ML20052A392

Suppl to Petition to Intervene,Listing Conditions to Be Litigated.Certificate of Svc Encl

ML20052A392
Person / Time
Site:	Seabrook
Issue date:	04/21/1982
From:	Bernabei L, Curran D, Jordan W ECINECNP, HARMON & WEISS, NEW ENGLAND COALITION ON NUCLEAR POLLUTION
To:	Atomic Safety and Licensing Board Panel
References
NUDOCS 8204280293
Download: ML20052A392 (79)

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2 2 RECEIVED 73 '82 FPR 23 P12:12 APR 27 I -

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1. 4 UNITED STATES OF AMERICA .

9, NUCLEAR REGULATORY COMMISSION

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0) FORE THE ATOMIC SAFETY AND LICENSING BOARD

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In the Matter of )

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PUBLIC SERVICE COMPANY OF NEW ) Docket Nos. 50-443 HAMPSHIRE, et al., ) 50-444

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(Seabrook Station, Units 1 and 2) )

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NECNP SUPPLEMENTAL PETITION FOR LEAVE TO INTERVENE Pursuant to 10 CFR 2.714 (b) , the New England Coalition on Nuclear Pollution submits the following contentions for litigation in this proceeding. They are based on information available to date in the FSAR and the files of the NRC Public Document Room. To a degree, their development depended upon the extent to which the Applicant submitted an FSAR that was coherent and could be followed clearly. We discovered instances in which issues that should have been discussed in one section were covered in others and various other instances in which the FSAR was difficult to follow. We have attem.pted to ferret out as Fany of these hidden points as possible.

We will continue to do so and will file additional contentions where warranted.

l S%}s 8204280x13 __ _

' b' As a clarifying point, we have relied in several instances'on Regulatory Guides as the. basis for contentions.

Our point is not that Regulatory Guides constitute NRC re-guirements, but that the' Regulatory Guides themselves constitute factual bases for our contentions. In part-icular, they provide a benchmark against whi'ch the Board-may judge compliance with the regulations.

Respectfully submitted, dA}fht</3 %~

William S. ,qofdan, III

[. 2t -n 9J ' An n 9A Diane Curran

/f [ L~(' ~ C- t c K L9pne Bernabei

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Dated: April 21, 1982 1]ARMON & WEISS-L725 I Street , N.W.

Suite 506 Washington, D.C. 20006

-(202) 833-9070 Counsel for the New England .

Coalition on Nuclear Pollution

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Contentions Indbx

-PAGE I. Technical Safety Contentions . . . . . . . . . . . . .

, 5 A. EnvironmentalQualification-f:lectrical. Equipment . 5

1. . . . . . . . . . . . . . . . . . . . . . . . . 5

2. . . . . . . .. . . . . . . . . . .- . . . . . . 8

3. . . . . . . . . . . . . . . . . . .-. . . . . . 9 B. Environmental Qualification-Mechanical Equipment. . 10

1. . . . . . . . . . . . . . . . . . . . . . . . .- 10

2. . . . . . . . . . . . . . . . . . . . . . . . . 12' C. Environmental Qualification Emergency Feedwater Pump House'HVAC. .. . . . . . . . . . . . . . . 13 D. Testing of Equipment I

1. . . . . . . . . . . . . . . . . . . . . . . . .

2. . . . . . . . . . . . . . . . .- . . .. . . . .

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3. 7

4. 18 E. Reactor Coolant Pump Flywater Integrity . . . . . .

19 F. Diesel Generator Qualification. . .. . . . . .- . .. ._ 21 G. Pressure Instrument Reliability . . . . .. . . - . . . 22 H. Decay Heat Approval . . . .. . . . . . . . . . .- . 23-f I. Inadequate Provisions for Achieving Cold Shutdown. 25 J. Sabotage .. . . . . . . . . . . . . . . . . . . . 27 K. Instrumentation for Monitoring Accidents. . . . . .- 30 L. -PORV Flow Detection Monitoring System Inadequate. . 31 M. Fire Protection . . . . . . . . . . ._ . . . . . . . 32

{ N. Solid Waste Disposal System . . . . . . . . . . . . 33

0. . Emergency Feedwater . . . . . . . . . . . . . . . .- 34

1. * - - . . . . . . . . . . . . . . '34 a

2. - - . . . . . . . . . . . . . . . . . . . . . . 36-s P. Human Engineering . . . . . . . . . . . . . . . . . 37 Q. Systems Interaction . . . . . . . . . . . . . . . . 38-

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PAGE R. Ilydrogen Control System , ,,,,,,,,,,,,,,

42 S. Loose Parts Detection System ,,,,,,,,,,,,

45 T. Steam Generators , , ,,,,,,,,,,,,,,,,

47 U. Turbine Missiles , , ,,,,,,,,,,,,,,,,

49 V. In-Service Inspection of Steam Gt nr:rator Tubes , , ,

51 W. Seismic Qualification of Electrici Equipment 53 II. Quality Assurance Contentions ,,,, , , , , , , , , ,

55 A. Design and Construction 55

1. 55 57 11 Operations , , , , , ,,,,, ,,,,,,,,,,

62 III. Emergency Planning Contention , ,, , , , , , , , , ,

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• t I. TECllNICAL SAFETY CONTENTIONS A. Environmental Qualification -  ;

';ical Equipment

1. Environmental qualificati' m the requirement that all safety systems must be able to F. atinn in the accident environment, is " fundamental tc ;i> U tegulation of nuclear power reactors." CLI-78-6, 7 NPC inn, 412 (1978).

NECNP contends that the Seabrook facilit' .mnot be licensed because it does not meet the Commission',<+ ndards for environmental qualification of electrica! 'inipment under 10 CFR Part 50, Appendix A, General Design  :

crion (GDC) 4.

Furthermore, in light of the Three Mile intand accident, GDC 4 requires more rigorous environmental qu 11 i fi cation testing than was previously the case in order to provide reasonable assurance that electrical equipment will function for the entire time period-in which it is needed. The FSAR's discussion of environmental qualification is deficient in four respects:

1) the parameters of the relevant accident environment have not been identified; 2) the length of time the equipment must operate in the accident environment has not been included as a factor; 3) the methods used to qualify the equipment are not adequate to give reasonable assurance that the equipment will remain operable; and 4) the effects of aging and cumulative radiation exposure on the equipment have not been adequately considered.

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6 T Basis:

The Applicant's.FSAR states that the Applicant has complied with. Regulatory Guide 1.89 in qualifying electrical, instrumentatien and control equipment. FSAR at l'. 8-33.

' Regulatory Guide 1.89, however, is not the applicable standard for environmental qualification. The Commission has set DOR Guidelines and NUREG-0588, which are more detailed and specific'than Reg. Guide 1.89, as-the standard for compliance with GDC 4. CLI-80-21, 11 NRC 521 (1980).

The applicant must show compliance with CLI-80-21 in order to obtain an operating license.

A rule has recently been proposed which would extend the compliance deadline from. June 30, 1982 (as established in chi-80-21) to the second refueling outage after that date. 47 Ped. Reg. 2876 (January 20, 1982). However, even under this proposed extension, applicants for operating licenses would be required to submit an analysis ensuring.

that the plant can be safely operated pending environmental qualification. PR 10 CFR 50.49(k). The applicant has offered no such analysis or assurance that it will be able to meet the current environmental qualification standard.

Furthermore, the accident at Three Mile-Island, in which theoretically qualified equipment did not function for the time period in which it was needed, showed that the Commission's standards at that time were ingdequate to t

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. 3 provide a reasonable assurance that plants may be operated safely.*/ Although CLI-80-2L was issued after the accident, the Commission stated specifically that it had not attempted to incorporate the lessons of TMI into the decision. In light of the TMI-2 experience, to provide a reasonable assurance that the Seabrook plan can operate safely, the Applicant must show that safety-related equipment can withstand accident conditions for substantially longer than the matter of minutes currently required. The Applicant's environmental qualificiation information is also inadequate to support any finding that environmental qualification is complete because it has completely omitted data on the duration for which it is qualified. It should, therefore, be rejected as insufficient support for the granting of a license. NECNP reserves the right to amend itn environmental qualification contention when and if thin information in submitted.

./ Memorandum from S. II . 11anauer, Ass i stc.nt Director for Plant Systems, DSS, re: EnvironmentaI oualification and Instrumentation (April 6, 1979).

i s I. A.2. The Applican t has not complied wi t h Commission standaras a i ij.u ili tig <]ts.il i l i e.il e ott t e.:l : of e lis lr ti- v.s l vi o[ >t i .i t o r . . an-stalltd inside the containment..

Basis :

The PSAR indicates environmental qualit'ication of these components is in compliance with IEEE Stanaards 382-1972 and 323-1974. FSAR at 1.8-28, 29 Ilowever , the Applicant must, at a minimum, comply with DOR quidelines .uid NilRl:G -05 88 as required by the Commission in CLI 21 and with the more r igo rous environmental qual i i i cal ion t e :t. u1 ihat the TMI accid en t demonstrated is necessary to a;sur. saiety.

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i i 3 1. A. 3. The Applicant has not c.s: pt : eu wit h GDC 4 in that it h as not environmentally qual i t ied elect s i ca equipment inside the cont ainment io wit hat and ihe e fi ec t - ot a hyd rogen release such as occurred at Three Miie Island thii t. .

Basis :

The accident at Three Mile Island: ih wed that the accident environn'nt which must be considered i n det e rmin ing whether equipment is environmentally qualified:ine'.udes the generation of hydrogen g as . At Three M i le Island, h yt t r og en apparently burned or o therwise disabled elect rical equ ip r o n t inside the containment and thus prevented it from funetioning ulusing the accidenL. 11yd r ug en burn must therefore be considered a par t of t. h e accident environment for purposes of environmental qualification of electrical equipment inside the containment. The Applicant h as failed to do so here.

a w I. B. Environmental Qual i t ical i on - 51< cli . al i c a l Equipment

1. The Applicant has not mit is1 t < f i t u. requirement of GDC 4 that all equ ipmen t important to . . : f e t.y be environmen-tally qualified to survive .uid function in the accident environ-menL, In pariicular, the Applicant do " not classity as saioty-grade all systems t h a t. m ay be required ta remove heat from the steam generators during an accident: e .g . , steam dump valves, turbine valves, and the steam durap control system. This omission also violates GDC 3, which requires the Applicant to establish a reliable decay heat removal s yn t.em .

Basis :

The Three tiile Island accident demonst rated that sys t.em s that have not. previously been clahsified as " trapo r tan t to saf et.y "

or subjected to environmental qualification requirement.s may be called upon in accident conditions to mit ig at.e the effect of the accident. NUR EG -O'2 7 H at 18. There, non-saluty grade equipment was used to convey heat away from the core. The Three Mlle Island acc id e n t prorapt ed a rethinking within the agency of the scope of the environmental qualification " envelope . "*/ The Offico of Nuclear Reac t or Regulat ion has found that "one o f the most im po r t.an t factors in the safety of nuclear reactors is the reliability of the systems used for decay heat removal following the shutdown of the reactor for any reason . " NU R EG -07 05 at A-1. The Three Mile Island accident confirmed that " loss of capabi .1 i t y to remove heat th roug h

• / Memorandum from S. 11 . llan a ver , Ass i s t.an t Director tor planL Systems, DDS, re: Environmental Quali f ic ation and Ins t r ume n ta t i on ( Apr il 6, 1979).

