. *,. - for biological prosses monitoring and control, including analysis of physical, clumaical, and biological test resuks, process Suld and mass flow management, micro bianasical analysis of activated aimles, and maintenance of 4e=3 tad process logs for impet 'anto stats and federal reports on treatseet process and aNiumar quality. Received certi$ cation froan the Commonweakh of Penmeylvania as a wastownser treatment plant operator. Member of Water PoDution Control A==aciaeta= of 7-C'. Central Secuaa,1930. July 1977-July 1973 w- - T-- -pu o,.,,. ' 1 ort - _.1-- = ;-- - c. - Wasteweter treatment plant operator at 2.0 =h sanon per day --tary, salvated sledge, westewater treas==r plant. Perforened tanks as assigned by supervisors, laciuding sinspie e phyncal and chassical toets on westownter streams, maintenance and operation of plant agaipment, and mainteanece of the collection systesa. Septeseber 19'/5-June 1977 e- = T '. west w=m_. m.e e.-- Hut. P--- i '- Taash Earth and Space Sdemos at aimah grade level. Developed and imple== rad new course materiah on plate tectonics, envira=== sal geology, and space science. Sernd as Assistaat l consk of the dioria gymnastica team. yember 1975-June 1976 e - _,_ _ m c _. gm _ _. m2 e,,u_i_ 7 -- _ m - _._ Taught Earth and Spect Sdence and Envirrmmental Science at siath grade level Developed and impicmented new course materials on plate taaonia, environmental geology, noise pollution, water aaW. and energy. Served as Advisor to the Soence Projects Club. 1

2.

• I u.

y 4 EDUCATION: 34, Edesation, majors in Earth and Space Scioece and Ocaeral Sdesce, minor in Enviremmental Education, *;; - * -a State Couage, Sheppensburg, Pennsylvania,1975. Graemas soares=ork in land Use Pleasins, Shipp: asberg State Counge, Shi;; i--,, Pennsylvania,19771978. PUBL2 CAT 10NE: 1. t_--- ' ' Mersabi Intensities from th; ;:: Reports,* lantanL6 Geningical Educatina. Vol. 25,1F77.

2. '

A M *-- d. Am 1 '-- - ' ^A-- - - d 6 e'4-h h um t *- ' y- '- - Paner Plant. Three Mile laland Pelic Interest Resoures Center, Hardsberg, Pennsyhenia January 1981. 3. A "' ' 9 ' ' rM== of the EsAr' ' r: ^, * ' * '- ' ' 9- - -- 7, : - - ^ ^ - Elsa, Union of Comoerned Scientists, prepared for Rockland Comaty Emergsacy Flamang Pernemaal and the Chairman of the County Legislature, Washington, D.C., August 17,1981. 4 h S'- - '- far a n- -7*"- A*- - c - .. la the. Pathway - er m. Actaca, and New Y Pubhc terest Reneerek Groep, Washington, D.C., Augast 27,1981.

• 5.

'tJaies of Commersed Seisesists, Inc., ca====e= on Notice of Proposed Rulemaidag, Amesubasat to 10 CPR S0. Appendia E, Sectica IV.D.3,' Unica of Concerned $destists, Washiagon, D.C, Octaber 21,1981.

• 6.

"The Evolution of Emergsasy Plasmin.g Rules,"__in N Mme Ph naan" A " ' = on the .. _ _ g g,_, _ 7_ 3,_. -. 7__ g,, g, pg,g Concerned Senatists (Washington, D.C.) and New Yark Pebbc Interest Reneerch Groep (New York, NY),1981. 7. " Union of Concerned Sdestists Comssents, n@ Rule,10 CPR Port 50, Essergency Flamaing and T.~; M Easrcises, Clarihmeta= of 8- '"--. 46 P.R. 61134," Union of Concerand Scientists, Washinyon, D.C., January 15,1982. = 8. Testimesy of Robert D. Po5ard and Steven C. Sholly before the Suh===lew on Energy and the Envirmassist, Cosamistes on laserior and lasular Affairs, U.S. House of Represostatives, Pennsylvania, March 29,1982, available from the Union of Commerned Scientists. 9. " Union of Concerned hiaatiara Detailed Comments on Petition for Ruh==W b T y Citiaea's Tank Parea, Emersoney Plemming,10 CPR Pans 50 and 70, Dodet No. PRM 50 31,47 F.R. 12639,' Union of Concerned Sdentists, Washington, D.C., May 24,1952. 10. Supplements to the Testimony of EUyn R. Weiss, Esq., Gescral Counsel, Ualon of Comoerned Scicatists, before the Subcommittee on Eastgy Conservation and Power, Committee on Energy 3-

and Commerca, U.S. Hoase of Reproacatatives, Unica of Concerned Sdesdats, Washington, D.C., Augut 16,1982. Tesdmony of Steven C. Shouy, Unica of Concerned Wm. Wasinaggon, D.C., os behalf of 11. the New York PubEc laterest Research Group, Inc., before the Special Comaktes em Nedaar Power Selsey of the Aansably of the State of New York, hearmy on Legisladve Overnight of the I Emergency F "" ? Preparedness Aa, Chapter 708, Laws of 1981, W-- i _ 2,1982. 12. 'Casamaats on 'Dratt Supplencet to Plaal Environmental Statsacat Related to Construction and: Operation of Chach River Breeder Reactor Plaat',' Docket No. D537, Unica of Comentand Scientists, Washington, D.C., Septessber 13,1982.

• 13.

' Union of Concerned Scientists C====** on ' Report to the Couary ca===ia la=,as*, by the Advisory Committee on Radiological Esecrgency Plan for Columbia Conaty, Pcansylvania," Unica of Cwd Scientists, Washington, D.C., 6=a-" 15,1982.

14. -

'm ? ' t # 8 ^:

- Planaims for Nuclear Reactor Accidents," prescated to Kwf.c Ontamentald Congress, Rotterdam, 'line Netherlands, Union of Commersed bie=eien Washington, D.C., October 8,1982..

i 15. "Neclear Ronctor Accident e- _r n = f=pnearia== for Radiological Emergsacy Planalag,* presented to the Citissa's Advisory Pa==inar to Review Rocidaad Comery's Own Nuclear Evacuation and h 7 :'---- Plan and Ocacral Disaster n; _ _ ' Plan, Union of Commersed Seisatists, Washington, D.C., Nosenber 19,1982. Testissomy of Steves C. Shady before the Sal-r===irtes os Ovesight and Investigatkm3 16. r-ia=- on Istariar and lasalar A5 sirs, U.S. House of Repressaa%s, Washington, D.C, Union of comesraad Sdestists, December 13,19t2. 17. Testimony of Gordon R. Thosspoon and Steven C. Sho0y as ra==ia=ia= Question Two, ra=*==ria=a 2.1(a) and 2.1(d), Unien of Concerned Scientists and New York Pubbc letsrest Research Groep, before the U.S.Naclear Regulatory fa===* ala= Atomic Safety and Llosasing g Board,is t,'4 Matter of ra==alidanad Edison Cosspany of New York (fadina Point Unit 2) and tbs Power Authoney of the State of New York (ladian Point Unit 3), Dodet Nos. 50 247.SP and 50 206.$P, Demeber 28,1982.

• 18.

Testimony of Stevu C. ShoDy on the Consequescas of Aaade=*, at Indian Point (ra==i=la= Questics One and Board Queados 1.1, Unica of Cometraed teie=*i=*= and New York Pebbe laterest Rammerch Oroup, before the U.S. Nuclear P1 '"=y ra==l-laa Atomic Safety and Lissasing Board, la the Matscr of ra==aadarad Edison Casapany of New York (Indian Point Unit 2) and the Power Authority of the State of New York (ladian Point Usk 3), Docket Nos. so.247 8P and 50 236.SP, February 7,1983, as corrected February 16,1983.

• 19.

Testinear of Steven C. ShoDy os Commisanos Question Five, Unico of Concerned tri==tiate and New York Public Interest Rasmarch Group, before the U.5, Nacisar Regulatory ca==W Atomic Safety and Licensnes Board, in the Matter of enanandated Edison Compeer of New York (Indian Point Unit 2) and the Power Astborky of the State of New York 3 (Indian Point Unit 3), Docket Nos. 34247 SP and 30 286.SP, March 22,1983.

• 20.

" Nuclear Reactor Aaeide=8= and Accidcat Conseqocaces: Planning for the Worst,* Union of Concerned Scientists, Washington, D.C., prescated at Critical Mass '83, March 26,1983. l 4

21. Testanoar of Steven C. Shouy on Emergsocy Flamming and Preparodaces at Comunercial Nasisar Power Plants, Union of Coacorned Scisadsts, Washengton, D.C, befors the $=h a==amaa on Nuclear Regulation, r= 'w on Environment and Pabin Works, U.S. Seasta, Apru 15,19R5, (with ' Union of Concerned Scicatists' Response to Questions for the Record base Senator Ales K. Simpeaa,' Stewa C. Shelly and Micheal E. Fedea). 22. "PRA: What Ces it Randy Tell Us Aboat Public Risk bom Nedear Acaldants?,' Unica of Conserned $ dentists, Washington, D.C, pressatation to the 14th Anneal Meeting, Seacoast Anti-Pollution laages, May 4,1983. 23. "Prak=hu1=*ie Risk Assessment: The Im Plamains and.L _ '-- - e- r='= pact of Uncertainties on Radiological Emergency - f Union of Concerned taiaastaan Washington, D.C, Jane 28,1983. 24. " Response to GAO Questions on NRC's Uec of PRA,' Union of Concerned Scientants, Washington, D.C., October 6,1983, attacherat to letter desed October 6,1983, frosa Steven C. Shelly to John E. Bagnolo (GAO, Washington, D.C). 25. The '- " af 'R= J F-- ' am e ^'-

• 1 *-

a-- -- Pt Gemidualissa Union of Comoormed haatiara, Washiagion, D.C., Deossaber 22, 1983, attachamat to letter dated December 22,1983, from Steven C 3hoDy to NRC c--i.aaname Jannes K. Aa==h*'=

26. -

SiseecS T Public laquiry, Proof of Evidence on: Safety and Wasta Management 1malications of_thrJiisaanDWR. Gordos Dompson, with sapporong evidamos by Stevea Shouy, on behalf of the Toes and Country Planalag A=a*Lat== February 1984, including Aasex G, 'A review of Probabilistic Risk Analysis and ks Applicatics to the Siaowe5 PWR," Stena ShoDy and Ourdon Tha=paan (August 11,1983), and Amnes 0, "Emergsacy Flamains la the UK and the US: A Cosepanson," Steven Sho8y and Gordon Dampaas (October 24,1983). 27. Testimony of Steven C. ShoDy on Esseresacy PlaW ca='a=rian Number Eleven, Union of Consermed Scientiaan, Washlagton, D.C, on behalf of the Pahmetto Alliance and the Carolias Environmental Study Group, balbre the U.S. Nucinar Ragalatory camadanian Atomac Safet - and 1Jonssing Board, is the Mattar of Deks Power Company, et. al. (Catsuba Nueimar Station, Units 1 and 2), Dodat Nos. 30 413 and 30 414, AprE ~6,1984.

• 28.

Risk ladbestors Reicvent to Assesslag Nuclear Accident Liability Preaniums,'la Frannalmary a , to th. = > _e _ er ^ - to the NAIC M='- n:=k Task h December 11,1984, Steven C Shady, Union of Concerned u-='a. Weslungton, D.C. 29. ' Union of Cosenraad Sdentists' and Naclear Information and Resource Service's Joint Commaats en"NRC's Proposal to Bar hom 13 ceasing Proceedings the Consideration of Earthganho ESocas on Emergemey Flamaing,' Union of Concerned Scieetists and Nuclear leformanian and Resource Serves, Washington, D.C., Diane Curras and EDya R. Weiss (with lapet frosa Stessa C Shelly), February 28,1985.

• 30.

" Severe Aasident Source Terms: A Presentation to the=--!='=-2. on the Status of a Renew e of the NRC's Souros Term Reammassment Stady by the Union of Ca W %=ia ' Union of Concerned Scientists, Washington, D.C, April 3,1985. * -____-_- - _ -

31. " Severe Aasident Sowca Twas for IJaht Water Nuclear Power Flants: A Praamarath o the t Illinois E:. ^ of Nuclear Safety on the Status of a Review of the NRC's Source Term a--======* Study (STRS) by tbs Unica of Concerned We Union of Cooceroad tai==85% Washington, D.C, May 13,198$. j 32. 'rha * - - T==== r ' JA * "- - d the C_._ ' "- far E " h "-. ; A- ' ' - ^*==_ T_ __ =dsk " " 2-on the NRC " =- Tar = 8 - - - 7_s (NURFC. 33Q, Union of Concerned Saentists, Cambridge, Massacheneus, Steven C. Shouy and Gordon Thompson, Jamnary1986. 3 33. Direct Testimony of Dale G. Bridenbeugh, Oregory C. Minor, Lynn K. Price, and Steves C. Sk3 as behalf of State of e--- --% E - r of Pablic Utnity Control, Prosecmarmi % sion and DMaion of Consumer Coannel, regarding the pruslance of expendamres on M detone Unit III, Febevery 18,1986. 34. Impikations of the Chernobyl4 Accadent for Nuclear Emergsacy Plashing for the State of New York, prepared for the State of New York Consumer Protection Board, by MHB Technical Amanada,ma Jgne M 3$. Rawlaw of V r = e Y " t h e'- at * ' ^. e- ^- " A - ' - *-. d r> = ^ - " V " - r- - - 1b, the V_ _.e y s-x==--- T-F'== prepared for New England Coalition om Nuclear PoEstion, Inc., December 26,1906. 36. AfEdevit of Steven C. ShoEy before the Atomic Safety and Ucassing Board, in the maner of Public Service Casapany of New Hampshire, et al, regarding Seabrook Stados Units 1 and 2 Off site Emergsscy Planning Issues, Docket Nos. 50 443 OL & 50444 01, January 23,1987. 37. Direct Tanel=any of Richard B. Hubbard and Stanea C. Sho8y on behalf of California Pubbe UtiEties V- ' *. regarding Diablo Canyon Rate Case, PGdLE's Pailure to Estabhsh Its r ra===*tod Desiga OA Program, Appucation Nos. 84 06 014 and 854W25, Eskibit No.10,935, March,19g7. 38. Testinomy of Oregory C. M1mor, Steven c. shony et. al. on behalf of safreak Cousy, regarding LIILO's Recepoca Centers (Planning Basis), before the Atomic Safety and f transing Board,la the saatter of long Island Lighting Company, Shoreham Noclear Poner Statica Unit 1, Dockat No. $0 322 Ole 3, AprB 13,1987. 39. Robettal Testissomy of Oregory C. Minor and Steven C. Shogy on behalf of Suffolk County regardag LILCO's Reception Centers (Addressing Testimony of Lewis G. Hulanan), Docket No. 30 322-01-3, May 27,1937. Review of selsdod Aspan of NURso 1150, Renaar Risk Reference Document," prepared 40. for the lRisois Deperament of Nuclear Safety by MHB Technical Nataa September 1937. Direct Technomy of Richard B. Hubbard and Steven C, shouy os behalf cf the Pennsylvania 41. OfEco of Consumer Advocate, before the Pennsylvania PubEc Utility cr===ianlaa, Evaluation of Beaver VaRoy Unit 2 Plam Costs, OCA Statement 6, Docket No. R 870651. October 23,1937, 42. Flaal Reaart: Cl-a:Aema* Facenen Ahedan the Coat of n.. war Valley ir,; ; trada= Unit 1 prepared for Pennsylvania Office of Consumer Advocate, by MHB Tochacal Annociates, OCA Exhibit 6A. October 1987. +

b 43, &aam -'7 - - p.aon A-e the t w g - u y_._, p m e,.,ta= Unk_1 prepared Ier Pennsylvania Of5cc of Consunwr Adecate, by MHB Technical Aaaae araa, OCA Baddiit 44,Ostaber1987, 44. serretuned Tamh= Peas Usky e-y of Richard B. Hubbard and sesves C shony before the Pennsylvania , on behalf of the Pennsylvania omos of Consumer Ad catr, regnadagIhaluation of Beaver Valley Unit 2 Maat Costs OCA Statassent 61, Docket No. R. 810651, Desember 7,2937. 45. Testimony as Diablo Canyon Rate Case, Daniga Onality Assarance. Supplemental and Rebuttal Testimosqy of Richard B. Hubbard and Steven C Sho8y, ce behalf of the California Public ' U tE kies P ~ '^ * -- Divisias of Ratspeyer Advoestas, Applicarw= Nos. 84 06 014 and 854 025, Euhlbk No.16,a00, '"r**" 18EB-46. Testimony on Diablo Cearom. Rate Case, = = ef QA B ^ - And M U '- --- - - Rv %s N _ ' - T r' "' F - AsA13 - - " Taal Vohames I and II, *.,' "M and Rebuttal Testianomy of Richard B. Hubberd and Steves C Sho8y on behaK of tbs California Pcblic UtDities th=W Divinica of Ratepayer Advocate, Application Nos. 845014 and 85 08 025, Esbibk No.16,650, September 1988. 47. OE Raed Report Safety Issue Reviews, Issues 1, 6, and 14, prepared by h05 Technical Associates for The Ohio State Universey Nuclear 8*-j -- '.-; Program Espert Review Panel, Pulde Utility r%==i=6a= of Ohio, October 1938. 48. Direst Testimony and Ealdbits of Dele O. Bridsabaugh, Oregory C Minor and Steven C sholly on Behalf of Massachusetts Departacet of the Attorney General, Re: Pilgrim Nucicar Power Station, Isvestignalos of Pilgrim Outags, DFU 8g.28, November 30,1988. 49. Supplameetal Testbacey of Dale G. Bridsabasgh, Oregory C. Minor and Steven C Shouy on BehaK of Massachusetts Department of the Attorasy General, Re: Pagrian Naclear Power Stadua, lavestigation of PBgrum Ontags, DPU E28, January 20,1989, Eskibit AG 2. $0.' Surrebuttal Testissosy of Dale G. :" * - f Oregory C Minor and Steven C shony as BabaK of Massachusetts Department of tbs Attoasey General, Re: Pilgrian Naclear Power Station, LM ^'= of Pilsnm Outags, DPU 88 28, February 13,1991, Eskibit AO-74. SL Surrebettal Taarima=y of Dale G. BrW Gregory C Minor and Stseen C ShoDy on Behalf of Maa mehe ee= Department of Attoracy Osecrat, Re: Pilgrim Nuclear Power Stanon, Investigation of P9prins Ou: age, DPU 86 28, February 17,1999 Eskibh AG43. g g,,,,, f __ o _..,, n, - r - > Unir 1. sc. Ar e>=- c'h=== ch for g _ 2.. '*"-- -- h - c -- Plan r-prepared for Institute for Resource and Sasurity Stades,Pubruary1989. Available froan the U.S.Nucacar Regulatory (%==saaman Public INeumcat Room, lobby,1717 H Street, N.W., Washington, D.C.

