ML20213E680

Proposed Findings of Fact & Conclusions of Law & Proposed Form of Decision Re Procedural History of 861030 Order. Perspective on Handling of Onsite Issues Before Board Prior to Reconstitution Provided.Certificate of Svc Encl

ML20213E680
Person / Time
Site:	Seabrook
Issue date:	11/07/1986
From:	Backus R BACKUS, MEYER & SOLOMON, SEACOAST ANTI-POLLUTION LEAGUE
To:
Shared Package
ML20213E631	List: ML20213E628 ML20213E680 ML20213E701 ML20213E716
References
OL, NUDOCS 8611130254
Download: ML20213E680 (22)

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s DOLMETED 3 Dated: Novembe r T,RCl986 UNITED STATES OF AMERICA I NUCLEAR REGULATORY COMMISSION 6fflCt CF ' . ;u 00CnUn;s q .. ,v n p before the ATOMIC SAFETY AND LICENSING BOARD In the Matter of PUBLIC SERVICE COMPANY OF Docket Nos. 50-443-OL-1 NEW HttMPSHIRE, et al 50-444-OL-1 Onsite Issues (Seabrook Station, Units 1 and 2)

SEACOAST ANTI-POLLUTION LEAGUE'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW AND PROPOSED FORM OF DECISION Procedural History The Seacoast Anti-Pollution League has no quarrel with the chronology of events set out in Appiicants' deseription of the procedural history of this case i i) fyp;: cants' Proposed Findings of Fact and Conclusions of Law and Proposed Form of Order filed on October 30, 1986 as far as it goes. SAPL would like to make clear i

herein its per spective on the handling of the onsite issues in which

, SAPL had an interest that were before this Board prior to its i

l reconstitution in September 1985.

8611130254 861107 PDR ADOCK 05000443 G PDR

On November 13, 1981, SAPL moved to become an intervenor in 4

this case and on April 5, 19 82 'SAPL submi t t ed f our content ions. The three which dealt with onsite issues were as follows:

. SAPL 2: The operat ion of the proposed condenser cooling system will have an unreasonable adverse effect on the quality of the aquatic environment.

SAPL 3: The operat ion of the proposed nuclear plant will have an unreasonable adverse effect upon the economic wellbeing of the seacoast area.

SAPL 4: The decommissioning of the Seabrook Plant, should it receive its operating permit and actually operate will have a major long term negative impact on the health and wellbeing of the citizens in the area of the facility.

On April 20, 1982, SAPL filed six supplemental contentions on onsite issues as follow:

I SAPL Supplement 1: The Applicant has not established reasonable assurance that the saf ety sys tems of the proposed plant can withstand a worst- case accident analysis because of interactions with components presently classified as non-saf ety, contrary to the requirements of 10 C.F.R. Part 50.

SAPL Supplement 2: The Aaplicant has not provided the assurance that safety related equipment will be able to perform adequately in an accident environment over the projected lifetime of the plant.

SAPL Supplement 3: The appl i cable requ i remen t s o f t he Commi s s ion's Interim Policy Statement issued June 13, 1980, 45 Fed. Reg. 40101 on Nuclear Power Plant Accident Considerations Under the National Environmental Policy Act of 1969 have not been met.

SAPL Supplement 4: There is no need for the electricity hoped to be produced by the proposed plant and consequently this Board should find that the costs, including the risk of station operation, outweigh the benefits.

1

SAPL Supplement 5: The lead Applicant and certain other Applicants including Uni ted Illuminat ing and Bangor Hydro, cannot demonstrate reasonable assurance that they are financially qualified to meet the cost of operating and decommission the proposed faci 1ity.

SAPL Supplement 6: SAPL hereby joins in and adopts as its own the contentions and the bases therefore set forth by the State of New Hampshire and Attorney Gregory P. Smith nos. 4 through 10, and 12 through 16.