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. b the st eam generat or is a significant contributor to the probabtljty of a core . mel tdown . " M. at A-2. The N14C ; tait asserlod that

" alt ernat. i ve mean:, o f dec ay heat removal could :ubs tan t ial ly increase the plants' capabi1ity to deal with a broader spect. rum of t ransi en t :, and accident , and, f liere fo re , could potent ially significanLly reduce the overalI risk to t he publi c . " M. tJulens those components which may be called upon to remove residual heat during an accident are environmentally qualified, they cannot be relied'upon as parL of a reliable system for removing decay heat trom the core. The Three Mile luland accident has proved that st.eam generator heat- removal equipment for normal operations m us t. be cons i dered sa f e ty-g r ade .

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m a r I. 13 , 2. The Applicant has not satistned t. h e requirement.

of GDC 4 t hat. all e qu i pme n L i mpo r i an i Io ,aiety be environnentalJy qualified because it has not specilied the time durat ion over which Ihe equ i p me ri t i t. qualiiied.

11asi s :

Equip men t i mpo r t an t to safety must be abic to withstand the accident environment for the period in which it i s, required to function. This was made clear by the Three Mile Island accid en t , during which accident conditions pi:rs i s t ed for an un-expectedly leng thy period . The Appl ic an t h an n o t. speci fied in its l'S A R what l e nq t h o f t. ime the tal e t y-s ' 1 i t . ri mechan i ciil equ i p me n t is qualitied to operate. It inuoI L' required t o do so before mec h an ic al equipment can be accept " t .n. environmentally qualifled. Ni:CNI' reserves the r iqht to c lui l ! "ng e the adequacy of the dur.it ion.il parameler* i l' and wil< n ~h4- liit or Iiat ion i:.

submitLed to the NRC by Lhe ApplieanL.

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i I. C. Environmental Qualification-En;crgency reedwater Pump House IIVAC l

According to Table 1.3 -2 , sheet 14, of the FSAR, the Appl ican t I

has added a new heating, ventilating, pnd air conditioning (HVAC)

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system for the emergency feedwater pump house . Only parts of the ilVAC system are considered safety related and environmentally qu ali fi e d . Table 3.l(B)-1, Sheet 4. NECNP contends that the entire system and'its function must beienvironmentally qualified, and that the environmental qualification must t ak e into account the likely duration of an accident during which the !!V A C system would be relied upon.

Basis :

Since the emergency feedwater pump house and its equipnent are capable of functioning and can be relied upon to function only within a particular temperature r ang e , the llVAC system is required to maintain conditions within that range. Acco rd ing ly , it must be environmentally qualified to assure that it will be able to function when needed . The environmental qualification must td< e into account the f act that the equipment may be required to operate for a considerable length of time in the event of an acc id en t .

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1. D. Testing of Equ i p men t

1. The Applicant has not compiied wi t h GI)C 1 as implemented by Regulatory Guide 1.150, requiring ultrasonic testing of reactor vessel welds during preservice and inservice examinations.

Basi s :

The Appl ic an t has stated that it does not intend to comply with all the terms of Regulatory Guide 1.150, yet it does not ind icat e any alternalive w ay:, i ti which l. jp - l equ i rt men t.n of the Rcg. Guide will be sat is f ied . The Applir ut 's statement that it aqroes with the "in t en t i ons " of R eq . G 2i- . ..t50 does not co n u t. i t u t e sufficient compliance with the Reg . G u i 1 ince the specific implementation plan for Reg. Gutde ].IV t). not been submitted, NECNP reserves the r14hl to amend thi: '.n 'ntton to challenge the sufficiency of its provisions.

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. o I.D. 2. The A pplican t h as not coni:l i ert with CDC 21 as implemented by Regulator y Guide 1.27, recluiring periodic testing of protection system actuation functions. In par ticular , there are twelve safety functions which the applicant does not plan to tes t at power:

1. Generation of a reactor trip by tripping the main coolan t pump breakers .

2. Generation of a reactor trip by tripping the turbine.

3. Generation of a reactor trip by use of the manual trip switch.

4. Generation of a reactor trip by manually actuating the safety injection system.

5. Generation of safety injection signal by use of the manual safety injection switch.

6. Generation of containment spray signal by use of the manual spray actuation switch.

7. Full closure of main steam isolation valves.

8. Full closure of feedwater isolation valves.

9. Full closure o f f eedwater control valvet, .

10. Main feedwater pump trip.

11. Closure of reac tor coolant pump component cooling water isolation valves.

12. Closure of reactor coolant pump seal water return valves.

FSAR at 1.8-9. The design and operation of the facility should be changed to provide for testing at power in all of these cases.

Basis :

GDC 21, as implemented by Regulatory Guide 1.22, requires that actuation of safety functions be tested at power. Otherwise suf-ficient assur ance canno t be provided that it will be able to func-tion while the reactor is operating . This is a fundamental require-n

men t that cannot be waived by an unsupported assertion that the probability of f ailure at power in too' low. The design of the Seabrcok facility should be revised, if necessary, to allow testing at power for these necessary safety system actuations.

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I.D. 3. The Applican t has not provided a reasonable assurance that the leakage detection system for the Seabrook reactor will operate when needed because not all of time system is to be testea during plan t operation, as required by GDC 21, and as inplemented by Reg . Guide 1.22. Only the airborne radioactivity detector h as the capacity to be tested during power operation . FSAR at 1.8-17.

The Applican t thereby also fails to satisfy GDC 30, which requires the development of adequate leak age detection systems.

Basis :

GDC 21, the Staf f 's Standard Review Plan and IEEE Standard 279 -71 all require that safety systens be capable of being testcd at power. Otherwise, there c an be no reasonable assurance that the equipment will operate under the cond i t. ions in which it wi]l be called upon to function . The Appl i c an t han failed to justify its failure to comply with GDC 21 e r to demonstrate that it will provide protections equivalent to those provided by compliance wit h Reg . Guide 1.22.

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• b I .D . 4. The Applicant has not complieu with the terms of Regulatori Guide 1.118, Rev. 2, requiring periodic testing of electric power and protection systems. In particular, the Applican t. indicat.es compliance with an outdated standard, IEEE 338-1975, which has been superseded by IEEE 338-1977. Furthermore, the Appl ic an t improperly asserts that it need not comply with IEEE 338-1975 whenever the standard states that an action "should "

be taken or a requirement "should " be met . PSAR at 1.8-41. All the provisions of the IEEE standard should be treated as mandatory unless the applic an t c an show an alternative means of satisfying them.

Basis :

Regulatory Guide 1.118 Rev. 2 incorporates IEEE 388-19 77 as the standard to be applied in determining the requirements for com-pliance with the Rcqulatory Guide. The applicant may not sub-stitute another standard of its own choosing without demonstrating that it h as provided equivalent protection. It has made no such showing.

The f ac. t that IEEE asserts that act ion: or requirements "should "

be met constitutes a factual basis for N!:CNP s contention that they must be met in order to assure the saf ht v of the seabrook facility.

At the very least, the applicant shoulfi show that it has satisfieo the requirement by some other means 11 it does not met the IEEE standard itself. By stating that its conpl iance with these pro-visions is discretionary, the Applicant w i t holds any commi tment to con tinue to comply with them in the! fnture. Thus, it.does not i

provide the requisite reasonable assurance t hat the plant c an be i

operated safely.

• - * 'I. E. Reactor Coolant Pump Flywcel Integrity The Applicant has not complied with GDC 4 as implemented by Reg . Guide 1.14 in that it has not met all the requirements of Reg.

Guide 1.14 or provided for suitable alternative means of satisfying the requirements. In particular, the Applicant should be required to perform post-spin inspections of thofflywheel, should identify the design speed of the flywheel and test it at 125% of that speed, and should specify the cross-rolling ratio. Furthermore, the flywheel should be environmentally qualified under GDC 4 because iticonstitutes equipment.important to safety.

Basis :

GDC 4 requires that equipment important to safety be pro-tected from missiles. In addition, it requires that equipment important to safety be able to function when called upon to' mitigate the effects of an accident. The flywheel is both a potential source of damaging missiles, and a component important to safety-because it provides inertia to ensure a slow decrease in ' coolant flow in order to prevent fuel damage as a result of a loss of power to the pump-motors. According to Reg . Guide 1.14, Overspeed of the pump rotary assembly during a transient increases both the potential for' failure and the-kinetic' energy of the flywheel. The safety consequences could be significant because of possible damage to the reactor coolan t system , ' the containmen t or o ther equipment or systems important to safety.

The Applicant 's FSAR is therefore inadequato because the flywheel has not been environmentally qualifiect; and because not all of the requirements of Reg . Guide 1.14 have een met or a reasonable alternative implemented. Under R eg . Guidy 1.14, the Applicant I

. o should be required to iden tif y the design speed.of the Seabrook flywheel, which should be at least 125% of normal speed. Appl ic an t should be required to test the flywheel at that speed, and to per fo rm post-spin inspections, as required by Reg. Guide 1.14.

Applican t 's assertion that calculations demonstrate a suf ficiently low probability of flaw growth is not an adequate justification for waiving the post-spin inspection. FSAK at 1.8-6. Finally, as required by Reg . Guide 1.14, flywheel plate material should be cross-rolled at a ratio of 1 to 3. 'r he Appl ican t cannot waive this requirement by relying on material selection, particularly in lig ht of the inadequacies in vendor surveilance that have occurrea at Seabrook .

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. . I. F. Diesel Generator Qualification According to the FSAR discussion of R(gul atory Guide 1. 9 at page 1.8-4, the " Load Capabili ty Quali fication " test for the Seabrook diesel generators was per formed according to the re-quirements of IEEE 387-1977. NECNP contends that the diesel generators c an no t be co nsid ered to be qualified for use at Seabrook on that basis. They must also, at a minimum, meet the requirements of IEEE 323-1974.

Basis :

The basis for this contention is t he MI: ' Staff p,sition set out in Regulatory Guide 1.9, which provirles t hat the qualification testing requirements of IEEE 323-1974 should be used in section i

5.4 of IEEE 387-1977. Based on the FSAR, the Applicant has failed to do so, and it has f ailed to demonstf at e that it has provided pro tections equivalent to those provid d by Rog . Guide 1. 9 .