.m J / t a. b EXHIBIT E s 1 l l I J

E i . :( [ ' I IRT NHY MANAGRENT EFFECTIVENESS ANALYSIS REPORT i ITY Performance During and Following The Natural Circula1. ion Test of June 22, 1989 ISSUED: July 10, 1989 l l j >. 6

g TABLE OF CONTENTS IIT ERY * "*BEENT EFFECTI W M 5 Am*1YSIS REPORT SECTION _ PAGE 1.0 Executive S"==ry 1 2.0 Analysis and Review 2 A. Ca===nd and Control Pollev_ 2 1. Background and Existing Policy 2. Application of the Command and Control 2 Policy. 3. Conclusions 3 4. Command And Control Policy Recommendations 5 6 B. Procedure Comoliance Policy 7 1. Background and Existing Policy 2. Application of the Existing Procedure 7 Compliance Policy 3. Conclusions 8 4. Procedure Compliance Policy Recommendations 10 11 C. Post Trio Review Polley 12 1. Background and Existing Policy -2. Application of the Post Trip Review 12 Policy 3. Conclusions 13 4. Post Trip Review Policy Recommendations 24 27 D. Steam Du== Valve Incounlete Work Recuest 29

===1. Background=== 2. Conclusions 3. Roccamendations \\ ( i l

TABLE OF CONTENTS IRT NHY MAMAGEMENT arrmariv e M S m '-YSIS REPORT l-ATTACHMENT: NHY MANAGEMENT EFFECTIVENESS CHRONOLOGY, SECTION PAGE Thursday.-June 22. 1989: 1 1219--- Initiation of Natural Circulation Test (1-ST-22) 1 NRC Communications To NHY In Main Control Room 1 1236 -- Manual Reactor Trip 3 1350 --- VP-NP Communications To President and CEO, VP-Engineering, Licensing and Quality Programs 5 1500 -- Post Trip Review Critique In TSC 7 1800 -- NHY Telecon To NRC Projects Branch Chief 9 2315 -- VP NP Briefing Of President end CEO 14 Friday. June 23. 1989: 15 0730 -- NHY Telecon To NRC Projects Branch Chief 16 1140 -- NHY/NRC Meeting 20 1350 --. NRC/NEY Confirmatory Action Letter Telecon 23 1430 -- NHY Meeting To Plan and Schedule NHY CAL Response Activities 25.

1 NHY Mansaament Effectiveness Analysis RecoA

l..

L p 1.0 Executive Summary New Hampshire Yankee (NEY) conducted a review and analysis of the policles, . procedures and implementing actions taken by NHY personnel to assess the adequacy and effectiveness of the NNY Management response during and subsequent to the Natural Circulation Test conducted at Seabrook Station on June 22, 1989. Short and longer term corrective action recommendations have been developed and are presented in this repcrt based on the review and analysis conducted by the NHY Independent Review Team (IRT), a staff group not functionally associated with the NHY groups that conducted the Low Power Test Program. The scope of the IRT review was specifically f:c4 sed on, but not limited to three NHY policy areas as they relate to this event: Command and Control Procedure Compliance Post Trip Review i i NHY has established written policies and haplementing procedures and has conducted training for NHY personnel as part of the normal administrative process for providing guidance on the conduct of operations for Seabrook Station. The results of the IRT review and analysis form the basis for recommending enhancements in each policy area. Each of the following analysis and review sections of this report provides a background and summary of the NHY policy in effect at the time of the event, provides a discussion of the application of that policy during and subsequent to the event, provides the IRT management effectiveness analysis conclusions and provides recommended corrective actions for each of the above three policy areas.

? l l 2.0 ' Analysis and Review ~ A, 'Co==and and Control Policy 1. Backaround and Existina Policy [ A series of hierarchical NHY manuals describe the lines of authority delineating responsibility for command and control at Seabrook Station. The Seabrook Station Management Manual, Operations Management Manual and Startup Iest Program Description provide the guidance applicable'to the operation of the Station and conduct of testing. The Operations Group is responsible for ensuring the safe operation of Seabrook Station in accordance with-regulatory guidance, technical specifications and approved procedures. NHY uses six. shift crews of Operations. Group personnel to provide continuous manning of the station. Each shift consists of a Shift Superintendent. Unit Shift Supervisor, a Supervisory control Room Operator, Control Room Operator and Auxiliary _ Operators. The Operations Management Manual details the responsibilities, qualifications, authority and source of direction for ~ each shift position. Each licensed operator retains the authority to either order the shutdown or to shutdown the. reactor. The Unit Shift Supervisor is the shift position normally responsible for maintaining a comprehensive perspective on operational conditions and in an emergency or transient situation becomes the sole authority in charge of the ' Control-Room unless relieved of that duty, t The Startttp Test Program Description describes the organization, responsibilities and procedures which ensure that components, structures and systems are tested to verify that they function as designed. The Startup Test organization is structured to provide parallel counterpart positions to the operations Group. Under the overall direction of the Startup Manager, each Startup test shift is comprised of a Shift Test Director, Test Director and-Test Engineers. These personnel are trained and qualified to perform assigned responsibilities, but are not licensed operators. The Startup Test Program consists of specific individual tests performed during and subsequent to initial fuel loading. These tests 2

vsrify the prepar dosign of the reactor core and the dynamic characteristics of the plant during both normal and anticipated abnormal operating conditions at power levels up to and including 100% design reactor power. Individual test procedures for Low Power Testing wete written by the Startup Test organization, were reviewed and approved by the Startup organization and the Stat: ion Operation Review Committee. In addition, all FSAR listed Startup Test procedures were reviewed by the NRC and selected 3tartup Test procedures including 1-5T-22, were reviewed by the NSSS supplier (Westinghouse). The plant operation for each test was done by the Operations Group with technical assistance and direction provided by the Startup Test organization to ensure the technical adequacy and satisfaction of test parameters. The Startup Test organization retain 1 the authority to interrupt or tenninate tests based on the conduct of the test or plant parameters. The authority and responsibility for overall plant conditions, including the conduct of tests, and reactor shutdown is vested only in licensed operators.. The Operations and Startup Test organizations developed a working relationship, policy and practice whereby Startup personnel provide the technical direction for conducting the test and operations personnel implement the test. 2. Acolication of the Command and Control Policy The conduct of Startup Test-22 Natural Circulation, was initiated using the standard practice where the Startup Test Director briefed the operations shift on the conduct of the test, expected plant performance and limiting conditions for the test. The Test Director provided copies of Attachment 9.3 to 1-ST-22. Expected Plant Response and Manual Trip Criteria (which included pressurizer level less than 172), to the Shif t Superintendent, Unit Shif t Supervisor (US$), and licensed board ope rat or s. During the actual conduct of the test the following events occurred: (a) One of two primary board operators focused primarily on pressurizer level, letdown and charging. This operator monitored and audibly informed the USS of pressurizes levels several times throughout the test. 3

(b) The Unit Shift Supervisor was cognizant of pressurizer level I throughout the test. The USS conferred with the primary side board operator several times and informed the ' Director that pressurizer level was approaching and later o selow 172. The Test Director did not subsequently recall receiving this information. (c) A NRC inspector informed the Startup Manager that pressurizer level was below 172. (d) A NRC inspector informed the Test Director (second NRC to NHY notification) that pressurizer level was below 17%. The Test Director informed the USS. The USS acknowledged this communication. (e) Two NRC inspectors subsequently informed the Assistant operations Manager (AOM), (third NRC to NHY notification) that pressurizer level was below 17%. The AOM confirmed the test trip criteria with the Test Director and that the pressurizer level status information had been conveyed to the USS. The AOM began to more fully assess plant status to determine the need to personally intercede with the USS actions being taken. Prior to completing the assessment the USS had directed the second of the two primary side board operators to shut.down the reactor. (f) After tripping the reactor, the operators entered their post trip emergency procedures and brought the plant to the HOT STANDBY condition. The following fact' ors influenced the application of the Command and control policy. The operations Manager and Assistant operations Manager were aware that the test procedure contained trip criteria but were not aware of the specific trip criteria (172 pressurizer level) cited in the startup procedure. The Shift Test Director and Test Director were involved in detailed in-process data gathering and analysis to ensure that certain test parameters were maintained. The Shift Superintendent heard the report that letdown had isolated l 4

(pressurizer level loss than 177) but continued with his personal efforts to identify the cause of the unexpected cooldown. He was focused on the Tavs trend and on the condenser steam dump valve performance. The Shift Superintendent was not aware of the communications between the NRC. Test Director and Unit Shift Supervisor and was not aware of their substance. Due to the unique nature of this test and a FSAR commitment for-operator training, a total of fifty-seven people were in the Control Room to conduct or observe the test. The Control Board was manned with supplemental primary and secondary side operators. Three additional full shift crews of operators were also present to observe the test as part of an FSAR training commitment. Sumary: 6 - NHY. Management 15 - Operators (On-Shift) 8 - Test Personnel (On-Shift) NRC 16 - Operators (Observation training to satisfy FSAR Commitment) 4 - Training Department Personnel 3 - QA/QC 1 - ISEG _J,- SAT 57 - TOTAL 3. Conclusions Analysis and comparison of the existing Command and Control Policy with the events of June 22, 1989 provides the basis for the following conclusions: (a) Operations Department Command and Control Policy functioned as designed if consideration is ' trictly 1Laited to the Limiting s Conditions for Operation sections of Seabrook Station Technical Specifications and the Operationu Department Operating procedures which provide'the bounding criteria for safe plant operation and transient recovery. However, the Operations Command and Control Policy did not function as designed with respect to adherence to 5 L -~

[f s o oy L

tho Sosbrook StCtion Administrative Controls. Policy and Guidance-lt

-associated with procedure' compliance. (b)' The members of the' assigned Operations shift crew should have exercised their authority to either direct a shutdown or to L shutdown the reactor when' the pressurizer level fell below the 17I test'specified trip criteria. (c) The Startup Test 1 Director shoald have strongly and immediately recommended shutting the reactor down, in a clear precise manner to the operations individual in command (USS), when the test trip criteria was reached. (d) The.Startup Manager and Test Director should have been immediately responsive to the information provided by the NRC inspector. ~ (;; The' addition of supplemental board operators could have been more fully' supported through more specific training'and simulator exercises designed to practice command and control functions unique to the 1-ST-22 test and the actual crew assigned to conduct the test. 4~ ' Cn-nd and Contro-1 Poliev Recomsnendations The following corrective actions are rec mmended: a (a) Revise the existing Command and Control policy to clarify the integration, participation and input of the Startup Test organization and other groups that interact with the shift operatore concerning station operations. -(b) Revise the existing Command and Control Policy to delineate responsibility and authority when supplemental operators are assigned on shift. (c) Revise the existing Command and Control Policy to specifically encourage non-shift. licensed Operations personnel to provide 6 l

points of clarification or information when ar, assigned operator's actions appear to be inappropriate or not understood by the observer. B. Procedure comoliance Poliev 1. Background and Existing Policy The Seabrook Station Technical Specifications NHY Operational Quality Assurance Program Seabrook Station Management Manual and Operations Management Manual collectively delineate the criteria for issuing approved procedures and assuring that implementing activities are conducted using approved procedures. In addition, the.Seabrook Station Management Manual (Chapter 2 Section 1.5) and Operations Management Manual (Chapter 3, Section 2) provide specific guidance regarding adherence to procedures. The Operations Management Manual provides instructions for Operations procedutes including the following: " Plant oparation should be conducted in accordance with applicable procedures. If procedures are deficient, a procedure change should be initiated. An exception to this policy is that in emergency conditions operators may take whatever action is necessary to place the plant in a safe condition, and to protect equipment, personnel and public safety without first initiating a procedure change.' The recent Self Assessment Team review of existing Operations procedures indicates that procedure revisions are approved and issued per aristing administrative guidance. Ch.nges to procedures are - prepared and issued to correct identified procedural deficiencies. A further review of recent (1989) Station Information Reports indicates that'the Operations Group has experienced two incidents where additional policy clarification (valve and component position verification) in implementing procedures required incorporation. Corrective actions to incorporate the enhanced policy and provide initial anr ntinuing training were taken. 7

NHY personnel recently attended a regional forum where the NRC { discussed adherence to procedures as a-key industry factor for necessary performance improvements I More recently (March), the Operations Group. Station Management and Self Assessment Team each proposed'the formation and staffing of a group of personnel within the Operations ^;aup specifically dedicated to perform periodic procedure reviews, reducing the number of approved changes not incorporated in procedure revisions, procedure preparation, and a one time special consistency review for all Operations procedures. The March proposals also recomended the addition of one senior level operator position on each shift, who in turn would be assigned responsibilities designed to further relieve the USS of administrative duties. The NHY Management oversight Comittee of the Self Assessment Team accepted the recommendation. NHY has established a schedule for implementing the recommendations by the end of 1989. The Startup Test organization prepared and issued test procedures several years ago to support the proposed low power testing scheduled to be conducted subsequent to the 1986 core load.. These FSAR designated Startup test procedures were prepared per the Startup Test Program Description format, were reviewed by 'he NRC, Westinghouse, NHY t Quality Assurance, and were approved by SORC. Prior to conducting the 1989 Low Power Test Program, the technical content of these procedures was re-reviewed by Startup and Quality Assurance, and recomended changes were approved by SORC. 2. Annlication of the Existina Procedure Come11ance Poliev Startup Test 1-ST-22. Natural Circulation, was revised, approved by SORC and reissued on April 14, 1989. Field Change Number 1, which provided additional instructions regarding references, initial conditions for secondary plant warmup and typographical corrections, was issued June 21, 1989. 8 I

4 I Discussions with Operations Group personnel indicated that they had the opportunity to read the entire test procedure and that they were aware of the manual trip criteria listed on Attachment 9.3 of Startup Test Procedure 1-ST-22. Prior to conducting 1-ST-22 the Test' Director briefed the assigned crew and made individual copies of the manual reactor trip criteria (Attachment 9.3) for the operators. The attached chronology indicates that the Control Room operators were aware of the f stated manual trip criteria and, associated with one of those criteria, had started timing a 15 minute period to trip the reactor unless they could recover from a Tavg of less than 541*F. In addition, they were calling out pressurizer levels auring the conduct of the The USS specifically informed the Test Director of approaching test. the procedure trip criteria of pressurizer level less thsn 172. The Test Director did not appear to have acknowledged these communications regarding the approach to trip criteria nor the subsequent communication by the USS that pressurizer level was below 17Z. The USS subsequently directed the primary board operator to inform him when pressurizer level approached 15Z. Concurrent with interactions among the Control Room operators, there were a ;, aries of communications between the NRC and NHY Startup and Operations personnel regarding the level of the pressurizer below the 172 manual trip criteria setpoint. Discussions with Operations personnel subsequent to the event indicate that they recognized the non-Technical Specification test parameters as conditions for Startup to either interrupt or terminate the test. The Operations personnel also indicated that based on their own operations procedures and training they did not consider the non-Technical Specification. Startup test procedure trip limit at 172 pressurizer level to be a itniting criterion because of the bounding nature and' inherent conservatism for safe operation incorporated in i the more familiar Excess Cooldown and Loss of Letdown Recovery system operating procedures. Prior to leaving the Control Room, the Vice President - Nuclear Production. Station Manager. Operations Manager and Assistant operations Manager were aware of the procedure violation. Within one and one half (1 ) hours of the event all of NHY's Executive Management was aware of a procedure violation and the need for corrective actions 9

on procedure adherence. During the course of the day. NHY management discussed the procedure violation and the need to revise the policy regarding adherence to procedures. This was the sole topic of a one and one-half hour meeting on Thursday evening where the revised policy and corrective action options were discussed. On Friday morning a l revised. policy was established and it was directed that the revised policy be communicated to the Operations Group personnel through verbal presentations by Station Senior Management. NHY Executive and Station management met at 0645 on Friday June 23, 1989 to determine the proposed policy and schedule for corrective action. NHY subsequently related the revised policy and schedule for proposed corrective actions to the NRC at 0730 on Friday, June 23. 1989 and began implementation later that day. 3. Conclusions Analysis and comparison of the existing NHY procedure adherence policy with the events of June 22 and 23, 1989 provides the basis for the following conclusions: (a) The assigned operations shift crew should have followed the manual trip criteria in ST-22 and should have shut down the reactor when pressurizer level went below 171. (b) New Hampshire Yankee did not fully comply with its own Administrative Controls, Policy and Guidance associated with procedure compliance. 1 (c) Yhe Station and Operatiens Management policy on procedure adherence requires additional clarification to clearly delineate those circumstances where it would be acceptable to deviate from i a SORC approved or Quality Assurance related procedure. 1' (d) The Operations personnel believed that the Startup Test Procedures, by their nature, were inherently flexible and did not require verbatim compliance. 10 I

[. (e) Startup Test personnel should have immediately considered and taken appropriate actions to recommend shutting the reactor dcwn i after being informed by the NRC that the plant condition had exceeded the indicated test trip criteria. 4. Procedure Compliance Policy Recommendations The following corrective actions are recommended: (a) Revise the Policy regarding adherence to procedures. This policy should clearly delineate under what conditions it might be acceptable to deviate from an issued approved procedure. These conditions must be based on compelling safety reasons: (1) Protecting the public health and safety. (2) Preventing personnel injury or life threatening situations. (3) Preventing plant system, component, or structure damage. (b) Incorporate the revised Policy in established administrative and management manuals. (c) Ensure that all NEY personnel receive initial and continuing l training on the revised policy for adherence to procedures. (d) Revise the Startup Test Program and procedures to (1) Rewrite and issue those Startup procedures, that would not ordinarily exist as operations procedures, as Special Procedures that support infrequent evolutions and testing. The preparation, alpha-numeric designation, review, approval and distribution of these procedures should be brought into full compliance with the existing guidance in the Station and operations Management Manuals. 11

s (2) Provide additional guidance for terminating the test'and exiting the test proc.tdure when plant. transients or equipment malfunctions occur during the conduct of tests. (3) Establish a program to conduct additional review and rehearsals for the test crews in the classroom and on the simulator, as necessary, when complex or unfamiliar test procedures will be conducted. C. Post Trio Review Polley 1. Baektround and Existing Poliev New Hampshire Yankee reviews and analyzes unplanned reactor trips, Engineered Safety Features (ESP) actuations and other similar significant operational events through a series of integrated procedures performed by the Operations Group, an independent Event Evaluation Team composed of NHY individuals with the requisite experience and training to provide immediate analysis, a Root Cause analysis conducted by the Reliability and Safety Engineering Department and two other overall mandatory reviews by the Station Operation Review Committee (SORC) and the Nuclear Safety Audit Review Committee (NSARC). The specific requirements and conduct of activities are delineated in the followings (A) Technical Specification 6.4.2.7 'The NEARC shall be responsible for the review of e. Violations of codes, regulations, orders, technical specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.' g. ALL REPORTABLE EVENTS * (B) Technical Specification 6.5 - Reoortable Event Action 12

The following actions shall be taken for REPORTABLE EVENTS: The Commission shall he notified and a report submitted a pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b. Each REPORTABLE EVENT shall be reviewed by the 50RC and the results of this review shall be submitted to the NSARC and the Vice President - Nuclear Production. (C) Operations Procedure OS 1000.08 Post Trio Review. (D) NHY Procedure 12830. Event Evaluation and Reduction Program. (E) NHY Procedure 12810. Root Cause Analysis. (F) NHY Reporting Manual Station Information Reports. The intent of these activities is to provide a thorough, detailed analysis of the physical plant, as well as the procedural and personnel factors that vera the root cause or contributing factors to the event. By design, portions of this analysis must be initiated and the entire Operations Post Trip review (OS 1000.08) must be completed to the satisfaction of Station management prior to an authorized restart of the reactor. 2. Aeolication of the Post Trio Review Poliev The attached chronology provides a detailed account of the substantive activities and communications relevant to the post trip review process i and NHY involvement in reviewing and analyaing data, developing resolutions to identified problems, and implementing corrective actions. The following narrative addresses certain key events and activities and incorporates the perspectives of NHY personnel interviewed as part of this report. 13 l l

was in a stable condition. Making use of the increased number of operators present, the Assistant Operations Manager directed the on- . coming shift Superintendent and Unit Shift Supervisor to begin the post trip review data collection in parallel with the assigned shift performing the notification process. At 1320 the assigned Shift Superintendent completed the NRC offsite notification. This communication reported the following: l "While performing low power physics testing, Startup test procedure 1-ST-22 Rev. #2 (Natural Circ.), Tavg had reduced to below 541*F, pressurizer pressure at 2340 PSIG and pressurizer water level less than 172: these were manual trip criteria per procedure. MS-P3011 stuck open causing initial cooldown." The INPO Nuclear Network (PS section) contained the following summary of the June 22, 1989 Seabrook reactor shutdown based on this communication. Facility. Seabrook Event Date 06/22/69 Time 12:35 EDT Unit

Unit Emergency

Not Applicaole Reactor Trip (1)

Auto with T.od Motion Reactor Trip (2)

Reactor Trip (3) Reactor Critical

Y Prior Mode Startup Prior PWR 003 Current Mode Hot Standby Current PWR W1TH THE UNIT AT 32 POWER AND ATTDiPTING TO ESTABLISH NATURAI, CIRCULATION AS PART OF A LOW POWER PHYSICS STARTUP TEST PROCEDURE. A MANUAL REACTOR TRIP WAS INSERTED DUE TO A TRANSIENT INITIATED BY A FAILURE OF A STEArt DUMP TO THE CONDENSER. DURING THIS TEST A STEAM D W FAILED OPEN.

