ML20050C014

Amend & Suppl to 811117 Petition to Intervene.Addl Contentions Specified.Certificate of Svc Encl

ML20050C014
Person / Time
Site:	Seabrook
Issue date:	04/05/1982
From:	Kinder E NEW HAMPSHIRE, STATE OF
To:	NRC COMMISSION (OCM)
References
NUDOCS 8204080016
Download: ML20050C014 (52)

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(, 7 7pg [2 UNITED STATES OF AMERICA PI[gD JUCLEAR REGULATORY COMMISSION '82 TPR -6 P12:20 0 \

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In the matter of: ) 7,

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PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE,) Docket Hos.: 50-443 ET AL. ) and

, ) 50-444 (Seabrook Station, Units 1 and 2) )

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AMENDMENT AND SUPPLEMENT TO THE PETITION FOR LEAVE TO INTERVENE AND REQUEST FOR HEARING OF THE STATE OF NEW HAMPSHIRE AND GREGORY H. SMITH, ATTORNEY GENERAL OF THE STATE OF NEW HAMPSHIRE H-NOW COME the State of New Hampshire and Attorney General Gregory H. Smith (hereinafter " Petitioners"), and amend and supplement their Petition to Intervene pursuant to 10 C.F.R. 2.714 upon the following basis and terms:

'l. On November 17, 1981, the State of New Hampshire and the Attorney General filed a Petition to Intervene and Request for Hearing.

2. 'On March 17, 1982, the Petitioners received an order of the Atomic Safety and Licensing Board which appeared to require the. filing of contentions by April 6, 1982.

3. The Petitioners immediately attempted to determine whether they could reasonably-comply with the April 6 deadline and, finding that the development of.a complete set of contentions in this time frame was nearly impossible, on March 24,_1982, filed a Motion for Additional Time to File Contentions.

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4. The Petitioners cubrait the following contentions without prejudice to their right to additional tirae for filing contentions as set forth in the March 24 Motion referred to above. It continues to be our position that it is unrea'onable s to expect the Petitioners to file contentions in this raatter on the short notice provided by the Board's Order.

5. Subject to the above reservations of right, the Petitioners sublait Contentions 1 through 21, which are attached hereto and made a part of the Petitioh'to Intervene of the State of tiew IIaiapshire and Gregory 11. Srai t h , Attorney General.

Respectfully submitted, THE STATE OF IJEU llA!!PSHIRE A11D GREGORY H. S t1 I T H , ATTORt1EY GEITERAL By E. Tupper kinder Assistant Attorney General Environ:aental Protection Div.

Attorney General's Office State IIouse Annex Concord, New Haupshire 03301 Tel. (603) 271-3679 -

Dated: April 5, 1982 CERTIFICATE OF SERVICE I, E. Tupper Kinder, Esquire, hereby certify that a copy of the foregoing Addendula and Supplement to the Petition for Leave to Intervene and Request for llearing of the State of Iteu Haupshire and Gregory II. Srai t h , Attorney General of the State

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of ilev llampshire, has been mailed this 5th day of April, 1932, by first class mail, postage prepaid, to:

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• o elen Hoyt,' Chm. Dr. Emmeth A. Luebke Administrative Judge Administrative Judge Atomic Safety and Licensing Atomic Safety and Licensing Board Panel Board Panel U.S. NRC U.S. NRC Washington, D.C. 20555 Washington, D.C. 20555 Dr. Oscar H. Paris , Paula Gold, Asst. AG Administrative Judge Stephen M. Leonard, Asst. AG Atomic Safety and Licensing Jo Ann Shotwell, Asst. AG Board Panel Office of the Attorney General U.S. NRC One Ashburton Place, 19th Floor Washington, D.C. 20555 Boston, MA 02108 Lynn Chong Nicholas J. Costello Bill Corkum 1st Essex District Gary McCool 'Fhitehall Road '

Box 65 Amesbury, MA 01913 Plymouth, NH 03264 Roy P. Lessy, Jr., Esquire Tomlin P. Kendrick Office of Executive Legal 822 Lafayette Road Director P.O. Box 596 U.S. MRC Hampton, NH 03842 Washington, D.C. 20555 Robert A. Backus, Esquire William S. Jordan, II, Esquire 116 Lowell Street Ellyn R. Weiss, Esquire P.O. Box 516 Harmon & Uciss Manchester, NH 03105 1725 I Street, N.W.

Suite 506

. Washington, D.C. 20006 Phillip Ahrens, Esquire Assistant Attorney General State House, Station #6 Augusta, ME. -04333 Paul A..Fritzche, Esquire Donald L. Herzberger,.MD Public-Advocate Hitchcock. Hospital State House, Station #12 Hanover, NH 03755 Augusta, ME 04333 Wilfred L. Sanders, Esquire- Edward J. McDermott, Esquire Sanders and-McDermott Sanders and McDermott 408.Lufayette Road 408: Lafayette Road llampton, NH 03842 -Hampton, NH 03842

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Thomas G. Dignan, Jr., Esquire Senator Robert L. Preston Ropes and Gray State of ticu Hampshire 225 Franklin Street Senate Chambers Boston,14A 02110 Concord, !!H 03301 Docketing and Service Sec. Robert L. Chiesa, Esquire Office of the Secretary Wadleigh, Starr, Peters, U.S. NRC Dunn & Kohls Washington, D.C. 20555 95 Market Street Manchester, llH 03101 lis . Patti Jacobson 3 Orange reet Newburypot. 14A 01950 E. Tupper Kinder

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t A. TECilNICAL SAFETY CONTENTION

1. Interim Reliability Evaluation Program.

A thorough, plant-specific interim reliability evaluation program using probabilistic risk assessment techniques to find risk dominant sequences, consider multiple failures and assess the reliability of systems which may be called upon to mitigate an accident, is necessary to assure that the Seabrook Plant safety reviewNhss considered the I appropriate high-risk accident sequences to ensure coupliance with 10 C.F.R. 50.46.

Basis:

The accident at Three-Mile Island Unit II demonstrated that serious reactor core damage accidents are a real possibility and that consideration of core degradation and melting beyond design basis accidents should be conducted.

See, NUREG. 0660, NRC Action Plan Developed as a Result of the Three-Mile Island II Accident, May 20, 1980, p.II-1, NUREG.

0737 Clarification of Three-Mile Island Action Plan; Item I.C.l.

The Three-Mile Island Action Plan identifies the need for a systems-oriented approach to safety review. This approach, called Interim Reliability Evaluation Program (IREP),

uses probabilistic risk assessment' techniques to find risk-dominant sequences, considers multiple failutes, and assesses l

l the relicbility of systems which may be called on to uitigate F

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4 an accident but are not normally considered as ECCS. In order to recommend necessary plant specific safety improvements, this program should be implemented prior to the issuance of an operating license to ensure that modification of the facility to accommodate additional safety features is not foreclosed.