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I.. G. -Pressure' Instrument Reliability The Reactor Cooling System wide range pressure instruments for the Seabrook f acility cannot be relied upon for accurate in-formation , and thus may lead to inapproprinte operator-actions jeopardizing the cooling of the reacto'r . Iflp pressure-instruments therefore do not' provide the requisite reasonable assurance of safe 7

operation of the plan t. ,

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Basis -

According to IE Information Notice No'. .R2-ll ( April 9, 1982),

qualification tests on Westinghouse-manuf actured RCS pressure instruments have shown " ambiguities in their accuracy which could 1

result in inappropriate operator actiopn . " In particular, West-inghouse told the NRC that :

post-acciden t accuracy ambiguitieh for RCS wide range pressure instruments under certain plant accident conditions have the potential for maximum ack:umulated inaccuracies of-

+ 390 psig indication . According the: inaccuracy of RCS wide range pressure measuremelLy',

nts could lead to pres-surizer power operated relief -valVen being lifted prior to theterminationofsafety_injectihn (SI) and to a greater number of valve challenges , therchy increasing the pro-bability of a small break loss-of;-coolan t accident due to a valve failing to close. Likew ine , the inaccuracy ofLthe wida range pressure instruments could lead to the termina-tion of SI without adequate' reactor coolant subcooling. In add iti on , the inaccuracies could-lead to premature or late tripping of the RCS pumps during the course of. a small break loss-o f -coolan t accident .

Because the pressure instruments may provide inaccurate information leading to the exacerbation or f ailure to correct accident condi-tions, their use constitutes a threat to the public health and saiety and cannot support a license for the Seabrook reactor.

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I. 11 . Decay lleat Removal l l One of the lessons of the Three M!ile I sl and accident was that heat exchanger capacity in nuclea r power plan ts should be expanded and improved. This is particularly true with respect l

to the unexpectedly lengthy period it Joh to cool the TM1 reactor and the need to assure effective heat hxchange at high pressures .

I However, the Applicant 's proposed hearl exchanger capacity is even l-i lower than a number of older plants ofl comparable design . FSAR i

Table 1.3 -1, Sheet 2. The Applic an t should be required to install additional heat exchanger capacity to allow for more rapid cooldown of the facility in the event of an accident.

Basis :

The current inadequacy of decay heat removal systems has been noted as an unresolved safety issue :

Even though .it may generally be. considered safe to main tain a reactor in a hot standby condition for a long time, experience shows that there have been events that required eventual cooldown and long-term cooling until the reactor coolant system was cool enough to perform inspection and repairs . For this reason the ability to transfer heat from the . reactor to the environment after a shutdown is an important safety function for both PWRs and BWRs.

NUREG -0510 , Identification of Unresolved Safety Issues Relating to Nuclear Power P lants (January 1979 ) at A -15 . The issue was further recognized by the formal addition of Task A-45 to the list of unresolved safety issues as a result of the TMI accident.

NUREG -0705 at A -1 . The Reactor Safety Research Review Group found in 1981 that :

A major ef fort should be undertaken to develop and evaluate improved or alternate approaches to more reliable shutdown heat removal systems, for both the reactor vessel and the con tainmen t .

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Report of the Reactor Safety Research ,Heview Group, Sep tera ber 1981, at 11 -3. The principle of expanded aiul imjiroved heat r eraov al i

capacity is violated by the reduction in .zu of the heat exchanger i

capacity for Seabrook.

1. 1. Inadequale P rov i s i otu, for Achi-vinq cold Shuf down Rcquiat ory Guide 1.139 es t abl i she., speci l' ic design require-ments that address the various system functions required to achieve and maintain a safe hot standby and cold shutdown position. Ac-cording to I&E Bulletin 79-01B, Supp. 3, October 24, 1980, it is the position 01 the NRC S t.a l l that "the licensee must identify and environmentally qualify the equ i p men t. needed to complete one method (path) of achieving and main tain ing a col o shu tdown cond i-tion . " NECNP con tend s that Seabrook does not. con fo rm t. o NRC re-quirements dnd constitutes a threat to the public health and safety bec au se , as described in the FSAR, par ticular ly at pag es 1.8-52-54, it is not capable o f being brought d i rec t. l y to cold shutdown in the event of an accident. Further, the Applicant has failed t. o demon st ra te that all systems, structures, and components necessary to bring the facility to cold shutdown have been environmentally quali fied or that they have met. all of the design criteria applicable to systems, structures, and components important to safety, including but not limited to G C 3,4,15,17,18,20-25,34,35,44.

Dasis :

The basis for this contention is the position of the NRC Staff as set out in ISE Bull. 79-OlB, Supp.3, and Ruj. Guide 1.139, and the Applicant's admission in the PSAR that it does not meet the R o;i . Guid e . The Applicant has provided only " systems with the capabil i t y to place and maint ain the plant in a sale ho t. stanoby condition, " such that a restoration of some dujree of systems capability would be required to bring the plant to cold shutdown in the event of an accident. The Appl ic an t has failed t.o demonstr ate n

that it has provided equivalent. protection. In addition, this con tentio n is based on the fact that the Applicant has failed I

to demonstrate compliance with the exceptions to lug . Guid e 1.139 discussed at pag e 1.H-53 oi the l>S Alt , the fact that t he Tf41 accident demonstrated the need to llave Ihe capabilit.y of achieving cold shutdown in the event of an accident, an d the need t.o a,sure the environmental quali fication of all syst ems, structures, co rnpo n en ts ,

and functions necessary to achieve cold s hu tdow n . To the extent that operator actions are relied upon to achieve cold shutdown, the function is not environmentally qualified and does not meet the applicable requiremen ts .

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I. J. Sabo t ag e 10 CFR Part 73, and particularly Sect i ons 75.40-73.% , require the Applicant to develop and impiement a plan that would e t Iect.1 vel y protcet Ihe Seabrook r cac t or:. aij a i n si i n hist r s a l . a bo t a< J e . losju l at ory Gu id e i.17, Pev. I, iusued in J une 197.1, e.; L a til I s he., L h'. requiroments and procedures that the NRC Stalf belteven would be suffictent.to comply with the regulations and provide the necessary protections.

NECNP con tends that the Seabrook reactors are seriousl y vu l ner abl e to industrial sabotage by virtue of their design and that the Applican t 's securit y pl an is inadequate to prevent actions of industrial sabo tage at Seabrook that would tbreaten the public health and s a f et y .

Basi s :

The Appl ic an t 's PSAR itself provide, a L71 ma facie basis for this contention when it states that the Smbrook security plan " meets the in ten t et Rujulatory Guide 1.l7, Rev. )." in this r eg ard ,

f NECNP has obtained t.hu assistance 01 Ro be r t "ollar d , lumer ly 01 the I

NRC Staff, now with the Union of Concernec Scientists, who has found in his many years of reviewing PEAR: 1. h a t a statement that an aspect of a nuclear pl an t meets the "in tent " of the Regulat ory Guide constitutes an admission that ths' plant cannot meet actual i

specifics of thc Reg . Guide . It is also 2nstructive to compare the PSAR language on this issue with otho" FSAR statemen ts in which the Applicant asserts that it is in tuli co mpl i ance wit h Regulatory Guides . While Rujulatory Guides do not consLitute firm require-ments, they do indicat e what the NRC Stati believes to be one man ner of compl y i ng with the roju lat i ona , and t hey pr ovlue a L,ania for j udg ing the adequacy of o ther met hods by comparing then to

.- . the Staff position. Failure to comply wi,th a R(gu l atory Guide in this case constitutes a factual basic for this contention.

In addition, the hist ory of industrial sabotage in recent years at the Surty, Midland, Beaver Val l ey , and Indian Point IJni t 2 reactors demohutrates that securi t.y pl an: I hat were presumably developed and approved as compiyinq wiih oi r cet . ng Lhe intent of Regulatory Guide 1.17 failed to providi' I he required protection.

It is not ponsEble to provide l ur' t lo i 1 n:i n or specificity

) since the Appli cant 's securit y plan han not n. en made availabe for review . Ilowever , Mr. Pollard i.as ponfirned on the basis of his expertise the vulnerbility of the Tabrook reactors to in-dustrial sabotage and the iact that prev.ous incidents of industrial sabo tag e demons tr ate that the same or wor.;e could occur at. Seabrook.

Mr. Pollard has considerable experienen ;n addressing the issue i

of industri al sabotag e both while he w ~.u on the NRC Staff and since he le f t . Ilis expertise in in reac tor th . ion as it r e l a t.es to so-botage, rather than in t. he technical a,tpoet s of security planning itsel f , alt hough h is expert i se is essential to an adequate plan, lie is prepared to examine the Seabrook design in connection with the security plan in order to identify the im pr ovemen ts that are required to assure safety. An example of an aspect of the design t h a t- is vulnerable to industrial sabot a le is any valve that is intended to remain open in the event of an accident, but that could be manually closed without detection prior to an accident, with the result that recovery from t.he accident would be hindered or prevented . This occurred at Beaver Valley. N1;CNP would also at t em pt to obtain the assistance of an expert in securi ty plans to evaluate whether the impr ovemen t s identified by Mr. Pollaro can be accomplished , and if so, how.

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f41:CNP and Mr . Poll ard are preparc<t in pr ov iele any at:,surances igainst disclosur e t h a t. may be r e qu i r ee t t o. 1,t o t e c t the i n teq r i t. y o f the security plan and the public heal t h . rut , , a f e t. y . We will also provide additional speci fity and bas i:, art er we have had an op-portunity to review the security plan.

I

• , 1. K. I nstr umen t.a t i o n for Monitorinh ?cc): fonts The Applic an t has not sa tisf ied Gi>C l 's , 19, and (2 4 , as im-plemen ted by Rcg . Guide 1.97. The Gen.ral I)esign Criteria and the Hrquiat ory Guide relat e to the innt rument it ion required to mon itor plan t conditions both during anci ca r t er an accid en t . The instrumentation should be environmentally rpiali f ied .

Basis:

The basis for this cont.cntion is that t.he Applicant has given no in fo rmat ion t o show sat. isf act ion o f the r equ i r ene n t.s . The Applican t inst ead indicates that it is still in the procebo of selecting t. he Post-Accident Monitoring inutrumentation for the facility. PSAR 1.8-36. NF.CNP reserves the r ig ht to amend this contention t.o challenge the adequacy of the PAM if and when it is submitted to the NRC by t.he Applicant.

4

1. L. PORV Plow Detection Monit orinti nv tom I nad equ at.e According to Table 1.3-2 of 1he l'. : A" , ibe Applicant h at, added a three channel acoustic acceler omet.cr sv n t om t o comply with the requirement to detect flow from PORV ' and safety valves. NECNP con tends t hat. this monitoring system i: inadequate to protect the publi c heal t h and safety and la11s to conply wit h t.he min uaum I requirements of the NRC. Further, whatever monit_oring system is provided must be sa f e t y-g r ad e and environmentally qualified.

Basis :

One of the major lessons learnnd from the accident at Three Mile Island is that the instrumentation and mon itoring systems used to measure the various conditions in the reactor during an I accident are both crucial to reactor saf et y and seriously inade-quate i n m an y r espec t s . In par Lieu 1ar , lhe NRC :st af f rtquire, the use o f a posi t ive, d irect i nd i cal i on o f valve positton rather than the indirect measurement previously used. This is based on both the TMI accident experience and on IEEE 279, which " requires that, to the extent f easibl e and practical, prot.ection system input shall be derived from signals that are direct measures of the desired var iable . " NUREG -05 78, p. A -9 .