THE STEAM DUMP WAS THEN CLOSED BY CYCLING IT. THE TE&T WAS CONTINUED. HOWEVER. THIS SEQUENCE WHILE ATTEMPTING TO ESTABLISH NATUeAf-CIRCULATION RESULTED IN TfAVE) BEI!vG REDUCED TO 541 pf9REES. PRESSURIZER LEvil BEING REDUCED TO JUST LESS THAN 17! AND THEN PRESSURIZER PRES,jg 1 INCREASING AGAIN TO 2340 PSI. THESE WERE ALL CRITERIA [9J MANUALLY TRIPPING THE UNIT PER THE STARTUP TEST PROCEDURE. ALl4 SYSTEMS FUNCTIONED AS REQUIRED FOR THE TRIP AND UNIT IS NOW STABLE IN MODE'3.. PRESSURIZER LEVEL HAS BEEN RETURNED TO NORMAL. FOUR RFACTOR COOLANT PUMPS WERE RESTARTED WITHIN ABOUT 30 - 45 MINUTES. 14 w_

~(Emphasis add:d) The three trip criteria indicated in this report were three of several trip criteria-listed in Attachment 9.3 of 1-ST-22. Comparison with actual events, the procedure and technical specifications indicate that (1) The Pressurizer level went below 17% for approximately five minutes but had been recovered to 21% at the time of the manual reactor trip. Pressurizer level below 172 is a non-Technical Specification trip criteria listed in 1-ST-22. (2) The Tavg less.than 541*F criteria has a 15 minute window to correct the condition (Te:hnical Specification 3.10.3 &q11gg Statement b) or to place the reactor in Hot Standby within tha next 15 minutes. Tavg went below 541'F at 1232:50 and was noted by Operations personnel. The reactor was shut down prior to the action required by this criteria. (3) The USS directed the shut down of the reactor in anticipation of reaching the pressure trip criteria of 2340 PSIG. The actual peak pressure was 2310 PSIG and therefore this trip criteria was not exceeded. At 1423 on June 22, the Test Director initiated a Station Information Report (SIR), which is an additional method for NHY to document an event or condition for further review and analysis. The SIR becomes the document which encompasses or includes the Event Evaluation Team Report and the Root Cause Analysis. The SIR includes recommended 15

,1 corrective actions which are reviewed by SORC and NSARC. SORC and NSAAC review the entire package for the adequacy of the review. analysis and recommended corrective actions. y The on-coming Shift Superintendent and Unit Shift Supervisor began implementing Operations Procedure OS 1000.08 Post trip Review, at 1250 on June 22. This procedure provides a detailed checklist, with required sigroffs, that focuses on gathering Post Trip /SI data, evaluating the data and making a recommendation for plant startup..The assigned SS, USS and SCR0 completed the post trip review and. submitted it to the Operations 1 tanager at approximately 1700. The Station Manager reviewed the past trip review document package at 1755 and indicated that reactor restart would be contingent upon his personal approval. The Operations notification process includes contacting the assigned Site Emergency Director (SED) for any reportable event. NHY Procedure 12830. Event Evaluation and Reduction Pronram, further requires the SED to assign an Event Team Leader to begin data gathering and analysis, assist the Shift Superintendent in the completion of OS 1000.08 and to cossnunicate this information to the

• Director of Corporate Communic tions
• NSARC Event Reduction Committee Members
• Station Maceger
• Vice President - Nuclear Production
• Vice President, Engineering, Licensing and Qur.lity Programs
• President 16 9

o

l I

• NRC Resident Inspector l'
• Regulatory Services Manager

.i

• ISEG. Supervisor t

I

• Event Evaluation Team Members j

$'l'l This procedure also' integrates the parallel activities associated with j ( the Post Trip Review and Station Information Report processes and serves as an input to the Root Cause Analysis Procedure 12810. The chronology indicates that the assigned Site Emergency Director initiated the Event Evaluation Team (EET) at approximately 1330.- The Event Team Leader began the data gathering and analysis and' assisted the assigned Operations shift crew in completing the Post Trip Review. The'EET debriefed the assigned shift crew and had the sequence of physical plant events and responses completed prior to the Station Manager's. post trip review critique at 1500 in the Technical Support Center. The Event Evaluation Report will be attached to the SIR and reviewed by SORC and NSARC. At approximately 1515 on June 22. the. Station Manager commenced the verbal debriefing of the events that caused the reactor shutdown. The preliminary results discussed at this meeting were (1) The sequence of events, starting with the initiation of 1-ST-22, ' based on various plant data collection systems. (2) The primary side of the physical plant responded to the events and reactor shutdown as anticipated. 17

5 (3) The secondary side of the plant responded to'the events as anticipated and designed with the' exception of the condenser steam dump valve. (4) A work request was initiated to correct MS-PV-30ll, the condenser steam dump valve that caused the excessive cooldown. (5) A work request was initiated to investigate what evolved to be an unrelated 'D' Main Steam Isolation Valve position problem (an air maintenance pump sticking pilot valve lead to component

eplacement to resolve the problem).

(6)- The Startup Test Procedure, (1-ST-22), trip setpoints were adequais and required no revisior.s. (7) The Operations Post Trip Review, EET and SIR had been initiated. The Station Manager reviewed the actions necessary for completion of the event evaluation to support a tentative restart the next day. Discussions regarding the violation of procedures and procedural corrective actions were intentiona77.y excluded in order to focus the discussion on physical plant. This critique ended at approximately 1630. At approximately 1630 the Vice President - Nuclear Production and Station Management met to prepare for an 1800 conference phone call with the NRC Region I, to explain the events that transpired during the 18

-, = - ,+ i Natural Circulation Test and to discuss the corrective actions that NHY j i either had in progress or anticipated.to be necessary prior to restart. I A discussion of the Startup procedure and Station policy on procedure adherence' concluded with the recognition that the reactor should have been shut down. the recognition that the procedure adherence policy required resolution prior to restarting the reactor and that Operations personnel should be' indoctrinated on any policy changes prior to restarting the reactor. The discussion also included consideration of the operator's actions based on normal operating parameters and procedures which would have explained the operator's actions from a technical and operational basis. This brief period, prior to the 1800 telephone conference call, did not provide sufficient time to fully develop and finalize the proposed policy revision. At 1800..NEY, on-site NRC inspectors and NRC Region I personnel conducted a telephone conference call to discuss the reactor shutdown. NHY described the chronological sequence of events and actions currertly in progress to determine and correct the failure mechanism for the condenser dump valve and the position indicator for the 'D' MSIV. NEY also indicated that the physical parameters of the plant responded as anticipated and designed and that based on current maintenance activities. It was estimated that the plant could be made ready for restart by 0700 on Friday, June 23, 1989. NHY proposed to place the plant in a reactor critical, standby mode in anticipation of correcting the physical impediments prior to resuming low power testing. 19 w-_ _.

'During the discussion, NHY described the basis for the operator's I actions'taken during the event. It was explained that the USS was in i command and the operators felt that they had control over the cooldown -{ ) sequence. The operators recognized that they were below the test ] procedure pressurizer low level trip limit, but knew that they were within the parameters and criteria listed in the Technical Specifications and Operations procedutas. Notwithstanding their procedure vio'lation error, the operators had taken otherwise appropriate actions to restore the pressurizer level and to re-establish the test parameters. Onc6 they became aware of the cause, the operators took corrective action to close the condenser steam dump valve and thereby tenninate the unanticipated excess cooldown. Four minutes and forty-eight seconds (4:48) after isolating the condenser steam dump valve the USS ordered the reactor shut down when he determined that the plant might possibly exceed pre-specifi. parameters before normal operating system lineups could be fully. restored in a conservative, controlled fashion. With regard to test procedure adherence, the NHY position was that although tt.c JSS and board operators had taken actions to operate the plant in a safe and technically sound manner, they had inappropriately i violated the specific manual reactor trip criteria listed in the test procedure. NHY's position was that the USS should have immediately directed the reactor shut down when the pressurizer level reached the 172 test limit. 20 =aL

p l p .NHY and the NRC agreed that a follow-up conference call would occur at 0730 on Friday morning and that NRC concurrence would be achieved prior 'to restarting the reactor. 1 At approximately 1830, NHY Production and Station Management continued their discussion of procedural adherence begun at their previous 1530 meeting. This meeting continued until approximately 2100 and adjourned with the following conclusions and' corrective actions: The 1-ST-22 Manual Trip Criteria, indicated on Attachment 9.3 to the procedure, was appropriate and should remain in the procedure. The USS should have ordered the shut down of the reactor. NHY needed stronger policy and guidance regarding procedural adherence. At apprezimately 2315, after having arrived home from Washington, DC, the NEY President and CEO phoned the NHY VP-NP for a status report. The VP-NP provided a brief summary of the 1800 conference call with the NRC Region I (Wiggins) making the following points: The NRC 'seemed to be pretty well satisfied" with the NHY explanation of the details of the event. A couple of hardware corrective actions and some relatively minor procedure changes needed to be taken care of, but once these were concurred with by the NOC, the plant should be ready for restart. 21

3 NHY had committed to review with the NRC. its schedule of --l I completion of hardware corrective actions such as the corrective j maintenance being performed on the steam dump valve that had malfunctioned during the test. ~ l t.nother conference call with NRC Region I had been scheduled for 3 0730 Friday morning to review these items. The VP-NP then spent a few minutes focusing on the procedure compliance issue and reviewed the NHY management group's proposed position that had been developed following the 1800 conference call with the NRC. The President and CEO indicated that he would participate in the 0730 conference call on Friday. NOTE: The President und CEO was not made aware, either during this . briefing or during the next morning's 0645 pre-NRC telecon briefing, of a number of key facts associated with the 1800 NHY/NRC telecons i.e. the defense of operator actions taken or not taken, the suggestion that the operator actions were more conservative than strictly adhering to a test procedure, the possibility thst the NHY procedere compliance policy was essentially adequate as written and the proposal that reactor restart be allowed to occur in parallel with NRC/NHY event evaluation conclusions and corrective action determinations. 22 L_,.._. m_._.

i v LAt 0730 on June 23.1989' NHY.NRC' Region I and on-site NRC; inspectors / participated'in a. telephone. conference call that was a continuation of the discussion:fr 'he previous evening. The topics discussed-I included: The 1-ST-22 procedure, including-the manual trip criteria, was appropriate and that revisions were not required. NHY had adequate existing procedure adherence guidance that . applied-to static conditions but that this guidance was not fully satisfactory'for unanticipated transient or emergency conditions. The NHY_ policy recognized that strict, verbatim compliance to procedures would not always be appropriate during such transient or emergency. conditions. Otherwise. NHY personnel are expected to. follow procedures. NHY recogniand the netd to revl the procedure adherence policy to accommodate an unantac:y ated transient cr amargency - situation. NNY proposed that the procedure adherence policy would be revised Land issued through SORC within two weeks. The Station Manager would brief each operating crew prior to their assuming their'next assigned shift. NNY would conduct ongoing procedure adherence training. NHY related the status of the prerequisite actions considered to be necessary prior to a reactor restart and estimated that they t could be resolved prior to 1030 that day. 23 L L.

Th3 NRC indicated that they had no additional questions but expressed the need to further brief their management. The NRC t# committed-to respond to NHY by 1000 and requested that NHY not restart until after the return telephone call. The 1000 NRC return phone call did not occur, and at approximately 1030 the Vice President - Nuclear Production phoned the NRC Region I to indicate that the restart readiness target had been slipped to 0700 on Saturday, June 24, 1989. Reference to the attached chronology also indicates that numerous additional internal (NHY - NHY) and external (NHY - Others) communications occurred between individuals and at impromptu meetings. 3. Conclusions Analysis and comparison of the existing NHY Post Trip Review Process with the actions taken in response to the events of June 22 provides the basis for the following conclusions: (1) NHY should have allowed the existing, conservative and deliberate approach to. analyzing the e<ents of June 22, 1989 to be completed before discussing restart schedules with the NRC. The following actions should have been completed: (a) Completion of a thorough and detailed analysis of the physical events that had initiated the unanticipated 24 w=-__---____

f transient as well as the operator actions taken in responding to the transient. l: j '(b) Completion of corrective maintenance and post maintenance testing for the specific component that failed'and a ~ determination that similar components did not exhibit the same condition. (c) Completion of the determination and dissemination of revised corporate policy on procedure adherence. (d) Completion of Operations and Startup personnel specific training on the revised policy. (e) Completion of the Event Evaluation Team Report. l (f) Completion of a NHY Executive Management meeting with the NRC to brief them on the results of the completed analysis and the status of in-progress corrective actions. (2) NHY should have more fully utilized the Event Evaluation Team leader as the communications focal point within NHY. This is currently described as a primary function for this position. (3) NHY should have designated single points of contact to provide information to individuals or groups, both internal an? external to NHY, for this events i.e. NHY to NRC, NEY to State. NHY Production to the rest of NNY. 25 I _m_.

-. _ ~ l: . Communications with the-.NRC should have more accurately ccnveyed ( (4) that the total NHY post trip revitv process was still'in progress. that-the event conclusions were still being determined, that restart decisions were still being made and that even though plant equipment corrective actions were being taken in order to establish physical plant readiness to restart..no restart determination would be made until all NHY and NRC concerns had .been fully addressed. (5) The NRC required (10 CFR 50.72) report should have been more accurate. The NHY four-hour reportable event communication to the NRC (Bechesda) on Thursday at 1320 was inadvertently incomplete, inaccurate and may have been a source of misconstunication regarding the actual event. (6) The event evaluation review prscess should have more adequately addressed the procedure.or personnel contributions to events. (7) NHY should not have conveyed an inappropriate focus on resuming the Low Power Test Program and should not have engaged in inappropriate justifications of the operator actions taken in responding to the event. (8) NHY personnel contacted by the NRC during the performance of 1-ST-22 and informed of the pressurizer low level trip criteria should have more appropriately conveyed this information. in a timely and effective manner, to NHY management. 26

s, ~ ' Post Trio Review Policy Recommendations 4 lr -(l) NHY should' reinforce its policy of using a deliberate, cautious and conservative approach when responding to future unplanned reactor shutdowns or ESF actuations. NWY should emphasize the completion of the full post trip review process including a discussion of the results with the NRC prior to recommending the rectart of the reactor or the resumption of testing. During power ascension testing the Event Evaluation Report should be completed prior to recommending the restart of the reactor. (2) NHY Operations should review the 10 CFR 50.72 reportable event notification process to determine the need to use a preparer and reviewer methodology prior to actual transmittal of the report. NHY Operations should also consider confirming the reportable event communication by verification on the same day or by a next day comparison of the INPO Nuclear Network data base. (3) NHY.should review the effectiveness of the current internal and external post event communications processes and the assigned responsibilities to ensure that management expectations will be met in terms of timely, focuse'd, accurate cosununications appropriate to the significance of the event. s.. NNY should re-evaluate the overall post trip review process for .(4) unplanned reactor shutdowns and the assignment of personnel by porition (i.e. Operations Manager or Assistant Operations Manager) l to the Event Evaluation Team. These key management positions a, 27

retain-line management responsibilities irrespective of the current post' trip review processes. The direct participation, by these individuals, in the event evaluation process will assure that the individual (s), with the decision making responsibilities leading up to restart of the unit, will be fully aware of the post' trip review status and conclusions. (5) NHY should further review the event evaluation process and should evaluate the adequacy of its present criteria that concern procedure and personnel contributions to events. NHY should: (a) consider incorporating the post trip review critique meeting (held in the TSC) as a standard practice, (b) require each key participant in the event to prepare a written, chronological report as part of their post trip report debriefing, and (c) upgrade the event evaluation process with respect to human factors and performance analysis and corrective action determinations. (6) NHY should aggressively pursue resolution of these recommendations and complete implementation prior to Power Ascension Testing. i 28

D. Steam Dumo Valve Incomplete Work Reouest 1. Backaround: The steam _ dump valve MS-PV-3011. that was the cause of the unplanned plant cooldown, had an open work request pending completion of a normal. operating pressure, norum1 operating temperature dynamic flow and stroke time retest. This retest was a NHY identified and conservative option selected to further determine valve operability and functionality following some corrective maintenance performed subsequent to the second Seabrook Station Hot Functional during February of 1987. Although it was being tracked by the Station Work Control Planning and Scheduling process, this retest was not conducted prior to the Natural Circulation Test. Several requests had been made by the retest implementation department (Instrumentation and Controls) but plant conditions did not allow retest at those times. It can not be determined at this time (post event) whether the specified retest would have.successfully identified the MS-PV-3011 impending failure or not. The retest called for a one-time, rapid stroke of the valve to verify freedom of movement over one full cycle i from full shut to full open to full shut and to collect valve stroke timing data. Because the valve appeared to operate correctly during the preparations for and the onset of the Natural Circulation Test. it most probably would have successfully pasPed the retest without the subsequent valve binding and positioner feedback linkage problems becoming evident. i 29

2. -Conclusions: 1. .Seabrook Station Technical Specifications do not require that this valve be OPERABLE during MODE 2 operation of the plant. 2. By-the Seabrook Station Maintenance Manual operability test criteria and the specified retest criteria determined to be appropriate subsequent to the corrective maintenance performed on-this valve, MS-PV-3011 had not been determined to be OPERABLE nor had the operability retest been waived through the applicable'NHY justification / documentation procedures. 3. The Startup Test Program Description requires that temporary modifications and tag outs be evaluated but does not require that open retests be identified or addressed prior to the conduct of a Startup Test. 1-ST-22 did require, as a prerequisite, that the Steam Dump System be available for use. System availability was specifically considered and was determined to be satisfied by virtue of the system use during the Emergency Feedwater System testing completed on June 12, 1989 and use during the June 22, 1989 pre Natural Circulation Test preparatory activities just prior to 1-ST-22 performance. 30

3. Recommendations: 1. Resolve the Steam Dump Valve positioner feedback linkage and valee binding problems on the twelve steam dump valves and-other Seabrook Station valves that are of similar design and provided by ' the manufacturer of these valves. 2. Revise the Startup Test Program' Description to require consideration of any outstanding ratests associated with equipment to be used in any specific Startup test. Outstanding ratests so identified should either be performed or waived through an acceptable, documented process. 31 m_.____.____-