These additional safety features include items which are not designed in Seabrook, such as molten core retention devices and groundwater interdiction measures.

At present, the applicant h5'tnot s presented an l accurate assessment of the plant specific and site specific accident risks posed by operation of Seabrook contrary to the requirements of 10 C.F.R. 51.21 and 50.47. Neither the accident analysis presented in Environmental Report (ER) 1 nor Final Safety Analysis Report-(FSAR) 13 is adequate. Further, the NRC staff has recently indicated by letter of October 1, 1981 to the applicant that its evaluation of accidents and transients appear inadequate under the requirements of NUREG.

0737. - The FSAR discussion does not clearly identify the safety systems required to function to provide the safety actions necessary to mitigate the consequences of the transient or accident. Also, there is no discussion of long-term effects.of an. accident on plant operation.

The IREP--concept is now:being instituted by the probabilistic analysis staff of.the NRC. In all likelihood,

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this-type of evaluation-will be required for all operating A

6 reactors in the near future. Public Service Company itself has determined that a probabilistic risk assessment (PRA) should be i done and that "the PRA results could be useful for the ,

licensing hearings." (Seabrook Station Newsletter, Feb/ March 1982 at 4.)

The State of New Hampshire and the Attorney General contend that the Seabrook applicant should perform an i

integrated reliability evaluation prior to the operating license for the following key plant's'p'ecific systems:

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I sub-criticality systems, reactor core cooling systems, ECCS

, injection and recirculation systems, safety features actuation I

systems, and auxiliary systems on which these depend (alternating and direct current, compressed air, essential service water or cooling systems, and heating, ventilating, and air conditioninj systems). Particular emphasis should be given to determininc potential failures that could result from hunan errors, common causes, single point vulnerabilities, and test and maintenance outages.

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2. Systems Interaction.

The applicant has not performed an adequate analysis of systems interaction and thus, there is no assurance that the appropriate interactions, failure combinations and accident sequences have been considered in assessing the ability of the systems design to meet 10 C.F.R. 50 Appendix A. This contention relates to both the consideration of the interaction of safety and non-safety systems and the interaction and multiple failure of safety systems. "Ehere are systems and i

components presently classified as non-safety related which can have an adverse effect on the integrity of the core because they can directly or indirectly affect temperature, pressure, flow, and/or reactivity. The interaction between non-safety and safety systems may create demands on the safety systeus that exceed their design basis.- Not only must the applicant perform fully an analysis of systems interaction, but-also it must identify all systems and components which can either cause or aggravate an accident or can be called upon to mitigate an accident and thus should be classified as important to' safety and' required to meet.all safety grade design criteria.

Basis:

The staff and the applicant.ha~ not sufficiently accounted for systems ~ interactions in their-safety reviews'and.

conclusions. .The accident at Three-!!ile Island demonstrated that there are systems and components-which are. classified as

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non-safety related but which can have an adverse effect on the integrity of the core. This issue is discussed in Section 3.2

" systems design requirements" of NUREG. 0578, the Three-Mile Island II Lessons Learned Task Force Report (short term). The following quote from page 18 of the Report describes the

- problem:

There is another perspective on this question-provided by the TMI-2 accident. At THI-2, operational

problems with the condensate purification system led to a loss of feedwater and. initiated the sequence of i events that eventually. result'ed in damage to the' core. Several nonsafety systems were used at'various times in the mitigation of the accident in ways not considered in the safety analysis; for example, long-term maintenance of core flow and cooling with the steam generators and the reactor coolant pumps.

The present classification system does not adequately recognize either of these kinds of effects that nonsafety systems can have~on the safety of the plant. Thus, requirements for nonsafety systems may-be needed to reduce the frequency of occurrence of i~ events th'at initiate or adversely affect transients and accidents, and other requirements may be needed to improve the current capability for use i of nonsafety systems during transient or accident situations. In its work in this area, the Task Force will include a more realistic assessment of the interaction between operators and systems.

- Also, in UURUG. 0572, " Review of Licensee Event Reports," the ACRS found a~ number of systems interactions which only were revealed by'their actually occurring. Sign'ificantly,' with regard tofSeabrook, the examples cited ~were. Westinghouse reactors. The Three-Mile TslandLAction Plan, NURUG. 0660, Item 2.C.3. describes: approaches for1the_ analysis.of systems

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finteraction. (It is.noted that systems-interaction-has also been the subject.of-an unresolved safety issue-(A-17). See,

- UUREG. 0606.)

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.I Further, the Three-Mile Island Action Plan (11UREG.

0737, Item I.C.1) requires the applicant to further evaluate transients and accidents (see, Contention 1-IREP). In October of 1981, the staff notified the applicant that it had serious doubts concerning whether the applicant's present proposal could address the full range of initiating events and subsequent failures so as to comply with the requirements of 110 REG. 0737. The PSAR is inadequate to comply with 10 C.F.R. 1 50.34 and 50.46, and 10 C.F.R. 50 App 2ndix A. I l

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3. Class 9 Accidents.

The applicant has not presented, contrary to the requirements of 10 C.F.R. 51.20(a), (d), a complete assessment of the risks posed by the operation of Seabrook . The environmental report and the environmental impact statement should adequately address and evaluate the impact of a greater than design basis accident or " Class 9" accident on the environment. Unless the so-called " Class 9" accident is adequately considered, there can be no reasonable assurance that Seabrook can be operated without endangering the health and safety of the public.

Since the draft Unvironnental Impact Report was not available for the preparation of this contention, the State reserves the right to amend this contention at a later date.

Basis:

Reactor Safety Study (Wash. 1400) attempted to demonstrate that the actual risk from a class 9 accident is very low. Ilowever, the Comnission has stated that it "does not regard as reliable the Reactor Safety studies' numerical estimate of the overall risk of reactor accident" (llRC Statement on Risk Assessment and the Reactor Safety Study Report (Wash. 1400) in Light of the Risk Assessment Review Group Report, January 18, 1979). Given that the accident at Three-!!ile Island demonstrates that Class 9 accidents are not of such low probability that they need not be considered, such

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accidents must be thoroughly described and evaluated for environmental impact in the applicant's Environmental Report and in the NRC Environmental Impact Statement prepared under 10 C.F.R. 51. . Of course, the draft Environnental Impact Statement was not available when the contention was filed. The applicant's Environmental Report does discuan Class 9 accidents. (ER Chapter 7) However, this discussion is based in large part on the Wash. 1400 methodology which has been rejected. Additionally, the Environa'ntal e Report does nob consider the impact of human factors on the-probability of an event oCCurence.