The Board, by Memorandum and Order of September 13, 1982 denied all of the above contentions save SAPL Supplement 3 and SAPL Supplement

6. SAPL subsequently filed an Objection and a Motion for Reconsideration and a Motion to Cer t i f y Obj ect ions to the Appeal Board with regard to SAPL Supplement 4, but its motions were denied by the Board's Memorandum and Order of November 17, 1982. On February 11, 1983, SAPL filed a motion for summary disposition relative to SAPL Supplement 3 citing facts which showed that the Final Environmental Statement (FES) for Seabrook f ailed to comply with the Commission's Interim Policy Statement in numerous respects.

Applicants also filed a motion for summary disposition of SAPL Supplement 3, on the same date to which SAPL objected on March 21, 1983 on the basis that the Applicants' motion did not address the legal and f actual issues of compliance and adequacy that SAPL raised in its motion. After argument at a prehearing conference in Boston on April 7 and 8, 1983, the Board ruled on May 11, 1983 that the Applicants' motion was granted and SAPL's denied. SAPL filed an appeal or in the alternative, a request for certification of the denial of SAPL Supplement 3 on May 26, 1983. On June 20, 1983, the Appeal Board dismissed the appeal and denied directed certification and s tated that SAPL had to await the Board's initial decision before presenting its grievance for appellate consideration.

i Additionally, SAPL participated in hearings conducted by the Board beginning on August 16, 1983 on the New England Coalition on T-Nuclear Pollut ion (NECNP) cont en t ions III.12 and III.13. The Board rephrased those contentions by order of June 30, 1983 as follows:

NECNP III.13 Evacuation Time Estimates The evacuat ion t ime es t imates provided by Applicant s in Appendix C of the Radiological Emergency Response Plan are deficient in failing to include an estimate of:

1) The t imes for evacuat ion during adverse weather condi t ions developing on a busy summer weekend; and

2) The time for simultaneous evacuation of beach areas lying NE to SSE of the Seabrook site. ,

SAPL filed its proposed findings and conclusions with regard to those evacuation time estimate (ETE) contentions on October 26, 1983. SAPL holds that ETE is both an onsite and an offsite issues and that therefore no license can be granted until the Board has made f indings on evacuat ion t ime es t imat es f or Seabrook S tat ion. In f

accord with this position, SAPL filed " Seacoast Anti-Pollution

! League's Motion for Board Decision on Applicants' Evacuation Time Estimate (ETE) as a Condition to Issuance of Operating License" on July 16, 1986. The Board denied SAPL's motiion by Memorandum and Order of August 14, 1986.

l Contention SAPL Supp. 6 (Formerly NH-10) 4 Introduction 1

SAPL Supp. 6, as admitted, read as follows:

I I

The Seabrook Station Control Room Design does not comply with General Design Criteria 19 through 22 and 10 C.F.R. Part 50, Appendix A, and NUREG-0737, Items I . D.1 and I.D.2.

By Memorandum and Order of September 15, 1986, the Board summarily disposed of that portion of SAPL Supp. 6 having to do with the Detailed Control Room Design Review (DCRDR) and ordered that:

... SAPL may and the Staff and/or Applicants shall present wr i t ten explanatory test imony upon the issue of whether or not, in light of the fact that the SPDS is not currently at an optimum, i.e., incomplete, because of the aforementioned deficiencies, there is reasonable assurance that, in deferring improvements to the SPDS until the first ref ueling outage, the safety of the population in the immediate vicinity of the plant will be protected.

SAPL assumes that the Board intends to follow its Sept. 15, 1986 order that the s tandard for j udging the adequacy of the Saf ety Parameter Display System is whether there is reasonable assurance that, in deferring improvements to the SPDS until the first refueling outage, the safety of the population in the immediate vicinity of the plant will be protected. SAPL holds that that is not the appropriate legal s tandard to be applied and instead holds that the proper legal standard is set forth in NUREG-0737, Supplement No. 1, which required that implementation dates for the SPDS negotiated in 1983 were to have been translated into a binding legal requirement.

