ML20214P463

Proposed Findings of Fact & Conclusions of Law Re Onsite Emergency Planning & Safety Issues

ML20214P463
Person / Time
Site:	Seabrook
Issue date:	11/26/1986
From:	Perlis R, Sherwin Turk NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:
Shared Package
ML20214P447	List: ML20214P444 ML20214P463 ML20214P482
References
OL-1, NUDOCS 8612040223
Download: ML20214P463 (36)

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g 11/26/86 COLKETED UNITED STATES OF AMERICA '

NUCLEAR REGULATORY COMMISSION

'86 DEC -1 A10:43 BEFORE TIIE ATOMIC SAFI;TY AND LICENSING BOARD GFFiCt E ,.

00CKET1.% t In the Matter of )

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PUBLIC SERVICE COMPANY OF ) Docket Nos. 50-443 OL-1 NEW HAMPSHIRE, et al.

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) 50-444 OL-1

) On-site Emergency Planning (Seabrook Station, Units 1 and 2) ) and Safety Issues NRC STAFF PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW Pursuant to the Licensing Board's order issued during evidentiary hearings on October 3,1986 (Tr.1024-26), the NRC Staff hereby flies its proposed findings of fact and conclusions of law. II In accordance with the Licensing Board's instructions (Tr.1025), the Staff identifies below the Applicants' findings with which the Staff does not really disagree.

As to all other matters, the Staff's proposed findings and conclusions of law are set forth seriatim below.

I. PROCEDURAL HISTORY As a general matter, the Staff has no quarrel with the procedural history set out by Applicants in their proposed findings at TT 1-25. The Staff does request that the following additional findings be made:

-1/ The Staff has requested a two-day extension of time for the filing of these proposed findings and conclusions of law , due to an unprecedented fcilure of the Staff's word processing equipment which (FOOTNOTE CONTINUED ON NEXT PAGE) 8612040223 861126 PDR ADOCK0500g3

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0 23a. On July 25, 1986, Applicants filed a motion for summary disposition of SAPL Supp. 6. The Staff filed a response in support of Applicants' motion on August 18, 1986; SAPL filed responses in opposition on August 4 and 28,1986.

23b. In its response to Applicants' motion, the Staff identified 11 items with respect to the Seabrook SPDS that might require corrective action, but further took the position that the uncorrected items did not pose a threat to the safe operation of the plant and that final resolution of the 11 items could safely await the completion of the first refueling outage. Staff Response at 12-14; Eckenrode Affidavit (attached to Staff Response) at i 15. The Staff also pointed out that no party, including 4

SAPL, had ever pointed to any adverse safety effect that might result from deferring resolution of these items until the completion of the first refueling outage. Staff Response at 14, 23c. In denying summary disposition of that portion of Applicants' motion that involved the SPDS, the Licensing Board agreed with the Staff that SAPL had never identified any safety problems that might result from deferring resolution of the 11 items, but went on to schedule for hearing the very issue of whether deferral of resolution of the identified items would have an adverse effect on the public health and safety.

Memorandum and Order of September 15, 1986, LB P-86-30, Slir Op. at l 16, 18, i

(FOOTNOTE CONTINUED FROM PREVIOUS PAGE necessitated rewriting much of this document and re-typing it in full. To our knowledge, the Board has not yet ruled upon the Staff's motion.

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II. CONTENTION NECNP I.B.2 Electrical Equipment Environmental Qualification Time Duration

1. This contention reads as follows:

"The Applicant [s have] not satisfied the requirements of GDC 4 that all equipment important to safety be environmentally qualified because [they have] not specified time duration over which the equipment is qualified."

2. At the August 1983 hearings, witnesses were presented on this contention on behalf of the Applicants (Messrs. Maidrand and Anderson),

App. Dir. , ff.1-Tr. 970, and the Staff (Messrs. Lagrange and Walker), <

Staff Dir. , ff. 1-Tr. 990. In the 1986 hearings, the Applicants offered a panel of witnesses composed of J.M. Salvo, G.S. Thomas and N.K.

! Woodward. App. Dir. ff. Tr. 357. The Staff submitted further testi-many by Mr. Walker. Walker, ff. 494. No other witnesses were offered by any party.

3. General Design Criterion (GDC) 4,10 C.F.R. Part 50, Appendix A, requires as follows:

Criterion 4 -

Environmental and missile design bases.

Structures, systems , and components important to safety shall be designed to acecmmodate the effects of and to be compatible with the environmental conditions associated with normal operation, irtintenance, testing, i

and postulated accidents, including loss-of-coolant accidents. These structures, systems, and components

shall be appropriately protected against dynamic pipe whipping, and discharging fluids, that may result from equipment failures and from events and conditions outside the nuclear power unit.

l 4. 10 C.F.R. I 50.49 specifies the requirements that must ba met to demonstrate compliance with GDC 4, relating to the environmental quali-fication of electrical equipment important to safety that is located in a l

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t potentially harsh environment. In conformance with 10 C.F.R. I 50.49, electrical equipment may be qualified in accordance with the acceptance criteria specified in Category I of NUREG-0588. In addition, guidance as j l

to the means by which 10 C.F.R. I 50.49 may be satisfied is provided in Reg. Guide 1.89. Walker, ff. 494, at 2. The Board took official notice of these two publications. Tr. 676.

5. The Staff's review of the Applicants' environmental qualification program led it to conclude that the Applicants have included in that program all electrical equipment important to safety that is located in a potentially harsh environment and required to be environmentally qualified under GDC 4 and 10 C.F.R. I 50.49. Walker, ff. 494, at 2-3; SSER 5 (i,taff Ex. 5), 5 3.11.3.1.

6. The staff's review further demonstrated to the Staff's satisfaction that the Applicants have specified the qualified life for electrical equipment important to safety that is located in a potentially harsh environment. The qualified life for such equipment includes pre-accident, accident, and post-accident time durations. App, Dir. ff.

1-Tr. 970, at 8; Walker, ff. 494, at 3-4.

7. The Applicants indicated that all such equipment is qualified to be run under pre-accident (that is, normal) operating conditions without losing its ability to withstand a harsh environment after a postulated accident has occurred, (a) for the lifetime of the plant, or (b) for shorter periods for particular items of equipment, in which case such equipment is to be replaced, refurbished or requalified before the period elapses. App. Dir. ff.1-Tr. 970, at 8; Walker, ff. 494, at 3-4.

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8. With respect to post-accident time durations, which is the focus of this contention, Commission guidance typically provides that electrical equipment important to safety that is located in a potentially harsh environment should be environmentally qualified for a period of 100 days after an accident. In some instances, regulatory guidance allows for post-accident qualification periods of substantially less than 100 days; in those instances, an applicant is expected to satisfy the post-accident time margin requirements specified in Position C.4 of Reg. Guide 1.89, Revision 1. See SSER 5 (Staff Ex. 5), 8 3.11.3.4, at 3-24; Walker ff. 494, at 5-6; Walker, Tr. 499.