The applicant and the NRC staff must accurately evaluate the probability of a Class 9 accident occurring at Seabrook and what aeasures can be taken to reduce that probability. Additionally,-the consequences of a Class 9 accident occurrence.at Seabrook and the measures which can be taken to mitigate those consequences must be evaluated. This is consistent with the requirements of NUREG. 0737 which requires an evaluation of the full range of accidents'-and t'ransients for a reactor site. ..The information contained in-the FSAR is insufficient to provide the assurance required'by-10-C.F.R. 50.40.-

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4. ' Anticipated Transients Without Scram (ATUS).

The applicant and the NRC staff have not demonstrated that the risk from an ATUS event is sufficiently reduced by interim measures to provide reasonable assurance that the Seabrook Station can be safely operated prior to the resolution of the generic issue.

Basis:

NUREG. 0460 (" Anticipated Transients Without Scram for-Light Water Reactors") set forth the~2taff position that "the reliability of current scram systems cannot be shown to'be adequate to meet the safety objective considering the rate at which these systems are challenged by anticipated transients" (p.39). The staff position articulated in 1973 (HUREG. 1270) was that the likelihood of nn ATUS event was acceptably small

-(10-6) given existing conditions, including the small numbers of reactors operating in 1973. Since that time, the number of reactors has increased dramatically.

The NRC's more recent evaluations of ATUS suggest-that the probability may be higher than 10-6, owing largely to the

-June 28, 1980 ATUS event at Brown's Ferry Unit 3 and the Lincreasing number of design basis transients-in operation

. nuclear reactors. The-applicant's FSAR regarding the Reactor cProtective-System does not include adequate provisions for reducing lthe; likelihood of an ATUS1 accident. Seabrook should' be required.to comply with any-forthcoming NRC-regulations for-t

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ATUS concerns. Additionally, operators should have training to recognize ATUS events (and have the necessary instrumentation to indicate such events) and to respond to ATWS events. FSAR 15.8, which relies on a Westinghouse response prepared prior to

, , 1974, is inadequate in addressing these points.

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5. Liquid Pathway Impact.

The applicant has not adequately considered the consequences of a nuclear accident resulting in releases of radiation and exposure to.the public by the liquid pathway, i.e., into groundwater which can contaminate aquifers, rivers, and streams. The failure to consider adequately liquid pathway accident impacts and corrective measures results in the inability to satisfy the standards of 10 C.F.R. 50.40 and 10 C.F.R. 51.21. ll* I Basis:

Core melt accidents which may result in liquid pathway.

releases must be carefully. studied. The NRC instituted a research program at Sandia Laboratories on this point which resulted in a report being released in Augustoof 1981. (Sandia Laboratories Study for-U.S. URC, "Effect of Liquid ~ Pathways on Consequences of Core Melt Accidents.") The dispersal of

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radiation through liquid pathways is more complex than the air

- pathway dispersal mechanisms because of the variations-in hydro-geological conditions and occause of the parameters of interaction between a molten-core and surrounding soil ~and 4

water table. However, interdiction 1or prevention.of liquid pathway-releases at the source may be possible-if adequate

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' design and . control'iaeasures .are taken. A nuclear-accident at

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Seabrook would have significant potential for liquid pathway.

contataination since radioactive materials may leak into .the

. groundwater.andiinto the estuarian and marine systems along the

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4 The FSAR 2.4 Hydrologic Engineering sets forth the hydrologic description of the site. However, in evaluating the impact of-a radioactive release, the FSAR considers only minor

.sp'lls and not a major release which would occur under major accident conditions. Under such conditions, dilution, which is apparently' relied upon for protection, may not be sufficient to protect the public. (See, FSAR 2.4.13.3) The Environmental Report (ER 7.4.1.2) does not study the liquid pathway because it was believed by the applicant to 6e slower than atmospheric pathways, and thus not important.

Evaluation of liquid pathway impacts should be undertaken before an operating license is issued, such that modifications or corrective measures can be implemented before such measures are foreclosed. Safety systems to flood runaway reactor cores with cooling water, " core catchers" to contain a melting core for several days, and interdiction mechanisms.

should be evaluated. The radiation dose to the nearby population should not be the only-factor in consideration. At-the Scabrook site, contalaination of the estuarian and marine systems through the liquid pathway would have an adverse economic'effect on the-New Hampshire seacoast and the State as a whole, regardless of the actual radiation dose to the human population. The information contained in the FSAR is inadequate to evaluate properly the impact of liquid pathway releases and modifications to_ mitigate the impact of such releases.

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6. Environmental Qualification of Safety-Related Equipment.

The applicant has not demonstrated that all equipment

'important to safety will be able to perform its function in the environment resulting from accidents. The environmental qualification of safety-related equipment at Seabrook.is-inadequate in four respects: (1) the parameters of the t

relevant accident environment have not been identified; (2) the length of-time the equipment must ope' irate in the environment has been underestimated; (3) the methods used to qualify the

-equipment are not adequate to give reasonable assurances that the equipment will remain operable; and (4) the effects of aging and cumulative radiation on the equipment has not been-adequately considered. All safety-related equipment must be demonstrated to be qualified to operate as required by Appendix A, G, and K of Part 50 and Criteria III and XI of' Appendix B, Part 50 and 10 C.F.R. 50.55a. In the absence of this demonstration, the standards by 10 C.F.R. Section 50.40 have not been satisfied.

Basis:

The accident at Three-14ile. Island demonstrated that the' severity'of the environment in which equipment important:to safety must' operate was underestimated and equipuent previously d deemed to be environmentally qualified. failed. .One. example was the pressurizer level. instruments. Further, the t1RC has-i i

proposed to amend its regulations to strengthen.the criteria

' for environmental qualification of electrical equipment. (NRC Proposed Rule 47, Fed. Reg. 2876, January 20, 1982.) In its Federal Register Notice, the NRC stated that " Experience has shown that qualification of equipment without test data nay not be adequate to demonstrate functional operability during design basis event conditions." See, 47 Fed. Reg. 2877. January 20, 1982.