SAPL does not intend to argue that quest ion in these proposed findings but in order to assure preservation of this key issue for appeal purposes, includes below a proposed finding and ruling "A-A" as its f i r s t a l t erna t i ve f indi ng wi t h r espect to NUREG-3 737, I t em I . D. 2 and respectfully prays that the Board make a specific ruling with reference to it. SAPL's position on this issue is preserved in this record Tr. 723-726 and 924-829.

A-A The Board rules that, as a matter of law, the Applicants have not satisfied the requirement of NUREG-0737, Item 1.D.2 for prompt implementation of a fully compliant SPDS pursuant to a schedule to have been negotiated between the NRC Project Manager and the Applicants in 1983, to have been translated into a binding legal requirement in accordance with the requirements of NUREG-0737, Supplement No. 1.

Findings of Fact

1. SAPL incorporates by reference Applicants' proposed finding No.

t 55.

2. The Seabrook Safety Parameter Display System (SPDS) is not at t h i s t ime in comple t e compl iance wi th I t em 1. D. 2 o f NUREG-0737, Supplement 1 [ Staff Dir. Post Tr. 822 at 4, T_r. at 833.] The NRC Staf f has identified a minimum set of approximately 20 plant parameters it believes to be sufficient to provide plant operators with information about the critical saf ety f unct ions i specified in NUREG-0737, Supp. 1. [ Staff Dir. Post Tr. at 6, l

l. H. at 860]. The Seabrook SPDS does not currently display 5 of these minimum parameters [T_r. r at 860]. Therefore, under specific circumstances, the SPDS does not supply some of the minimum information needed to assess the critical safety l functions (CSF's) [Tr. at 861].

I

3. NRC Staf f witness Eckenrode and Applicants' witness Walsh agreed that prompt implementation of an SPDS is of primary importance

[Tr. at 745, 920]. Prompt implementation of an SPDS provides

an important contribution to plant safety [T_r. r at 920]. NRC Staff witness Eckenrode stated, in regard to the defini t ion of

" prompt" that, of the items in NUREG-0737 (post-TMI action items), the SPDS was probably the one that should be done first

[Tr. at 920]. It is always bet ter to have the SPDS done before 1

, plant start up [Tr. at 836].

4. The purpose of an SPDS is that there be one concise display in the control room where all the critical safety parameters are displayed [R . at 931]. The SPDS is to show a general picture of the safety of the plant. That is the reason that NUREG-0737, Supp. 1, states that an SPDS is supposed to be in addition to other control room equipment [Tr. at 931]. It is to bring together a lot of dispersed instrumentation into a single picture of safety [Tr. r at 840]. The SPDS is supposed to be "the eyes" beyond the si tuat ion. [Tr. at 807]. It is an aid to the control room operator in rapidly and reliably assessing the safety status of the plant [ T r_. at 925]. It therefore decreases the potential for operator error in the event of an abnormal occurrence [Tr. at 929]. React ion time is also one process the SPDS could serve to improve [T_r. at 1000].

5. Laur ence A. Walsh, one o f Appl i can t s ' wi t nes s es , has neither a college nor any Masters or higher level degree [Tl. at 732].

His only f ormal t raining in human f actors engineering was a one week course, non-credit, at the University of Wisconsin [Tr.

at 732-733]. Mr. Walsh's only experience in designing safety parameter display systems or similar control systems was his i involvement at Seabrook Station. His resume lists only one course in psychology, taken at the University of New Hampshire, which did not, insof ar as he can recall, deal with the interf ace between man and machine [Tr. at 733]. Mr. Walsh had primary responsibility for the design of the SPDS [Tr. at 787]. The term Inadequate Core Cooling Monitor (ICCW) was " foreign" to Applicants' witness Walsh [Tr. r at 788]. The ICCM is referred to in the NRC's contractor audit report for the Seabrook SPDS

[H. at 918]. Mr. Walsh said he had read through the NRC's contractor audit report [TR. AT 752].