-,,.----,----,----r--w-,--,-- 4 C EXHIBIT F

[ 88k%8AA RESULATORY Ccesh88440N 6 ST&fte g nomewa 4Fe Au,ge9AL4 ROAS mha or mussiA. psNwevi,vAwu stees Ocekets No.: 50-443 M CAL No.: 49-1) Pubite servite of New waeschire Mr. Edward A. Br:wn, :aes(tcentPSNH) AT7N: and Cnief Executive Cfficer New Hampshtce Yankee D1yisten Post Office Box 300 5esertok, New Hampshire 03874 Gentlemen: $vbject: CONFIRMATORY ACTION LETTER (CAL) 89-11 Tbts letter confirms our understanetng of these actions you intend to take in response to the reactor manual trip which occurred on June performance of the natural circuistion staetup test. 22, 1949 during the 23, 1949 These actions were discussed during a June Thomas T. Martin, Deputy Regional Administrator, NRC Region 1. phone Specifically, we understand taat, prior to startup of the unit, P$NH will: (1) Complete and document the results of the post-trip review process

ssociated with the June 22, 1989 event; (2)

Establish those short-ters corrective actions to be completed prior to resta-t of the un't to address the specific deftctancies identified during your post-trip rev'ew; (J) Detenntne those longer ters corrective acttens and their respective schedules, to address any potestf ally broader impittations associated with the spectftc deficiencies identified as a result of your review; and, (4) Review the results of ttees (1), (2) and (3), above, with the NRC staff. We further understand that the agreement of the Regtenal Administrator, Region 1, would be ettatred prior to restart of the unit. g j g #N N

Public Service of New Hampshire 1 I* If your understanding. differs from that **t forth abcve immediately. , please call me $1nceral. William I. Russell llegional Administrator cc: J. C. Dwffett. President. and Chief fuscutive Officer T. C. Feigeebaue, Vice President Engineer ag, Ltcens PSNH J. M. Peschel, Regulatory Services Manager Station Manager, NHY , NNY ing & Quality Program, NHY D. E. Moody P. W. Agnes,, Jr., Assistant Secretary of Public Safety Commonwe41th of Massachusetts Public Document Acom (PCR Local Pubite Deccant Room) (LPOR) NRC Resident inspectr,rNuclear Safety Infomation Center (h3[C) State of New Hampshfre Commonwealth of Ma sachusetts $oebrook Hearing $4rvice List 9 9

7..s._.---.---.,.. i;' g e EXHIBIT G J )h '. ' " -i 'v

U.S. NUCLEAR REGULATORY COMNISSION f REGION ! j Report No. 50 443/88-09 Docket No. 50-443 License No. CPPR-135 Priority Category C Licensee: Public Service Company of New Hampshire P. c. sox 330 Mancnester, New Hampshire 03105 Facility Name: Seabrook NL. clear Power Station Inspection At: Seabrook, New Hampshire Inspection Conducted: Juns 27-27,1988 Inspeetors: U b!tfDRSS C. Amato EPS C. Gordon, EPS S. Paleschat EPS D. Ruscitto, RI, Seabrook D. Perrotti, NRR J. Jamison, PNL 1 Approved By: e_ b

• ) 4 /k

u. J.,. Lazarus, chter, LPs, date FRSS8, DRSS Inspection Summary:

Inspection on June 27-29. 1988 (Report No. 50-443/88-09) Areas Inspected: Routine, announced emergency preparedness inspection and observation or the licensee's annual full-participation amargency exercise performed on June 28-29, 1988. The inspection was performed by a team of 1 { seven latl: Region I, headquarters and contractor personnel. ( Results: No violations were identified. Emergency response actions were I Requate to provide protective measures for the health 74 safety of the public. i 4 Q h{ e

b r n' OETAILS 1.0 Persons Contacted June 29, 1985..The following licensee representatives attended the exit m E. Brown. President and Chief Executive Officer D. Bovino, Exercise Coordinator P. Casey, Senior Emergency Planner T. Feigenbaum Vice President Engineering / Quality -G. Gram, Execu,tive Director, Emergency Preparedness and Commu Affairs T. Harpster, Director, Emergency Preparedness Licensing j D. Moody, Station Manager P. Stroup d G. Thomas. Director, Emergency Implementation and Response I Vice President Nuclear Production J.MacDonald,RadiologicalAssessmentManager The team observed and interviewed several licensee emergency response personnel, controllers and observers as they performed their assigned functions during the exercise. 2.0 Emeroency Exercise -The Seabrook Nuclear Power Station full-participation exercise was conducted on June 28, 1988 from 9:00 AM to 7:00 PM. The State of New Hampshire 11 local towns and the State of Maine participated. The 1 Commonwealth of Massachusetts and 6 local towns in New i participate. The State of New Hampshire compensated for tie local l non-participants. The New Hampshire Yankee Offsite Response Organization (191Y 000) compensated for the Commonwealth non-participants. The licensee, New Hampshire, Maine ar.d NHY OR0 conducted field monitoring activities, an in Management Agency (y activities on June 29, gestion pathw recovery and reentr 1988. The Federal Emergency FEMA) observed all off-site activities. 2.1" Pre-exercise Activities Prior to the emergency exercise NRC Region I and FEMA . representatives held meetings an,d had telephone discussions with licensee representatives to discuss objectives, scope and content of the exercise scenario. As a result, einer changes were made in order to clarify certain objectives revise certain portions of the scenarie and ensure that the scenarlo provided the opportunity for the licensee to demonstrate the stated ob octives as well as those l, areas previously identified by NRC and FEin as in need of corrective action. _ _ _ _,,, _,, _, _ _ _ _ - _ - - - - - - - - - - - - - - - - - - - - " " - - ' - " - - - - - ^

p L t I 3 NRC observtrs attended a licensee briefing on June 27, 1988, and participated in the discussion of emergency response actions expected during the various phases of the scenario. The licensee stated that controllers would intercede in exercise activities to prevent scenario deviation or disruption of normal plant operations. The exercise scenario included the following events: - Fuel damaged by loose parts; - Damage to a turbine driven eenrgency feedwater pump; - Large break Loss of Coolant Accident (LOCA) due to a total weld failura; - Venting of the containment into the containment enclosure building with a subsequent elevated, filtered release to the atmosphere; - Declaration of Alert, Site Area Emergency and General Emergency Classifications; - Calculation of offsite dose consequences; and - Recommendation of protective actions to off-site officials. 2.2 activities Observed During the conduct of the licensee's exercise, seven NRC team members made detailed observations of the activation and augment-ation of the emergenc response facilities, y organization, activation of emergency and actions of emergency response personnel during the operation of the emergency response facilities. The following activities were observed: 1. Detection, classification, and assassment of scenario events; , 2. Direction and coordination of the emergency response; 3. Augmentation of the emergency organization and response facility activation; 4. Notification of licensee personnel and offsite aBencias of pertinent plant status information: l ~ 1 5. Communications / information flow, and record keeping; 1

4 6. Assessment and projection of offsite radiological dose and consideration of protective actions; 7. Provisions for inplant radiation protection; 8. Performance of offsite and inplant radiological surveys; 9. Maintenance of site security and access control; 10. Performance of technical support, repair and corrective actions; 11. Assembly, accountability and evacuation of personnel; 12. News Center; andPreparation of information for dissemination 13. Management of recovery and reentry operations. 3.0 Exercise Observations 3.1 Exercise Strenoths The NRC team noted that the licansee's activation and augmentation of the emergency organization, activation of the emergency response facilities, and use of the facilities were generally consistent with their emergency response plan and implementing procedures. The team also noted the following actions that provided strong positive indication of their ability to cope with abnormal plant conditions: 1. Very good command and control of all emergency response facilities (ERF's)wasdemonstrated; 2. Plant conditions were quickly recognized and classified; 3. shift turnover was accomplished smoothly and with no apparent loss of control of the situation; 4. The ERF's were activated in a timely manner; and 5. ProtectiveActionRecommendations(PAR's)werepromptand conservative. utilized in determining the PAR's. Evacuation time esttmates we

I 1 5 3.1 Exercise Weaknesses The NRC identified the following exercise w I be evaluated and corrected by the Itcensee.eaknesses which needs to an adequate The licensee conducted these areas.self critique of the exercise that also identified 1. The Technical Support Center (TSC) and Emergency Operations! Facility (E staff displayed questionable engineering judgement a i or did not recognize or address technical concerns (50-443/88-08 01). 1 For example: - Neither the EOF or TSC staff questioned a release of greater than 7000 curies per second with only clad damage and no core uncovery; - Efforts continued to restore the Emergency Feedwater Pump after a large break LOCA; - A questionable fix for the Containment Building Spray system; - A lack of effort to locate and isolate the release path; and - No effort was noted to blowdown Steam Generators to les the heat load in contalissent. 2. The TL and Operational Support Center entrances and exits that are not control (OSC) have multiple led. As a result contamination controls were ineffective at times as person,nel entered without frisking and it couldn't be determined if continuous accountability was, or could be, maintained (50 443/88-09-02). 3. No apparent consideration was given to the departing first shift to account for possible Bose when leaving the plant during the release, as they were not given dostmetry (50-443/88-09-03). 4. The response to some q9estions in the Media Center were not adequate such as: the NRC's role in an eneroency; and wh reactor trip wasn't performed earlier (50-443/88 09 04). y a 4.0 Licenses Actions on Previously Identified items The followi items were identified during a previous inspection (Inspection rt No. 50-443 the NRC taani ring the exerc/87-15. Based upon observations made by ise t following opens itsa were acceptably demonstrated and are closed:

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L 6 (CLOSED 3 47-25 01 IFl: classification procedures and failed to recognize the los Radiation Monitoring Systems trains as an Unusual Event. (CLOSED) 87-25-02 IFI: prevented dose assessment personnel from estimating t atmosphere todine concentration. -5.0 Licensee Critious The NRC team attended the licensee's post-axercise critious on June 1988, during which the key licensee controllers discussac observations of the exercise. evaluated and appropriate corrective actions taken.The Itcense 6.0 Exit Meetino and NRC Critioue of this report at the end of the inspection.The NRC team me the observations made during the exercise. The team leader summarized The licensee was informed that previously identified items were adequately addressed and no violations were observed. Although there were areas identified for corrective action, the NRC team detemined that within the scope and limitations of the scenario, the licensee's performance demonstrated that they could implement their Emergency Plan and Emergency Plan laplementing Procedures in a manner which would pubite.adeountely provide protective measures for the health and safety of the Licensee management acknowledged the findfees and indicated that a priate action would be taken regarding the identified open items. ppro-At no time during this inspection did the inspectors provide any written information to the licensee. l o

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y ( .3 .a *rcJ .i ~ 'c, , UNITED STATES ' l [n ' [$ NUCLEAR REGULATORY COMMIS$10N

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475 ALLENDALE ROAD KING OF PRUS$1A. PE.NNSYLVANIA 19406 Docket,No. '50-443-IIE E

License'No. NPF-67 EA No. 29-153 t Public. Service; Company of New Hampshire ~ ATTN: 'Mr. Edward A. Brown, President b and Chief Executive Officer m . New Hampshire Yankee Division ' Post Office Box 300 Seabrook, New Hampshire 03874 Gentlemen:

Subject:

NRC Region I Augmented Inspection Team ( AIT) Inspection (50-443/89-82) - of the Natural Circulation Test at Seabrook Station, Unit No.1 This letter refers te the June 28-30, 1989 AIT review of the June 22, 1989 natu-- .ral circulation test at-Seabrook Station, Unit No. 1. The AIT inspection, led by P. W. 'Eselgroth of this effice, was a f act findi.ng and causal factor deter-mination effort. At. the_ conclusion of the inspection, an exit interview was held with you and members of your staff to discuss the inspection findings. The AIT, report is attached 'as Enclosure L Confirmatory. Action' Letter (CAL) 89-11 stated yor agreement to review correc-tive actions and post-trip review results with the NRC staff. and to obtain the agreement' of 'the Regional Administrator prior to restart of the unit. You should be prepared te discuss the findings and conclusions of this inspection report -and your response to CAL 89-11 at a public meeting planned for Septem-ber 6,1989 at the New England Center at the University of New Hampshire in Durham, New Hampshire, Following this meeting, elected officials'and inter- ~ ested members of the public will be offered an coportunity to provide comments to-the NRC staff on the results of the AIT inspection and on the adequacy of .your corrective actions. If, af ter review of the enclosed report, you identify additional corrective actions not discussed in your July 12, 1989 response to CAL 89-11, please provide those in writing no later than August 25, 1989. The expression of concern in Section 5.3.3 of the report for the failure of certain managers observing the natural circulation test to assure adherence to test procedure requirements should not be viewed as establishing new expecta-tions of performance. NRC encourages licensee managers to tour their facilities and observe significant activities and to be alert to conditions that could adversely affect safety. In general managers, not on watch, should not direct activities of licensed operators but rather should make their concerns known to shi't'supervisio-However, we remain concerned with the lack of action by r.snagers in the to,ttrol room on June 22, 1989, during the five minutes a test criterion was exceeded, particularly since this condition -was identified to licensee management by an NRC representative.

Public Service of New Hampshire 2 JgM} 17 ggg The failure to trip the reactor when required and the failure to promptly re-view and resolve any associated personnel performance implications associated with the failure to trip are potential violations of NRC requirements. To dis-cuss these and other matters identified in Enclosure 2 to this letter, we have scheduled an enforcement conference in the NRC Region I office at 1:00 p.m. on September 7, 1989. At that conference, for each item, please be prepared to present your assessment of safety significance, root cause(s), and your interim and final corrective actions. You will be informed in writing of the NRC deci-sion on enforcement action when that decision is reached after the conference. In accordance with 10 CFR 2 Appendix C, the enforcement conference will not be open for public observation. Your cooperation with us is appreciated. Sincerely, ORU31NAL StGNED BY William T. Russell Regional Administrator

Enclosures:

1. NRC Region I Augmented Inspection Team Report No. 50-443/89-82 2. Enforcement Conference Issues and Related Regulatory Requirements cc w/encls: J. C. Duffett, President and Chief Executive Officer, PSNH T. C. Feigenbaum, Senior Vice President and Chief Operating Officer, NHY J. M. Peschel, Operational Programs Manager, NHY D. E. Moody, Station Manager, NHY P. W. Agnes, Jr., Assistant Secretary of Public Safety, Commonwealth of Massachusetts Public Document Room (PDR) local Public Document Room (LPDR) Nuclear Safety Information Center (NSIC) NRC Resident Inspector State of New Hampshire Commonwealth of Massachusetts Seabrook Hearing 3ervice List l i_ . __ _____.___ ____-____________ A

U.S. NUCLEAR REGULATORY COMMISSION REGION I Report No. 50-443/S9-82 Docket No. 50-443 License No, NPF-67 Priority Category C Licensee: Public Service of New Hampshire New Hamoshire Yankee Division Post Office. Box 300 Seabrook, New Hampshire 03874 Facility Name: Seabrook Station Unit No.____1 Inspection At: Seabrook, New Hampshire Inspection Conductea: June 28-30, 1989 s/ Inspectors: ])/ ,7/2 5/89 P W. Eselgydth, Team Leader, RI .date (See attached sheet) N. F, Dudley, Sr. Resident Inspector, R1 date (See attached sheet) L. Loi s, Team Member, NRR date (See attached sheet) J. M. Trapp, Team Member, RI date (See attached sheet) F. Guenther, Team Member, NRR date Approved: _llt.t/u m ~~)!ol3 ,WiggigTeamManager,RI date Inspection Summary: See Executive Summary

4 ~ ~~~ ~ - ~ El +&_ p. 3, M.,, .q.; - -a y ,-;,p, e 3; m t j,, ao b,* U.S. NUCLEAR. REGULATORY COMMISSION REGION-I e x 50-443/89-82 Report No'. y

Dodket No.

50-443' L. ' License.No. NPF-67 Priority Category C' Licensee.: Public Service of New Hampshire New Hampshire Yankee Division-Post.0ffice Box 300 Seabrook, New Hampshire 03874 Facility.Name:,Seabrook Station Unit No. 1 j Inspection At:.Seabrook, New H'ames' hire . Inspection Conducted: June 28-30, 1989 Inspectors: .P. W. Eselgroth, Team Leader, RI date O.c u ?.) A 'is, 7 -H - ?'l N. F..Ducley, Seniov Resident, RI -date L. Lois. Team Member, NRR date hb 7-: r - 69 J.p.Trapp,idamMember,-RI date F.. Guenther, Team Member, NRR date - Inspection Summary: See Executive Summary i _____._--.E

U.S. NUCLEAR-REGULATORY COMMISSION REGION I Report No. E0-443/89-82 Docket No. 50-423 License No. NPF-67 Priority Category C Licensee: Public Service of New Hampshire New Hampshire Yankee Division Post Office Box 300 Seabrook, New Hampshire 03874 Facility Name: Seabrook Station Unit No. 1 Ir.spection At: Seabrook, New Hampshire Inspection Conducted: June 28-30, 1989 Inspectors: P. W. Eselgroth, Team Leader, RI date N. F. Ducley, Senior Resident, RI cate La.~ 6 r Y[ 'l /7f fI L. Lois, Team Member, NRR date 1 J. M. Trapp,. Team Member, RI cate F. Guenther, Team Member, NRR date Inspection Summary: See Executive Summary 1 s ~ ~ ~~

,f~~. <o L I i p 1 L I' 1 1 TABLE OF CONTENTS Page 1.0 Introduction 5. 1.1 Scope of Inspection 1.2 ' Team Composition 2.0 Executive Summary. 5 2.1 Event Summary 2.2 Assessment Summary 3.0 Event Description 9 4.0 Plant and Equipment Performance. 10 4.1 Introduction. 10 4.2 Plant Response.. 10 4.2.1 RCP Trip to Steam Dump' Valve liS-PV-3011 Failure Open 4.2.2 5 team Dump Valve MS-PV-3011 Failure Open to Closure of All Steam Dump Valves 4.2.3 Steam Dump Valve Closure to Reactor Trip 4.2.4 Summary of Plant Equipment Response 4.3 Steam Dump Valves............. 13 4.3.1 Introduction 4.3.2 MS-PV-3011 Failure to Modulate 4.3.3 Steam Dump Valve History 4.3.4 Valve Failure Cause 4.3.5 Licensee Short-Term Response 4.3.6 Licensee Long-Term Response 15 5.0 Personnel Activities and Performance 5.1 Operating Crew. 15

TABLE OF CONTENTS Page 5.1.1 Organization and Responsibilities 5.1.2. Training 5.1.3 Pre-Test Briefing 5.1.4 Creve Response 5.1.5. Performance Assessment 5.2. Startup Test Group. 22 5.2.1 Organization and Responsibilities 5.2.2 Test Procedures 5.2.3 Training 5.2.4 Pre-Test Briefing 5.2.5 Test Group Response 5.2.6 Performance Assessment 5.3 Management and Support Staff......... 26 5.3.1 Management and Other Support Personnel 5.3.2 Management Responsibilities 5.3.3 Management Response 5.3.4 Performance Assessment 29 6.0 Safety Assessment. 6.1 Reactor Safety Significance of Event 6.2 Safety Significance of Personnel Performance 7.0~ Exit Interview. 30