9. The Applicants indicated that all electrical equipment important to safety that is located in a potentially harsh environment is environmentally qualified to perform its required safety function for a period of one year following the occurrence of a postulated accident, with the exception of 11 equipment item types which were determined to be qualified for periods of less than one year. The Applicants' selection of a one-year post-accident qualification period is very conservative, as compared to the typical regulatory guidance provision for a post-accident time duration of 100 days. See Walker ff. 404, at 5-6; Walker, Tr. 499, "23A.

16. The 11 equipment item types which the Applicants determined to be qualified for periods of less than one year following an accident are set forth in SSER 5 (Staff Ex. 5), Table 3.1, at p. 3-34. Walker, ff.

494, at 4; Tr. 493-94. These 11 equipment item types, the respective periods of time for which they are required to be and have been T

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,s determined to be environmentally qualified following the occurrence of an accident, are reproduced from SSER 5, Table 3.1, as follows: 2_/

Qualified Time Post- required accident to perform EQ file operating safety number Description time function 174-00-01 Foxboro transmitters 100 days 90 days 252-38-01 ASCO temperature switches 160 days 30 days 600-01-04 Raychem 8-ky motor connector 100 days 30 days kits 173-05-03 Maisonellan E/P converter 204 days 100 days 248-36-01 Borg-Warner Feedwater 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Approximately isolation valves 1 minute 600-06-01 NAMCO EC 210 series conduit 318 days 30 days seals 113-03-01 Okonite 600-V power cable

• 30 days 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 113-17-01 Anaconda 600-V control cable

• 30 days 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 113-18-01 Anaconda 300-V instrumentation 30 days 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> cable

• 113-20-01 ITT Suprenant 300-V instrumen- 30 days 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> tation cablet 172-01-01 General Atomic Radiation 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Approximately Monitors and Detectors 2 minutes (inside containment)

• Only the cable that is subject to submergence is limited to 30 cuys' qualification. This applies to all cable listed here.

2_/ Table 3.1 of SSER 5 included one additional equipment item type which is not required to be environmentally qualified at Seabrook, and it has been deleted from the Table as it appears here. See Walker, Tr. 493-94.

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11. The Applicants determined that these 11 equipment item types are environmentally qualified for the period of time required for them to perform their safety functions plus a margin of at least one hour, as specified in Position C.4 of ReF. Guide 1.89, Revision 1. Walker, ff.

494, at 4-5; SSER 5 (Staff Ex. 5), 6 3.11.3.4, and Table 3.1. This determination was based upon testing and engineering analyses, which demonstrated that each of these item types is qualified for time periods longer than the actual required post-accident operability duration in the harshest environment that they will face. App. Dir. ff. 357, at 8-12, 13-14, and 16-17; Walker, ff. 404, at 6. b

12. The Staff reviewed the Applicants' determination that the electrical equipment important to safety at Seabrook which is required to be environmentally qualified under GDC 4 and 10 C.F.R. I 50.49, is 4

qualified for a period of one year following a postulated accident or, in the alternative, for the time required to perform its safety function plus Qt margin , as specified in Position C.4 of Reg. Guide 1.89, Revision 1.

The Staff found this to be acceptable, based upon its reviews of the 1

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Following the conclusion of the Staff's review as presented in SSER 5 (Staff Ex. 5), the Applicants conducted a detailed examinaticn of the operability requirements for Foxboro transmitters and Raychem 8-kv motor connector kits. The Applicents determined that these items are not required to face as harsh an environment as had previously been believed, and that they are actually qualified for a one-year l period following an accident. App. Dir. ff. 357, at 12-13, 15-16.

Documentation as to this longer period of qualification will be placed in the EQ files and will be available for Staff audit in the future.

Salvo, Tr. 477. This determination does not alter the Staff's conclusion that the equipment in question is environmentally qualified in accordance with Commission regulations. Walker. Tr. 721-23 A.

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s Applicants' submittals and the results of the Staff's audit of the Seabrook environmental qualification program. Walker, ff. 494, at 5, 6.

13. The Staff normally conducts an EQ audit only when it is satisfied, on the basis of a pre-audit review, that the program is sufficiently complete to be approved. The Staff does not require an EQ program to be 100 percent complete prior to conducting the audit. The determination as to when a program is ready to be audited is a matter within the discretion of the individual Staff reviewer, and is based upon the nature and significance of the identified deficiencies rather than the number of such deficiencies. Walker, Tr. 702. The Seabrook EQ audit was conducted when its program was about 75 or 80 percent complete.

The program was substantially complete at that time, sufficient to permit the audit to be conducted. Walker, Tr. 692-93, 696, 698, 701-02.

14. Prior to conducting the Seabrook audit, the Staff met with the Applicants and discussed the deficiencies which the Staff had identified during its pre-audit review. The Staff itientified a number of such deficiencies, all of which the Applicants agreed to correct. Walker, Tr.

696. One such deficiency, which affected many of the files, involved the

, manner in which the Appifcants applied the Arrhenius equation (establishing the demonstrated post-DBA operability times for certain equipment) . Walker, Tr. 679; NECNP Ex.12, at 6; Woodward, Tr. 482.

At the Staff's request, the Applicants reviewed each of the EQ files in which the equation was utilized, and made corrections wherever the equation was not applied as the Staff required. Walker, Tr. 682-83, 716.

15. The Staff conducted its EQ audit at Seabrook in February 1986 in conjunction with the Idaho National Engineering Laboratory ("INEL").

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s Walker, Tr. SIC, 677. During the audit, the Staff reviewed 12 EQ files, six of which were found to contain deficiencies. Walker, Tr. 676. For certain files, the Staff found that additional information or clarifying language was required to be added to the files , and this was subse-quently done by the Applicants. Walker, Tr. 685-88. None of these deficiencies led to detrimental findings by the Staff with respect to I

pre-accident or post-accident qualification time durations. Walker ff. 494, at 7; Walker, Tr. 517. The Applicants subsequently resolved all outstanding items which were identified during the audit. Walker, Tr. 709.

16. The Staff has concluded that the Applicants satisfactorily specified the time duration over which electrical equipment important to safety that is located in a potentially harsh environment is environ-mentally qualified, in conformance with GDC 4 and 10 C.F.R. I 50.49.

Walker ff. 494, at 7. Accordingly, the Staff has determined that the Seabrook environmental qualification program satisfies GDC 4, as implemented by 10 C.F.R. I 50.49, with respect to the time duration over which the electric equipment important to safety is environmentally qualified. Walker, ff. 404, at 5.

17. NECNP did not present any direct testimony on this contention, but sought to discredit Applicants' and Staff's testimony through cross-examination.