6. George S. Thomas , the other witness brought by the Applicants, also did not claim particular expertise in the field of human factors engineering and his purpose on the panel was primarily to provide a management perspect ive on the r at 735].

issues [T_r.

7. Applicants hold that the present SPDS design should be suf ficient for the life of the plant, though they have agreed to add the radiation parameters [Tr. at 895]. Applicants have agreed to have an operational SPDS (as described in PSNH's submittals dated January 6, 1986 and April 2, 1986 and as modified by the S taf f's audi t findings) that is acceptable to the NRC prior to restart following the first refueling outage [Tr. at 798]. A period of time of 12 to 18 months typically elapses between initial start up and the first ref ueling for a pressurized water reactor such as Seabrook [Tr. at 834]. The Applicants do not take the position that there is any more assurance of safety during the first year of plant operation than there is during subsequent years [Tr. at 768].

.I

8. NRC Staff witness Eckenrode stated that, with regard to_the SPDS, the period of plant operation from initial start up to the first refueling is not considered by the Staff to be any safer than any other similar period of plant operation at full power [T_r.r at 837]. Staff witness Eckenrode had no statistics on what period of operat ion accidents were more likely to happen

[Tr. a t 975 ]. Mr. Eckenrode was the person primarily responsible for the recommendat ion that the modi ficat ions to the SPDS sys tem, as well as others, could be deferred until the first refueling and handled in license conditions [H. at 868]. The Staff issued the safety evaluation report for Seabrook's SPDS that stated that corrections for any deficiencies identified during the audit would not need to be implemented until the first refueling outage a month earlier than the NRC audit of the Seabrook SPDS (in April 1986) [Tr. at 930]. The Staff witness did not know the f ull extent of the deficiencies of the Seabrook SPDS until the onsite audit [Tr. at 929-930].

9. The Three Mile Island accident happened during the period between initial plant start up and the first refueling at the TMI Unit 2 reactor. [Kemeny Commission Report]

10. Aoplicants' wi tness s tated that the SPDS is not a saf ety sys tem because no operator actions are taken at the SPDS. Yet, Applicants' witness conceded that there are many important safety components of the plant at which operator actions are not taken [T_r. at 791].

11. SPDS shall be a continuous display [ Staff Dir., Post Tr. 822 at 2]. Applicants propose that until the continuous display is produced to an acceptable degree to the NRC that one screen will be kept selected to the panel display at the primary SPDS location [Tr. at 764, 804-805]. However, Applicants have not adopted any administrative procedures to ensure that that display is continuous [Tr. at 805].

12. Applicants' witness Thomas did not know how long it would take to implement a continuous display because he did not know when the software would be developed [Tr. at 764-765]. Applicants' witness Walsh stated that achieving a continuous display was neither a hardware nor a software problem but just a transfer problem [Tr. at 805]. Applicants' witness Walsh expects that the continuous display would be in place prior to proceeding to 5 percent power [Tr. at 805].

13. The Seabrook Station SPDS does not display Residual Heat Removal (RHR) Flow var iables [Sta f f Dir. Pos t Tr. 822 at 5, Tr. at 777-778]. RIIR Flow is considered by the Staff to be part of the minimum information required to assess the CSF's [Tr. at 861, 938-939].

14. Applicants are not commit t ing to have RilR Flow displayed on the SPDS and have yet to determine the need of that indication and are still conferring with NRC on this item. Applicants hold that RilR Flow may never be added to the SPDS [Tr. at 768].

15. RilR Flow is an important indicator so the plant operators will quickly know i f the RilR pump is f unct ioning properly [Tr. at 777].