I TABLE OF CONTENTS APPENDICES Appendix A: Chronology of Events Appendix B: Chronology of Communications Appendix C: Individuals Interviewed Appendix D: Entrance Interview Attendees Appendix E: Exit Interview Attendees Appendix F: NRC.0 observations Regarding Seabrook Natural Circulation Test Appendix G: Augmented Inspection Team (AIT) Charter Appendix H: Plant and Equipment Performance Figures Appendix I: Acronyms and Initialisms l

i l 1.0 Introduction 1.1 Scope of Inspection In response to the performance of a natural circulation test at the Seabrook Station Unit No.1 in a manner contrary to the test procedure reactor tripping criteria on June 22, 1989, the NRC formed an Augmented Inspection Team (AIT) to determine the event sequence, causes and safety significance. This was accomplished by establishing a chronology of the event (Appendix A) and accompanying communications (Appendix B), and reviewing equipment performance, plant staff actions relative to this occurrence and applicable station procedures. The NRC Team held an entrance interview with plant management and support personnel on June 28, 1989 and performed the inspection during the period of June 28-30, 1989. An exit interview was conducted with plant management on June 30, 1989.. Individuals interviewed during the course of the inspection are listed in Appendix C. Attendees at the entrance and exit interviews are listed in Appendixc.s D and E. Appendix F contains the statement of NRC observers present during the June 22nd test. Appendix G is the memorandum of assignment of the AIT to this Seabrock Unit 1 event. Appendix H contains plant and equipment performance figures. 1.2 Team Composition The team was ctrposed of a team leader and four headquarters and regional specialists with expertise in plant operations, reactor core and plant systems, operator training, test programs and management controls. 2.0 Executive Summary 2.1 Event Summary On June 22, 1989, the plant conducted the natural circulation test of the primary system which is part of the reactor testing program. This test gathers primary system data under controlled conditions to demonstrate the ability of the reactor coolant system to remove decay heat usiffg natural circulation. At tne initiation of the test, the reactor was operating at about 2% power and heat was being removed from the plant by the steam dump l. valves. Shortly after the point in the test where the reactor coolant pumps were turned off to establish natural circulation of the primary system coolant, one of the three controlling group steam dump valves (MS-PV-3011) malfunctioned and went to the fully open position, resulting in a rate of heat removal from the primary system beyond what was planned for the test. At this time none of the predefined criteria in the natural circulation test procedure

l (1-ST-22) for termination of the test were exceeded nor was the MS-PV-3011 position problem recognized. The presence of this equipment proolem and the accompanying rate of heat removal resulted in a primary coolant average temperature transient that resulted in the coolant level in the pressurizer decreasing towards one of the I test's reactor trip criteria at the pressurizer 1794 level. l Prior to pressurizer level reaching the 17?; point, the Unit Shift Supervisor (USS), a Senior Reactor Operator, informed the Test Director (TD) that one of "your limits" is being approached. When pressurizer level reached 17?e (at which letdown is automatically isolated and pressurizer heaters are deenergized) the Senior Control i' Room Operator (SCRO), who is the primary side reactor operator, informed the USS of this, but did not mention the associated reactor l trip requirement. At this point the USS conferred with the TD; steam dump valve MS-PV-3011 had been shut; the pressurizer level decrease had been halted; and, pressurizer level had begun to increase. It was on the basis of the pressurizer level recovery taking place that the USS decided to allow the reactor to continue to operate in support of the test. However, the USS did not correlate the isolation of letdown indication with the loss of pressure control and the need to trip the reactor in accordance with the pressurizer 17'4 level criterion. An increasi.ig reactor coolant pressure transient was now developing dua to the closure of the malfunctioning steam dump valve (MS-FV-3011) with the subsequent recovery of pressurizer level, and the US$ directed tFat the reactor be tripped due to primary plant pressure gproaching the test procedure trip criterion. The shift crew then carried out the emergency operating procedures for a reactor trip and the natural circulation test (1-ST-22) was terminated. 2.2 Assessment Summary The conclusion of the AIT staff regarding the licensee's response to the plant transient resulting from the malfunction of one of the steam du;np valves is that reactor plant safety was never in question, and with the exception of the significant error of not tripping the reactor at the point first called for by the test procedure and loss of pressure control due to letdown isolation and pressurizer heater deenergization, the operating staff performed well. The following summary of assessments is provided with references to the sections of the report where further details are documented: o The actual plant dynamic response was reviewed and compared to the post trip review predicted response. The plant responded as predicted in the June 22nd natural circulation testing including the very mild overcooling event which resulted from steam dump valve MS-PV-3011's failure to properly modulate. (Section 4.1)

p Plant equipment was not ready to support the June 22nd test.

o Prior to commencing the test, a test prerequisite to confirm the availabil'ity of the steam dump system was signed off. However, there was an open work order for post maintenance stroke testing of steam dump valve MS-PV-3011. (Section 4.3.3) The interviews of the' Unit Shift Supervisor (USS), Senior o Control Room'0perator (SCRO) and Control Room Operators (CRO) found them to be highly competent' individuals, clearly aware of their assignments for safe operation of the plant. In particular, the USS communicated that he had no doubts about being the one responsible for conduct of the test in a safe and controlled manner. (Section 5.1.1). Training relative to the conduct of the natural circulation test .o which covered details of the expected plant response had been accompi.ished about a year prior to the test. The AIT found no evidence that such training had oeen repeated or refresher training given since that time. (Section 5.1.2) o A review of the pre-test briefing that was conduc';ed for the operators by the Test Director determined that it was inadequate with respect to covering the details of the testing to be performed and thoroughly reviewing the reactor trip criteria. (Section 5.1.3) The operating crew was observed to be conducting. plant o operations in a' controlled, unfrenzied manner prior to, during the test and following the reactor trip when the' applicable ' emergency operating procedures were entered and carried out appropriately. (Section 5.1.4) 9 During the Low Power Testing program prior to the June 22nd event, as well as during this event, there was no evidence of pressure applied by management or anyone else to complete testing at the expense of controlled, safe operation of the plant. In fact, the NRC has been aware of personnel assigned to shift operating and test responsibilities having received direction from management to proceed with testing in a controlled manner and specifically to not permit themselves to feel rushed into completing evolutions. (Section 5.1.4) The USS did not trip the reactor at the 17% pressurizer level, as o called for in the test procedure (1-ST-22). He stated his reason was that the decreasing pressurizer level was under control and turning around. The AIT concluded that a cause of this event was the lack of importance and/or sense of ownership placed on test procedure requirements by the USS as compared to his other operating 1 )

L e _g. requirements such as those contained in Technical Specifications r and plant operating procedures. Two other operators interviewed also indicated the perception of a hierarchy of importance for procedural requirements between test procedures and plant cperating procedures. These misunderstandings on the part of the operators demonstrated an absence of recognition of test procedure criteria as controlling requirements for operation under testing conditions. (Section 5.1.4) o The Shift Superintendent (SS) did not provide effective supervisory involvement in the conduct of this test. (Section 5.1.5) From the interviews of operating crew personnel it has been o concluded that these personnel now recognize and understand that the proper action was to have tripped the reactor before the 1-ST-22 trip criterion on pressurizer level was exceeded. (Section 5.1.5) The startup test group had responsibility to interrupt or o terminate the test in the event that required plant conditions were not maintained. However, no such recommendation was made to the shift operating crew by the test group even though the Startup Manager was made aware of the NRC's concern about the plant being below a manual trip criterion. The overall direction given by the test organization during the performance of this test was inadequate. (Section 5.2.5) From the interviews of startup group personnel it has been o concluded that these personnel now recognize and understand that the proper action was to have terminated the test and recommend to the operating crew that the reactor be tripped before the 1-ST-22 trip criterion on pressurizer level was exceeded. (Section 5.2.5) During the ccnduct of 1-ST-22 and at the time when plant o conditions had reached the reactor trip criterion associated with pressurizer level, there were several plant management representatives in the control room with the responsibility and authority to terminate test and plant operations when approved procedures are not being followed. When members of management having specific responsibility and authority relative to safe operation of the plant are present in the control room, their presence in no way dilutes the responsibilities of the operating crew and test group personnel assigned to shift. However, by virtue of the particular responsibilities and authorities that they do possess relative to safe plant operations, there is a responsibility particularly during unique testing situations - to keep themselves informed of key limits for operation and plant status relative to those limits and to take appropriate action relative to plant operation whenever others they have

u -g-. . assigned to.do this have not done so. This was not done by the' management-members present. (Section 5.3.3) The initial' management. thrust following this event appeared to o be to resolve any equipment problems necessary to resume testing. An in-depth review of the cause er causes leading to the improper conduct of the 1-ST-22 natural circulation test _ apparently did not take place prior to an initial management i decision to resume testing. A thorough review of this. event L was not completed by the licensee until after the NRC raised. l this issue with licensee management. (Section 5.3.3) . 3.0; Event Description The following_ description of the event was determined through observations, interviews with the operators and review of the plant computer traces and printouts. A chronology of the event is presented in ' Attachment A. On' June 22, 1989, the plant was at about 2% rated power in. preparation for the performance of tne natural circulation test, which was intended to demonstrate the ability of the reactor coolant system to remove decay heat e using natural circulation At approximately 12:19 p.m., the reactor coolant pumps were tripped. The loop average temperatures began to increase, as expected, and the-pressurizer level and pressure began to increase. At'12:25.p.m., the steam dump valves began.to modulate open and one valve failed full open resulting in a rapid cooldown of the primary system. During the cooldown, pressurizer level dropped below 17% at 12:29 p.m. This caused an automatic isolation of letdown and deenergization of the pressurizer heaters. The steam dump valve was manually shut at 12:31 p.m. and the cooldown was terminated. Level in the pressurizer did not go below 14%. Pressurizer level increased above 17% at 12:34 p.m. and a corresponding pressure overshoot occurred. At 12:35 p.m. the reactor was manually tripped due to primary plant pressure approaching the test procedure trip criteria. The pressure rise was terminated prior tc reaching the auto-matic trip set point due to the manual reactor trip. A re, actor coolant pump was started and primary plant temperatures were stabilized at 12:50 p.m. At no time during the transient was a reactor protection or engineered safeguards features actuation setpoint reached. The natural circulation test contains a manual trip criterion which states that the reactor must be tripped if pressurizer level decreases below 17%. NRC inspectors recognized that a manual trip was not initiated when pressurizer level dropped below 17% and informed the Startup Manager, the Test Director, and the Assistant Operations Manager of the requirement to trip the reactor. However, no apparent steps were taken to direct the tripping of the reactor prior to the manual reactor trip for increasing primary pressure.

L , 4.0 Plant and Equipment Performance 4.1 Introduction This section covers plant dynamic response including the steam dump valves. In accordance with the AIT charter the objective is to " determine the expected plant response during a transition to natural circulation cooling and compare it to the actual plant dynamic response' observed during the event." In addition " assess the scope and quality of... licensee identified concerns and corrective actions." Information was collected through interviews with PSNH employees and from GETARS (General Electric Transient Analog Recorder System). This segment of the report is divided into two major parts: 1. Plant response to an overcooling transient, and 2. Mechanical and electrical instrumentation aspects of the first steam-dump valve-bank, before, during and after the test. 4.2 Plant Response For this test the reactor was heavily borated at 1150 ppm boron with all control rods fully withdrawn except for bank D rods which were positioned at 130 steps out of the core (full out equals 228 steps). There were no safety systems or safety functions bypassed for this test,however, the plant was being operated under special test conditions which allowed the reactor to be critical at power without reactor coolant pumps operating. All low reactor coolant flow trips are automatically blocked below the P-7 permissive setpoint (approximately 10*4 power). There is adequate data collection by the plant's GETARS system computer to reconstruct the vital plant parameter behavior with the exception of the valve MS-PV-3011 response. The reason for the lack of valve MS-PV-3011 data is that a connecting link of the (Bailey) positioner of the valve feedback mechanism had become disconnected which affected both valve operation as well as the computer indica-tions. For the primary coolant system transient there are three distinct time segments, that is: 1 From the RCP trip to the steam dump valve opening, 12:18:50 to 12:25:56 (5465 to 5891 sec in GETARS indication) 2., From the opening of the steam dump bank to their closing 12:25:56 to 12:31:06 (5891 to 6202 see in GETARS indication) 3. From the steam dump valve closing to reactor trip 12:31:06 to 12:35:54 (6202 to 6489 sec in GETARS indication) ______-_---__-_-___________a

y . Total transient time 12:35:54 - 12:18:50 = 17 min 4 sec. The following three report subsections discuss each of the distinct time segments in detail. 4.2.1 RCP Trio to Steam Dumo Valve MS-PV-3011 Failure Open After the RCPs were tripped, the total heat input into the p0imary coolant system decreased by about 12 MWt, the total heat input from the primary pumps. The reactor was already at about 2.2% power and the steam dump valves were in manual because they were used to dis-pose of the reactor total heat input of about 86 MWt, (i.e., 74 MWt of nuclear heat and 12 MWt of primary pump input). With reactor power at about 2.2% of rated power, Th and Tc loop temperatures showed the initiation of natural circulation with Th rising to about 570 F and Tc dropping to 545 F cnd Tavg rising by c few degrees in all loops. (See Figures 1.1 to 1.4, Appendix H). Pressurizer level and pressurizer pressure increased as the average reactor coolant temperature increased due to loss of forced circulation, see Figure 2 (Appendix H). Steam generator level stayed constant as well as the charging and letdown flows, see Figure 3. Steam generator pressure began decreasing due to cooldown, see Figure 5. Decreased primary circulation rate caused coolant and fuel temperature in-creases in the core which in turn decreased core power due to doppler feedback, see Figure 4. The core configuration for the test (control rod position and boron concentration) were such that the moderator temperature coefficient was about zero thus, doppler was the only feedback. About 5 minutes into the test, core flow was removing the generated heat, thus Th stopped rising. However, Tc continued falling due to steam dumping. Therefore, Tavg began to decrease. ' Steam generator pressure was also decreasing due to the mismatch between steam dumping level and power and heat production before the RCP trip (at 86 MWt) and after the RCP trip (at 74 MWt). With decreasing reactor power and Tavg, pressurizer pressure and level began to decrease and at 12:24:56 (6 min 6 sec into the test) the (condenser) steam dump valve control was lost due to the Lo-Lo Tavg interlock at 550 F (P-12). This occurred because this inter-lock operates on the narrow range Tavg signal which is located on a bypass loop and without forced circulation cools faster than the reactor coolant. P-12 was bypassed through control room switches and the operator regained steam dump manual control through the first valve bank on the steam pressure mode. As soon as P-12 was bypassed the valves attempted to return to their existing demand position at about 5%. However, MS-PV-3015 was blocked due to a pre-existing excessive air leak, valve MS-PV-3011 went open (i.e., probably failed to modulate) and only valve MS-PV-3019 operated properly. Within 40 seconds valve MS-DV-3019 closed, but MS-PV-3011 most likely stayed open. In this brief time interval steam demand increased, charging flow continued to increase and pressurizer level l

.p: ' f.g ,.m continued to decrease. The operator responding to decreasing pres-surizer level, further decreased the letdown flow, see Figures 2, 3, and 5. 'As soon as valve MS-PV-3011 closed, steam generator level showed a small rapid increase for a few seconds, see Figure 2. y 'In this first time segment the reactor-responded as expected. All major parameters varied in the expected direction and within ex-pected ranges.. Post-event inspection showed that MS-PV-3011 probably failed to modulete. 4.2.2 Steam Dump Valve MS-PV-3011 Failure Open to Closure of All Steam Dump Valves Six seconds after MS-PV-3011 went closed the operator manually be-gan to open the first steam-dump valve-bank to initiate energy dis-posal.. The valve-bank valves are supposed to modulate in unison with instruments in manual pressure control. The control board sig-nals for valves MS-PV-3015 and MS-PV-3019 were correct and as ex-pected. However, valve MS-PV-3011 went fully open and stayed in that position as was verified a few minutes later by actual obser-vition. As steam dumping continued, pressurizer pressure and. level decreased and the operator responded by increasing charging flow, decreasing (to almost zero) the letdown flow and closing the main steam drains. In addition Th began to decrease and Tc showed a sharp downturn, resulting in decreasing Tavg. At this time nuclear power generation shows a slight upturn from a minimum of about.1.4% due to excess heat removal and fuel cooldown. Steam generator pressure decreased due to excessive cooldown see Figures 2, 3 and 5. This plant behavior, that is, excess cooling of the RCS, was caused by valve MS-PV-3011's failure to modulate and being fully open. The valve failure was established by visual inspection during the transient. As this trend continued, at 12:28:54 the pressurizer level fell below 17% (which is a procedural reactor trip level), and, as a result, letdown isolated and pressurizer heaters de-energized, causing the loss of normal pressurizer pressure con-tro3. At this point in time the pressurizer pressure was 2192 psia and reached a minimum value of 2179 psia. Pressurizer level continued to fall and reached a minimum of 14.5% at which point the operator closed the dump valves. Valves MS-PV-3011 and MS-PV-3019 went fully closed. In this time period the plant responded as expected in view of the excess cooling of the RCS. Each valve when fully open discharges about 3.3% of total steam load, thus, with reactor power at about 1.5%, valve MS-PV-3011 fully open and valve MS-PV-3019 partially open the heat loss was at times over 4.0% and the primary system heat loss exceeded the heat input from the reactor. However, valve MS-PV-3011 failed to modulate and the operator failed to trip the

i L p reactor as required. (Note: the valve f ailure' to modulate is-E discussed in paragraph 4.3.) 4.2.3 Steam Dump Valve Closure to Reactor Trip L Within a few seconds of steam dump valve closing, pressurizer pres-sure and level began rising. Charging rate was at 122 gpm (about 1% level / minute) and letdown was isolated. Likewise, steam gene-rator pressure and level began to rise after.a small dip'in the level and an upturn in the pressure, see Figures 2, 3 and 5. Reactor power leveled off at about 2.5%. As pressurizer level and l pressure increased rapidly the operator realized that pressure was y. getting close to 2340 psia (another procedural trip requirement). At 12:35:54 the. reactor was tripped at a reactor pressure of about 2310 psia'and the operators entered emergency operating procedure E-0 in response to a. reactor trip. The rise of pressurizer level-to 17% was recorded at 12:33:55. Therefore, the reactor. stayed below 17% for about 5 min. In this time segment, the reactor coolant system responded as would be predicted due to reduced cooling at high chart.ing rate and zero Jetdown. 4.2.4 Summary of plant Eouipment Response The plant response during operations related to the natural circu-lation test was as would be predicted, and all plant parameters be-haved normally. A steam dump valve MS-PV-3011 failure to modulate caused an unanticipated cooling of the reactor coolant system. All phenomena were explainable and no unexplained parameter values were observed. 4.3 -Steam Dump Valves 4.3.1 Introduction The origin of the primary cooling transient was the malfunction of steam dump valve MS-PV-3011, which stuck open and failed to modu-late. This section reviews valve performance, operating record, failure root cause and licensee short term and long term response. Most of the information regarding the valves was obtained from post-event examination. 4.3.2 MS-pV-3011 Valve Failure to Modulate Post-event examination revealed that a connecting link nut to a positioner arm fell off. This mechanism was providing the feedback and the disconnected link explains the failure to modulate. Never-theless, valve MS-PV-3011 was able to respond to the final fully-closed signal from the control room.