18. With regard to Anaconda Sky power cable (EQ File No.

113-01-01), NECNP sought to demonstrate that the cable was not environmentally qualified, on the grounds that the manufacturer's qualification test involved a single conductor cable rather than the 1

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• l The single conductor cable multiple conductor cable used at Seabrook.

tested by the manufacturer was described as being " exactly similar" to the single conductor of the multiple conductor cable used at Seabrook.

Woodward, Tr. 367; NECNP Ex.1, Reference 6, Attachment at 2.

19. While the term " exactly similar" is somewhat ambiguous, the two cable types are sufficiently similar for environmental qualification purposes such that the test results adequately represent the test results that may be expected for the Anaconda Sky cable at Seabrook.

Woodward, Tr. 369. See Walker, Tr. 503-04. U

20. With respect to ITT Suprenant instrument cable (EQ File No.

113-19-01), the evidence indicated that one of these, RG-58 coaxial instrument cable, had not been subjected to qualification testing but had been qualified only by comparison to other, tested cable. Woodward, Tr.

378-79; NECNP Ex. 4, Reference 4. NECNP sought to demonstrate that the manufseturer's reason for not testing the RG-58 cable was based only upon economic considerations, and that no other valid reason existed for not conducting the test. See NECNP Ex. 4. Reference 4.

4/ NECNP unsuccessfully sought to introduce two reports prepared by Sandia National Laboratories under contract to the NRC. These were (1) NUREG/CR-3538, SAND 83-1258, "The Effect of LOCA Simulation -

Procedures on Ethylene Propylene Rubber's Mechanical and Electrical Properties" (October 1983), and (2) NUREG/CR-4536, SAND 86-0450, "Superheated-Steam Test of Ethylene Propylene Rubber Cables Using i

a Simultaneous Aging and Accident Environment" (June 1986).

Whatever relevance there may be between the first of these reports I and Seabrook was not demonstrated by NECNP (See Tr. 370-71); and i the second of these reports was shown to be irrelevant, in that it involves cables with different materials than those used at Seabrook.

l Woodward Tr. 373-75. In any event, the two reports were inadmissible and were excluded from evidence for lack of a proper sponsoring witness qualified to submit to cross-examination on their j contents. Tr. 472.

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21. There was no evidence to dispute the manufacturer's deter-mination that the tested cables are representative of the cable types used at Seabrook. Id. Further, the Applicants conducted an independent review of the manufacturer's environmental qualification submittal, and determined that the tested cables were sufficiently similar to the Seabrook cables to adequately represent the expected performance of the cable at Seabrook. Woodward, Tr. 380-83.

22. With respect to ITT Suprenant 300-V instrument cable (EQ File No. 113-20-01), NECNP sought to demonstrate that the cable was not qualified to withstand submergence for 30 days following an accident, contrary to Applicants' representation. See Woodward, Tr. 384. NECNP pointed to the Impell Corporation's determination that instrument rack MM-IR-12 was not qualified for submergence, and should be relocated l

I above flood level unless it can be shown that operability for accidents associated with submerged conditions is not required. The affected equipment included items connected to the ITT 300-V cable. NECNP Ex.

5, Reference 12, Attachment at 4-6; Woodward, Tr. 386.

23. The Applicants did not relocate the instrument rack in question, even though they determined that it may encounter submerged conditions following an accident. The Applicants determined that the r

equipment performs no safety function related to mitigating the accident or putting the plant in a safe condition, and its failure following the accident would not affect the safety of the plant. Woodward, Tr. 386-87.

Accordingly, the instrument rack was downgraded from operability Code A (required to perform a safety function during the accident) to operability Code C (not required to perform a safety function during the accident).

Salvo, Tr. 387-88. The Applicants' EQ file contains supporting documentation justifying this change. NECNP Ex. 5, Reference 16; Salvo, Tr. 388-91.

24. NECNP also sought to demonstrate that the 300V cable had not been tested for submergence for 30 days, and therefore was not shown to be qualified. There is no question that the cable ws.s not tested under submerged conditions for 30 days. Rather, it was thermally aged and irradiated, and then subjected to pressurized steam, high temperature, chemical spray, and humid environments, after which it was submerged briefly. Woodward, Tr. 395-402, 405-06; NECNP Ex.5, Reference 14 l at 1. The test conditions summarized a worst case environment, combining the anticipated environmental conditions fer a loss-of-coolant (LOCA) accident with those for a steam line break. Woodward, Tr. 397.

These conditions exceed those which the cable would be exposed to in the plant. Woodward, Tr. 399-402. On the basis of the tests conducted, the i Applicants concluded the cable may be expected to function in a submerged conditions for 30 days. Woodward, Tr. 396, 400-02.

25. No evidence has been presented before the Board which would indicate that the Appl.eants' conclusions are invalid. On the contrary, Staff witness Walker confirmed that environmental qualification for l

submergence may be shown either by submergence testing or by a demon-stration that the equipment in question can be operated in a submerged condition. Walker, Tr. 501, 504-05; see generally NUREG-0588, Revision 1, 12.2 at p. 12. The E3ard accepts Applicants' determination that the cable has been demonstrated to be qualified for submergence for 30 days.

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26. NECNP also challenged the submergence qualification for Transamerica Delaval water level transmitters (EQ File No. 174-15-01).

The Applicants initially determined that these transmitters are qualified for a one year period of subniergence based , in part , upon a submergence test of 30 minutes and use of a silicone-filled riser device intended to prevent moisture intrusion. Woodward, Tr. 429. The Applicants subsequently determined that the 30-minute test was inadequate to support the one year period of qualifleation, and the device was then redesigned to be qualified for one year. Woodward, Tr. 480.

27. NECNP sought to demonstrate that an unsupported engineering change authorization had been made for the riser device , deleting a pressurized leak test prior to its installation. NECNP Ex. 7. Reference 7; see Salvo, Tr. 430-31.

28. The pressure test was found to be unnecessary since it was Intended to test for pressure conditions that would not be present in an o

accident scenario; and it was determined that a visual examination of the installed device for leaks would suffice. Thomas , Tr. 435; Salvo ,

Tr. 480-81, 483. Pressure testing of splice boxes on the transmitter's level element was deleted upon determining that it could potentially damage the equipment, and that visual inspection would be adequate to verify the leak-tightness of the system. See NECNP Ex. 7 Reference 7, at 2; Salvo, Tr. 431, 450-51,

29. No reason has been provided to the Board to discredit the Applicants' testimony concerning these matters, nor has it been shown

that pressure testing of the equipment in question was deleted without sound engineering justification. -

30. On the basis of the evidence presented, the Board is satisfied l

that 'he Applicants have adequately specified the environmental qualification time durations for electrical equipment important to safety at Seabrook.