16. RHR Flow is displayed on the main control board at two locations to the left of the main control board front facing panels [Tr.

at 775]. It is approximately 26 f eet 6 inch to the board and 3 feet down the board from the primary SPDS/CRT location [Tr. at 776].

17. The basis of NRC Staff witness Eckenrode's determination that deferral of the addition of the RHR Flow parameter to the Seabrook SPDS would not post a threat to public health and safety was the advice of his technical receiver and not based on his own personal knowledge [ T_r . at 941]. Mr. Eckenrode admitted that he personally did not understand the safety significance of requiring RHR Flow [ T_r_ . at 940]. He did not know where the information on RHR Flow is located in relation to the primary SPDS/CRT [R . at 862]. He did not know how the operators ' at tent ion would be drawn to RHR Flow displays in the control room r . at 942].

[ T_r It was beyond Mr. Eckenrode's technical competence to know whether the RHR system is used when i t is no longer practical to use steam generators for decay heat removal when reactor coolant system pressure has dropped

[Tr. at 1002].

18. Containment Hydrogen Concentration is considered by the Staff to be part of the minimum information required to assess eritical sa f ety f unct ions (CSF's) and is not displayed on the SPDS [Staf f Dir. Post Tr. 822 at 5, Tr. at 861].

19. Applicants have not yet determined that containment hydrogen concent rat ion should be added to the SPDS. Applicants are s t ill j

in negotiations or conferences with NRC on this matter, and l

Applicants' position is that it may never be necessary. A l containment hydrogen concentration display would not be added l

4 if it were solely up to Applicants' witness Walsh to decide

[Tr. at 769-770].

20. Containment hydrogen concentrations would give an important indication that fuel cladding is being damaged [T_r. at 862].

High levels of hydrogen concentrat ion in the containment is one factor that could lead to a hydrogen explosion [Tr. at 945].

One of the problems during the accident at Three Mile Island was that the operators were not aware of the significant hydrogen concentration in the containment [ T_r r . at 944].

21. The basis of NRC Staff witness Eckenrode's determination that deferral of the addition of the containment hydrogen concentration parameter to the Seabrook SPDS would not pose a threat to public heal th and saf ety was the advice of his technical reviewer and not based on his own personal knowledge [Tr. at 941-943]. Mr. Eckenrode did not know whether there could be a potential for serious offsite consequences in the event that the operator were not aware of hydrogen concentration variables i

[Tr. at 943). He did not know where on the main control board i

information with regard to hydrogen concentration is located

[H. a t 862, 944].

22. Since RilR Flow and containment hydrogen concentration are not on the SPDS, those parameters also do not appear on the hard j copies of the SPDS logic diagrams. The SPDS system in no way directs operators to look at parameters which are not on the SPDS even if hard copies are used [Te. r at 897].

24. The NRC Staff testifled that the containment isolation display

, is not sat is f actorily readable f rom the prime SPDS location to l

be considered part of the SPDS [ Staff Dir. Post T_r. 822 at 5].

A containment isolation display has not been added to the SPDS r at 771, 883].

screen Qr, The containment isolation display is 26 feet away from the primary SPDS location [Tr. at 965].

25. Applicants allege that corrective actions have been taken and completed with regard to the containment isolation display on the main control board [Tr. at 771, 782]. Applicants state that spare light boxes on the display have been grouped into one location to create a situation where the operator will no longer be confused as to whether they are spare indicators or indicators of actual components [Tr. at 772, 783]. Applicants have had no conference with the NRC to see if it meets all their requirements [Tr. at 771]. No documents attesting to the corrective actions had been sent to the NRC staf f as of the time of the hearing (in early October) [Tr. at 782-783]. Tnere had been no review of the installation by the NRC as of the time of the hearing [Tr. at 783, 856].