b 1 4.3.3 Steam Dumo Va ne History Interviews with the system support personnel, revealed the following with respect to the steam dump valves: MS-PV-3011 was stroked after the natural circulation test and failed to operate properly due to mechanical binding. Examination after removal of the valve mechanism revealed that the binding was caused by stem misalignment and interference with a guide bushing. At the begtnning of the preparation for the natural circulation' test, valve MS-PV-30Il was not ready to support the test'since work order WRB7 WOO 5592 was still open for a stroke test at NOP/NOT. In spite of this, a test pre-requisite to confirm the availability of the steam dump-system was signed off. There'is no. indication as to when the linkage in valve MS-PV-30Il failea. It had been tested earlier from the control room for close1/open positioning, however, this test would not reveal the linkage problem. After'the' June 22nd event, binding was also found in valve MS-PV-3019 but not enough to prevent open/close motion or modulation. ~ Post-event testing of cll steam dump valves revealed that seven of the twelve valves showed binding, scored stems loose linkage or tight linkage. In general, the history of steam durp valve system work orders indicates that there is a valve r.. intenance or design problem. !.3.4 Valve Failure Cause It is concluded that the MS-PV-3011 steam dump valve failure cause is apparently inadequate valve maintenance or design. Licensee personnel failed to follow through on a pending work order and failed to recognize and resolve a maintenance problem with the steam dump valves. In addition the licensee failed to adhere to test procedures by failing to assure that the required test prerequisites and initial conditions were met before commencing the test. l' 1

-iS-4.3.5 Licensee Short-Terra Response The following actions were taken or initiated by the licensee while the AIT was on the site: dismantled the va)ve MS-PV-3011 maxhanism for shop testing called a vendor representative to the site initiated extensive diagnostic testing for all steam dump valves replaced the valve MS-PV-3011 actuator with a new unit from storage, and performed a comprehensive logic circuit test. These actions envelope an appropriate review of the behavior of the steam dump valves and constitute a technically sound, prompt and adequate response to the specific valve problem. 4.3.6 Licensee Leno-Term Response Licensee personnel expressed their intent for a complete, detailed and in-depth investigation of the valve problem so as to be able to take the appropriate corrective action. It is the team's understanding that the licensee will investigate: generic failure rate data base for this type of valve seek to verify whether valve usage (including surveillance testing) is related to failure frequency, and review (and if necessary revise) the current valve maintenance and surveillance program. These actions appear to be appropriate. 5.0 Personnel Activities and Performance 5.1 Operatino Crew 5.1.1 Organization and Responsibilities Seabrook Station's normal control room shift crew composition and the crew composition that existed during the day shift on June 22, 1989 were reviewed. Normally, while in mode 2 (startup) operations, the Unit I control room operations staff would consist of a Shift Superintendent (55), a Unit Shift Supervisor (USS), a Supervisory Control Room Operator (SCRO) and a Control Room Operator (CRO). Both the SS and the USS possess a senior reactor operator license and the SCR0 and CR0 must be licensed as reactor operators or senior reactor operators. This is consistent with the minimum requirements for licensed operators per shif t for on-site staffing of nuclear power units l l l-

_ _ _ specified in 10 CFR 50.54 and in the facility's Technical Specifications. In anticipation of performing the natural circulation test on the morning of June 22, 1989, the normal shift complement was augmented with additional CR0s to assist in performing various control room functions: one operator was held over from midnight shift to assist in acknowledging secondary alarms; one was assigned to control steam generator level and reactor coolant system temperature; a third operator was assigned responsibility for turbine shell and chest warming (he was never used, however); and a fourth additional operator monitored the radiation monitor panels. This crew augmentation allowed the operators normally assigned to the shift to concentrate on the reactor and the primary plant. The inspectors reviewed a number of facility lictnsee documents in an effort to determine the operating shift crew's responsi-tilities curing normal operations and upset conditions and during the startup test program. Section 2.3 of the Seabrook Operations Management Manual (OPMM) discusses the control room command function and states that the SS is the senior on-shift manager and is responsible for the control room command function. It goes on to state that the SS may, and normally will, delegate this responsibility for each unit to its respective U55. The SS, under Section 3.3.2 of the CPMM. retains the authority to assume command of the control room, or to order the shutdown of the reactor when, in his judgement, such action is required to protect the safety of the unit or the health and safety of the public. Furthermore, the 55 is responsible for the safety and operation of the unit equipment, in accordance with approved Station procedures. Section 3.1 of the OPMM provides an overview of shift operations and states that the SS maintains a broad perspective of conditions affecting the status and safety of the unit, while the USS maintains a comprehensive perspective of operational conditions affecting the safety of the unit and is in charge of the control room during emergencies. Section 4.2.4 of the Startup Test Program Description (STPD) states that the station staff will perform its normal job func-tions as required to support plant operations and the startup test program. Although the Test Director has the primary responsibility for the execution of the test, the station i operating crew has the responsibility for the proper operation I of equipment, systems, and the plant and reserves the right to take appropriate corrective actions whenever unsafe or unsatis-factory conditions exist. A determination by the 55 or USS that l a test would place the plant in an unacceptable condition is identified in Section 4.3.5 of the STPD as an event which constitutes grounds for a test interruption. ..___..______._._._______.___J

-. Section 1.5 of the Seabrook Station Management Manual (SSMM) addresses the issue of procedural adherence and states that where a procedure exists, it shall be considered guidance regarding the method of performing a function. Procedures shall be followed, but not without question. If a procedure directs an action contrary to what is considered proper, the operator should question the procedure and seek resolution with appro-priate supervisory personnel. It states, however, that a Drotecure being questioned should not be deviated from on the basis that it is being questioned. 5.1.2 Trainino The inspection team reviewed the operators' startup test program training completed in preparation for low power testing and other aspects of the licensed operatcr training program which may have had a bearing on this event. During the period from April 14 to May 23, 1986, the facility's plant reference simulator was used to train all the operators on the tests that would be run during the startup program. The SS, USS and the two CR0s having primary plant responsibility during the natural circulation test were verified as having completed that training. Additional claisroom training on the low power test program was conducted as part of the licensed operator requalification training program curing the period from September 12 to October 21, 1988. This course was observed by an NRC inspector and was addressed in Incpection Report 88-13. The four-hour course, which was conducted by the Assistant Startup Program Manager, provided a detailed description of the startup testing program. The course topics included program administration, organization, test equipment, and applicable procedures, including 1-ST-22, the Natural Circulation Test. The training also provided the operators with an awareness of the startup test program structure. The licensed operator initial and requalification training prog. rams were reviewed to determine whether deficiencies in diagnostic and team training or in command and control and procedural compliance training may have contributed to the event on June 22. It was determined that these subjects are addressed in classroom and simulator training during the initial and recualification training programs. Operations Training Standard Number 3, dated January 1989, states that procedural adherence is required with deviation allowed enly af ter procedure changes have been made or in the case of the emergency response procedures by invoking 10 CFR 50.54(x). This training standard is endcrsed by both the Operations and Training Managers. Interviews with a Training Department representative indicated that the operators are trained to comply with all approved station procedures, regardless of whether they are operations

- _ -__ - _ -___________ _ procedures, administrative procedures or test procedures. The CR0s are instructed te advise the USS when anc if reactor trip criteria are approached and/or exceeded and to trip the reactor unless directed otherwise by the U$S. 5.1.3 Pre-Test Briefinj The licensee's requirements for operating crew briefings were reviewed to determine whether those conducted in preparation for the natural circulation test were adequate. Section 1.8 of the OPMM addresses shift evolution briefings and states that they shall be conducted for individuals involved in the performance i of the evolution. The detail of the briefing depends on the complexity, logistics or number of people involved in the evolution. Evolutions involving many individuals, especially from two or more departments or disciplines, may require large formal briefings or planning sessions. It goes on to state that complex evolutions requiring clase coordination of individuals should include the following five elements: review of the appropriate section of the procedure by key individuals; examirztion of each in(ividual's specific involvement and respouibility; discussion of expected results or performance; review of precautions, limitations, emergency actions to be taken if contingencies arise; and assurance that everyone understands the required interface and communications required. The inspection team interviewed the operators involved with the conduct of the Natural Circulation Test, 1-ST-22, on June 22, 19S9, and it was determined that the operators were not briefed as a crew prior to commencing the test procedure. The operating crew members were individually briefed by the Test Director (TD) during the early hours of their shift. Copies of the procedure had been distributed to the operators the preceding day but not all the operators had taten the time to review it in detail; the USS reviewed the procedure on the morning of June 22nd. The primary plant CR0s were given copies of the manual reactor trip criteria, Attachment 9.3 of the test, just prior to commencing the test. One of the CR0s and the SS never received an inq1vidual briefing. The SS did, however, read the procedure three days before the test was attempted, but he cid not have a copy available to him at the time it was being performed. Immediately prior to commencing the test, the TD provided a general overview briefing of the test objectives and procedure i geared for the management observers and operators from other crews present in the back of the control room. This briefing provided a brief overview of the test and was not directed to the operators performing the test.

a I , d lE Discussions with NRC inspectors who were present during earlier phases of the' low power testing program indicated that the pre-test briefings for the natural circulation test were less. thorough than others had been; previous tests'had also generally included.some sort of pre-shift group' briefing rather than. relying solely on individual briefings. The SS did not find out L until after the test was aborted that the operators in his crew had not!been properly briefed, and the USS indicated during his interview that while other pre-test briefings have been short, Lthey have generally been more thorough than what was done for the natural circulation' test. 5.1. 4 Crew Response-

A detailed description of the event is provided in Section 4.0.of this inspection report and a chronology of significant events is provided in Appendix A.

A chronology cf conmunica-tions-during the event is provided in Appendix B. It became evident during the operatur interviews that the primary CR0s and the USS 'were aware that pressurizer level was decreasing and approaching the 17% manual. reactor trip criterion specified-in Attachment 9.3 to 1-ST-22. The SS, not being as familiar with the test trip criteria as the rest of the operating crew and not having a copy of the procedure to which 'he could refer, suspected that level had decreased to less than or equal to 17% when he heard letdown isolate, but he did not associate the letdown isolation with a manual trip requirement. The primary CR0s and the USS were aware that letdown had isolated at 17% pressurizer. level and that the manual reactor trip' criterion had been satisfied. The question of why the operators did not promptly trip the reactor when they realized that pressurizer level had decreased below the 17% trip criterion was pursued by the inspection team in the interviews. The primary CRos knew that the USS was aware of the level control problems and that he was also aware, as they were, of the requirement to manually trip the reactor. Howaver, the CR0s never actually recommended to the USS that the reactor be tripped. Interviews with-the CRD2 indicated that they were generally aware of discussions taking place between the USS and the Test Director (TD) regarding the loss of pressurizer level. The USS informed the TD that pressurizer level had decreased below "your limit." In the interim, the USS directed the primary CR0s to monitor level and to report when it reached 155 At about this time the control room received a report froc an operator in the plant that one of the condenser steam dump valves had failed full open. The valve was promptly closed, terminating the cooldown transient and reversing the pressurizer level decrease at approximately 14.5%. Both level and pressure began to recover quickly after closing

t .. the. failed open steam dump valve, and the operators quickly tried to restore pressurizer pressure control capability. L' Without pressurizer spray or letdown. capability, pressure rapidly increased past 2300 psig and was approaching the high-pressure reactor trip setpoint of 2385 psig. Realizing that pressure was continuing to rise,.the USS directed that the reactor be manually tripped.' The total elapsed time from the point when pressurizer level decreased below 17% until the operators - manually tripped the reactor was approximately five minutes. i It was apparent.from the-operator interviews that there was no doubt in their minds that the command and control function in ' the control room rested with the USS and not with the TD. The USS informed the TD that level was at 17% and decreasing but he failed to recognize that the test procedure 17% pressurizer level-trip criterion required him to direct shutdown of the reactor at this point without further discussion or deliberation. During his interview, the USS indicated that-he did not trip the reactor because other operatir.g procedures do not require a trip until a lower pressurizer level. Since pressurizer level appeared to be stabilizing as it passed through 17%, he made the decision not to insert a manual reactor trip at that time. It was only after the steam dump valve was closed and pressurizer pressure began to rapidly increase toward the automatic trip setpoint that the USS decided that recovery from the transient was not feasible and a manual trip was necessary. The NRC inspectors who were present during the natural circula-tion test witnessed the crew's response to the reactor trip and their performance of the emergency operating procedures. No performance deficiencies were noted during this post trip response. Through observations and interviews the inspectors determined the Emergency Operating Procedures (EOPs) were adequately implemented following the manual reactor trip. The Emergency Operating Procedures are normally implemented with two operators at. the control panel, however during the natural circulation test there were four operators at the control panel. No prior discussion had been held by the USS as to how the operators were to implement the E0Ps. At the inception of E0P implementation the operator's recognized the need to adjust to the situation and reached an unspoken agreement that only two of the four operators would conduct the E0P procedure. As a result one of the additional operato s who was designated as the Shift Technical Advisor (STA) performed the E0P control board manipulations. If the E0P recovery had been extended, the inspectors were uncertain whether this operator would have been free to perform his STA responsibilities. Forty-five minutes after the manual trip the NRC was notified in accordance with 'r

. 10 CFR 50.72 by the Shift Superintendent. This was well within the four hour reporting requirement. 5.1.5 performance Assessment The operating crew did not comply with an explicit procedural requirement to manually trip the reactor even though they were fully aware that the established trip criterion had been exceeded. The CR0s should have recommenced to the USS that the reactor be tripped before level exceeded the 17% pressurizer level criterion. _The operator interviews revealed an apparent tendency by some of them to place higher priority on satisfying some procedural requirements than others. Some of the operators had attached a greater safety significance and importance to complying with a 7 chnicel Specification or emergency operating t procedure requiremev. 1546, for instance, a test procedure requirement. Subst:;# erd to arriving at this conclusion from the operator interview., the team viewed a video tape of the Jane.22nd natural circulation test in which the USS, when discussing the pressurizer level problem with the TD, referred to the 17% pressurizer level reactur trip criterion as "your limit". The USS apparently felt comfortable tnat the situation was under control since he had not yet approached a lower level trip criterion established in the emergency operating pro-cedures. This hierarchical approach to procedural compliance is not endorsed by the facility's administrative policies nor by the operators' licensing and continuing training programs. The pretest briefing conducted for 1-ST-22 appears to have been inadecuate in that the operators were never formally briefed as a group to address the five elements identified in Seabrook Station's 0PMM. All complex evolutions, particularly those involving new or infrequently performed tasks, should be thoroughly briefed. The fact that the natural circulation test simulator training had been performed over three years earlier and the classroom training was almost a year old should have provided added incentive to ensure that the operators receive some refresher training and were thoroughly briefed prior to commencing the test. The fact that the operators on shift that morning had not routinely worked together as a crew shculd have emphasized the need to examine, during the pretest briefing, each operator's specific involvement and responsibility and understanding of the required interfaces and communications. These observations and the assignment of the STA function to a i panel operator, discussed in the previous section, indicate the need for more thorough planning and preparation to have been j done prior to this test. Although the $$ normally serves an oversight function in the l control room, his level of awareness, knowledge and involvement of shift evolutions was not commensurate with the significance m_-a.m_ _ _ _ m .m m _m._

- _ - I l and complexity of this test. The fact that the natural circu-lation test is one of the first evolutions performed with a critical reactor and the fact that the test involves abnormal operating conditions should have been sufficient to raise the SS's level of awareness and involvement. NOTE: The video taping referenced above was done by the licensee for use by the training department in future training sessions. The inspection team found the video tape to be supportive of the information obtained from the interviews and the resultant conclusions. 5.2 Startup Test Grouj 5.2.1 Organization and Responsibilities The organization and responsibilities of the startup staff are delineated in the Startup Test Program Description (STPD), Rev. 2. The Startup organization is led by the Startup Manager. The Startup Manager has the overall responsibility for the initial startup program and reports to the Station Manager. The Startup Supervisor reports to the Startup Manager. The Startup Supervisor is responsible for detailed coordination of the startup test program. Reporting to the Startup Supervisor are the Shift Test Directors. The Shift Test Director's responsibilities include in part to insure required test conditions are established in a safe and prudent manner, and maintained as necessary for test performance. The startup staff normally present in the control room during startup test performance are the Shift Test Director and the Test Director. The Test Director reports to the Shift Test Director and is responsible to perform individual startup tests. At the time of the Natural Circulation Test Performance, the Startup Supervisor was the Acting Shift Test Director. During the performance of the Natural Circulation Startup Test, 1-ST-22, the Startup Manager, Shift Test Director, and Test Director were all present in the Control Room. The responsibilities of the Startup Staff and the Station Operating crew for specific activities are provided in Table 1-1 of the Startup Test Program Description. Test coordination and direction activity is designated as being the responsibility of the,Startup Test Department. The responsibility for systems and equipment operations is delegated to the Station Operating crew. Section 4.3.5 of the Startup Test Program Description states that the Startup Supervisor or the Shift Test Director will determine if a startup test should be interrupted. An example i of events which may warrant a test interruption provided in the Test Program Description is the inability to maintain plant L

u . i conditions as specified in the startup test. Section 4.3.6 of thelStartup Test Program Description states that.the Startup Manager, Startup Supervisor, or Shift Test. Director will determine if a test will be terminated. An example of a test termination event is if the performance of a test procedure reveals design or equipment deficiencies which prevent the cajettives and/or acceptance criteria from being met. During ti.ie. natural circulation test, a plant condition (pressurizer 'hvel greater than 17%) was not maintained due to the steam dump. valve problem which prevented test objectives from being met. However,-no interruption or termination action was'taken by the Test Organization. 5.2.2 -Test Procedures . Methods to change Startup Test Procedures are described in the ~ Startup Test Program Description. Test procedure changes may be made utilizing two methods. For major changes, a procedure revisi.on is required. Procedure revisions undergo extensive review and comment cycles including review by Westinghouse. The procedure 'is recoa. mended for approval by the Startup Manager and reviewed by the Station Operations Review Committee (SORC) prior to being approved by the Station Manager. Field procedure changes fall into two categories: intent changes and non-intent changes. Procedure changes which involve a change of intent must be reviewed and approved by the Startup Supervisor, SORC, and the Station Manager prior to being implemented. Non-intent procedure changes (e.g. editorial changes) must be reviewed and approved by the Startup Supervisor and the Unit Shift Supervisor (or anoti.er SRO) prior to implementation. In the event that the Startup Supervisor is unavailable the Shift Test Director may provide this review and approval. Non-intent changes are reviewed by SORC within 14 days of implementation. A review of the Seabrook startup test procedures is documented in NRC inspection reports 50-443/86-31, 86-48 and 88-13. Each of the. inspection reports describes a small number of minor procedure changes which, if incorporated, would more clearly or. correctly state procedural steps and test objectives. The applicable changes described in the inspections were incorporated into the startup test procedures prior to implementing the Zero Power Test Program. The inspections concluded that the startup test procedures were well prepared and technically sound. Also, the number of test procedure changes made during the test program thus far appears to be less than other comparable facility test programs.

. 1 5.2'.3 ' Training. Training requirements for the Shift Test Director and the Test Director are provided in the Startup Test Program Description, Rev. 2, Section 5.2, " Personnel Training." The Shift Test Director, and Test Director were provided training in those aspects'of the program applicable to' procedure compliance,. test performance, and test documentation. This training was provided to-students, as a formal classroom lecture. In adaition, the Shift Test Directors and Test Directors were also provided training on selected transients which might be expected as abnormal occurrences during various startup tests. This training covered general transient conditions which could occur and did not explicitly cover cooldown transients during natural circulation. The transient training was provided on a self study basis, without formal training handouts or lesson plans, with an examination given at the end of the self study period. Both the Shif t Test Director and the Test Director met the training requirements described in the Startup Test Program Description. In addition, both the Shift Test Director and Test Direccer attended an additional course on transient analysis -which was conducted in the simulator. No members of the Startup Organization are operator license holders at the Seabrook Station, nor are they required to be. 5.2.4 Pre-Test Briefing Procedure 1-ST-22 Rev 2., " Natural Circulation Test," Step 3.2 states that " Personnel involved with the performance.of this procedure have been briefed on the procedure content and informed of their respective duties." The Test Director provided information copies of test procedure 1-ST-22, Rev. 2, to the primary desk, the Unit Shift Supervisor's desk, and the Shift Superintendent's desk a few days prior to initiating this procedure. The actual execution of the pre-test briefing and sign-off of the procedural step occurred a few hours prior to initiating the test by the Test Director speaking with the licensed operators individually on shift. The briefings were very short according to the operators, but did cover the manual trip criteria. The Test Director supplied copies of the manual trip criteria and expected plant response, Attachment 9.3 of 1-ST-22, to the primary operator, reactor controls operator, and Unit Shift Supervisor just prior to the test. The control board operators and the Unit Shift Supervisor responsible for shift operation, stated in interviews following this event that they were made aware of the 17% pressurizer level trip criterion during the pre-test briefing with the exception of one of the two control board operators assigned to assist the shift crew who stated he was not briefed and was not aware of the manual trip criteria.

ma<

g; L 5.2.5 Test Group Response The.startup test group crew respo'nse was derived from observations made by the inspectors during,the event and.