Contrar y to these lessons and requirements, the Appl ic an t is relying upon an indirect measure of pro t.ec ti on syst em input by measuring noise rather than measuring t.he actual flow from the power oper ated relief val ves and the saf et:y val ves . Isafe reactor operation requires t.h a t the acoustic accelerometer system be replaced with a monit.oring syst em that directly measures the f]ows.

I. M. Fire Protection The Appl i c an t 's fire protection system does not satisfy GDC 3 as imp 1emen t ed by Reg . Guide 1.120 and the Counission 's dec ision in CL1 21. In par ticular , the Applicant indicates that it does not comply with all aspects of the Standard Review plan, Sect ion 9.5 -1, but does n o t. specify what those i tems are or any alternative means of satisfying tA requirements. Furthermore, the info rm ation submitted to the NRC in 1977 is outaated and should be revised and made available as par L of the PSAR .

Basi s :

The Commission 's decision i n CLI 21 regiaires that Ap-plicants whose construction permit. applicat. ions were docketed before J uly 1, 1976, demonstrate compliance with Appendix A to BTP 9.5 -1 and the requirements set forth in the proposed rule, which was I in al i v.ed at 45 P ed . R aj . 76602.

The Applicant i ndicat ed that it doen not. comply with all re-quirements of the Standard Review Plan at 9 .5 -] . PSAR at 1.8-43.

This in fo rmation , which was submit ted in 1977, is outdated and shoula be revised to reflect more recent developments, including chang es in SRP 9 .5 . The Applicant should be required to apecify which items are not complied with and to specify alternative means of satisfy-ing the requirements. NECNP reserves the r ig ht. Lo comment. on the sufficency of the Appl ic an t 's tire protection plan when thi.s information is made available.

- 0 1. N. Solid Waste Disposal Syst em The Applicant has not submitted a complete notid waste man ag emen t plan for radioactive waste a:, required by GDC 60 and implemen ted by Rcq . Guide 1.143.

Bas i s :

General Design Criterion 60 requires that the nuclear power unit design include means to control suitabl y the release of rad ioac tive mater ials i n g aseous and liquid ef fl uents and to handle radioactive solid waste produced during normal reactor operation, including an ticipated oper at tonal occurrencen. The Applicant should be required to compl o t .- i t t, solid wante m an ag em en t plan for radioactive waste in satisfaction of this requirement. NCCNP reserves the right t o challenge t he .;uitici.ncy of the plan it and when it is submitted in f inal form.

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I. O. Emergency Peedwater

1. The emergency feedwater sy s t e.a i t, inadequate in that a single failure in the common discharge heaoer, in conjunction with del ayed operat or act ion or no act. ion t o correct it, would result in a loss of feedwater to all the steam generators . The seabrook design must be revised to provide an emergency feedwater system that is single failure-proof with respect t.o a rupture of t. he high-energy piping in the discharge header. Even if the common discharge header is not considered to be covered by the single failure criterion, the Applicant has not adequately considered the factors necessary to protect against passive system fai]ure.

Itasi s :

The emergency feedwater syst.em desian for the Seabrook f acility provides one common discharge hea Jer for all the steam gen er at ors . This system is placed under :t t "u n and pre:.sure when the emergency feedwat er system is ac t. ) i a t ed . In the event.

of a r upture in the common discharge headcr, feedwater supply to all the generators would be jeopardizerl . Nah a r upture should meet i

the Single Failure Criterion of Appenoi: A. t o 10 CFR Part 50.

Even where systems are not speci f ft c a l] ', required to meet the i

Single Pailure Criterlon, the Applican! nui t consid er the possibilit.y o f a a ing le failure. Footuoto 2 to Aplmn:ii x A stales that "the i

conditions under which a single failure at c. passive component in a fluid system should be considered inide .D;ning the system ag ai n s t a single failure are under development . " tiowever , this s t a t.em e n t "does not relieve any appli can t fren considering these mat ter s in the design of a specit)c !.nctlity and sat.i s f y ing the necessar y requi remen ts . Thes6 at ters inclur e :

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i (i) Consideration of the nelid th>' design against single failures of passive ch)mpohOnts in fluid-systems important to safety.

( ii ) - Consideration of redun :lancy 'and diversity re-quirements for fluid systems important to safety.

(iii) Consideration of the jtype , size, . and orien-tation of possible breaks in; the ' components of the reactor coolant pressure bouhdary-in determining design requirements to suitably protect ag ainst postulated loss of coolant a'ccidents.

(iv)- consideration of the possibility of systematic, nonrandom, concurrent failurbn. of redundant elements in the design of the protect llon ' system and reactivity, control systems. l 36 Fed . Rog . 3 255 (Febr uary 20, 1971), preamble to 10 CFR P ar t ' 5 0 Appendix A.

At Seabrook, in the abse nce of prompt operator action to correct a loss of feedwater, all of th'e steam generators would be threatened by loss of coolant. Reliance on such operator action is unacceptable. To satisfy the Single Failure Criterion and the considerations listed in the' preamble to Appendix A, the Applicant should redesign the facility to provide redundant-feedwater capacity or institute automatic initiation of the emergency feedwater system.

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I .O .2. The emergency feedwater dy;; ten is inadequate in that a break in the common discharge headt r between Valve 65 and Valve 125 (see FSAR, Figure 6.8 -1 ) , coupled with a loss of offsite power, would result in a loss of feedwl.it et to all steam generators.

The Seabrook design must be revised to provide an emergency feed-water system that complies with GDC 17 wit h respect to the hig h-energy piping in the discharge header .

Basis :

The applican t has failed to meet its own criteria with regard to loss of power and single failure in high-pressure systems. In Section 6.8.3, the PSAR states: "The system has been designed to provide the required flow following a singic active failure coupled with a passive f ailure in the high or moderate energy piping and a loss of of f site power . " ilowe ver , the PSAR im-properly f ails to include the common discharge header in the class of "high or moderate energy piping " because it is "not pressurized during normal plant operation . " In analyzing the adequacy of safety functions, normal oper at ing conditions cannot be the basis for classification of the type of equipment which must be considered in the analysis, because normal operating conditions are irrelevant to how the equipment will behave under accident co nd i ti on s . The classification of the common discharge header as unpressurized is invalid because the header will be pressurized when it is called upon to function. The appl i can t therefore has no basis for claiming that the system is designed to provide needed feedwater flow in the event of a single active failure and a moderate or high energy piping failure coupled with a loss of offsite power.

. . ~ . .

I. P. Human Engin eering According to Table 1.3 -2, Sheet 6 of the FSAR, the Applicant h as added a 0-2300

• P multipoint recorder on the backf of the main con trol panel . Its purpose is to record temperature . at four core locations. NECNP contends that the location of this' recorder on the back of the control panel constitutes poor human engineering that would detract from the operator's ability to take prompt, correct actions in the event of an accident.

Basis :

This contention is based on the fact that information that may become of major interest to the operator will be available only on the back of the. control panel . The operator will be required to leave his station and divert his attention from on--

going events in order to determine the temperature in the core as stated on the multipoint recorder. The information should be readily available such that the operator need not raove to the back of the control panel.

I, Q. Systems Interaction The Applicant has not demonstrated that it has adequate-ly evaluated systems interaction to confirm that Seabrook has been designed against'all potential undesirable interactions between and among systems in order to meet the requirements of 10 CFR Part 50, Appendix A.

Basis:

One of the so-called " generic unresolved safety issues" is the interaction between non-safety and safety systems, which creates demands on the safety systems that exceed the latter's design basis. This problem is listed and described as A-17 in NUREG-0510.

The NRC, in investigating an accident at the Zion plant on July 12, 1977, discovered a design defect in the Westinghouse plant in that Westinghouse designs provide'for a large number-of, and different types of, interactions between control systems and safety systems. The Applicant has not demonstrated that these design deficiencies have been modified or remedied at Seabrook.

On October 12, 1979 the ACRS recommended with respect to the Indian Point plant and all light water reactors that the NRC study systems interactions by investigating, inter alia, sub-system failures within interconnected electrical i

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. . i or mechanical complexes and potential interactions between nonconnected systems. The latter investigation may require

. in-situ examination of the plant.

On March 9, 1982, the ACRS again recommended to the NRC chairman that a walk-through systems interaction study be developed for all light water reactors to detect obvious interactions.

The Standard Review Plan states that a systematic metho-dology is necessary for evaluation of systems interactions ,

Sandia Laboratories, under contract to the NRC,has stated l that the methodology used by the NRC staff is inadequate to identify all interactions important to safety. In one study of systems interaction for carthquakes at.Diablo Canyon, about 600 previously undetected systems interactions were i found. The Applicant was required to make plant modifications.

The Report of the Reactor Safety Research Review Group, i

issued September 1981, recommended intensified research to define better the role of plant control systems in light water reactors. The recommendation was prompted by severe transients initiated by control systems,in recent years in light water reactors such as Seabrook.

According to indications in the Applicant's letter to the Commission dated January 8, 1982, it has not satisfied the Staff's questions about the safety consequences of interactions i

between control systems and safety systcaa at Seabrook.

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. - The recent " state of the art" review, conducted by Brookhaven National Laboratory, Battelle Columbus Laboratories and Livermore Laboratories concluded that no single method currently exists to perform an adequate review of adverse systems in-teractions. NUREG/CR-1901, Review and Evaluation of System Interactions Methods, A.G. Buslick et al, Brookhaven National Laboratory, January, 1981; NUREG-1896, Review of Systems Interaction Methodologies, prepared by P. Cybulskis et al., Battelle Columbus Laboratories, January, 1981; NUREG/CR-I 1859, Systems Interaction, State-of-the-Art, Review and Methods Evaluation, J.J. Lim et al., Lawrence Livermore Laboratory, January 1981.

The Applicant indicates that it meets only the current Regulatory Guides, which the Commission itself believes to be inadequate to deal with the safety effects of systems interac-tion. Therefore the Applicant's analyses of systems inter-actions, contained in Chapter 15 of the jFSAll, are inadequate.

Furthermore the Appeal Board in Virginia Electric and Power Company (North Anna Nuclear Power! Station, Units 1 and 2), ALAB- 4 91, 8 NRC 24 5 (1978) rulchthatasarequirement for the issuance of an operating license the record must demonstrate that each applicable generic! nafety issue has been resolved for the particular reactor, or demonstrate the exis-tence of measures employed at the plantito compensate for the i

i

lack of a solution to the problem. The Applicant has failed to demonstrate that it has resolved this generic safety problem or that it has devised compensating measures.

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I. R. Ilydrogen Control System The design of the hydrogen control'systemiat Seabrook is inadequate to protect the public safety {in that it would protect against the hydrogen produced by a metal water reaction i

involving only 1.5% of the fuel cladding. 'FSAR 1.8-3.

i Basis: i i The accident at Three Mile Island emor strated that as I

much as 50 percent of the zirconium cladding in the TMI core reacted chemically to produce hydrogen,- an anount greatly in

'l excess of the design assumptions of 10 CPR 50.44. The NRC, I

in its Proposed Rule of December 18, 1981, " Interim Require-ments Related to Hydrogen Control," staked that two years I

after the effective date of the rule:all operators would be required "to perform and submit . . janalysestoassure thatduringadegradedcoreaccidentcoiftainmentstructural integrity will be maintained. . . ." The Commission a'ssumed I

that a degraded core accident would involve a 75% metal-water-

. reaction, with the resulting massive hydrogen release to the containment.