26. The Applicants' witness did not know precisely how many valve lights were associated with the containment isolation display.

lie gave an estimate of 15 rows on top, 7 down the side and approximately 10 blanks [ T r, . at 784]. The Staff witness described the problems with the containment isolation display as human engineering problems [Tr. at 861]. The Staff witness did not know how many lights there were on the containment r at 863, 965].

isolation display [Tr.

27. A containment isolation valve lef t open could compromise a good degree of the containments intended function [Tr. at 784].

_= __ =- , .

l

28. S ince containment isola t ion is not par t o f the SPDS sys tem, the hard copies of the logie diagrams would not direct plant operators to look at the containment isolation display [Tr. at 897-898].

29. The NRC Staf f considers steam generator (or steamline and stack radiation to be the minimum status indicators of the radiological control f unct ion [H. at 865, 967]. These radiat ion parameters are not yet displayed on the SPDS [Tr. at 865].

30. Information on steamline radiation is available only on the Radiat ion Data Management Sys tem (RDMS) and nowhere else in the control room [Tr. at 779]. The stack monitor is displayed on the RDMS screen and it also has a recorder and viewer on the back of the main control board [Tr. at 780]. The RDMS i s loca t ed J

behind the Shift Technical Advisor (STA) [Tr. at 866]. The j STA's attention would be drawn t o t h e RDMS by auditory alarms j [Tr. a t 969]. However, Seabrook has about 100 hard wired alarms and a number of different alarms could go off at once [T_r. at 1000).

, 31. The NRC Staff is requiring that steamline radiation and stack j radiation parameters be added to the SPDS because they are two r

impor tant parameters indicat ing the general saf ety of the plant

[H. at 969].

32. Applicants say they would not disagree with a license condition which would prohibit them f rom bringing Seabrook Stat ion above

! 5 percent of power before steamline and a stack monitor radiation 1

parameters have been included on the SPDS [Tr. at 774-775]. An l

l additional screen has been developed for the SPDS. There will be a select ion button to pick up that screen to show the status j

of all the radiation monitors [Tr_. at 806]. Applicants have not, however, planned to add the radiation monitoring function of the top level continuous display [Tr. at 816]. The Staff witness could not imagine what would draw the operator's attention to the lower-level displays if radiation parameters were added only to those lower-level displays [T_r. at 970].

33. Steamline radiation can let you know when you have primary to secondary system leakage. Normally, there should not be radiation in steam generation systems [Tr. at 781]. Steamline radiation is one indication of a steam generator tube rupture event and several other events and would be an important variable in dealing with a scenario including steam generator tube rupture

[Tr. at 967, 867].

34. Stack monitor could be of assistance in letting an operator know that a containment isolation valve may have been lef t open

[Tr. at 781]. Stack radiation indicates radiation to the environment [Tr. at 967-968). Stack monitor readings are of value when performing population dose projections during accidents [R. a t 781, 867].

35. The NRC staff has not yet been offered verification of the Applicants' correction of the mode dependency problem related to the suberiticality status display [H. at 837-838].

36. The NRC staff has not yet been offered verification of the Applicants' correction of the mode dependency problem related to the core cooling status display [H. at 838-839].

1 4

1

37. The Seabrook Station SPDS data validation algorithms may not be sophisticated enough to ensure valid data is displayed to the operator [ Staff Dir. Post Tr. 822 at 5, Tr. at 806, 839].

38. The Lawrence Livermore audit report states that: "The use of high or low values provided by redundant instrumentation may 4

result in a conservative estimation of the status of Critical

Safety Functions but it also ensures that the operator will be misled about safety function statu3 in the event of large Instrument errors or on scale instrument failures. Use of i

average values without additional validation checks does not 4

guarantee the operator will consistently be misled in the conservative direction. [ Staff Dir. Post Tr. 822 Appendix at 13, Tr. at 841-842, 971].

39. The Seabrook SPDS does not currently make use of interchannel comparison of redundant ins t rumentation in the data validat ion l scheme [Tr. at 842-843].