Structures, systens, and co5p'onents important tol ^

safety must be' qualified to demonstrate their ability to withstand natural forces, such as earthquakes, and accident environment and still perform their safety functions. In analyzing the ability of'such equipnent to survive, insufficient account is taken of aging which can progressively weaken components. Brand new equipnent is tested and no systematic effort made to determine for how long the results are valid. For example, testing of reactor relief and safety, valves is essential for verification of their capability to perform safety-related functions. Without such testing, the_

applicant cannot demonstrate compliance with General Design Criteria:1,'14,-15, and 30. See, NURUG. 0737. Thus, the Commission cannot:say with reasonable assurance thatta sufficient margin exists to' maintain equipment qualification

- over_Lthe 40-year lifeLapan of1the plant.

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. Clearly, environmental qualification of safety-related equipment is critical to the safe operation of Seabrook. Until all. safety-related equipment has been demonstrated by appropriate analysis and testing to be qualified, the application does not comply with General Design Criteria 1, 2, 4,.21, and 23 of Appendix A. There is insufficient information to evaluate the overall adequacy of Seabrook's compliance with the environmental qualification requirements for safety-related equipment. ll - l b

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7. Instrumentation.

The Seabrook design does not provide adequate instrumentation to monitor variables and systems over their

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anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety. Included are those variables affecting the fission process, the integrity of the-reactor core, the reactor coolant pressure boundary, or the containment and its- associated systelas. Thus, the PSAR is.not in compliance with General Design Criteria 13 of Appendix A, 10 C.F.R., Part 50, and the requirements of NUREG. 0737, and i Regulatory Guide 1.97.

Basis:

The results of the investigation at the Three-Mile Island II accident indicated a need for more direct indications of low-reactor coolant levels, reactor-vessel water level, inadequate cooling, and hydrogen generation. The Three-Mile Island II accident also demonstrated the inadequacy of

_ post-accident monitoring in terms of the parameters' monitored, the range and accuracy of instrumentation, and the qualification of the instrumentation for the accident-and post-accident environment. The concerns over this issue were clearly pointed out by the Kemeney Commission in its Report of the' President's .Coramission on the Accident :at Three-Mile Island

.(1979) at page 72, 73. Instrumentation taust be ' considered

' safety-related as perhaps its greatest significance is.

Loperation under accident conditions.

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The lack of adequate instrumentation affects public

' health and-safety because accurate information is required by

- public officials to provide bases for decision-making related

. to emergency actions. The Seabrook design does not comply with

- Regulatory. Guide 1.97. " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant. Environs During and

- Following an Accident" (see, FSAR at 1.8-31).

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8. Hydrogen Control System.

The applicant has not demonstrated that in the event of a loss-of-coolant accident at Seabrook (1) substantial quantities of the hydrogen (in excess of the design basis of 10 C.F.R. 50.44) will not be generated; (2) that in the event of generation, the hydrogen recombiner can process adequately the hydrogen generated; and (3) that in the event of combustion, the containment and key safety systems within containment have the ability to withstand pressure, thereby preventing releases of off-site radiation in excess of Part 100 guideline values.

Basis:

The Three-Mile Island accident has raised a safety issue regarding hydrogen control measures following a loss-of-coolant accident. Sea, Metropolitan Edison Company (Three-Mile Island Nuclear Station, Unit 1, CLI-80-16, 11 NRC 674 (1930)). The Three-Mile Island accident resulted in hydrogen being generated far in excess of the hydrogen generation design basis assumptions of 10 C.F.R. 50.44. The hydrogen control issue is directly related to the question of coupliance with 10 C.F.R. Part 100. In effect, the applicant must demonstrate whether or not the generation and combustion of hydrogen and the following failure of reactor containment to withstand hydrogen combustion would result in public radiation exposure in excess of that permitted by Part 100. Since pressures from hydrogen explosion could threaten the structural

g, integrity.of. containment'and since purging of containment to release pressure may result in unacceptable leve).s of 1' -

radionuclides, a credible accident scenario exists with regard ~

to'the Seabrook plant involving hydrogen production resulting inLoff-site doses in excess of 10 C.F.R. Part 100 limits.

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- 9. Radioactivity Monitoring.

The'Seabrook design does not provide.an adequate program for monitoring the release of radioactivity to the plant and its environs either under normal operating conditions or in pre- and post-accident circumstances. Thus, the application'is not in compliance with General Design Criteria 63, 64 of Appendix A, 10 C.P.R. Part 50, and the requirements

-of IlUREG. 0737 and NUREG. 0800.

Basis:

The Three-Mile Island II accident demonstrated the inadequacy of post-accident monitoring in terms of the parameters monitored, the range and accuracy of instrumentation, and the qualification of the instrumentation for the accident and post-accident environment. The concerns over this issue were clearly pointed out by the Kemeny.

- Commission in its Report of the President's Commission on the Accident at Three-Mile Island at-page 75. The lack of adequate rtonitoring capacity' in terms of the- range of monitoring equipment and. the location and' number of monitoring sites places the health and safety of plant-personnel and the-public at'significant risk because such information is required by public officials to provide bases.for decision-making related to. emergency actions. Adequate radiation monitoring-for.

radioactivity'which may be. released due to anticipated

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operational occurrences'is necessary to adequately protect'the public health and safety..

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a p - The applicant must' provide sufficient radiation.

monitoring capacityLin containment spaces which could contain

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required by General Design, Criteria 64.

Additionally, the applicant must assure that the health physics division-at the plant is qualifled and properly staffed totperform its function. See, Regulatory Guide 1.97 " Instrumentation for

-Light Water Cooled Nuclear Power Plants to Assess Plant Environs Conditions During and Follb idg Accidents." I I

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10. Control Room Design. ,

" The control room design for the Seabrook Plant does

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potfprovide-adequate controls and-instrumentation to monitor variables _.as appropriate to comply with General Design' Criteria 13.. All~ operator actions necessary to take the plant from normal operation to cold shutdown should be capable.of being performed-_from the-control room. The control room.-panel must

'be adequate to provide the appropriate and necessary information to operators in the even6}'of an accident. I-Instrunentation must be provided for an adequate number of

. parameters and' additionally, that such instrumentation be

l environmentally. qualified. Further,=an adequate system must be-provided to inform the operator regarding the status _of safety

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Lsystems, _ i.e., whether a safety system.has been deliberately disabled.

Detailed' Control; Room Design' Review (DCRDR) should be carried out in conformance~with the guidelines of NUREG. 0700-

andEUUREG. 0737o(Item.l.D.1 and 2).. .

Finally,-the'Seabrook control-facility must be designed to provide adequate 7 equipment:outside the^ control room to promptly put the reactor in hot! shutdown and' maintain'it.