Furthermore, in the Commission's current Proposed Policy Statement related to Safety Goals for Nuclear Power Plants (Feb. 11, 1982) the Commission explicitly recognized the importance of mitigating the consequences of a core-melt accident through assuring the integrity of the containment in the event of a hydrogen release.

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The Commission in Metropoli tan l'di non Comp.iny (Three Mile Island Nuclear Station, Unit. No. 1) , C L 1 '3 0 - 16, 11 NRC 674 (1980), appeared to require t-hat an intervenor contesting t.he adequacy of a plant.'s hyd rogen control system hypothesize a credible accident scenario in which the hydrogen generated would cause release of off :it.o radiation levels in excess of 10 CFR Part ]00 1 ; m i L:.

and denied, in that case, the intervenor'n r t . p t > <: t .o waive the c1carly inadequate design ba s i t. conta'rtd in 10 CFM 50.44. Yet. in a moie recent Comm:suion inion, tb -

Comm us. ion i t s.' ) f recoqnized the ii ade m m , f '4.44 anr:

required an applicant to install and uto an 1 7 niter hydrogcn mitigation system t'o mitigate the pos: It']" < lease o f hydrogen in ( xcess of the assumpt ion? of '30. 4 4 , and to continue a research program on hydre o;- n control measures and the effects of hydrogen burns on s..fof.y functions.

Duke Power Company, (McGuire, Units 1 aml :9 CLI- 81-15, 14 NRC 1 (1981).

Another licensing board admitted .s limilar contention on the inadequacy of a plant's hydroget co'itrol system on the basis that " Commission utterances,' proposed and tenta-tive though they may be, [ appear] t o l'< - incon :Lat ent wi t h the TM1 deci--i' x which we relied. Tho U" i'ssion now appears to be of the view that the assipet i c o' o .r 50.44 are unrealistic and that some additioia) -

ray need to

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i I

i t

I be taken.". Cleveland Electric Illuminating Corp. (Perry-Nuclear Power Plant, Units 1and2's,Nof;.50-440, 50-441, i'

March 3, 1982, slip op, at 8.

The Applicant has stated that Seabrook is designed on the assumption that no more than 1.5 percent of the cladding will react to produce hydrogen, FSAR at 1.8-3, and that it has met the requirements of Regulatory Guide 1.7 only with regard to the Westinghouse-supplied components of its hydrogen control system. Thus the Applicant has failed to demonstrate that Seabrook's' hydrogen control system-is adequate to control the amount of hydrogen created in loss-of-coolant accidents in light of the experience from the TMI accident.

j I. S. Loose Parts Detection System i

The Applicant has not yet designed or developed a l

loose-parts detection system.for the' reactor's primary system,

and therefore does not satisfy' Criteria 1 and 13.of i Appendix A to 10 CPR Part.50, 10 CPR 50.36, or 10 CFR 2 0. l(c) , as implemented by Regulatory Guide;1.133, and does-not provide an adequate alternative to satidfy the require- l t

I monts.

Basis:  !

Regulatory Guide 1.133 describes an acceptable method to implement NRC requirements with respbet to' detecting potentially safety-related loose parts n light' water cooled reactors during normal operation. By complying with-1 1 Regulatory Guide 1.133 an Applicant will satisfy the NRC i

staf f that Criteria 1 and 13 of 10 CFR Part 50, section 50.36 of 10 CPR Part 50, and Paragraph 20.1(c) of 10 CFR Part 20 I have been met.

The Applicant has not designed'or developed a loose partsdetection system, and thus has notjmet the NRC's design, instrumentationandtechnicalshocificationsrequire-

! monts for safe operation of Seabrook. In addition, the Applicant has not demonstrated, with respect to the handling of loose parts, that it can meet Part 20 criteria which i

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mandate that licensees make every reasonable effort to maintain exposures to radiation as far belov Dart 20 limits as reasonably achievable.

Moreover, the Applicant has not derronstrated that it has designed or developed an adequate alt ern:1tive method of implementing the above requirements.

After the Ginna accident, a TV-optics examination identified foreign objects and tube fragments inside the i r B-steam generator. An examination of the A-ateam generator also revealed the existence of some small foreign objects.

The NRC report states that nei ther the int i1 i ty nor the NRC has conclusively determined the cause of the tube rupture, which raises the possibility that these 100:5e parts, probably left from recent repair work on the steam generators, caused the accident. NUREG-0909.

Therefore, the Applicant should, at a minimum, demonstrate that its loose parts detection s; tem for Seabrook meets the minimum requirements of Regulatory Guide 1.133, especially since the history of Westinghouse steam generators has shown the need for frequent repair work and thus the likelihood of loose parts being left near the generators.

I. T. Steam Generators  !

The Applicant has not demonstrated that the Seabrook l

steam generators are capable of resistihg degradation or that the new Model F Westinghouse generittom have been designed to solve the degradation prob 1brr and maintain their integrity during normal operation and during a credible accident scenario such as the accident which occurred at Ginna on January 25, 1982.

Basis:

The Report of the Reactor Safety Ecsearch Review Group, issued September 1981, advised the NRC that so-called steam generator tube degradation is a problem which "has not been considered sufficiently using recent accident analysis codes" in order to estimate "the consequences of a transient or some other failure that might lead in turn to the failure of a significant number of tubes. Such failures could lead to the degradation of ECCS function. "

Among the problems with Westinghouse steam generators since June, 1977, are the following:

1) Tube decay and support plate cracking (related to denting) at Indian Point Unit 2;

2) Tube denting and discovery of support plate cracking at San Onofre Unit 1;

3) Tube denting at Surry Units 1 and 2, and Turkey Pont Units 3 and 4.

The rupture of a large-scale generator tube in the Ginna accident combined with the failure of a pressurizer PORV created a situation in which operatorn had to choose between possible exposure of the core and release of radioactive steam.

Westinghouse claims, but has not actually demonstrated, that its improved Model F design has features that reduce the potential for denting and that it has changed its chemistry progran to ensure tube integrity against thinning, However, the Ginna accident demonstrated that serious problems with Westinghouse steam generator tubes continue, and that the Applicant should be allowed to operate with such tubing only after extensive testing and analysis of the steam generator tubes, and an affirmative demonstration that the probability of their thinning, denting or rupture causing an 11CCS steam binding emergency or other serious event is acceptably low.

-v I. U. Turbine Missiles The Applicant has not demonstrated that it meets GDC 4 of Appendix A to 10 CFR 50, as implemented by Regulatory Guide 1.115, which requires that structures, systems, and components important to safety be protected against the-efeects of turbine missiles whose launching might occur as a result of equipment failure. Nor has the Applicant demonstrated that it has designed an adequate alternative method to meet GDC 4.

Basis:

The Applicant states that it does not comply with Regulatory Guide 1.115, Rev. 1, and that the probability

-3 of damage due to low trajectory missiles is greater than 10 .

The Applicant has failed to demonstrate in subsection 3.5.1.3 of the FSAR that the Seabrook plant has an acceptable alternative method to meet GDC 4, or that it has met Regulatory-Guide 1.115 which provides an acceptable method to comply with GDC 4.

As can be,seem from Figure 3.5.1 of the FSAR, certain low-trajectory missiles resulting from turbine failure l

could severely harm the containment of one cri both of the two Seabrook plants.

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The Applicant, furthermore, has admitted that it cannot demonstrate, as the NRC requires, that the probability of damage due to such low-trajectory missiles is lower than

-3 10 .

In-service inspection of steam generators has been demonstrated historically to be inadequate to prevent their degradation and resulting accidents due to this degradation.

The Applicant has not stated that it fully meets all require-ments of Regulatory Guide 1.83, which sets forth one acceptable method of ensuring that in-service inspection of steam generator tubes complies with General Design Criteria 14, 15, 31, and 32 of Appendix A to 10 CPR Part 50.

Therefore, the Applicant must redesign the Seabrook plant to ensure full compliance with GDC 4.

I. V. In-Service Inspection of Steam Generator Tubes The Applicant has not demonstrated that it meets General Design Criteria 14, 15, 31 and 32 of Appendix A to 10 CPR Part 50, as implemented by Regulatory Guide 1.83, in order adequately to reduce the probability and consequences of steam generator tube failures through periodic in-service inspection for early detection of defects and deterioration. Nor has the Applicant developed an adequate alternative program for in-service inspection of steam generator tubes.

Basis:

The Applicant has not demonstrated, in subsection 5.4.2.5 of the FSAR, that its program for in-service inspection of steam generator tubes meets Regulatory Guide 1.83, which provides one acceptable method of complying with the requirements of General Design Critoria 14, 15, 31 and 32 of Appendix A to 10 CFR Part 50. Nor has the Applicant designed an adequate alternative method.

The Applicant says merely that the inspection program "will be performed in accordance with the requirements of Regulatory Guide 1.83, Re v. 1," and that the steam generators are designed to permit access to tubes for inspectjon or plugging repairs. See FSAR 1.8-31 and 5.4-19.

Given the lang history of serious problems with Westinghouse steam generators, and the recent accident at

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Ginna caused by a rupture in a steam generator tube, it is imperative that the Applicant demonstrate that its inspection program is adequate to provide some assurance that problems will be found and remedied in steam generator tubes prior to another Ginna-type accident. Even though Ginna and a number of other plants have satisfied the NRC staff of the adequacy of their in-service inspection and repair programs, the

( continuing problems with Westinghouse steam generators and l

the serious accident at Ginna demonstrate that the requirements of Regulatory Guide 1.83 are not sufficient to reduce the probability and consequences of steam generator tube failures through periodic inspection for early detection of defects and deterioration.

The NRC found after the Ginna accident that the ruptured tube had been inspected as late as April 1931, and found not to have experienced any signi ficant degrmlation. The Ginna in-service inspection program also appeared to comply with the requirements of Regulatory Guirle 1.B3. HUREG-0909.

Accordingly, the April 11, 1981, inspecti on iai. led to accomplish its most important goal--prevention of an accident that might result from a defective tube.

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. l I. W. Seismic Qualification of 1;lert r r t al Equipment The Applicant has not demonstrat -a ! hat it has adequately assured the scismic qualificati]n of electrical equipment at Seabrook, as required by Critoripn III, " Design Control" of Appendix B to 10 CPR Part 50 and implemented by Regulatory Guide 1.100, Rev. 1. The Applicant also has not shown that it has developed an adequate! alternative method l

ofseismicallyqualifyingelectricaleqjti.pment.

Basis:

According to FSAR 1.8-36, the Applicaat has not demonstrated that all NSSS safety-related electrical equipment or BOP electric equipment has been seismically qualified to meet all requirements of Regulatory Guide 1.jo0, Rev. 1. However, in a letter to the NRC the Applicant sL6 ten that qualification of electrical equipment and instrumentation complies with the guidelines of Regulatory Guide 1.100.