~

40. Applicants do not agree that more sophisticated methodology is needed because operators are required to validate any SPDS l conclusions prior to implementing any corrective actions [Tr.

at 806]. However, the operator would not have any indication i that the SPDS indication should be checked (i.e., validated) i f, due to the data validat ion problem the SPDS display was not showing that anything was wrong [Tr. at 807].

41. The NRC's audit team observed a simulation drill conducted by PSNil to demons t rate use of the SPDS under plant upset condi t ions.

The audit team observed that during the entire course of the drill critical safety function (CSF) status was monitored by

~

the STA using hard wired instrumentation and hard copies of the CSF status trees rather than the SPDS lower level displays  ;

[ Staff Dir. Post Tr. 822 Appendix at 14, Tr. at 760,845].

42. The NRC staff has not evaluated why the lower level displays j at Seabrook are not useful and is asking the utility to make j an evaluation [Tr. at 972].

43. Applicants' witness Walsh said he did not believe the NRC staf f unders tands the process by which Applicants use their emergency procedures [Tr. at 761]. Mr. Walsh said he did not believe i that the NRC knew the operators were using hard copies of the lower level t rees in their hands [Tr. r at 759]. Staff's witness Eckenrode said that he did see they were using hard copy [T_r. r at 972].

44. Applicants' witness Walsh stated that the computerized portion i of the SPDS could be eliminated without defeating the goal of 4 safe and efficient power operation [Tr. at 789]. NRC staff l witness Eckenrode stated, however, that hard copy is not part of the SPDS [ T_r r , at 972].

i 45. The lower level display formats aid the operator in assessing abnormal conditions [Tr. at 971]. Operators are instructed to 4

l use the lower level SPDS displays when the SPDS indicates something they should react to [H. at 815].

46. The report of the Design Verification and Design Validation l

Audi t o f the Seabrook S ta t ion SPDS s ta t es tha t no documen ta t ion t

was available at the time of the audit to support the Applicants'

conclusion that the SPDS reduced the time required to respond 1.

i

[

to plant upset conditions [ Staff Dir. Post Tr. 822, Appendix at 5].

47. NUREG-0737, Supp. 1, calls for continuous, concise and rapid SPDS displays [T_r. r at 859]. The NRC staff does not yet have accura t e availabili ty calcula t ions f or Seabrook's SPDS [T_r. r at 973]. The SPDS sys tem response t ime has only been tes ted dur ing a lightly loaded sequence [Tr. at 974]. Applicants have not demonstrated that the SPDS update and response times will not be unacceptably af fected by the high main plant computer loading conditions expected to occur during response to a severe plant upset [Tr. at 857].

48. Applicants' witness Walsh said the SPDS is one of the loads that the main computer would always try to show. He could not explain more clearly its priority in the process of the main plant computer [Tr. at 785-786]. The NRC Staf f wi tness did not know what priority the SPDS has in the process of the main plant computer [Te. r at 911]. The manufacturer of the main plant computer is not still in business [Tr. at 786-787]. There is ,

therefore, no assurance that the SPDS update rate will be in the time that is desirable [Tr. at 857]. There was an availability problem at the Davis Besse plant with regard to its SPDS [H. at 990].

i

49. The main plant computer receives inputs from plant instrumentation via nine intelligent remote terminal units (IRTU's ) wh ich t hen t ransmi t the data to 2 host computers which in turn perform SPDS calculations and develop SPDS displays

[S ta f f Di r. Pos t Tr. 822 a t 2, Tr. a t 912 ]. Some of the IRTU's r

. _ . . - _ . , _,._..__-.-_._..__.-...,.,_,,y _ _ . , , _ _ , _ . . . - . _ , . , . _ _ , , . _ . , . _ _ _ _ . _ _ _ , , , . , . , , . . - - _ _ - . ~ . _ _ . - . , _

have to be functional as part of the opecation of the SPDS, though they are not part of the SPDS. The issue of the reliability of the IRTU's is related to the availability of the SPDS and IRTU's were in the calculation for availability [H. at 915].