III. CONTENTIOliS NECNP III.1 AND NEW HAMPSHIRE 20 Emergency classification and action level scheme

31. Both New Hampshire and NET raised contentions addressing the adequacy of Applicants' emergency classification and action level scheme. NECNP Contention III.1 states:

The emergency plan does not contain an adequate emergency classification and action level scheme, as required by 10 C.F.R. 550.47(b)(4) and NUREG-0654, in that (a) No justification is given for the classifi-cation of various system failures an unusual events, alerts, site area emergencies, or general emergencies.

(b) The classification scheme minimites the potential significance of transients.

I 5/ NECNP also challenged the level to which the riser was filled with oil, asserting that the flood level of the plant had been recalculated without a corresponding change in the oil fill level; this allegedly calls into question the riser's ability to withstand the pressure head which might exist under flood conditions. See NECNP Ex. 7, Reference 7, at 3, 5; Woodward, Tr. 454. IIowever, the Applicants satisfactorily explained that the riser was filled with oil to the specified level (the top of a " tee" configuration), and the device was properly qualified regardless of the change in flood level.

Woodward, Tr. 455, 457.

(c) The Applicants' classification scheme fails to include consideration of ~ specific plant circumstances, such as the anticipated time lag for evaluation due to local problems.

(d) The classification scheme fails to provide a reasonable assurance that Seabrook onsite and offsite emergency response apparatus and personnel can be brought to an adequate state of readiness quickly enough to respond to an accident.

(e) The emergency action level scheme fails to identify emergency action levels or classify them according to the required responses.

(f) The scheme is incapable of being imple-mented effectively to protect the public health and safety because it provides no systematic means of identifying, monitoring, analyzing, and responding to the symptoms of transients and other indicators that transients may occur.

New Hampshire Contention 20 states:

The accident at TMI 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 tahc to protect the health and safety of the public. The Applicants in these proceedings have not adequately demonstrated that they have developed and will be able to implement procedures necessary to assess the impact of an accident, classify it properly, and notify adequately their own personnel, the affected government bodies, and the public, all of which is required under 10 C.F.R. 550.47 and Appendix E and NUREG-0654.

32. The applicable legal requirements pertaining to emergency classification and action level schemes are set forth in 10 C.F.R.

I 50.47(b)(4) and 10 C.F.R. Part 50, Appendix E. Section 50.47(b)(4) requires that emergency plans meet the following criterion:

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(4) A standard emergency classification and action level scheme , the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local

, response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures.

Appendix E, in pertinent part, states:

B. Assessment Actions The means to be used for determining the magnitude of and for continually assessing the impact of the release of radioactive materials shall be described, including emergency action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The ert.ergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring. These emergency action levels shall be discussed and agreed on by the applicant and State and local governmental authorities and approved by NRC. They shall also be reviewed with the State and local governmental authorities on an annual basis.

33. The Licensing Board held evidentiary hearings on these contentions on August 19 and 23, 1983. The Applicants ct that time presented as witnesses on these contentions George S. Thomas, James A.

MacDonald , and Peter L. Anderson. Mr. MacDonald prepared the emergency classification scheme for Seabrook; Messrs. Thomas and Anderson reviewed and provided comments on the scheme.

1-Tr. 1481-82. Their direct testimony on this issue is incorporated into 1

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the record ff.1-Tr.1482. b The Staff presented as its witness in 1983 John R. Sears; Mr. Sears was then the NRC staff reviewer of the Seabrook classification scheme. His direct testimony is incorporated into the record ff.1-Tr 1891. Neither NECNP nor New Hampshire presented any witnesses on these contentions; both parties relied on cross-examination for their cases.

34. After the Board reopened the record with respect to these contentions (see Applicants' Proposed Findings, i 22), the Staff and Applicants filed additional testimony for the reopened hearing. At the hearing, NECNP and New Hampshire indicated they had no cross-examination on the additional testimony; the testimony was incorporated into the record by stipulation and no examination of the witnesses was conducted. See Tr. 485 g seq. The Applicants additional testimcny was prepared by Messrs. MacDonald and Thomas and was incorporated into the record ff. Tr. 487. The Staff testimony was prepared by its later reviewers of the Seabrook classif! cation scheme, Donald J. Perrotti and Gorden R. Bryan; their testimony was incorporated into the record ff.

Tr. 489.

35. An emergency classification and action level scheme is designed to enable responsible personnel in the control room to recognize and declare an emergency of a particular category or severity so that onsite l and offsite emergency response organizations can be contacted and so that l

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In this section of the findings, the direct testimony submitted by l Applicanta in 1983 is cited to as " App. Dir.", Applicants' 1986 i testimony is referred to as " App. Dir.1986." Similar terminology is used for the Staff's direct testimony.

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corrective actions can be taken to restore the reactor to normal (or stable) conditions. 1-Tr 1495-97 (MacDonald), 1700-03 (Sears).

36. The emergency classification and action level scheme for Seabrook was first transmitted to the NRC and to all the parties to this proceeding on June 27, 1983. The scheme, incorporated into the record l

as Applicants' Exhibit 1 (fol.1-Tr 1483), utilizes a symptomatic approach to emergency recognition and classification. App. Ex. I at 5-1; 1-Tr 1486-87 (MacDonald). Subsequent amendments were made to the scheme, and incorporated into the record as attachments to Applicants' September 19, 1986 testimony (ff. Tr. 487).

37. The Seabrook emergency classification scheme categorizes a variety of component or system failures into four classes: unusual events, alerts, site area emergencies, and general emergencies. An unusual event is defined as a condition indicating a potential degradation of station safety margins not likely to affect personnel on-site or the public off-site. An alert indicates a substantial degradation of station safety margins which could affect on-site personnel safety, could require off-site impact assessment, but is not likely to require off-site public protective action. A site area emergency is an event which involves likely or actual major failures of station functions needed for the protection of the public. A general emergency indicates substantial core degradation or melting with potential for lose of containment integrity.

App. Ex. I at 5-1 and 5-2. The four classes of events included in the Seabrook scheme are consistent with the classes identified in NUREG-0G54, Appendix 1.

38. The symptomatic cpproach used at Seabrook is a result of three years work performed by the Westinghouse Owners Group. This approach relies on the monitoring of five critical safety functions and the recog-nition of various degrees of challenge to said functions. App. Dir. ff.

1-Tr 1483 at 15-16. The five critical safety functions are: suberiti-cality, core cooling, heat sink , reactor coolant system integrity, and containment integrity. Color-coded status trees, based on plant events i which pose a threat to the safety status of the plant , have been developed for each of the critical safety functions. These trees will assist the operators of the plant in emergency classification and direct them to procedures to be used to mitigate the situation. Each safety function will be displayed to the operator as green (safety-function satisfied - no operator action indicated), yellow (function not fully satisfied - action may eventually be needed), orange (function under severe challenge - prompt action necessary), or red (function in jeopardy

- immediate action required). App.1986 Ex.1, Figure 5.6. The classi-fication scheme at Seabrook relates the status of the critical safety functions to the four emergency action classifications. App.1986 Ex.1, Figures 5.1 through 5.5.