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-until' cold shutdown from:outside the control room as required-byL. General Design Criteria.19y 20, 21, and.22 of:AppendixEAlto

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Because the accident at Three-Mile Island II was substantially aggravated by the confusion created in the control ~ room due to inadequate instrumentation and monitoring of reactor core conditions, a thorough evaluation of the Seabrook control room design is essential. For example, at Three-Mile Island II, two auxiliary feedwater system valves were closed when they should have been open, and no adequate system was provided to inform the reactor operator that the safety system had been deliberately disabled. Any evaluation must necessarily consider the human factor in responding to an accident. The applicant must provide a system which meets the specifications of Regulatory Guide 1.47.

The purpose of DCRDR is'to assure that displays and controls added to the control room do not increase the potential for operator error. It is especially critical at Seabrook that accident-monitoring and control room design be the optimum because of the difficulties inherent in carrying out protective actions for the population in the immediate vicinity of the plant.

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11. Deviation from Current Regulatory Practice.

The applicant has not j'stified and the NRC staff has failed to require documentation for all Seabrook deviations from current. regulatory practices. The Seabrook facility, due to its long licensing history, has in many instances been reviewed by the staff against guides and standards which have subsequently been updated or modified. Neither the applicant in the PSAR, nor the NRC staff, has systematically described standards against which Seabrook has b,een reviewed and the

i. basis for acceptability of any deviations from any current regulatory practices. This circumstance iu not acceptable, particularly since the Board nust make findings based.upon the 1

applicable regulatory requirements.

2 Basis:

The Three-Mile Island II accident demonstrated the need for documenting deviations. It a.so demonstrated that past staff practices were not suitably conservative to protect the health and safety of the public. See, Kemeny Commission Report at 20, 35, 65-66. The applicant and the NRC' staff have (1) not documented in the FSAR where the Seabrook design structures and components do not conform with current

-regulatory practices (i.e., regulatory guides, branch technical positions,'and. standard review plans).and.the basis for an acceptability of those deviations, and'(2) not set forth in a Jaafety evaluation report the standard against which Seabrook

1

, has been reviewed and the basis for any deviations from current regulatory practices approved by the staff. Absent such documentation, there is no basis for a Board finding that the i

level.of safety equivalent to current regulatory practices does exist as required by 10 C.F.R., Section 50.34(b) and 50.40.

This is-of particular concern at the Seabrook site because of the difficulty in implementing emergency protective actions, should an accident occur.

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11 2 .. Quality Assurance.

The applicant has not demonstrated that a quality

assurance program for both the construction and operation phase has been established, executed and documented in accordance with'10 C.F.R. Part 50, Appendix B. It is not clear that

. adequate control measures have-been observed during construction of the facility to assure that unsuitable 1 materials, parts and equipment are not used, and that unsatisfactory construction practice'sifare promptly identified i and corrected. On at least several occasions during construction, there has been inadequate supervision of the construction work force and insufficient documentation of quality assurance tests. As a result of this failure, applicants have not demonstrated adequate assurance that a J

.particular structure, system, or component of the' facility.will perform satisfactorily.in service. Of particular concern are e compliance of construction with General-Design Criteria for (a).

-Protection and Reactivity Central Systems-(General. Design i

Criteria 20-29); (b) Fluid Systems (General. Design Criteria

30-46);'and-(c) Reactor ~ Containment'-(General Design Criteria 50-57).

Basis:

This contention is supported by, but..not limited to,

, Lthe followingLexamples of the-applicants-failure 1to meet-quality assurance' standards:-

,- ,v-.7 . - -.- , ,, 4 ~ ,

(a) In February of 1982, the applicant failed to discover'for several weeks faulty velding techniques used in the construction of the containment building by three employees of the contractor, Perrini Construction Company. This failure violates Paragraphs VII, IX, X, and XVI of Appendix B.

(b) In January of 1981 several parts of the-reactor were-delivered to the site by ship in such a way as to damage metal. internal reactor parts by exposing them to salt water.

This demonstrates a failure to contr'ol'Ishipping and preservation practices to prevent damage or deterioration to parts and materials as required by 10 C.F.R. Part 50, Appendix B,-Paragraphs IV and XIII.

(c) Because of the unusually large work force at the applicant's construction site, the applicant has not demonstrated that inspectors and supervisors were able to.

adequately control the quality of materials and labor used in-construction.

(d) On February 12, 1982, the NRC staff found that' the qualification requirements for the applicant's. Nuclear Quality Manager were unsatisfactory. See, FSAR 17.1-2.

Applicant luts f ailed to demonstrate the competence and training of~ quality assurance personnel as required by 10 C.F.R.

Part

.50,-Appendix B, Paragraphs I and XVIII.

13. Operations Personnel Qualifications and Training.

The applicant ~ has not demor.strated that. its operations

' personnel will have sufficient expertise, training, and

_ experience to test and operate the Seabrook Station under the standards of 10 C.F.R. 50.40 and-NUREG. 0660 and NUREG. 0737,

' the Three-!!ile Island Action Plan. As the accident at Three-Mile Island and subsequent investigations indicated,

~ staffing plants with fully qualified operations personnel is imperative for proper' operation of t$hI facility. -1 Operations 1-Personnel must be supplied with detailed _ training in analyzing and responding to emergency _ events. The applicant must demonstrate that the following, and all other operations personnel, are fully qualified and properly trained:

(a) Station llanager; (b) Assistant Station Manager; (c) . Senior Reactor Operator; (d) Reactor Operators; and (e) Shift' Technical Advisors.

1

-Basis:

~

The applicant has notLdemonstrated that on-going and I

~ future training programs for operations-personnel will

- adequately meet _the requirements specified inLNUREG. 0737, clarification of'the:Three-Mile Island-Action Plan 1

Requirements, Sections:I.A.l.1, I.A.2.1, I.A.2.3, II.B.4,Jand I.C.1 and FSAR 13.5.2. The. capabilities of:the training' center 4  %

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staff to provide instruction to reactor operators in technical engineering ~ topics have not been demonstrated. The applicant has made an inadequate showing that the shift' technical

- advisors will be adequately trained and qualified under the requirements of-NUREG. 0737, Appendix C. NUREG. 0737. requires further that the applicant develop a training program to teach the proper responses to accidents in which the core is severely damaged; this training has not yet been fully developed or On October 1, 1981, thelINRC Staff advised'the

~

implemented.

- applicant that.the_ staff "had serious doubts that the full range of initiating events and subsequent failures could be addressed ..." by the applicant's proposal for complying uith the Three-Mile Island Action Plan. Assuring that operations personnel have sufficient expertise, training, and experience to understand the initiation of an accident and to properly respond is necessary for safe operation of the plant.