Because of the conflict between the statements of the Applicant in FSAR 1.8-36, and those statements made in its January 8, 1982 letter to the NRC, the Applicant must demonstrate that its method of seismically qualifying electrical equipment at Seabrook fully complies with Criterion III of Appendix B. to 10 CFR Part 50.

1. 0 Seismic Qualification of Equipment'ia dn Unresolved Safety Issue listed as A-46 of NUREG-07(G. The Applicant has not demonstrated that this unresolv(a loroty issue has been resolved for Seabrook or that meacuro*; oxist and are employed at the plant to compensate for! t-3 ? lack of a solution to this unresolved safety issue, as required under the Appeal Board decision in Virginia ElpcPric and Power Company (North Anna Nuclear Power Station, Units 1 and 2),

ALAB-491, 8 NRC 245 (1978).

/

m II. QUALITY ASSURANCE CONTENTIONS A. Design and Construction

1. General Desiqn Criterion j of Iw endix A'to 10 CFR Part 50 requires the establishment on:i ;nplumentation of a quality assurance program. This and all general Design i i Criteria cover all aspects of the faciljty shak are "important 6

t.o safety." NECNP contends that theSe5hrodkQuality 1

i Assurance Program for design and constr0rt:ign has boon too narrow in scope, applying only to items cognidered to be

" safety-related," rather than to the brha:lhr category of aspects that are "important-to-sa fety. " Accordingly, the Applicant has failed to comply with GDC;l to Appendix A.

1 Basis: l According to paga 105 of the-lequlatory Agenda issued by the NitC on January 31, 1982, the original intent i

of the Commission in issuing Appendix Blto 10 CFR Part 50 was'to require that the Quality Assurande programs required by'that Appendix cover "the full range of safety matters,"-

rather than some subset that is considered to be " safety-related." The Regulatory Agenda cites examples of structures, systems, and components-for which Appendix B criteria have not been fully implemented, including in-core instrumentation, reactor coolant pump motors, reactor coolant pump power cables, and radioactive waste system pumps, valves, and storage tanks.

.n

~56-Section 17.1.1.2(b) of the FSAR describes the aspects

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of the Seabrook facility covered by the Quality: Assurance Program as the " safety-related structures,-systems, and I

components" listed in Tables 17.1-1, 17.1-2, and 17.1-3."

None of the examples cited by the' Commission as important to safety is covered by the Seabrook Quality Assurance Program. Another example of aspects of the facility that are important to safety, but are not included in the FSAR tables is all equipment that removes heat from the steam generators during shutdown. Such requirement is essential to assure adequate decay heat removal as required by GDC 24.

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II. A. 2. Under 10 CFR Part 50, Appendix B, the Applicant is required to develop and implement an effective Quality Assurance Program the provides the following assurances: (1) that the design of the Seabrook plant. complies with all applicable regulations and requirement; and assures the protection of the public health and safety, and (2) that all aspects of the construction of the facility are carried out in a quality manner in conformance with the design and the applicable requirements. NECNP contends that the Applicant has failed to meet the requirements of Appendix B with respect to either the design or construction of Seabrook.

The Quality Assurance / Quality Control Program for Seabrook has been subject to pervasive inadequacies in all areas such that there is no assurance that the plant has been designed or constructed in accordance with applicable requirements and consistent with the protection of the public health and safety. It may not be possible to determine whether the plant has been designed and constructed to assure safety.

At a minimum, a complete independent audit of all design and quality assurance documentation, a complete physical reinspection of all aspects of the plant, and a thorough conservative engineering analysis of all aspects of the plant that cannot now be inspected are required to provide a reasonable

. assurance of protecting the public health and safety.

.. . Basis:

This contention is based on the performance of the Applicant as reflected in reports of the NRC Office of Inspection and Enforcement, the inadequacies in nuclear industry quality assurance demonstrated by investigations or unforseen incidents at a number of plants, including Diablo Canyon, Zimmer, Midland, South Texas, Brown's Ferry, North Anna, Davis Besse, and Rancho seco, an,d on the fact that the NRC itself, in the person of Chairman Paladino, recognizes that Quality Assurance Programs to date have not been adequate. It is also based on the fact that the findings reflected in NRC I&E Reports are not the result of a thoroughgoing, systematic evaluation of all aspects of the plant, and reflect-only a small percentage of the failures and violations that exist at a nuclear facility. See, e.g.,

Inskeer., G. W., "The Cause and Effect at Three Mile Island,"

Quality, May 1980, pp. 22-26.

Specifically, with respect to Seabrook, I&E reports reflect the following inadequacies or violations:

1. Acceptance of deficient conditions through apparent oversight or incompetened of inspectors.

I&E Report Nos. 79-05, 75-07, 79-10, 80-06, 80-10, 80-01, 81-09, 81-12, 80-13, 82-1.*/-

1  :

Appendix B, Criteria II,!V, :h'XIV.

i  !

• /All I&E Reports will be identified by reference to the report number for Unit 1, Dockot No.,50-43, except as otherwise noted. l

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i I

2. Acceptance of deficient conditions as a result of inadequate or nonexistent Quality' Assurance procedures. I&E Report Nos. 80-06, 80-04, 80-11, 81-01, 81-02, 81-03, 81-05, 81-07, 79-07, 79-06.

Appendix B, Criteria II, V, XIV,

3. Failure to perform required inspections. I&E Report Nos. 79-06, 80-03. Appendix B, Criteria V, X.

4. Falsification of inspection record to show inspection was properly performed when it was not. I&E Report No. 79-06. Appendix B, Criteria II, X.

5. Failure to prevent deficiencies in pipe supports, pipe welds, and piping and tubing generally. I&E Report Nos. 80-06, 80-10, 81-03, 81-05, 81-14, 79-06. . Appendix B, Criterion V.

6. Failure to determine the! root capses of i

deficiencies or to assure that corrective I i i acctionsaretakentoprhvendbeficiencies j ii from recurring. I&E Report Non. 79-06, 79-09, 80-03, 80-11, 80-12, 81-03. 'AhpendixB, i

Criterion XVI.

7. Failure to assure proper denign. I&E Report Nos.

81-14, 81-05. Reports pttrauant to 10 CFR 50.55 (c) ,

. . dated 10-27-78, 12-6-79 (three reports),

12-1-80, 7-17-81, 1-15-81, 2-23-81, 6-18-81, 8-25-81. Appendix B, Criteria III, V.

8. Failure to assure proper repairs. I&E Report Nos. 79-07, 80-04, 80-11, 80-12. Appendix B, Criteria V, IX, X.

9. Failure to assure deficiencies are not caused by poor contractor interface. I&E Report Nos.

80-11, 80-12, 81-12, 82-01. Appendix B, Criterion V.

10. Failure to assure the' procurement of proper-materials and failure to; assure that procured items comply with all requirements. I&E Report Nos. 81-09, 81-12. Appendix lB,:CriteriaV, i

VII, XV. I i .

A t r ;

11. Failure to assure proper documhqt control such that required changes are!not made, and

'specifkcationsare incorrect procedures'and

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used. I&E Report Nos. 7 -06, 80-03, 80-04, 80-11. Report pursuant to 10 CFR 50.55(e),

dated 12-6-79. Appendix B,. Criteria II, III, V, VI.

12. Pervasive deficiencies in welding and weld i

repairs. I&EReportNosj 79-06, 79-07, 79-10, 80-03, 80-11, 80-10, 81-01, 81-03,'81-05, I

81-09, 80-04, 80-11, 80-12. NRC Stop Work Order in letter dated 12-22-80. Appendix B, Criteria V, IX, X.

13. Inadequate audit program and inadequate ,

commitment to and understanding of Quality Assurance. I&E Report Nos. 79-b8, 78-06, 80-05, 81-12, 80-09, 78-16. Append}x B, Criteria I, II , .

XIII, XVIII.

t ,

In each of the. areas cited above, i t ;is clearly necessarytoundertakeasystematicanalysidtodetermine j I thefullextentof'thedeficienciesor2nadegbaciesandthe proper remedial actions. However, this contention is not limited to those areas. Rather, particularly by: virtue of the broad ~

range covered by these items, they demon l strate that the entire

l. .

Seabrook Quality Assurance Program for design and construction-

.is deficient.

I

i i .

l 1 i.:

! I ii II. . B. Operations Appendix B to 10 CFR'Part 50 establishes the requirements that must be met by -the Ap?lican{indeveloping and implementing a Quality Assurance Prbgram for operation of I

the Seabrook facility. 10 CPR 50. 34 (b) {(6) (ii) requires I

that the FSAR include "a discussion of how the applicable requirements of Appendix B will be satiIfied." NECNP contendsthattheSeabrookQualityAssupanceProgramfor i

operations does not comply with_either Appendix B or 10 CPR 50.34 (b) (6) (ii) in the following areas:

1. The PSAR fails to address cach of the criteria in Appendix B in sufficient detail to enable an independent reviewer to determine whether and how all of the requirements of Appendix B and the guidance in all applicable regulatory guides will be satisfied.-

2. The Quality Assurance Program for. operations extends only to matters considered to be

" safety-related," and not to all structures, systems, and components "important to safety."

Examples are discussed in Contention II. A. 1.

3. The Quality Assurance organization does not have the independence required by Appendix B, Criterion 1.

m-I

4. The Quality Assurance Prbgrqm for operations asdescribedintheFSARldoesnotdemonstrate I

how the Applicant will at:sure that replacement materials and replacement parts incorporated i

into structures, systems [ or components important to safety will be equivalent to the original equipment, installed in accordance with proper procedures and requirements, and otherwise adequate to protect the public health and safety. Similarly, the Quality Assurance program does not assure or demonstrate how repaired or reworked structures, systems, or components will be adequately inspected and tested during and after the repair or rework and documented in "as-built" drawings.

5. The Quality Assurance Program for operations as described ir. the FSAR fails to assure the  !

presence on the operating staff of an adequate number of qualified QA/QC personnel, particularly during of f-shif ts.

Basis:

1

1. The basis for point 1 above is the language of Section 17.2 of the FSAR, which includes only a very general discussion of-the Quality Assurance Program, with scattered references to procedures,

but does not provide the detail necessary to determine how the program _will be implemented.

Examples are the discussion of design control at Section 17.2.3 and document control at Section 17.2.6. Although both of these are areas in which the NRC found deficiencies during I&E inspections, the discussion in the FSAR is not sufficient to allow a determination of whether design and dobument:' control will

! l actually be accomplished successfully. The same point applies throughout.;

1 ';i

2. The basis for point 2 is thc comparison l

i i between the scope of the SeabrhokQuality i

Assurance Program for operatio's n and the requirements of GDC 1 of Appendix A.

3. The basis for point 3 is the fact that the Nuclear Quality Manager reports to the Vice President - production onl'an equal basis with l

the Nuclear Production S0porintendent, rather I

than to the Executive Vihe President - Engineering and Production. Since the Vice President - Production I

is directly responsible for maximizing the

)

amount of power produced:by Seabrook, the quality assurance organization must report to a separate-but-equal level or a higher level in order to assure its independence and freedom of operation.