39. In addition to the status of the five critical safety functions, Applicants' scheme takes into account thirteen miscellaneous emergency conditions (listed on Figure 5.6 of App. 1986 Ex. 1). Each of these conditions le related to at least one of the four emergency classifications.

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40. The NRC Staff had reviewed the framework of the emergency action level scheme utilized at Seabrook and had found that framework to

be acceptable at the time of the 1983 hearings. 1-Tr.1699-1700 (Sears) .

Neither New Hampshire nor NECNP ever challenged the framework either at the 1983 hearings or in their proposed findings filed at that time. The Board finds that the framework as described in Findings 37-39 above fully meets the requirements of 10 C.F.R. I 50.47(b)(4) and Part 50, Appendix E.

41. At the time of the 1983 hearings, the Applicants' emergency classification and action level scheme was not yet complete. The testimony introduced in 1986 indicated that the system is now complete.

App. Dir.1986, ff. Tr. 487 at 3; Staff Dir.1986, ff. Tr. 489 at 4. The Staff completed its review and evaluation of the Applicants' scheme and provided its detailed evaluation of the EALs in SER Supplement 4, May 1986. (Staff Ex. 4). Subsequent Staff inspections verified that the corrective actions, identified in Section 13.3.2.3 of Supplement 4 have been completed. Staff Dir.1986, ff. Tr. 489 at 4. The Staff concluded in its review that Applicants' emergency plan provides an adequate planning basis for an acceptable state of emergency preparedness with regard to the emergency classification system planning standard of 10 C.F.R. 50.47(b)(4) and the guidance criteria of NUREG-0654. Staff Dir.1986, ff. Tr. 489 at 4; Staff Ex. 4, SER Supplement 4, I 13.3.2.3.

42. As indicated earlier, Applicants' classification system was not yet complete at the time of the hearings in 1983. Based on the testimony filed in 1986, the Board concludes that the open items that were discussed in the hearing in 1983 have now been satsifactority resolved.

In particular, the Board finds:

1,

. 43. All the Seabrook-specific act points for the critical safety function status trees have now been selected. App. Dir. 1986, ff.

Tr. 482 at 4; Staff Ex. 4; SER Supplement 4 at 13-10; c_f. f Tr.-1 1489-91, 1511-13,1544-45 (MacDonald), 1516-23,1545 (Thomas).

44. Applicants have now incorporated indications and alarms from six different condition monitors as emergency action levels. App. Dir.

1986, ff. Tr. 487 at 4; Staff Ex. 4, SER Supplement 4 at 13-10 and 13-11; f. Tr.-1 1717-20 (Sears).

45. Applicants have now performed an acceptable comparison between their emergency action levels and NUREG-0654. App. Dir.1986, ff. Tr. 487 at 5; App.1986 Ex. 2; Staff 1986 Dir. , ff. Tr. 489 at 5; cf.

Tr.-1 1717-20 (Sears).

46. Applicants have now completed the training of operators in the use of the emergency action levels. Staff 1986 Dir., ff. Tr. 489 at 5; cf. Tr.-1 1506-08 (MacDonald) 1711-13 (Sears).

47. Applicants have also revised their treatment of fire and control room evacuation events so that the treatment is now consistent with the guidance contained in NUREC-0956. Staff Dir.1986, ff. Tr. 489 at 5-7.

48. The uncontroverted (and indeed, unopposed) testimony in 1986 indicates that the Applicants' emergency classification and action level scheme is complete and meets the applicable requirements of 10 CFR I 50.47 and Part 50, Appeendix E. The Board accordingly resolves Contentions NECNP III.1 and New llampshire 20 in favor of authorizing operation of the Seabrook facility.

l 4

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IV. SAPL SUPPLEMENTAL CONTENTION 6 Safety Parameter Display System

49. The only issue remaining for litigation with respect to SAPL Supplemental Contention 6- after the Board's September 15, 1986 Memorandum and Order on Summary Disposition was whether, in light of the 11 items identified by the NRC Staff in Draft License NPF-56 and in Paragraph 15 of Richard J. Eckenrode's Affidavit attached to the NRC Staff's August 18, 1986 Response to Applicants' Motion for Summary Disposition of SAPL Supplemental Contention 6, there is reasonable assurance that in deferring improvements to the safety parameter display system (SPDS) until the first refueling outage , the safety of the population in the immediate vicinity of the plant will be protected.

Memorandum and Order of September 15, 1986, LBP-86-30, Slip op. at 18-19.

50. The Board took evidence on the remaining portion of SAPL Supplemental Contention 6 at hearings in Portsmouth, New Hampshire on October 1 and 3, 1986. The Applicants presented direct testimony from Lawrence A. Walsh and George S. Thomas (ff. Tr. 739); the NRC Staff presented the dir" cit' testimony of Richard J. Eckenrode (ff. Tr. 822).

l SAPL and Massachusetts presented no direct case, participating through the cross-examination of the witnesses presented by the Staff and Applicants. No other party participated in the litigation of this contention.

i l

l t

I i

51. The requirements for an SPDS are contained in Item I.D.2 to Supplement 1 to NUREG-0737. The SPDS is designed to p vavide a concise display of critical plant variables to control room operators to aid the operators in rapidly and reliably determining the safety status of the plant. Eckenrode, ff. Tr. 822 at 2. The SPDS primarily serves to accumulate important safety information in one centralized location.

Tr. 985-86, 995-96, 998,1001 (Eckenrode).

52. The SPDS is not considered a safety system; no operator actions are to be taken at the SPDS or based exclusively on information displayed on the SPDS. Eckenrode, ff. Tr. 822 at 2: Walsh & Thomas, ff.; Tr. 739 at 1-2; Tr. 808 (Walsh), 839, 978-79, 998 (Eckenrode). The SPDS is instead used to refer operators to various other displays and controls in the control room where corrective actions are to be taken if needed. Eckenrode, ff. Tr. 822 at 2; Tr. 808 (Walsh), 839, 979 (Eckenrode). Operators are trained to respond to emergencies both with and without the SPDS. NUREG-0737, Supplement 1, Item I.D.2 (h);

Walsh a Thomas, ff. Tr. 739 at 2; Tr. 812, 817 (Walsh).

l

53. Applicants submitted their SPDS report to the NRC Staff by i

l letter dated January 6, 1986 (SB N-920) . Eckenrode, ff. Tr. 822 at 3.