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~ 14 . , Reliable Operation Under On-Site Emergency Power.

(Diesel Generator Units)

The Seabrook design does not adequately insure reliable operation in the event of loss of off-site power and a LOCA at the plant. The_URC staff has recognized that a

. genetic, unresolved, safety problem arises from the unreliability of emergency on-site diesel generators at pressurized-water reactors of the Seabrook-type. That problem is addressed in Task B-56 of NUREG.'Olh0. l The FSAR does not indicate compliance of the on-site power system with General Design Criteria-2, 4,.5, and 50 (see, FSAR 8.1).'

Basis:

The NRC staff has recognized the generic, unresolved,-

safety problems arise from-the unreliability of emergency 1

on-site diesel generators. Obviously, the unavailability of power sources assential to' emergency power-would create a

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severe condition foripublic health and safety in the event 'of-

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accident conditions. In' order to insure reliable 1 operation, the' diesel generator systeu must be supplied with high m'  : reliability controls and monitoring instrumentation _for temperature and pressure, forlits_ cooling water _ system.and:

e'ngine lubrication system.- Inability of theLgenerator' unit-to-.

-respond'on demand (in an~ emergency-situation could cause'aL serious problem,.Jand its status should be-known'at-all times.-

The applicant'.s'FSAR 9.5 fails?toladequately. address _ problems

< associated'with.' diesel; generator-reliability in1the eventLof.

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loss oftoff-site. power-and;in the-event of.a>LOCAp

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'15. ' Unresolved Safety Issues.

-The applicant and the NRC staff have not adequately addressed certain unresolved safety issues nor justified a substitute approach for resolving these issues with respect to the Seabrook facility and thus, have not complied with 10 C.F.R., Part 50, Appendix A, General Design Criteria 2, 4 and' the standards of 10 C.F.R. 50.40. As a requirement for the issuance of an operating license, the applicant must demonstrate'either that each applicab'i' le generic safety isbue has been resolved for the particular reactor or the existence of measures employed at the reactor to compensate for the lack of a solution to the problen. Virginia Electric and Power Co.

(North Anna Power Station, Units 1.and 2, ALAB 491_(1978). LA finding that each unresolved safety problem applicable to "Seabrook has been addressed must be made.

The NRC staff uust revieu,- in the SER, any unresolved' safety problems.which might have an impact on the operation of.

the Seabrook facility. ~ In its decision in Gulf States ~ Utility Co.-(River Bend, Units 1.and:2 , ALAB. 444 (1977), the Appeal

Board stated:

- To'this.end, in our. view, each SER'.should_contain.

a sunmary description of those generic problems under

. continuing study which havelboth relevance to-ff acilities. of the type . underT review - arid potentially.

'significant public natetyiimplications .

This sunuary should include 'inforn:ation:of the . kind now contained-

~in nost: Task? Action Plons. . More'specifically, there

.should:beLan indication cf.the~ investigative prograu which:has been'or uill be undertaken'with regard to the problem,: the . program's at:ticipated Ltime' span, m

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whether (and if so, what) interim measures have been devised for dealing with the problem pending the completion of the investigation, and what alternative

. courses of action wight be available should the program not produce the envisaged results ... [t]his assessment might well have.a direct bearing on the ability of the licensing board to make the safety findings required,of it ....

Since the Petitioners have not received the staff's SUR, we reserve the right to raise contentions regarding the adequacy of that report from the point of view of permitting a finding that there is a reasonable assurance that the operation of the facility can be conducted with'out't endangering the health and safety of the public as required by 10 C.F.R. 50.57.

The generic safety issues relevant to Seabrook are those in NUREG. 0410 which have been designated by the NRC-staff as applicable to all PUR's or all Westinghouse reactors and currently presented in NUREG. 0606. These issues include the following:

Task'No. A-1, Water Hamner

. (a)

(b) Task No. A-3, 4, 5, Pressurized Water Reactor Steam Generator Tube Integrity (c) Task No. A-ll, Reactor Vessel Materials Toughnness (d) Task No. A-12, Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports

'(e) Task No. A-17, Systems Interaction in Nuclear

-Power Plants (f) Task No. A-24, Qualification of Claus 1-E Safety

~Related Equipment.

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~(g ) ~ . Task ' No. : A-31, Residual Heat' Removal Requirements (11) -Task No. A-36, Control of Heavy Loads'near Spent i .

Fuel'

~:(i) Task- No. A-47, Saf ety Itaplications of Control Systems

- (j). Task No. A-48, Hydrogen Control !!easures and I Effect'of Hydrogen Burns on Safety Equipment i.

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16. Ultimate Heat Sink.-

The FSAR 9.2.5. describes the-ultimate heat sink complex which consists of the Atlantic Ocean and the' service water cooling touer. The system is designed such that the service water: cooling tower would serve as the ultimate heat

- sink.for.the facility.in'the event that cooling water from the Atlantic.0cean is unavailable. The FSAR'does not assure that an adequate water supply and heat sink capacity would be available to maintain the plant in sif'e* shutdown in the event of the. unavailability of the Atlantic Ocean heat sink. The-PSAR-does not comply with the standards of 10 C.F.R. 50.40 and

{ General Design Criteria 44 of 10 C.F.R. 50, Appendix.A in-that suitable-redundancy _is not provided.

j Basis:

General Design Criteria 44 and Regulatory Guide 1.27

~

i' require a cooling _ water safety system which would have-continuous capability to maintain the plant in'. safe shutdown.

The PSAR does not demonstrate that-tl'e mechanical ~ draft cooling tower, which is conmon to both Units I-andsII, has the ..

. capability.to perform this function. :The'NRC' staff has. -

i observed that'the tower-make-up water is insufficient'to ,

.iaaintain the ' plant- in , safe shutdown--fort thirty days' asirequired' 1

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by Regulatory,. Guide-l.27. . See,1NRC Staff Request for 7Additio.nal Information, 410.25-(February-. 12, 1982).

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17.. Environmental Impact.

The applicant in its Environmental Report'and the

' staff in its Final Environmental Impact Statement have not demonstrated compliance with the provisions of 10 C.F.R. 51.20 and 51.26 respectively. Based on the information available, the applicant has not shown that a monitoring and surveillance program will be established which is adequate to satisfy the requirements of 10 C.F.R. Part 50, Apper. dix I. Additionally,

,, 9 it is not clear that Criterion 60 through 64 of 10 C.F.R.Ipart.

50, Appendix A will be complied with. At the time this contention was developed, the NRC draft Environmental Impact Statement was not available. For this reason, petitioners reserve-the right to provide amended contentions on the issue of environmental impact when that document becomes available.