__m. - -

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4. The basis for point 4 is the [p't c that the FSAR contains no discussion,yhatsoever of Quality Assurance for mainte ance, repairs, 1  :

or-rework, all of which will! occur during

! i.

the life of the plant. jj

5. The basis-for point 5-is theja'bsence of any i

discussion in the FSAR o'f mini' mum: staffing levels cn any. indication that sufficient:

Quality Assurance staffirig willibe' assured at all times.

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1 i

EMERGENCY PLANNING CONTENTION III.

NRC regulations require the license Applican t to submit with _ its FSAR a complete emergency'planbefhrealicensemaybe issued. 10 CFR .50.34 (b) ( 6 ) ( v ) . The pinn mus t be . " adequate and capable of being implemen ted , " providing a "reaso n abl e t

-assurance that adequate protective measurcs can and will be taken in the event of a radiological c morgency."- 50.54 (a)(1) , ( 2 ) .

The emergency plan submitted by the Applicant for the Seabrook

.I facility license is seriously deficledt in a number of. respects i

listed below, and f ails to provide ali the information required I

by _ Appendix E to Part 50. In its pre.1ont form, the plan is i

incapable of being implemented or providing any-assurance that in the event of an emergency adequate : measures can and will be taken, and therefore it cannot.be accepted as fulfillment of a licensing requirement under 10 CFR 50.47.

Specification and Basis :

1. The emergency plan does not contain an adequate emergency classification and action level scheme, as required by 10 CFR 50.4 7 (b)( 4 ) and NUREG -06 5 4 . No justification is given for the classification of various system failures as unusual events ,

aler t s , site area emergencies, or general emergencies. In general, the classification scheme minimizes the potential significance of transients, for example, " failure of a safety or relief valve.

in a safety related systna to close following reduction of applicable pressure " is classified merely as an unusual event, in spite of the f act that the accident at Three Mile Island Unit 2 was caused in par t by just such a failure. Although the a

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Appl i c an t 's classification scheme gen rally follows the scheme l

outlined in NUR EG -06 5 4 , NUR EG -06 5 4 1: n.> d i npos i ti ve on this, question. The Appl i c an L ':. ju(kynmt on iir. d n;9ificat. ion scine should include consider ation of speci f i c plant circumstances, such as the anticipated time lag for evacuat ion due to local pr obl em s . In its present form, the Applicant's classification scheme provides no assurance that the Seabrook onsite ano offsite emergency response apparatus and personnel can be broug ht to an adequate state of readiness quickly enough to respond to an accident.

2. The emergency plan fails to identify emergency action levels or classify them according to the required responses.

The symptoms of transients must be identified, monitored and responded to. The emergency plan provides no systematic means for analyzing these indicators and responding quickly to them.

Guidelines should be provided for choosing appropriate responses.

3. The emergency plan does not demonstrate the appl ic an t 's ability to respond to failures at both unitn of the Seabrook reactor, or a failure at one unit whicn af f ects the other's capacity to operate safely. A number of factors, such as ear thquak e , severe storm, loss of of fsite power, or deg raded grid voltag e, could simul taneously impair the oI>er ati on o f both units. There is no showing that the Applican t will have the requisite technical analysis capability or adequate emergency response personnel onsite or available within an adequate period of time in the event of an emergency aficcting both units.

4. Appendix E to Part 50 requires that employees be trained fo r familiarity with their speci fic emergency response duties .

-6lt-Ilowev er , the emergency plan does not pr ov ide for the tr aining of unit shift superviors t o enable then to deal with special problems involved in emergencies, including making choices among alternative responses under stress.

5. The emergency plan does not indicate that local conditions were taken into account in establishing the Plume Exposure Emergency Planning Zone (EPZ), as required by 10 CFR 50.54(s)(1).

As provided in that section, the 10-mile radius is only a recommendation, and constitutes the minimal bounds for an EPZ.

NECNP contends that, as established, the Plume EP Z is inadequate to protect the public health and safety in the Seabrook area, consid er ing such local factors as meterolog y , restricted access routes, evacuee directional bias , evacuat ion shadow , and the seasonal congestion of the area with summer tourists.*/

a) Meteorology: As shown by studies done for the state of Cali fo rn ia Office of Environmental Ser vices , the frequency and pesistence of winds has a great bearing on the probability of radioactive emissions being carried in any given direction .

The sh ape o f the EPZ may well be not perfectly round but elliptical due to these factors. Fur ther more , humidity, baro-metric pressure, and frequency of inversions may play a significant role in the dispersion of radioactivity throughout the area. The Applic an t should be required to perform a complete an alysis of the ways in which meteorology affects the size and shape of the EPZ.

• /Not all of the factors discussed in this conten tion are mentioned specifically as loc al factors, but the rajulation is not in tended to be limited to the listed factors, as indicated by the words "such as " preceding the examples of local factors.

e .

b) Access routes and directional biases of potential evacuees : The Seabrook area both within and beyond the LO-mile EP Z ,

is char acterized by very restricted access routes, especially along the seashore where there is very limited access from Route 1A to Route 1 or~the interstate. See PSAR Figure 4 .2 . These limitations should be considered in establishing the EP Z .

Furthermore, studies at Three Mile I sland have revealed direc-tional biases which have a potentially significant bearing on the size and shape of the Plume EPZ. The study has suggested that three factors affect the direction that evacuees will tak e :

(i) A strong directional bias in favor of upwind desti-nations from the reactor

( ii ) A "psy cho log ic al attraction" to mountainous areas in time of danger

( iii ) A reluctance to select destinations in more densely populated urban areas.

Zieg ler , Brunn, and Johnson, "E vac u at ion from a Nuclear Technologic al Disaster ", Geog raphic al Review, vo. 71 No. 1, J an u ar y 19 81, at 9.

All of these factors throw off calculations based on predesignated evacuation routes, and should be specifically considered as independent variables in the determination of the proper EP Z .

The information concerning evacuation preferences should be obtained directly by interviewing local residents and transients regarding their likely behavior during an accident.

c) Population Ch ar a c ter is't ic a : During the summer, beaches both within and beyond the 10 ntlo : idius are congested with people and the coastal road clogged wit 7 slow-moving traffic.

This condition would be aggravated if, in an emergency, everyone tried to leave the area at once. The it:TA:, ren 't based on a realistic model of how the evacuation network w(11 1 bb toadeu and therefore underestimate the time it would t ak e t;o evacuate the ar ea . See m i

j a

s Science Applications, Inc., "A Study of Postulated Accidents at Cali-fornia Nuclear Plants", prepared for the State of - California Office of Energy Services, PA101380-3811J p. 3-17. The ditficulty of moving traf >

fic expediently and sheltering larg e numters of people on the beach in the' summer months should be considered-in establishing the-size'and shape of the EPZ.

d) Evacuation Shadow: Another phenomenon which has not been considered by the Applican t in establishing its EP Z is the "evacua ti on sh adow, " or " tendency of an official evacuation advisory to cause departure from a much large area than was originally in tended . " Zieg ler , et al, " Evacuation from a Nuclear Technological Disaster ", _ Geog raphic al Review, supra, at 7.

At Three Mile Island, where pregnant women and preschool children within a radius of five miles were advised to evacuate and' everyone within a ten-mile. radius was advised to remain indoors, 39% of the population living within 15 miles of the plant evacuated (53% within 12 miles ) . Between 10 and 12 miles, some' 4 7% of louseholds sampled evacuated some of their members. Geographical Review at 6-7. An even greater response cou id be expected following an evacuation directive in an area with a rxlius of 10 miles or i

more. Such a response would have a significant impact' on the l  :

Applicant 's ability to plan for an omdrgency,_and thus on the size and dimension of the EPZ - Consideration of 1

the evacuation shadow, as based on exi n t ip'q atudies of the phenomenon and the population characteristics of thn area-surrounding the Seabrook  ;

facility, should be included in the Applicant's determination of the proper size and shape of the EP Z .

6. The emergency plan does not i ndicate that beyond design basis accidents were considered in thq entablishment of the EP Z .

i _ _ _ - . _

. , Thus, the Applicant has not planned for the type of accident involving the greatest risk to the public. Tite failure to consider such an accident as part of the EPZ is generally justified by reference to the low probability of a core meltdown determined in NUREG-0396. However, the foundation for this determination, WASH-1400, has been largely discredited.

As' stated by the NRC's own Risk Assessment Review Group Report, NUREG/CR-400 at 25, "when a great many rare events _.each have a small probability occurrence, the chance that at least one of them will occur can be rather high." In any event, NUREG-0396's determination of a 30% probability that if a core moltdown occurs it will result in doses above the Protective Action Guidelines constitutes sufficient justification for its consideration in establishing the EPZ. The need to expand the EPZ relates both-to improving emergency planning responses and limiting health effects of a major accident. As the government of California's Office of Emergency Services, which has adopted expanded EPZ's

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in consideration of beyond design basis; accidi3nts, has stated:

"[P]rudence dictates that the EPZ' . be e : tended so that advance planning c,n be' performed {o aid!in resolving the potential problems associated withlthe more severe types of accidents, the penetration leakage and major containment failure classification $...IL:did not seem

. prudent to restrict planning attention to: responding onlytothepotentialforincurrinyimmediatelife threatening radiation doses for sugh severe accidents.

Thus the extended zones were selected no that the potential for incurring health impacting dos s wan reduced not only for early fatalities, but also forjearly injuries and delayed cancer effects as well...[E]xtending the EPZ boundaries results in a prudent reduction in the probabilities of early health effchts and: a substantial reduction in the probability of' delayed health effects (associated with) . 5 to 1 Rem PAG dose limits."

i "Not only does the extension [of tihe Plume EPZ] improve the probability of limiting doses to the public, but it

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also substantially improves the base of procedural plans and f acilities for required. cmergency response ef forts with respect to evacuation, sheltering and relocation prepared-ness and capacity requirements . "

State of California, Of fice of Emergency Services , " Emergency

. Planning Zones for Serious Nuclear Power Plant Accidents ",

Novem ber 19 80, Alex R . Cunningham, Director.

7. The Applican t does not provide offsite emergency plans of state and local governments , nor does it indicate how the Ap-plicant 's emergency plan will coordinate with those~offsite plans. 10 CFR 50. 3 3 (g ) , Appendix E .III . 'The plan should show that all signals emanating from the plant -receive the same inter-pretation by state and local authorities m they are given by the utility. In addition, letters of agroevent between the Applic an t and the affected state and local covernments , reflecting-I their mutual understanding of each otder'd responsibilities, must be submitted to the NRC. NUR EG -0 654,-hpp.'3. NECNP reserves i

the right to amend its contention to challerne the sufficiency of the Applicant 's plans to coordinate -witIh state and local aut ho ri ties .

8. The FSAR does not show that all possible accident sequences I -

can be monitored , as required in orded to take ' the necessar y steps in responding to an emergency. For example,~there is no I

indication that process mon itors comply wi t h General Design Criteria 13, 19, and 64, as implemen ted by Reg . Guide 1.9 7, o r -

i are environmentally qualified as required by GDC 4.