. interviews held with the startup group staff following the event..The Startup personnel present in the control room during the event were the Startup Manager, Shift Test Director,. Test Director, and'other supporting Startup Engineers. Of the Startup Staff only the Test Director was positioned inside the operating area, other members of the Startup Group witnessed the test from inside the control room but outside. the operating area. prior,to the eventithe Test Director _was communicating test instructions with the operating staff and monitoring test data. After the' reactor coolant. pumps were tripped; per the test procedure, andLthe test initiated, the Test Director pritrarily monitored computer and panel indications. The Twt Director stated.that during the event he was aware that:the pressurizer level had decreased below the manual reactor trip criteria of.17% when it'was announced by the control' board operator that letdown had isolated. At this-point the Test Director did not recommend-to the. Unit Shift Supervisor to trip the reactor. He indicated to the Unit Shift Supervisor that he would monitor computer trends for Tavg to assure that the 15 minute Technical Specification on Lo-Tavg was not violated. Performing this task essentially removed the Test Director from the overview of plant status. During the period when pressurizer level was below 17%, an NRC inspector monitoring the test activities expressed a centern to the Test Director that the pressurizer level,was below the manual trip criteria. Following this communication the Test Director stated that he told the Unit Shift Supervisor that the NRC has a problem with being below the manual trip. criterion. The Test Director stated that the USS said he was handling it. At no time during this test did the Test Director recommend that the operators manually trip the plant. The Shift Test Director was monitoring test activities from outside the operating area. The Shift Test Director stated that he was aware of.the manual trip criterion for pressurizer level and that the value was exceeded during the test. He stated that he focused on attempting to analyze the on going transient and, therefore, did not provide an advisory role to the operating m W staff. At no time during this transient were recommendations-provided by the Shift Test Director to the operating staff. The Startup Manager was monitoring test activities from outside the operating area. The Startup Manager stated that he first became aware of the pressurizer level being below the 17% trip criterion when informed of such by the NRC inspector monitoring startup testing. The Startup Manager did not communicate the

-. inspector's concerns to the operating crew or other members of- -the Startup Staff. '5.2.6 Performance Assessment The inspection team concluded that the pre-test briefing performed by the Test Director was conducted in a fragmented and abbreviated manner. Due to the interactions which occur between plant systems and operator actions, it is important to perform the operator pre-test briefings as a group rather than in a piece meal ~ fashion. Three levels of the Startup organization were aware that the-pressurizer level was below the manual trip criterion during the transient. Only-after the NRC inspector voiced a concern did a startup-organization member (Test Director) indicate a concern. to the operating staff. At no time during the. performance of this test did any member of the startup group communicate to-the station operating staff a recommendation to interrupt or terminate the test procedure. The technical guidance provided by the startup organization to the operating staff during this event was inadequate. In general, the startup organization became more occupied with individual tasks at the expense of maintaining at least one individual with overall responsibility for overview and technical input to the Unit Shift Supervisor for conduct of the test procedure. 5.3 Management and Support Staff 5.3.1 Management and Other Support Personnel During the performance of the test there were approximately seven managers in the main control room. The Vice President of Nucitar Production was the most senior manager present. The Operations Manager and the Assistant Operations Manager were the only managers in the control board area. Approximately twenty licensed operators, in addition to the operating crew, were in the main control room to observe the natural circulation test to fulfill the commitment in Final Safety Analysis Report (FSAR) request for additional information response. These operators remained outside the control board area, did not become involved with p' ant operations and maintained a quiet presence throughout the test. 5.3.2 Management Responsibilities The responsibilities of the Station Manager and the Assistant Station Manager, both of,whom were present in the control room

p L D' l' l during the test, are delineated in the Nuclear Moouction Management Manual-(NPMM). The Station Manager.is responsible for ensuring the station is operated and maintained in accordance with applicable requirements and he serves as chairman of the Station Operation Revien Committee (SORC). The-Station Manager has the authority to direct reactor shutdown when conditions may endanger equipment st3tus or the health and safety of the public. The Assistant Station Manager is responsible for maintenance of the programs and procedures needed to operate the station in accordance with applicable requirements and he also has the authority to direct the reactor; to be shutdown. The responsibilities for the Operations Manager and the Assistant Operations Manager are delineated in the Operations Management Manual (OPMM). The Operations Manager is responsible to direct operating activities in a safe and reliable' manner, supervise the Assistant Operation Manager and he is a member of SORC. He has the authority to order the shutdown of the reactor when action is required to protect the safety of the station.or the health and safety of the public. The Assistant Operations Manager has the responsibility for safe operation of the unit's equipment and directs the activities of the members of the operating crews. He also has the authority to order shutdown of the reactor. S.3.3 Manacement Response Of the managers interviewed, two were aware, during the test, that pressurizer level had dropped below 17%. The Station Manager was the only manager interviewed that knew of the existence of a trip criterion on pressurizer level but was unfamiliar with the exact criterion. Most of the staff members interviewed are members of the 50RC which had reviewed and approved the natural circulation test procedure. Through interviews with the management staff and review of management responsibilities ir the NPMM and the OPMM it was determined that four of the managers interviewed had the authority to direct a reactor shutdown. However, none of these managers communicated to the USS a need to trip the reactor when pressurizer level decreased below 17%. The Station Manager stated he was not sure why the USS did not trip the reactor but believed it was due to the training the USS had received in the simulator. The Operations Manager stated that the USS did not trip the reactor because the US$ knew the cooldown was causing the pressurizer level drop and that the USS knew the cooldown was under control. During an interview conducted on June 24, 1989, the Station

_ Manager stated that he recognized soon after the reactor trip on June 22, 1989, that the failure to follow procedures was a significant problem but had been unable to conduct a full discussion of the problem with his management team prior to meeting with the VP Nuclear Production at approximately 5:00 p.m. During a conference call on June 22, 1989 with the Region I Branch Chief at 6:00 p.m. the VP Nuclear Production indicated that the procedural compliance issue would be looked at and put in proper perspective and that if the event occurred again he would expect the operators to trip the reactor. The VP of Nuclear Production initially indicated a desire to restart the reactor early the next morning but agreed to postpone reactor startup until after a follow-up conference call. During the follow-up conference call at 7:30 a.m. on June 23, 1989, the licensee outlined the planned modification to their management manuals that would provide additional guidance on the implementation of procedures and outlined the briefings that were planned with all shif t crews to present the new guidance. As a result of a subsequent phone call between the Deputy Regional Administrator and the President of New Hampshire Yankee Division a Confirmatory Action Letter was issued requiring that a complete review and analysis of the event be formally prepared and presented to the NRC prior to reactor restart. Immediately after the phone conversation the license's Event Evaluation Team, the Human Performance Evaluation System team, and the Independent Review Team were assigned to perform separate evaluations of the event. 4 5.3.4 Performance Assessment The initial management thrust following this emnt appeared to be to resolve any equipment problems necessary to resume testing. An in-depth review of the cause or causes leading to the improper conduct of the 1-ST-22 natural circulation test apparently did not take place prior to an initial management decision to resume testing. An extensive review of this event was not completed by the licensee until after the NRC raised this issue with licensee management. During the conduct of 1-ST-22 and at the time when plant conditions had reached the reactor trip criterion associated with pressurizer level, there were several plant management i representatives in the control room with the responsibility and authority to terminate the test and plant operations when j approved procedures are not being followed. This was not done. When a member of management having specific responsibility and i authority relative to safe operation of the plant is present ) in the control room, their presence in no way dilutes the I i i

!. I . responsibilities of the operating crew and test group personnel assigned to' shift. ' However, by' virtue. of the particular , responsibilities and authorities that they do possess' relative ~ to safe plant operations, there is a responsibility - particularly during unique testing situations - to Keep them- 'selves informed of. key limits for operation and plant status relative to those limits and to take appropriate action' relative to plant operation whenever others they have assigned to do this have'not done so. Plant.ma'nagement present.did not do this in the case of the 17*; pressurizer level trip criterion that was exceeded. The reactor was' subsequently shut down by the USS when the transient response of another parameter, primary plant pressure, caused the USS to take this action. 6.0 Safety Assessment 6.1 Reactor Safety Sigr.ificance of the Event The aspects of.this event which cause the plant transient to be different from the intended natural circulation test transient are the failure of valve MS-PV-3011 to modulate and the fact that the operators did not manually trip the reactor based on pressurizer level. The excess cooling of the reactor coolant system is of little or no reactor plant safety significance in that it is very minor by compgr f son to other analyzed events (steam line break, 3 inadvertent ini.t'fa d on'of a coolant loop, etc.) and these have been analy:ed add'ihown to be acceptable. The June 22nd event is, therefore, totally bounded by these other analyzed events. 6.2 Safety Significance of Personnel Performance The failure of the operating crew to trip the reactor when required by the test procedure during the June 22nd test; the failure of test group personnel to recommend tripping of the reactor at that point and the failure of management present in the control room to exercise their responsibilities in this situation, despite the fact the plant was being operated under a Technical Specification Special Test Exception, is safety significant. Also, the apparent willingness of management to proceed with testing following the June 22nd occurrence without first completing a thorough review and causal factor assessment is safety significant. Test procedures often involve placing the plant in unusual conditions for operation, conditions which are not routinely experienced nor necessarily adequately covered by normal operating procedures. Use of test procedures results in operation under an approved margin of safety only when strictly followed. These test procedures are carefully developed, utilizing industry experience and expertise, are carefully reviewed and only approved after confidence is established in their ability to assure plant safety. The conduct of tests such as the natural circulation test in which

. the reactor is critical without reactor coolant pumps operating is an example of testing under unusual conditions, conditions which call for a heightened sensitivity to plant status and attention to strict procedural adherence. Neither shift operators, the key test group personnel nor the managers present in the control room during the June 22nd test demonstrated an adequate understanding of this. As stated in the previous section, the particular plant transient ,which resulted from the combination of the steam dump valve equipment problem and failure to follow the test procedure reactor trip criterion did not significantly challenge the plant margin to safety. However, the operational practice exhibited by the personnel in the control room was unacceptable. The AIT concluded that all operations, test group and management personnel interviewed now recognize that testing can proceed only if done so in accordance with the test procedure requirements and that if testing should for any reason proceed otherwise the test procedure must first be formally revised. The AIT found no indications of uncertainty or equivocation about this during the site visit. 7.0 Exit Interview On June 30, 1989 a preliminary exit interview was held with licensee mar.agement to review the observations and assessments of the AIT. The licensee was informed at the time that this interview might not be the final exit for this inspection. During this inspection, the NRC inspectors received no comments from the licensee that any of their inspection items or issues contained proprietary information. No written material was provided to the licensee during this inspection. On July 5,1989, the team briefed regional management on the results of the inspection. The licensee was informed by NRC Region I management that the above exit interview would be considered the final exit.

y-( l 1' I.' e t.- W - EXHIBIT I l i _.._--__.-.--_------__a

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UNITED STATES OF AMERICA l L NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges: Sheldon J. Wolfe, Chairman Emmeth A. Luebke i Dr. Jerry Harbour ) In the Matter of ) ) pUBLIC SERVICE COMPANY OF ) Docket No.(s) NEW HAMPSHIRE, ET AL. ) 50-443/444-OL-1 (Seabrook Station, Units 1 and 2) ) On-site EP ) September 16, 19&& ) AFFIDAVIT OF ROBERT D. POLLARD I, Robert D. Pollard do make oath and say: 1. My name is Robert D. Pollard. Since February 1976, I have been employed as a nuclear safety-engineer by the Union of Concerned Scientists. My business address is 1616 P Street, N.W. Washington, D.C. 20036. Previously, I was employed by the United States Nuclear Regulatory Commission as a Licensing Project Manager for commercial nuclear power plants. 2. In May 1959, I enlisted in the United States Navy and was selected to serve as an electronics technician in the nuclear power program. After completing the required training, I became an instructor responsible for teaching naval personnel " O h$ ~q$$'$$ $b

both the theoretical and practical aspects of operation, maintenance and repair for nuclear propulsion plants. From February 1964 to April 1965, I served as the' senior reactor operator, supervising the reactor control division aboard the U.S.S. Sargo, a nuclear-powered submarine. In May 1965, I was honorably discharged from the U.S. Navy and attended Syracuse University, where I received the degree of Bachelor of Science magna cum 1Aude in electrical engineering of June 1969. 3. In July 1969, I was hired by the United States Atomic Energy Commission (AEC) and continued as a technical expert with the AEC and its successor, the United States Nuclear Regulatory Commission (NRC) until February 1976. After joining the AEC, I completed a year of graduate studies in advanced electrical and nuclear engineering at the Graduate School of the University of New Mexico in Albuquerque. I subsequently advanced to the positions of Reactor Engineer (Instrumentation) and project Manager with AEC/NRC. As a Reactor Engineer, I was primarily responsible for performing detailed technical reviews analyzing and evaluating the adequacy of the design of reactor 1 protection systems, control systems and emergency electrical power systems in proposed nuclear facilities. In Septernber 1974, I was promoted to the position of Project Manager and became responsible for planning and coordinating all aspects of the design and safety reviews of applications for licenses to construct and operate several commercial nuclear power plants. _m 4. In the course of-my six and a half years with the AEC and.NRC, I' performed technical reviews, analyses and 1 evaluations of designs of systems and components necessary for safe operation of reactor facilities under normal, abnormal and emergency conditions for the purpose of determining whether such systems complied with NRC rules and provided an acceptable. level of safety for the public. In particular, I was assigned to the agency's safety. review of the operating license applications for Indian point Units 2 and'3 which, like the Seabrook plant, were designed by Westinghouse. 5. For the past twelve years, I, along with other membets of the UCS's professional staff, have conducted numerous studies pertaining to the safety and reliability of nuclear power plants, both on a generic and plant-specific basis. I have provided technical analysis for UCS's participation in rulemaking proceedings before the Nuclear Regulatory Commission and for UCS's litigation against the NRC for failure to fulfill its responsibilities under the Atomic Energy Act. I testified before'the President's Commission on the Accident at Three Mile Island which investigated that 1979 accident. I participated as an expert witness in the NRC's adjudicatory proceeding on the restart of Three Mile Island Unit 1. I have also testified on matters pertaining to reactor safety before numerous committees of the United States Congress and various other state and local legislative and administrative bodies.

Thus, 3-

my 18 years of professional experience on the technical staffs of the AEC, NRC, and USC gives me a first-hand knowledge of NRC regulations and how they are developed, administered and interpreted. 6. On June 27 - 29, 1988 Seabrook Station conducted a FEMA /NRC graded exercise. In that graded exercise, objectives were defined for the Seabrook Station, the New Hampshire Yankee Offsite Response Organization and the States of Maine and New Hampshire. FEMA /NRC Graded Exercise, Chapter 2, Sections 2.2 -2.5. Classed under the Seabrook Station personnel who participated in this exercise are the Control Room / Simulator - Control Room, the Technical Support Center ("TSC") and the Emergency Operations Facility (" EOF") (hereafter referred to as licensee onsite emergency response personnel or onsite emergency staff, notwithstanding the offsite location of the EOF.) During an emergency, the EOF and TSC are responsible for, inter alia, making recommendations for protective actions that are carried out onsite. Thus, in assessing the adequacy of onsite emergency preparedness, the NRC evaluates actions taken by the TSC and the EOF. 7. Among the established objectives for the licensee's onsite Seabrook Station emergency plan was the following: " Demonstrate the ability to analyze station conditions, parameter trends and develop potential solutions for placing the unit in a safe, stable condition. The Control -4

N, Room, T[echnical) S[upport] C[ enter) and E[mergency) O[perations) F[acility) will demonstrate this objective." 1988-FEMA /NRC Graded Exercise at 2.2-2. One of the major objectives of an emergency response plan is to minimize the release of L I radioactive materials outside the plant. Thus, the emergency. I plan must provide for training and qualifying personnel on the emergency tasks for which they are responsible as specified in the plan. Among the most important functions for which trained qualified personnel are needed is to assess the plant condition to develop appropriate strategies for coping with the accident and to prioritize the various potential solutions-to the +. accident. B. The personnel responsible for assessing plant conditions must have adequate understanding of the plant's design, the identified design basis accidents and the effectiveness of each of the plant's discrete safety systems as they relate to the mitigation of those specific accidents. Without that understanding those personnel would be unable or unlikely to develop appropriate solutions and take the appropriate actions in response to a particular accident. 9. For example, the emergency feedwater system is one of Seabrook's eng'ineered safety feature systems. This system was designed to assist in mitigating some Seabrook design basis accidents such as loss of main feedwater and small break LOCA. However, the emergency feedwater system would have little or no i 5-

l ' l, potential-for mitigating a large break LOCA. Such knowledge of the benefits and limitations of each safety system in mitigating the effects of a particular design basis accident is one of the most fundamental criteria for accurately judging whether.the TSC and EOF staff have been properly trained and qualified to carry out the onsite emergency plan. 10. An exercise scenerio was developed to test the objectives established for the NRC-and FEMA graded exercise with regard to the state of the licensee's onsite preparedness. This accident scenario is described in Chapter.5 and in more detail in Chapter 6 of the document entitled 1988 FEMA /NRC Graded Exercise. The pertinent aspects of this scenario with, respect to the emergency feedwater system are as follows: a) The initial conditions of the scenario were that the plant is at 100% power and one of the emergency feedwater pumps is out of service; b) During a controlled shutdown of the reactor at 20% per hour another emergency feedwater pump is disabled; c) At this point the controlled shutdown is stopped 1 and attempts to restore to operability one of the l EFW pumps begin. d) A large break LOCA occurs. - - _ _ - _ _ _ - _ _ _ _ _ _ - _ _ _ _ - _

11. The scenario called for a halt in the controlled shutdown when the second EFW pump was disabled apparently .because continued shutdown of th9 plant could create the need for. operation of the emergency feedwater system. Thus, in my view, halting the shutdown and trying to repair the EFW pump would be the correct actions under those circumstances. However, as soon as the licensee's onsite emergency planning staff in the TSC and EOF recognized that a large break LOCA had

occurred, they should have then recognized that any further efforts to repair the emergency feedwater system were of little or no value in bringing the reactor to a safe stable condition and reducing the radiation release to the environment and the public.