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- 18. Health and Environmantal Monitoring.

The applicant has not provided an adequate surveillance and monitoring program for releases of radioactive material which complies with the provisions of 10 C.F.R 50 App. I and 10 C.F.R. 51.2., Thus, the applic_ation does not satisfy the standards of 10 C.F.R. 50.40.

Basis:

An appropriate surveillance and monitoring program of radioactive releases is important, both from the public health_

and-safety 9erspective (requiredbylbC.F.R. 50.47, 10 C.IF .R.

50 Appendix I) and from.the envirennental' impact perspective (required by 10 C.F.R. 51.23).

As the State has set forth in-Contention 9, we do not believe that an adequate monitoring system is available to deal with a serious accident.- The utility has not provided an adequate system; and at present, neither the State or_ local  ?

governments have the resources to-provide the balance of what is needed. .The need is particularly acute at.Seabrook because of the proximity-of intensely used beach areas where protective actions will-be difficult to implement.

The failure to. provide adequate monitoring.of normal operational releases is as important as monitoring of accident s

releases. There must'be full evaluation of allipossibla sources of releases so that the cumulative impact is:

considered. The monitoring program must give due. regard ~for the impact-ofLreleases on-the natural systems. .Seabrook's location at.the. edge of an-estuarine system,: which isLvital to New Hampshire's coastalfresources,Lrequires the presence of'an efficient and dffective monitoring;andLcurveillance program. _

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.19 . Financial Qualifications.

The applicant has s not demonstrated reasonable assurance of its ability to obtain financing necessary to cover the: costs'ofLoperating and shutting down both Seabrook I and II p.

as required by 42 U.S.C., Section 2232(a); 10 C.F.R. Sections 50.33(f], 50.40, 50.91; and 10 C.F.R. Part 50, Appendix C.

Basis

/, 'The lieu llampphire Public.. Utilities Commission's (PUC),

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> 'decisionjof January 11, 1982, reflects adversely on the

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> applicant's financial co'hdition in as'm;uch'as it allowed a 17%

r' ate of return on,the' Utility's equity, the highest rate ever i s allowed a utility in this country. (Report of the Public I

UtilitiesCommission)bR81-87,Januaryll, 1982. In reaching 3

its decision, the PUC fdund that if the applicant's bond rating I fell to BB (Standard & Poor's), it would probably not be able

tto raise $1.3 billion over the next five years. -$1.3 billion is the to,tal amount that the applicant plans to finance during the next five years from the sale'of stocks and bonds.. One I week after the PUC's decision, Standard & Poor's' lowered the-

k rating of the applicant's general and refunding bonds to BB.

Therefore, there is a question.of whether the applicant will be -

able:to. raise-the revenue.which'it requires to complete 0

construction lof Unit II, as well as to operate and shut down U , '3

~both units.

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y 4 The applicant's recently lowered bond rating is

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[g evidence of its financial condition. A rating of BB (Standard

& Poor's) is'below the level recommended for investment. Only if two other utilities in the nation (United Illuminating of  ;

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Connecticut and-General Public Utilities of New Jersey) have

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L f ;- bond datings below investment level. The applicant's problems are the result of its unusually low earnings and negative cash

([ow. '(See, Wall Street Journal, January 18, 1982.)

The PUC has questioned whether the applicant's low

! Mi , l bond, rating and poor financial condition will enable it to compete with other utilities in the money market and,

~

therefore, to obtain the financing it needs for construction of Unit II. (Report of the Public Utilities Conmission, DR 81-87, 1 January 11, 1982) Although the-PUC authorized the sale of 3 i million shares of common stock,-it prohibited the-proceeds of the;cale from being used touard any new commitments related to Unit II. ,If, by June 11, 1982, the applicant is unable to sell 7% of its 35% ownership interest in the Seabrook project, the PUC has stated that it will order a delay of further work on

• Unit-II until 1984. (Report of the Public Utilities

, Connission, DF 82-63, !! arch 24, 1972) Despite the applicant's

^

efforts to divest..itself_of the 7% ownership interest since

< 1 January;1982,~it has received little response from either~its

-, co-owners in-the Seabrook project ~or other utility investors.

- As a result o'f this, the' applicant may not receive'further rate J

relief frou the PUC or be able to raise the revenue required to meet the applicant's own forecast of its needs. (Report of the Public Utilities connission, DF 82-63, March 24, 1982, pages 16-17)

The company's inab,ility to sell its ownership in the Seabrook facility is partially the result of growing competition for utility investment from hydroelectric suppliers. A spokesman for the New England Electrical Systems (NEES) has stated that their company does not need the 1

additional energy offered by the applicant from the sale of its share of the Seabrook project and that Necs expects a better return from its investment in hydroelectrical power than it would receive from further investment in Seabrook. (See, Manchester Union Leader, March 24, 1982, Page 1A.)

In 1976, it was estimated that the total cost of the Seabrook project would be $1.5 billion. Decision of the NRC Atomic Safety and Licensing Board, In the matter of:

Public Service Company of New Hampshire, et al. (Seabrook Nuclear

/

Generating Stations Nos. I and II); Docket Nos. 50-443, 50-444 (June 29, 1976). Current estimates now range from $3.56 billion to $7.3 billion. These additional costs, together with the other factors referred to above, raises a question of whether the applicant can provide a reasonable assurance that it will be able to raise sufficient revenue to couplete construction of Unit II, operate both units for five years, and deconuission the entire facility.

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The applicant has not deraonstrated that it meets the financial qualificiations of 42 U.S.C. Section 2232, 10 C.F.R. 50.40, and 10 C.F.R Part 50, Appendix C.

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B. EMERGENCY RESPONSE CONTENTIONS t

20. Emergency Assessment, Classification, and Notification.

The accident at Three-Mile Island demonstrated the inability of all parties involved to comprehend the nature of the accident as it unfolded; communicate the necessary information to one another, to the Federal, State, and local governments, and to the public in an accurate and timely fashion; and to decide in a timely manner what course to take to' protect the health and safety of the public. The applicant in these proceedings has not adequately demonstrated that it has developed and will be able to implement procedures necessary to assess the impact of an accident, classify it 1

properly, and notify adequately its own personnel, the affected governnent bodies, and the public, all of which is rcquired funder 10 C.F.R. 50.47 and Appendix E, and NUREG. 0654.

Basis:

In support'of these contentions, the State contends as follows:

(a) The emergency classification and action' scheme required:by 10 C.F'.R. 50.47(b)(4) and'NUREG.

0654, App. 1, and outlined in Section 9 of.

applicant's Emergency Plan is inadequate.