50. A June 24, 1986 NRC meeting summary notes that during physical inspection of the IRTU terminal cabinet, No. CP-125, the staf f found a lack of suf ficient separation between electrical cables.

The Staf f witness did not know if the staf f has followed up with a check of other IRTU terminal cabinets to ensure that the cable rout ing will not cause a problem [T_r. at 913]. The Staf f witness also did not know if the identified cable routing deficiency has been corrected [Tr. at 913-914].

51. The draf t Technical Evaluat ion Report on which the Staf f witness relied, according to the Staff's response to SAPL's Interrogatory No. 9, stated:

"Although verification and validation reviews are not a requirement of Supplement I to NUREG-0737, the design problems identified by the NRC combined with the lack of a formal PSNH V& V process, led to the concern that additional design deficiencies may exist in the Seabrook SPDS that were not detected by the audit process [Tr. at 885-886].

52. The Applicants do not plan to do a field verification test of the Seabrook Station SPDS because of the difficulties of so doing [R. a t 788-789]. The NRC has not yet determined whether or not it will do a post-implementation audit on the SPDS as per Section 18.2 of NUREG-0800 [H. at 910].

53. The NRC Staff witness has not examined the range and reasonableness checks per formed by the RD11S as par t of the SPDS, though such an examination may have been done through another branch in instrumentation [H. at 917).

54. The contractor audit report states that the plant SPDS sof tware development had not proceeded by the time of the audit to the point where validation testing of the critical safety function status determination logic could be done [ Staff Dir. Post Tr.

822, Appendix at 4].

55. The contractor audit team's assessment stated that: "PSNil should reevaluate the adequacy of previous validation testing to insure that the usefulness of the Seabrook SPDS was thoroughly established." [ Staff Dir. Post Tr. 822, Appendix at 5].

Conclusions of Law

56. The Board finds that there is no basis for belief that the period of operation between initial start up and the first refueling outage is any safer than any other period of plant operation.

57. The Board finds that prompt implementat ion of a f ully complaint Saf ety Parameter Display System is a design goal and of primary importance.

58. The Board finds that Applicants' witnesses are t ' qualified, by virtue of their lack of background in human f actors engineering, to testify as to the matter of the safety of operation of Seabrook Station until the first refueling outage.

59. The Board finds that the NRC Staf f's witness was not qualified to tes tif y as to, nor did he purport to have personal knowledge of, the technical basis f or the NRC S ta f f's determinat ion tha t

RHR Flow, Containment Hydrogen Concentration and Containment Isolation displays need not be added to the SPDS.

60. The Board finds that the record does not support the conclusion that the license conditions will all be met by the first refueling outage.

61. The Board finds that the Applicants have not sustained the burden of proof of showing that there is reasonable assurance that the health and safety of the population near the plant will be protected pending the correction of the deficiencies j in the Seabrook Station SPDS.

Proposed Form of Decision For the reasons stated above, Applicants have failed to meet their burden of proof pursuant to 10 CFR 62.732 and the applicat ion for an operating license for Seabrook Station is dismissed.

Respectfully submitted, SEACOAST ANTI-POLLUTION LEAGUE By its attorney, BACKUS, MEYER & SOLOMON i

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f Ro b'e r t A. Backus P. O. Box 516 116 Lowell Street Manchester, N . l! . 03105 Tel: (603) 668-7272 DATE: November 7, 1986 L

i I hereby cer t i f y tha t a copy of the wi thin SEACOAST ANTI-POLLUTION LEAGUE'S PROPOSED FINDSINGS OF FACT AND CONCLUSIONS OF LAW AND PROPOSED FORM OF DECISION has been sent this date, first class, postage prepaid, to those on the attached service list.

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/ pew Robeft A, Backus l

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