In fact " efforts continued to restore the Emergency Feedwater Pump after a large break LOCA." Inspection Report 50-443/88-09 at 5. (Attached as Exhibit A hereto.) This ineffectual action is one example cited by the NRC staff in support of its conclusion that: '~*The Technical Support Center (TSC) and Emergency Operations Facility (EOF) staff displayed questionable engineering judgement (Exhibit A at 5) 12. As noted earlier, the exercise objective was to demonstrate the onsite staff's ability to analyze plant conditions, analyze parameter trends and develop potential solutions. The NRC Staff classed as an exercise strength that "[pilant conditions were quickly recognized and classified"

(Exhibit A at 4), i.e. apparently the onsite emergency staff recognized from plant parameters that a large break LOCA had occurred. The NRC labelled as an exercise weakness the questionable engineering judgment displayed by the onsite staff's continued efforts to restore the EFW pump to operability despite havi,ng identi'.ied the accident as a large break LOCA. In my view, a more fundamental flaw or deficiency is revealed by these actions than simply " questionable engineering judgment." The fundamental deficiency is that the exercise established that the licensee's onsite staff did not demonstrate an ability to develop potential solutions for placing the reactor in a safe stable condition. In this scenario the reactor was in the midst of a major accident with the potential for enormous offsite radiation doses but the onsite emergency personnel occupied themselves with activities that had little or no potential for preventing or mitigating such releases. Thus, rather than simply revealing questionable judgment such actions indicate a seriously deficient level of competency in developing " potential solutions for placing the unit in a safe stable condition". 1988 FEMA /NRC Graded Exercise at 2.2-2. No doubt the NRC Staff's finding that "the Licensee's performance demonstrated that they could implement their Emergency Plan and Emergency plan Implementing procedures in a manner which would adequately provide protective measures for the health and safety of the public" was based on the fact _-. _ _ _ _ -

that the inappropriate efforts to restore the EFW pump did not complicate the accident or exacerbate the consequences. However, under other accident scenarios the onsite staff's incapacity to " develop potential solutions" could complicate the accident and exacerbate the consequences. In this instance i an inadequately trained onsite response staff did no additional ~ harm, but there is no basis for concluding that the actions of an inadequately trained staff would be of no negative i consequences for the public in all accidents. 13. Another indication of the lack of adequate onsite staff training was that "[n]o effort was noted to blowoown Steam Generators to lessen the heat load in containment" (Exhibit A at 5). The NRC Staff labelled this observation an " exercise weakness." One of the goals of'the emergency reponse-y_ to an accident is to rapidly reduce containment temperature and - w pressure following a LOCA thereby lessening the magnitude of any radiological release. One of the sources of heat for the containment is the heat stored in the Steam Generators. In this particular accident scenario, blowdown of the Steam Generators would contribute to reducing the containment heat load thereby assisting in achieving the goal of rapid reduction in containment temperature and pressure. In my view, the l i failure to blowdown the Steam Generators stems from the same basic deficiency that resulted in the continued efforts to i restore the EFW pump, i.e., the onsite emergency response personnel do i l .J

not have a sufficient' level of knowledge of the potential solutions available to mitigate the onsite and offsite radiological consequences of an accident. In the case of the attempt to restore the'EFW pump, the emergency response personnel were expending effort which, even if successful, had little or no potential for placing the reactor in a safe, stable condition or reducing the radioactive release. In the case of the steam generator blowdown, the emergency response personnel made no effort to take action, which if successful, would have contributed to reducing the radioactive release. 14. A related aspect of the onsite staff's inability to. develop potential solutions for placing the reactor in a safe E-'I stable condition is the NRC Staff's conclusion that "[a] ~ questionable fix for the Containment Building Spray system" (Exhi' cit A a t 5) was used. In this particular case, the onsite emergency staff was taking action that had the potential for mitigating the radiological consequences, but the nature of those efforts give rise to questions, as the Staff found, about the engineering judgment of the personnel responsible for implementing the onsite emergency plan. 15. In summary, onsite emerge'ncy response personnel failed to take an appropriate action (Steam Generator Blowdown), expended efforts on inappropriate actions (continued efforts to restore the EFW pump) and implemented appropriate action with a " questionable fix" (Containment Building Spray System).

Thus,

contrary to the objective of the exercise, the onsite emergency response staff did not demonstrate an " ability to analyze station conditions, parameter trends and develop potential solutions for placing the unit in a safe, stable condition." 16. Two distinct objectives of the licensee onsite emergency plan are: 1) to recommend the appropriate offsite actions to mitigate the consequences which result from the amount of radioactive material being released; and 2) to take actions onsite to reduce or terminate the release of radioactive material. Adequate onsite emergency preparedness requires the capability to accomplish both objectives. 17. In this case, the NRC staff classed as an exercise strength its conclusion that that " Protective Action Recommendations (PARS) were prompt and conservative," and as an exercise weakness the onsite staff's " lack of effort to locate and isolate the release path." (Exhibit A at 5). The first step in attempting to reduce or terminate releases from the plant is to identify the location or path by which the radioactive material is escaping. The failure of the onsite staff to expend any effort in this regard is a fundamental I deficiency that is not and can not be counterbalanced by a capacity to recommend the appropriate offsite measures. Whether the failure to attempt to locate and isolate the release path was due to inadequate training, inadequate numbers of personnel or some other factor, it remains a significant and I 1 i

fundamental deficiency in the state of onsite emergency preparedness. 18. In addition, with respect to the issuance of a low l power license, the failure to attempt to locate and isolate the release path is of particular importance since low power operation does not require adequate offsite emergency planning. In short, the critical aspect of an onsite radiological emergency plan during low power operation is the capacity of the onsite staff to prevent any release that would require offsite emergency measures. Thus, the NRC Staff's claim that the of fsite PARS were " prompt and conservative" is - of no relevence to the issuance of a low power license. 19. The NRC staff classified the failure of both the EOF and TSC staff to question "a release of greater then 7000 curies per second with only clad damage and no core uncovery" as an exercise weakness in that the onsite emergency preparedness personnel "did not recognize or address technical concerns." (Exhibit A at 5). This failure of both the TSC and EOF staff is an indication that the onsite emergency response personnel's knowledge of the relationship between the magnitude and rate of a radioactive release and the amount of core damage is seriously deficient. 20. During an emergency such as a major accident, the onsite emergency response staff faces an unusual, complex set of circumstances with limited information and the potential for 12 -

'some information to be erreneous due to equipment failures. In attempting to analyze station conditions, the licensee's staff may be confronted with indications of a large radioactive release and little core damage or a small release with major core damage. Without a sound knowledge of the magnitude of releases possible under varying degrees of. core damage, the emergency response staff may not recognize that their analysis of plant conditions is incorrmet, leading them to take incorrect' protective actions or fail to take the correct protective ~ actions. Signed under the pains and penalties of perjury this 16th. day of September 1988. Robert D. Pollard 5 - - - - _ _ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ ._-- - _ _ a

1 m [. h. i G ' 5 l. l L. l ) l. l e EXHIBIT J ._-_mm_._.____._.__m_. _ _ _ _ _., _ _ _ _.

.h 1-7 UNITED STATES OF AMENICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges: Sheldon J. Wolfe, Chairman Emmeth A. Luebke Dr. Jerry Harbour ) In the Matter of ) ) PUBLIC SERVICE COMPANY OF- -) Docket Nos. 50-443-OL-1 OF NEW HAMPSHIRE, et al. ) 50-443-OL-1 ) (On-Site Emergency (Seabrook Station, Units 1 and 2 ) Planning and Safety ) Issues) November 8, 1988 SECOND AFFIDAVIT OF ROBERT D. POLLARD I,-Robert D. Pollard, do make oath and say: 1. I am a Nuclear Safety Engineer.for the Union of Concerned Scientists. A statement of my qualifications is contained in paragraphs 1. through 5. of my affidavit filed in this proceeding on September 16, 1988. 2. The purpose of this affidavit is to respond to the Board's Order, dated October 25, 1988, requiring, inter alia, affidavits "specifically and in detail responding to the affidavits attached to Applicants' response of September 28" and l "specifically and in detail address [ing]-pages 8 through 10 of the NRC Staff's Inspection Report No. 50-443/88-10. ORDER, October 25,-1988, at 2. 3. After observing the licensee's annual full-participation emergency exercise performed June 28-29, 1988, the e t-34((IV 0 nw] 3 ft

l t L NRC inspection-team reported that "[t]he Technical Support Center (TSC) and Emergency Operations Facility (EOF) staff displayed questionable engineering judgement and/or did not recognize or address technical concerns." NRC Inspection Report 50-443/88-09, page 5, attached as Exhibit A to my prior affidavit. The NRC gave five examples to support this conclusion. I have addressed each of these examples in successive paragraphs of this affidavit. 4. Since the NRC portrayed the five items as only examples, I conclude that there were additional instances during the exercise where conduct of the TSC and EOF staff provided further support for the NRC's conclusion regarding poor engineering judgement and failure to recognize or address technical concerns. However, since the NRC has not provided information about these other instances, I am prevented from providing an assessment of their significance. j Emergency Feedwatar System (EFW) 5. One of the examples cited by the NRC to support its j conclusion was that "[e]fforts continued to restore the Emergency Feedwater Pump after a large break LOCA." Id. 6. The licensee claims that, after a large break loss-of-coolant accident (LOCA) was postulated, efforts to repair the inoperable EFW pump were continued because "these efforts would not affect other ongoing LOCA response activities," and "to assure a backup heat removal method if a need for future use J l . ) .________-D

arose, even if a current need was not perceived." Affidavit of Gary J. Kline, page 4. 7. The NRC, relying on C. J. Conklin, a different Senior Emergency Preparedness Specialist than the one, E. Fox, who actually participated in the exercise inspection and observation, now says that the " licensee correctly stated that the EFW pump would be required to operate to support steam generator cooldown in the recovery phase and continued repair efforts were prudent." NRC Inspection Report 50-443/88-10, page 8, emphasis added. NRC also claims that "the inspector agrees and determined that the stated activity did not detract from the overall recovery effort, nor did it diminish other high priority recovery action in progress or planned, and that TSC judgments were made with long-term recovery in mind." Id., emphasis added. B. With respect to the claims by both the licensee and the NRC that EFW might be required in the long term, I believe that they are invalid. In a large break LOCA, there is little if any potential for usefully employing an EFW pump, in either the short term or the longer term recovery phase. In a large break LOCA, the emergency core cooling systems can refill the reactor vessel with additional water spilling out the large break in the reactor coolant system piping. Therefore, no reactor coolant can be circulated through the steam generator tubes and emergency feedwater would be of no use in removing heat from the reactor coolant or the reactor core. Steam generator cooldown is not required during long term recovery from a larga break LOCA and the new NRC inspector gives no explanation for his contrary - _.

belief. The steam generators would slowly cool down on their own' I by heat loss through the insulation. While efforts to restore i the EFW pumps would be required in the very long term, i.e., during the months prior to resuming operation, such efforts are of no utility during the time period following a large break LOCA which was covered by the exercise. 9. The new NRC inspection report does not indicate how the inspector " determined" that the irrelevant EFW pump activities did not detract from the overall recovery effort. In fact, as explained below, the licenses failed to make sufficient efforts to locate and isolate the release path, an important part of the recovery effort. 10. In sum, neither the NRC's revised inspection report nor Mr. Kline's affidavit explains or mitigates the conc 2.usion in the first NRC report that the continued efforts to restore the EFW pump were an example of questionable engineering judgment and/or the failure to recognize and address technical concerns. Containment Building Spray (CBS) 11. Another example of deficient technical judgment was the " questionable fix" for the Containment Building Spray (CBS) system. NRC Inspection Report 50-443/88-09 at page 5. 12. Gregg Sessler, the individual who coordinated the development of what the NRC originally called a " questionable fix," attempts to explain the NRC's conclusion as the result of "the constraints of the exercise on communicating with observers Affidavit of Gragg F. Sessler, page 5. He also claims,

understandably, that the flow path he developed was " technically sound" and that, in any event, it was a " contingency. plan" which would not have been implemented without NRC approval. Id., pages 5-6. 13. The licensee apparently convinced the new NRC inspector of the soundness of Mr. Sessler's judgment because the second NRC inspection report accepted the alternate CBS flowpath as a "last resort" measure. NRC Inspection Report 50-443/88-10, page 9. 14. As an initial matter, neither the original NRC inspection team nor the latest NRC inspector nor Mr. Sessler himsolf give any details about the equipment and flowpath used for this "last resort." Without additional information, no . assessment of its validity can be made. Moreover, the NRC's belated acceptance of the contingency plan as a "last resort" is unavoidably inconsistent with its first assessment. The NRC's first inspection team surely must have known that the contingency plan was.a last resort and yet they concluded nevertheless that it was " questionable." Finally, if the initia'l NRC conclusion is accurate, the safety issue is not resolved by dependence on the NRC ultimately to prevent the employment of a contingency plan 4 i founded on poor engineering judgment. Location and Isolation of Release Path L 15. Another example noted in the first NRC inspection report was the " lack of effort to locate and isolate the release path." NRC Inspection Report 50-443/88-09, page 5. l 16. Gary Kline, another member of the TSC who participated in the exercise, c1cims that "a concerted and planned effort was made to locate and isolate the source of the containment bypass leakage (i.e., the release path)," and that "the source of the. leak was initially isolated to the containment enclosure ventilation area. " Affidavit of Gary J. Kline, pages 5-6. He goes on to say that efforts were made to further isolate the release and that entry to the electrical penetration area was not made due to high radiation levels. Id., page 6. 17. The licensee told the new NRC inspector of these efforts leading him or her to conclude that the NRC's original seven-member inspection team "was unaware of these activities during the drill." NRC Inspection Report 50-443/80-10, page 9. The NRC further concludes that the postponement of entry into the containment enclosure was justified by the imminence of the restoration of the CBS pump. Id. 18. These explanations are not persuasive for several reasons. First, it is a significant overstatement to claim that the release path was " isolated" to the containment enclosure ventilation area because this area includes "the electrical penetration area, the enclosure building annulus, the enclosure area ventilation room, the mechanical penetration area, the equipment vaults and the charging pump cubicles." Affidavit of Gary J. Kline, page 6. 19. Second, the fastest and most effective way of terminating a release is to isolate the leakage path. l Contrary 1 to the suggestion of the NRC in its second report, restoration of 1 _ _ - _ _ _ _ - _ _ -

[ l the CBS pump could not have " stopped the release." NRC ( Inspection Report 50-443/88-10, page 9. The CBS system, by reducing containment pressure and washing radioactive material from the containment atmosphere, would reduce the rate of t radioactive material release, but not terminate it.

Thus, locating and isolating the release path should have received a higher priority than attempting to restore the CBS pump.

In any event, efforts to locate and isolate the release path were particularly important given the repeated delay (through controller intervention) in CBS pump restoration. 20. In addition, although Mr. Kline specifically cites high radiation levels outside only one area -- the door to the electrical penetration area -- he nevertheless claims that a review of the radiation data "for these areas" justified postponing entering other areas within the containment enclosure ventilation area. Affidavit of Gary J. Kline, page 6, emphasis added. However, the licensee does not provide sufficient information to judge whether its decision not to expend further efforts to locate and isolate the release was acceptable. 21. Finally, no explanation has been offered for why the original NRC inspection team present during the exercise noticed none of the claimed efforts to locate and isolate the release path. Steam Generator Blowdown 22. Another of NRC's examples of questionable engineering judgment was the lack of effort "to blowdown Steam Generators to ____________ -

ll p i lessen the heat load in containment." NRC Inspection Report 50-443/88-09, page 5. 23. Mr. Sessler states that the control room operators and TSC staff recognized that the Emergency Operating Procedures ' called.for blowdown of the steam generators. Affidavit of Gregg F.' Sessler, pages 6-7. He also acknowledges that these procedures were not followed and characterizes that failure as a temporary postponement to allow an assessment of the level of radioactive material in the steam generators. Finally, he offers a post-exercise rationale for not blowing down the steam generators that the rate of heat transfer between the containment atmosphere and tho' insulated steam generators was insignificant and "the potential reduction did not warrant immediate depressurization without further evaluation of potential radiological consequences. Id., pages 7-8. 24. .The new NRC inspector echoes the lic6nsee's claim that the steam generators were not blown down "because the TSC staff was unsure of the integrity of the S/G tubes because no sample was available due to blowdown system isolation." NRC Inspection Report 50-443/88-10, pages 9-10. The "NRC position" is that although " improved guidance to the operator may be warranted," the decision not to blowdown the steam generators " appears to have been reasonable and appropriate." Id., page 10. 25. These explanations are technically invalid and may indicate a fundamental lack of understanding of the behavior of the Seabrook Station during a large braak LOCA. The only possible way to have a level of radioactive material on the __.. _ _ _ _ _ -

secondary side of the steam generators that would justify postponing blowdown for fear.of the radiological consequences would be a significant primary to secondary leakage path through the steam generator tubes. However, if such a leakage path existed during a large break LOCA (the exercise scenario), the leakage flow would not be from the reactor coolant system to the seccndary side of the steam generators..Rather, the flow would be from the secondary side of the steam generators through the tubo leakage paths into the. primary system and out the large break. Thus, although tubes leaks could raise concerns about the radiological consequences of steam generator blowdown during some accidents, tube leaks do not raise such concerns during a large break LOCA. 26. Furthermore, assuming that the licensee's staff had genuine reasons to be concerned about the integrity of the steam generator tubes, blowdown was all the more important in order to reduce the containment heat load. Rather than being limited to the.small heat transfer through the steam generators' insulation to the containment atmosphere, the containment heat load would be substantially increased by the hot secondary water leaking through the steam generator tubes and out the large primary system break directly into the containment. l Radioactive Release Rate

27.. The fifth example of questionable engineering judgment and/or failure to recognize and address technical concerns is the failure of both the EOF and TSC staff to question "a release of -_

l -t l greater than 7000 curies per second with only clad damage and no core unenvery." NRC Inspection Report 50-443/88-09, page 5. 28. James MacDonald claims that the TSC staff did indeed question and discuss the lack of correlation between the release condition and core cooling indications. Affidavit of James A. MacDonald, page 3. 29. The second NRC report likewise says that the logs in fact revealed that several of the licensee's staff members questioned and/or commented on the mismatch. NRC Inspection Report 50-443/88-10, page 10. 30. There is no explanation given for why the first NRC inspection. team, who observed the drill and presumably had access to the exercise logs, reached a conclusion diametrically opposed to the second NRC inspector. In any event, I agree with the second NRC report that more effort should be made in developing an exercise scenario where the postulated core damage and release rates are consistent.- Id. However, the actual scenario used in the June 28-29, 1988, exercise postulated conditions which were mutually exclusive on technical grounds. Thus, it is of questionable validity to use the results of that exercise to determine whether the. licensee's staff ' demonstrate [d] the ability to analyze station conditions, parameter trends and develop potential solutions for placing the unit in a safe, l 1 stable condition." 1988 FEMA /NRC Graded Exercise at 2.2-2. l t

Conclusion 31. The affidavits of Messrs. Kline, MacDonald and Sessler and the NRC Inspection Report 50-443/88-10 are not sufficient to resolve the " weaknesses" identified in NRC Inspection Report 50-443/88-09. Furthermore, they do not alter my original conclusion that the exercise objective was not met because onsite emergency response staff did not demonstrate an " ability to analyze station conditions, parameter trends and develop potential solutions for placing the unit in a safe, stable condition." In my view, these matters irsvolve significant safety issues which have not been resolved. Signed under the pains and penalties of perjury this seventh day of November 1988. / District of Columbia City of Washinston as Sworn and subscribed,before me this 8th day of November, 1988, 4t ashington, D.C. '\\ HARI SINGH t 5 [ N Notary Public g, cguuna apes March 14.15 My commission expires _ _ _ _ _ _ _ - _ - _ _ _ -

3,_- L[' EXHIBIT K I l i E____-.-}}