Additional information was submitted to the Staff by letter dated April 2, 1986 (SBN-987). Id. at 4. The Staff and its consultants reviewed the information submitted by Applicants and conducted an onsite audit of the SPDS in May of 1986. Eckenrode, ff. Tr. 822 at 4. The results of the Staff's review are set out in Section 18 of Supplement 6 to the Seabrook Safety Evaluation Report. Id. ; Staff Ex. 6.

54. The Staff concluded in its review that the Seabrook SPDS was i not in complete compliance with the guidelines of Item I.D.2 of ;

NUREG-0737, Supplement 1. Item I.D.2 specifies that an SPDS contain the following eight elements:

a) SPDS should provide a continuous display; b) SPDS should display critical plant variables; c) SPDS should aid operators in rapidly and reliably determining the safety status of the plant; d) SPDS shall be located convenient to control room operators; e) SPDS shall be a continuous display; f) SPDS shall be suitably isolated from safety-related systems; g) SPDS shall be designed to incorporate accepted human factor principles; h) Procedures and training shall address plant safety status and accident conditions with and without SPDS.

Eckenrode, ff. Tr. 822 at 2.

55. In its review , the Staff identified the following items with respect to the Seabrook SPDS that might require corrective action:

1) The SPDS display is not continuous; l 2) Residual heat removal (RHR) flow and hydrogen concentration l variables are considered by the Staff to be part of the minimum information required to assess the critical safety functions and are not displayed on the SPDS;

3) The containment isolation display is not satisfactorily readable from the prime SPDS location to be considered part of the l

SPDS;

4) The SPDS does not display sufficient radiation variables;

}

i j 5) Several human engineering discrepancies were identified; l

6) Two critical safety function status trees are not mode dependent and have the potential to mislead operators; 1

_ 4 m- ~ J _a1-a __ .h

7) Isolation devices between the Reactor Vessel Level Instrumenta-tion System (RVLIS) and the SPDS had not been approved;

8) Data validation algorithms may not be sophisticated enough to ensure valid data are displayed to the operator;

9) The usefulness of the lower-level display formats to the operator is in question;

10) RVLIS and the Radiation Data Management System (RDMS) availability has not yet been factored into overall SPDS availability calculations;

11) System response time appears to be satisfactory, but a system load test is needed to verify the worst condition.

Eckenrode, ff. Tr. 822 at 5; Staff Ex. 6 at 9-10.

56. Before the hearing, the RVLIS isolators were successfully tested to the maximum credit fault voltage and current. The Staff there-fore considered the seventh item of noncompliance listed above resolved.

Eckenrode, ff. Tr. 822 at 9-10; Staff Ex. 6 at Appendix 18A; Walsh &

Thomas, ff. Tr. 739 at 4-5. No party challenged the resolution to this noncompliance either at the hearing or in proposed findings.

Accordingly, the Board finds that this item has been resolved.

57. As to the remaining ten items, the Staff found that these do not present any safety questions for operation at Seabrook. In particular, the Staff found that the SPDS in its current design should not provide erroneous or misleading information to plant operators and will therefore not increase the potential for operator error in the event of an abnormal occurrence at Seabrook; the current SPDS does provide useful information to the plant operators; and the information not presently available on the SPDS is available elsewhere in the control room.

Eckenrode, ff. Tr. 822 at 4; Staff Ex. 6 at 10; Tr. 834-35, 929-32 (Eckenrode). These findings, combined with the Staff's conclusion that i

I 4

., ,--,...m~, - - - . - - - . . . , - . - _ . -

NUREG-0737, Supplement i does not require that SPDS be implemented before full power operation, led the Staff to conclude that implementation j of the additions needed to bring the Seabrook SPDS into full compliance with NUREG-0737, Supplement I could be deferred until the completion of the first refueling outage. Eckenrode, ff. Tr. 822 at 4-5.

58. The issue for hearing was whether the deferral of the resolution of the remaining ten items until the completion of the first refueling outage would have an adverse affect on the public health and I safety. LBP-86-30, supra, Slip op. at 19. The Board discusses each item below seriatim. Before addressing each individual item, however, we l

turn first to general findings concerning the state of the Seabrook SPDS which are applicable to all of the items.

59. As noted earlier, the SPDS requirement is set forth in Item I.D.2 of Supplement 1 to NUREG-0737, the lessons learned from the accident at Three Mile Island. The Staff has found that the nuclear industry has experienced some difficulties in meeting the complete requirements of item I.D.2. Eckenrode, ff. Tr. 822 at 10; Tr. 982-83 (Eckenrode). In fact, although many operating reactors have an SPDS, no reactor has an SPDS which has received final approval from the NRC Staff. Tr. 921, 982 (Eckenrode).

60. While the Seabrook SPDS is considered incomplete by the Staff, Seabrook does currently have an operational SPDS. Eckenrode, ff.

Tr. 822 at 10; Tr. 785 (Walsh), 932 (Eckenrode). The purpose of the SPDS la to provide a general picture of plant conditions. Tr. 1001 (Eckenrode). The Staff reviewer of the Seabrook SPDS, Mr. Eckenrode, testified that the SPDS at Seabrook does a good job of providing a

general picture of safety. Tr. 932-33 (Eckenrode). The parameters missing from the SPDS are not called upon during the beginning of accident conditions. Tr. 933 (Eckenrode). The Seabmok SPDS was tested during various accident scenarios; the SPDS in its present condition successfully facilitated prompt identification of developing changes and directed the plant operators to appropriate recovery procedures. Walsh & Thomas, ff. Tr. 739 at 8; Tr. 751, 757 (Walsh);

see also Tr. 845, 848, 852 (Eckenrode).

61. In addition to the fact that Seabrook has a useful, functioning l

SPDS, the uncontroverted evidence of record indicated that the SPDS is not a safety system; no operator actions are taken at the SPDS or based upon information displayed on the SPDS. Eckenrode, ff. Tr. 822 at 2; Walsh a Thomas, ff. Tr. 739 at 1-2; Tr. 975, 978-79 (Eckenrode). The absence of an SPDS would not adversely affect the ability of operators to rapidly and reliably determine plant status; operators would be alerted to abnormal conditions through normal control room instrumentation and alarms. Eckenrode, ff. Tr. 822 at 4,11; Walsh a Thomas, ff. Tr. 739 at 2: Tr. 935-36, 985, 995 (Eckenrode). All the information available on the SPDS is available elsewhere in the control room. Eckenrode, ff. Tr. 822 at 4; Tr. 931 (Eckenrode). In fact, it is from these other sources that the information upon which the operators rely is obtained. Tr. 978-79 (Eckenrode). While the SPDS may improve operator actions, the absence of an SPDS (or portions thereof) will not adversely affect plant l operators. Tr. 834-835, 931 (Eckenrode). Nor will the Seabrook SPDS in its current state affirmatively cause operator errors in assessing plant conditions. Eckenrode, ff. Tr. 822 at 4.