Initiating conditions should include the postulated accidents'in~the PSAR and Emergency 5

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Plan.. The Environmental Plan should state the basis for selecting a certain emergency action level. Further, the responsibilities of the Unit Shift Supervisor and the Shift Superintendent

! relating to Emergency Operating Procedures and Emergency Implementing Procedures must be more l clearly delineated.

(b) The applicant's Emergency Plan does not include l

provisions for the adequate, continual staffing j.

l required by 10 C.F.R. 50.47(b)(2) and UUREG.

0654, Table B-1.

! (c) The Emergency Plan fails to demonstrate l establishment of a mechanism for recommending protective actions by the applicant to the appropriate State and local authorities, or for.

prompt notification directly to the off-site

authorities responsible for implementing l protective measures within the' Plume Exposure Emergency Planning-Zone, as required by HUREG.

0654, Criteria-J.7, page 60.- This general process outlined'in Section 3 of the applicant's Emergency Plan will involve unnecessary delay in implementing protective actions.. Further, the

-Plan fails.to.provideffor prompt-notification directly to all-off-site authorities.

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t (d) The Emergency Plan fails further to set forth the required bases for a choice of recommended protective actions for the plume exposure pathway under emergency conditions as required by NUREG.

0654, Criteri,a J.10(m). The state contends that the applicant must specify in the Plan an adequate mechanism-for making protective action recommendations to appropriate State and local authorities, and must establish adequate bases for such recommendations.

(e) The Emergency Plan fails to establish that information will be made available to the general public on a periodic basis on how they will be notified and what their initial actions should be in an emergency, or that procedures for coordinated dissemination of information'to the public have been established, as required by 10 C.F.R.E50.47(b)(7), in that no provisions for any information dissemination to the public appear in the Plan at all, nor are there plans for. programs to acquaint-news media with the emergency plans.

The State of New llampshire and Attorney General-reserve the1right: to review such provisions and procedures' when-they are cubmitted by the applicant and to file contentions based on the information contained therein.

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21. Protective Action.

The State contends that the applicant's Emergency Plan does not demonstrate how, in case of an accident resulting in a site area or general emergency, the large numbers of people in the zone of danger may be protected er evacuated. Until there is reasonable aesurance that adequate on-site and off-site protective measures can and will be taken, the Board should not issue an operating lice..re.

Basis:

Under 10 C.F.R. 50.47(a) and (b), 10 C.F.R. 50, Appendix E, and 11UREG. 0654, the applicar.t must demonstrate the adequacy of on-site and off-site protective measures in the event of a radiological emergency. The applicant has not deuonstrated with reasonable assurance the adequacy of the necessary protective measures on the basis of the following:

(a) At predent, the Energency plan does not contain any off-site preparedness plans of State or local euergency response organizations. Therefore, the applicant has not demonstrated compliance with 10 C.P.R. 50.33(g), which requires such plans. In the present applicant, there exista no basis upon which a finding under 10 C.P.R. 50.47 can be made that (1) the state of off-site emergency preparedness provides reasonable assurance that adequate protective measures can and will be

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o eP #1 taken in the event of a radiological emergency, l

and (2) that State and local emergency plans are

- adequate . and capable of being irapleraented.

The emergency plans for the State and localities of !!ew Hampshire are still being developed. The State suggests that this contention be adraitted for discovery purposes with additional specifics to be required af ter the plan has been raade available.

(b) The Emergency Plan does not deraonstrate that adequate arrangements have been, or will be, made for raedical service for contaminated injured individuals, as required by 10 C.F.R. 50.47(b)(12). Sections 10.4.4, 10.5.1, and 10.5.2 of the FSAR explain sirap1y that arrangements for medical care and transportation-of injured personnel have been made. The PSAR.

'does not sufficiently demonstrate how injured.

personnel'will be treated o'r the' adequacy of medical services that have been " arranged."

(c) Applicant-has not demonstrated _in its-Emergency-

~

-Plan whether,Lin case oflan accident, Lit will:be

possible'to protect or evacuate:the large numbers of people -who raay be' within the zone of danger at.

any 9iven moment. The applicant's. Emergency Plan' m -

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t also does not indicate sufficiently upon what bases protective action decisions will be decided b .and how protective actions will be implemented.

Until the. applicant has developed its Emergency

-Plan further and shown that it is able to and jf will implement adequate protective measures, it will not be in compliance with 10 C.F.R. 50.47(b)(10) and IIUREG. 0654 which require that protective actions be developed and in place.

To the extent that the adequacy of the applicant's emergency response measures depends on the development of the State and. local cuergency plans, the State.of liew Hampshire reserves the right to file contentions based upon the nformation contained in these plans after they are subnitted.

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22. Eraergency Planning Zones.

Applicant's acceptance without formal analysis or evaluation of circular 10- and 50-mile radius for the Emergency Planning Zones does not discharge the applicant's responsibility to ensure that adequate emergency response plans exist to protect the public health and safety in the event of an emergency at Seabrook. See, Section 4.3 of the Emergency Plan. Designation of circular 10- and 50-mile Emergen;y Planning Zonen is unjustified because such Emergency 21anning Zones does not consider local emergency response needs as they are affected by such factors as demography, topography, land characteristics, access routes, and jurisdictional boundaries.

Basis:

The types of issues to be considered in setting the size and shape of the Emergency Planning Zones are specified in 11UREG . 0654, 10 C.F.R. 50.33 and 50147(c)(2), and Appendix E, III. Spe,cifically, the applicant has not considered adequately the effect of the following factors specific to Seabrook on local emergency response needs and capabilities, and hence, on the appropriate size and configuration of the Seabrook Emergency Planning Zones:

(a) The proposed circular 10-mile Emergency Planning Zone does not account adequately for jurisdictional boundaries; evacuation of only a portion of many jurisdictions will lead.to severe confusion and inadequate preparation for an emergency.

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(b) The 10-mile circular Emergency Planning Zone does not take into account unique factors within the region, such as the rural-urban mix, automobile ownership, ownership of campers, vans and second homes, available public transportation, proportion of the population confined to institutions, location of friends and relatives, age of the population and proportion of children in the population, obstructions in the transportation network, the extent of cooperation among local governments, and the attitude of people toward evacuation. At various locations within and outside of the now-designated Emergency Planning Zone, these factors will vary greatly. A basic tenet of successful emergency planning is to avoid trying to change established patterns, attitudes and habits, to which people Will cling during an emergency. All of the above factors must be investigated in detail and tahec into consideration in deciding how large and what shape the Plume Exposure Emergency Planning Zone should be.

9

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