9. The FSAR does not provide for adequate _ radiological .

monit oring . Previous experience has shown that reliance upon individuals to take radiation samples in the field may lead to a

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e serious errors. See NUR EG -06 00 , at 11-3 -7, -97. Because weather conditions may impede utility personnel and human error f actors may detract from the accuracy of results, perm an en t radiological monitoring equipment should be placed at a number of locations surrounding the plant. These monitors should relay in f o rmat ion to a computer at the plan t wh 2.ch can plo t radiation levels and estimate the location of t. b e plume. The monitoring i

equ ip men t should be equipped with ind6 pendent backup power i

I supply and must meet criteria for withstanding adverse meteorolo-gical conditions.

10. The Applican t must submit and justify a dose assessment mod e l . App. E to P ar t 50. This model munt not be limited to l

a straight diffusion model, but must ak e into account the behavior of heated releases and other "so urce char acter istics . "

NUR EG -065 4 , App. 2 at 2-3. The computjer used for making dose i

1 assessments should have an ind ependen t' backup power source to assure that it will con tinue to oper atie it onsite or offsite i

i power fails. ,

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11. The emergency plan does not; provide for early notification and clear instructions to the local populace, as required by 10 CFR 50.4 7(a )( 5 ) . Detailed criteria for providing notification to the public, set forth in NUREG -06 5 4 App. 3, must be met before the Applican t may be licensed. The Applicant must provide a means by which all people within the EPZ will be able to hear a warning of a radiological emerg enc y , and they must be trained to understand the warning . In the Seabrook area, special provision must be m ad e for educating the thousands of transients who move through the

~74-? l-

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i area in - the summer months. 'In particUlar, Ihey should be apprised in advance of the location of shelterd evdct;d t' ion routes , .and congreg ation areas. j~!

12. The emergency plan does not provifeforthesheltering of the large numbers of people who. may be,onJbeaches during a radiological emergency in the summer, and who will not be close-to their own homes, motels, or public bu ild ing s .

13 . - The -emergency plan' does not indicate the basis for the-code it uses - to make evacuation time estimates. The plan does not indicate its bounds of error, or whet.her the model'is static -

t or dynamic. A' sensitivity analysis sl ould be performed for the 1

model and for its underlying assumptidns, and the Applicant should I

1 be required to disclose those assumptions which undergird the evacuation time estimates . Furthermore, the Applicant should indicate its reasons for substantially reducing the' estimates of evacuation time presented in its PSAR. NECNP reserves'the right to amend its contention to challenge the bases for' the Applicant's evacuation time estimates and their accuracy when this information is submitted.

14. The preliminary evacuation time estimates submitted by the Applican t assume conditions much more favorable-to rapid evacuation than are likely to occur. At the very least, the Applicant should consider a " worst case " scenario in order to assess the correctness of its estimates and assure that emergency planning is adequate to meet these circumstances . For example, the ' combination of weather and population conditions assessed by the ETAs did not include adverse weather conditions on a summer weekend, when the Seabrook population swells to its greates number .

i Mecond, Ihe 1;TA s .u.sume I hat no rm a l t r al I pattorso. wil1 provat1.

I The ETAs should inst ead t ak e into aceWm t ihe evacut e d irect ional bias discussed in these contentions at 1: 4, and t he poss i bil i t.y t h a t.

much of the traffic will move upwind of t he plant. Third, the l'TA a s hou ld t ak e into accoun t the del.n aused by subst antial evacuat1on from the "evacualion s h ad o v, . " l'our t h , Ihe ETA est imat es were based on evacuation by !4rivate vehtele, which excluden ihe more time consuming and conilix f act or of the use of buses to evacuate children in schools, chtluren in camps, tour groups travelling in buses, and people in institut. ions such as hospitals and nursing homes. In i t.s pr esen t fo rm , the Applican t 's preliminary evacuation time assessment in 1ar too simpl istic.

N!:CNP ener ve . Ihe ight io challenge Ihi ade<1uacy of the Appl i c an L 's ETas in their final form.

15. The Applicant has not sat ist act orily demonstrated the feasibi 1i t y of evacuat. ing the Seabrook EP Z wi thou L exposing evacuees to unacceptable radiation doses. The Commonwealth of Massachusetts has c al c u l a ted that, under cond i t i on s postulated by the NRC/ EPA Task l' orc e Report (which has been endorsed by the Commission),

a radioactive plume travelling at a speed of 10 mph over the two miles between the beach and the plant could reach flampton Beach within 15 minutes of a release, l'ur t her m o r e , a pl ume travelling to Sal i sbury neach at the same speed could reach the beach in about 30 minuten. Commonwealth of Massachusetta Memorandum in SupporL of SAPL 'a Request for an Order to Show cause (May 2, 1979) at 11, citing NRC/EP A Task I'orce Report, NtJR EG -03 9 6 at 20, App. I. The emergency plan 's evacua t ion t.ime est imat e fo r llampt on licach tu

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4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 20 minutes, and for Sal i sh'n v , i hourt, and 50 minutes.

As discussed above, NECNP considers t h- :o -stimates to be extremely opt omi s t i c .

Even if the Applicant could show t hat it h ad supplied enoug h shelters in the area for the large numbers of people who might be exposed on the beach or otherwise out; doors on a summer day, the sufficiency of shelters to pro tec t ag a i ns t atmospheric radiation is in grave doubt. See NtIR EG /CR -l l 31, "l:xamination of Offsite Rad iolog ic al Emergency Measures for Nuclear Reactor Acc iden t.s Involving Core Melt, Sand ia I,aboratories, June, 1978. ( "S he l te r i ng by itself, unlena the quan t it.y of rad ionucl ides inhaled can be substantially reduced, will also no t provide much pro tection . " Id . at 92.)

The Appl ican t should therefore be teluned a license because it has not provided a reasonable assurance t hat its emergency plan can and will pro tec t the public from excessive exposure to radiation.

16. In order to plan adequately for t hp protection of the public health and dalely, base 1ine dat a on 1ocaI healLh condiLions should be gat hered to determine the immediate ana l o ng -t er m health effects of radiation exposure.

Data to be collected should include al least measure-ments of thyroid hormone deficiency in newborns , fetal death rates, neonatal and infant morbidity and mortality rates, known exposure to carcinog ens , cancer incidence and prevalence withing cir cumscribed areas around nuclear facilities, occupational history, demographic characteristics, and symptoms of psychological distress of the population at risk, as well as the availability of medical staff support and heal th care facilities during a radiological emergency.

< s _.77-Gordon MacLeod, M .D . , "A Role for Public lioalth in the Nuclear Ag e , "

Amer i can Journal of Public Health, Vo . ~ ? tio . 3 (March 1982. )

The Applic an t has not shown that the l mal

• h ef fects of a radiolo-gical emergency can or will be monitored i. n the Seabrook ar ea .

It should be required to do so beforein i teense is issued.

4

- _ - , - , - 7_ - - - .-. , ,,p.,-w- - y m

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-4 , UNITED STATES OF AMERICA titlCl. EAR RI:Glit.ATO!(Y COMMISSIOtt ilCPOHE Till* ATOMIC S AFETY AND 1.lCI:NS!!M: HOAltD

)

In the Matter of )

)

I'U'11,1C SERVICE COttPANY OF )

N!M llAMPSillHE, e. .t al., --

) lucket tion.

50-43) o r, (Seabrook Station Unita 1 anit 2) )

50-444 01,

)

.. ._ __ _ _ . _ . _ _ _ _ .)

CEllTIFICATE OF SEllVICE I hereby cert i fy Ihat copien of the NUCNP SUPPI.EMI:f1 TAI.

PETITION FOlt 1. EAVE TO I NTEllV10NE in the above-captioned proceedinil have been nerveil on the foIlowirq by deponit in the Un i ted St aten ma il , firnt cla:,n, thin 21st day of Apri l , 19112.*

• llelen lloyt , 1:fiq . , Chairman Docketin't Service Section Atomic Safety and 1,1 censing Office of the Secretary incard Panel U.S. tiuclear flequlatory ll.S. Nucienr itegulatety Commission Commisnion Washington, D.C. 20555 Washington, D.C. 20555 Robert L. Chiesa, Estl. Paula Gold, Anst. Atty. General Hadleigh, Starr, Peters, Dunn Stephen it. l.conard, Asst. Atty.

s 1(t,hls .lo Ann Shotwell, Asst. Atty.

95 Macket Street Office of the Atorney General Pinchest er , Hl! 03101 1:nvironmental Protection Div.

One Anhurton Place, it)th Floor 1.ynn Chong non t on , MA 0210ft Bill Corkum Sary McCool Nicholan .1 Contello Dox 65 Int Untex District Pl yn out h, Nil 012 f.4 Whi t eha l l 1(nad Anerbory, rtA 01913

!:. Tupper Kinder, 1:sq .

Assistant Attorney General Tontin P. Kentir i ck 1:nvironmental Protection Divinion H22 1.afayette Road Oltice of the Attornoy General :P.O. H a x 'i%

State lioune Annex 'llanpt on , Nil 01842 Concord, Nil 0 3 301 b

• h

m ,

4 , Robert A. Backun, Ecq. Itep . Arnie Hight 116 Lowell Stieet State of New llampnhire p,0. pox 5)(, lloune of itep r eners ta t ives Manchester, Nil 03105 concord, tui o 11r)1 Phillip Ahrens, Esq. Paul A. F itzche, Eng.

Assistant Attorney General Public Advocate State !!ouse Station M titat.e !!ouse Sta tion 812 Augusta, itf: 04333 Augins t a, MI; 04 313 Wilfred L. Sanders, Esq. Donald I,. Ilerzberger , MD Sanders and McDermott lii tchcock lionpital 408 Lafayette Road llanove r , fHI 0~1755 llampton, Nil 03842

• • Thomas G. Dignan, Jr. , Esq. I'dwa rd .T . tielk rmott , Esq.

Hopes & Gray fi.uiders .ind McDermott 225 Franklin Street 408 Lafayet.te Road Dost.on, MA 02110 Ilampton, Nil 01842

!ien . 1(obe r t. I., Preston Mr. Robert :' . Preston State of New ilanpnhire 226 Winnacunnet Road tienate llampton, Nb 03842 Concord, Nil 0 3101

• Dr . Emmoth A. Luebke Admini strati ve J.udge

• Dr. Oscar 11. Paria Atomic Safety and Licensing Administrative . fudge ~ Board Panel Atomic Safety and Licensing U.S. Nuclear Regulatory Board Cornmi nnion U.S. Nuclear Requlatory Washington, D.C. 20555 Commission Washingten, D.C. 20555 Cooperat.ive tiembe rn for

• - IW 14 S%*e EM. Itm:ponnible I n ve n t inen t U.S. Nuclear Exjulatory Cantr.sim nox (,5 Wadiinjten, D.C. 20555 Plymouth, Nil 03264 At.omic Safet.y and Licensing Mn. Pa t.Li Jacobson Board Panel 3 Orange Street U.S. Nuclear hegulatory Commission Newburyport, MA 01950 Washington, D.C. 20555 Atomic Safety and Licensing Appeal Doard Panel U.S. Nuclear Regulatory Comminsion Washington, D.C. 20555 Date: April 21, 1982 -

Wi.11 i arl !i Jordiin , III

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• • Express Mail