62. Turning now to the individual items which might require corrective action, the Staff found that because the Shift Technical Advisor at the SPDS has the capability to call up display formats other than the SPDS at the SPDS terminal, the Seabrook SPDS is not a continuous display. Eckenrode, ff. Tr. 822 at 8; Staff Ex. 6 at 5-6.

The SPDS is continuously accessible at the SPDS terminal; the problem here is the operators have too much flexibility. Eckenrode, ff. Tr. 822 at8; Tr. 937 (Eckenrode). The Applicants have committed to dedicate the SPDS terminal to the SPDS display; Applicants' witnesses testified that this deficiency should be resolved before the plant exceeds 5% of rated power. Tr. 764-65, 804-05 (Walsh). Mr. Eckenrode testified that the existence of this deficiency would not affect operator reaction time in the event of an abnormal occurrence at Seabrook. Tr. 987 (Eckenrode).

. In view of the above, in conjunction with our general findings contained in if 60-61, supra, the Board finds that deferral of resolution of this matter until the completion of the first refueling outage will not adversely

affect the public health and safety.

63. The second item is the absence of parameters measuring RHR flow and hydrogen concentration on the SPDS. Information on RHR flow and hydrogen concentration is not needed during the early stages of accidents. Tr. 933, 978 (Eckenrode). Alarms and direct measurement J

indicators for" both parameters are situated on the main control board.

Tr. 775-78 (Walsh). Operators are stationed at the main control board and use of these parameters from the main control board is directed by the operator's emergency procedures. Walsh a Thomas, ff. Tr. 739 at 3; Tr. 862, 897, 977-78 (Eckenrode). As mentioned in f 61 above, no

mitigative actions involving RIIR or hydrogen concentration need to be taken at the SPDS location. Tr. 978-79 (Eckenrode). Mr. Eckenrode testified that the absence of these parameters on the SPDS will not increase operator reaction time in the event of an emergency. Tr. 986-87 (Eckenrode). Based on the above and the findings contained in TT 60-61, supra, the Board finds that the public health and safety will not be adversely affected if RHR and hydrogen concentration parameters are not added to the SPDS until the completion of the first refueling outage.

64. The third item involves the containment isolation display on the main control board. The focus of this noncompliance is not the fact that the display is twenty-six feet from the SPDS terminal, but rather that the I display is not readable from the SPDS terminal. Eckenrode, ff. Tr. 822 at 6, 9; Tr. 965 (Eckenrode). The critical factor when containment isolation is called for is whether the containment is or is not isolated; if containment isolation is achieved, individual valve information is not needed. Tr. 863 (Eckenrode) . The containment isolation display board I

is alarmed, and an operator is assigned to handle the board when needed.

Tr. 987 (Eckenrode). The present condition of the board would not adversely affect operator reaction time. Id. Not only is this information already available to the operators when needed, but Applicants testified at the hearing that this item has already been resolved so that the containment isolation display is now readable from the SPDS terminal.

Tr. 771-72, 782-83 (Walsh). Whether or not it 16 Pesolved at the present time, the Board finds that resolution of this matter can safely await the completion of the first refueling outage.

65. The fourth item involves the absence of steam line and stack radiation monitors on the SPDS. Information on these variables is currently available on the RDMS. Eckenrode, ff. Tr. 822 at 6; Walsh &

Thomas, ff. Tr.139 at 4. Applicants plan to add this information to the SPDS prior to exceeding 5% power. Tr. 774-75, 806 (Walsh). The RDMS is situated an arm's length from the SPDS terminal. Tr. 779 (Walsh),

866, 969 (Eckenrode). The WDMS has auditory alarms to inform operators when radiation exceeds a designated set point. Tr. 780 (Walsh), 969, 986 (Eckenrode). The absence of these radiation variables on the SPDS would not adversely affect the operator in timely assessing accident conditions. Tr. 968, 986 (Eckenrode). The Board finds that deferral of resolution of this item until the completion of the first refueling outage will not adversely affect the public health and safety.

66. As to the fifth item , the Staff identified three human engineering deficiencies in its review of . the Seabrook SPDS: the nonconformance of the heat sink display, the awkwardness of the method for calling up the lower level displays, and the difficulty in reading the containment isolation display from the SPDS terminal. Eckenrode, ff.

Tr. 822 at9; Staff Ex. 6 at 7. The containment isolation display is discussed in f 64, supra. Currently, the Seabrook display callup method

requires operators to position a cursor and press two buttons simul-teneously; the Staff would prefer a simpler operator action. Walsh &

' Thomas, ff Tr. 739 at 5; Eckenrode, ff. Tr. 822 at 9. The Staff found the current callup method adequate (Eckenrode, ff. Tr. 822 at 9; Tr. 855 (Eckenrode)); the requested improvement would do no more than eliminate a second or two in operator response time. Tr. 988 (Eckenrode). The l

Board finds this human engineering deficiency to be without safety significance.

67. The other human engineering deficiency involves an inconsistency with the heat sink display . Six of seven displays showed parsmeters below a decision block; the heat sink display reading was above the decision block. Eckenrode, ff. Tr. 822 at 9; Tr. 777 (Walsh).

The safety significance of this deficiency is questionable. In any event, Applicants have already changed the heat sink display so that it is now consistent with the other disnlays. Walsh a Thomas, ff. Tr. 739 at 4; Tr. 777 (Walsh). The Board finds that this deficiency is resolved, and further finds that even if unresolved, resolution prior to completion of the first refueling outage would not be necessary to protect the public heslth and safety.

68. The sixth item is the mode dependency of the suberiticality and core cooling status trees; the SPDS showed these functions under challenge during normal operating conditions. Eckenrode, ff. Tr. 822 at

7. Plant operators were made aware of this deficiency in their training.

Id,. In addition, Applicants' witnesses testified that the necessary enhancements have been made to resolve this deficiency. Tr. 730, 744 (Walsh) . This issue appears to have been rendered moot. If not moot, the Board finds that this deficiency would not need resolution prior to completion of the first refueling outage.

1

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69. The eighth item involves data validation algorithms. U The Staff audit determined that the data validation methodology included only range checking, averaging and auctioneering. Eckenrode, ff. Tr. 822 at 7; Staff Ex. 6 at 4, Consultant's Report at 13. It thus may be possible that , under certain circumstances, the SPDS would not immediately indicate that a limit had been reached. ,Id. ; Tr. 807, 809 (Walsh) . If the SPDS did fail to immediately indicate an abnormal situation, operators would be made aware of the problem by the main control room instru-mentation in a timely fashion. Tr. 839, 988 (Eckenrode). It must be kept in mind that it is the control room instrumentation that operators ultimately rely upon in taking protective actions. Walsh a Thomas, ff.

Tr. 739 at 6; Tr. 806-08 (Walsh), 978-79 (Eckenrode). - The Board finds that reliance on the current validation methodology until the completion of the first refueling outage will not adversely affect the rublic health and safety.

70. The ninth item questions the utility of the lower level displays on the SPDS. During Staff observation of an accident simulation at Seabrook, it was observed that the Seabrook operators did not use the SPDS lower level displays on the SPDS terminal, but rather used hard copy representations of the lower level display instead. Eckenrode, ff.

Tr. 822 at 7; Tr. 759-61, 815-16 (Walsh), 973, 979-80 (Eckenrode). The Staff did not identify any problem with the contents of the lower level displays. Tr. 970 (Eckenrode) . Furthermore, the observation of the

~

7/ The seventh item, involving the RVLIS isolators, has been resolved.

See T 8, supra.

accident scenario revealed that the operators, using the top level SPDS and lower level hard copies, handled the scenario well. Tr. 757 (Walsh),

. 845, 848, 852, 988-89 (Eckenrode). The Staff did not identify a safety problem; the Staff merely wants an evaluation of why the operators used hard copies instead of the lower level displays. Tr. 973 (Eckenrode).

Under the circumstances, the Board f*nds no adverse effect upon the public health and safety in deferring resolution of this item until the completion of the first refueling outage.

71. The tenth item involves the calculation of RVLIS and RDP.1S availability. The Staff found the availability of the SPDS is currently acceptable (over 99%). Eckenrode, ff. Tr. 822 at 7. There is no indication that there might be any problem here; the Staff has merely indicated that it wants the calculations redone when the RDMS and RVLIS are hooked up to provide input to the SPDS. Walsh & Thomas , ff.

1 Tr. 739 at 6; Staff Ex. 6, Consultant Report at 13; Tr. 928, 973 (Eckenrode). The additional calculations would only involve the availability of the RVLIS and RDMS input to the SPDS; the overall availability of the balance of the SPDS would remain unchanged. Tr. 974 (Eckenrode). As with all other input to the SPDS (see f 61, supra), the backup instrumentation could be called upon if information from RVLIS or RDMS were not available on the SPDS itself. Walsh & Thomas, ff.

Tr. 739 at 7. Absence of this information on the SPDS would not adversely affect operator response time. Tr. 989 (Eckenrode). The i

Board finds that resolution of this item can safety await the completion of the first refueling outage.

I 72. The final item involves an additional system response test. The Staff found that the tested system response time was acceptable, but f

noted that the test did not occur during a heavily loaded sequence.

Eckenrode, ff. Tr. 822 at 7. The Staff does not consider this a i

deficiency, but would like to see an additional test under heavier loads.

,Id. ; Tr. 928, 989 (Eckenrode). Applicants have agreed to perform such a test, but it cannot be performed prior to some level of plant operation.

Walsh a Thomas, ff. Tr. 739 at 7; Tr. 989 (Eckenrode). While there is no indication that the SPDS would become unavailable during heavy loaded

]

sequences, the testimony indicated that if the SPDS were out, the Shift Technical Advisor would leave the SPDS and monitor the control board instead. Tr. 790 (Walsh) . No operator actions would be affected.

I Tr. 975 (Eckenrode). The Board finds that resolution of this item can

safely be deferred until the completion of the first refueling outage.

73. In light of all the above findings, the Board concludes that final resolution of the ten remaining SPUS items identified by the Staff can be deferred until the completion of the first refueling outage without adversely affecting the public health and safety.

V. CONCLUSIONS OF LAW In reaching this decision, the Board has considered all the evidence of the parties and the entire record of this proceeding including all proposed findings of fact and conclusions of law presented by the parties. Based upon a review of that record and the foregoing Findings of Fact, which are supported by reliable, probative and substantial evidence, the Board, with respect to the issues in controversy before us, reaches the following conclusion pursuant to 10 C.F.R. I 2.760a:

i

1. The Applicants have specified the time durations for which

, electrical equipment important to safety at Seabrook is

,-- , .-,- ,- ,, . , . , - , . , , _ - . - ~ . . . . - , - . . , _ , . _ . - , - - . . . - . , . , , , , - , _ - - . . _ . , _ _ _ . - _ - - - . . . , - . -

. environmentally qualified, in accordance with 10 C.F.R. i Part 50, Appendix A, GDC 4, and 10 C.F.R. I 50.49.

l

2. The Seabrook emergency classification and action level scheme 4 meets the applicable requirements of 10 C.F.R. I 50.47 and Appendix E to Part 50.

3. The public health and safety will not be adversely affected if resolution of the 11 items identified by the NRC Staff with respect to the Seabrook safety parameter display system that may require corrective action (listed in f 55, supra) is 1

deferred until the completion of the first refueling outage.

With resolution of these issues, the Board has resolved all matters in controversy with respect to the Applicants' application for an operating license for the Seabrook facility, with the exception of matters pertaining to the adequacy of offsite emergency planning, in favor of authorizing operation of the facility. Pursuant to 10 CFR 55 50.57(c) and 50.47(d),

the Board therefore grants that portion of Applicants' motion of June 17,

1986 which requested that the Board authorize issuance of an operating license for Seabrook Unit 1 limited to 5% of rated power.

VI. ORDER Wi!EREFORE, in accordance with the Atomic Energy Act of 1954, as amended, and the Rules of Practice of the Commission, and based on the foregoing Findings of Fact and Conclusions of Law, IT IS ORDERED that:

Upon making the applicable findings required under 10 CFR I 50.57(a), the Director of Nuclear Reactor Regulation is authorized l

- 3 e

to issue a license authorizing low power testing and operation limited to 5% of rated power for the Seabrook Station, Unit 1.

Pursuant to 10 C.F.R. I 2.760(a) of the Commission's Rules of Practice, this Partial Initial Decision will constitute the final decision of the Commission forty-five (45) days from the date of issuance, unless an appeal is taken in accordance with 10 C.F.R. I 2.762 or the Commission directs otherwise. See also 10 C.F.R. Il 2.764, 2.785 and 2.786.

Any party may take an appeal from this decision by filing a Notice of Appeal within ten (10) days after service of this decision. Each appellant must file a brief supporting its position on appeal within thirty (30) days after filing its Notice of Appeal (forty (40) days if the Staff is the appellant). Within thirty (30) days after the period has expired for the filing and service of the briefs of all appellants (forty (40) days in. the case of the Staff), a party who is not an appellant may file a brief in support of or in opposition to the appeal of any other party. A responding party shall file a single, responsive brief regardless of the number of appellant briefs filed.

See 10 C.F.R. I 2.762(c).

Respectfully submitted,

$cbLrf ( O rlN g Robert G. Perlis Counsel for NRC Staff (bbtuN $l k Sherwin E. Turk Senior Supervisory Trial Attorney Dated at Bethesda, Maryland this 26th of November,1986