ML20214A463

Atty General Fx Bellotti Proposed Findings of Fact & Conclusions of Law & Proposed Form of Decision Re Onsite Emergency Planning & Safety Issues

ML20214A463
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Site:	Seabrook
Issue date:	11/12/1986
From:	Sneider C MASSACHUSETTS, COMMONWEALTH OF
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CON-#486-1546 OL-1, NUDOCS 8611200091
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UNITED STATES OF AMERICA 00LKETED NUCLEAR REGULATORY COMMISSION USNPI before the ATOMIC SAFETY AND LICENSING BOARD

• E6 NOV 17 A11 :31 GFFF:C-In the Matter of ) GCCyET .

) Docket No.(s) 50-433/444-OL-1 PUBLIC SERVICE COMPANY OF NEW ) On-site Emergency Planning HAMPSHIRE ET AL. ) and Safety Issues (Seabrook Station, Units 1 and 2) )

)

)

ATTORNEY GENERAL FRANCIS X. BELLOTTI'S PROPOSED FINDINGS OF FACT AND CONCLUSIONS OF LAW AND PROPOSED FORM OF DECISION Procedural History Attorney General Francis X. Bellotti does not take issue with the chronology of events set forth in Applicants' description of the procedural history of thi s case. See Applicants' Proposed Findings of Fact and Conclusions of Law and Proposed Form of Order, dated October 30, 1986, at 11 1-26. Attorney General Bellotti would only add to the Applicants' description of the procedural history that the Attorney General filed an Answer in Response to the Applicants' Motion for Issuance of an Operating License for Operation Not in Excess of 5% Rated Power, in which he raised certain issues in opposition to said license, including the issue that Applicants have not yet met all the pertinent emergency 8611200091 861112 PDR ADOCK 05000443 G PDR 3)So3

planning standards of S 50.47(b) and Part 50, Appendix E, which this Board summarily dismissed.

Contention SAPL Supp. 6 (Formerly NH-10)

Introduction Attorney General Bellotti adopts the description of this contention and the rulings thereon set forth in the

" Introduction" of the Seacoast Anti-Pollution League's (SAPL's)

Proposed Findings of Fact and Conclusions of Law and Proposed Form of Decision, dated November 7, 1986, at pages 4-6. The Attorney General also holds with SAPL that the contention as limited by the Board in its Memorandum and Order of September 15, 1986, does not apply the proper legal standard as set forth-in NUREG-0737, Supplement 1, and accordingly also adopts the proposed finding and ruling "A-A" set forth in SAPL's Proposed Findings at p. 6.

(a) Findings of Fact

1. Attorney General Bellotti adopts and incorporates by reference Applicants' proposed finding No. 55 and SAPL's proposed findings Nos. 1-55.

2. NRC Staff based its decision to allow Applicants to defer the correction of the Seabrook SPDS deficiences, in part, on the belief that there is no requirement that an SPDS be implemented prior to full power operation. Staff. Dir., Post Tr. 822 at 4, A.7.

a

3. NUREG-0737, Supplement 1, provides that "the fundamental requirements (for nuclear plant Emergency Response Capability] will be translated into binding legal requirements in the manner specified." NUREG-0737, Supplement No. 1, S 1,

p. 1.

4. NUPEG-0737, Supplement 1, specifies that licensees are required to furnish by April 15, 1983 a proposed schedule for completing actions to comply with the requirements of NUREG-0737, Supplement 1; each licensee's proposed schedule will then be reviewed by the NRC Project Manager, who will discuss the schedule with the licensee and mutually agree on schedules and implementation dates. The implementation dates will then be formalized into an enforceable document.

Applicants for operating licenses will develop plant specific schedules for the implementation of these requirements in a manner similar to that being used for operating reactors, taking into consideration the degree of completion of the power plant. Id. at pp. 1-2; Letter of D.G. E i se nhu t , forwarding NUREG-0737, supp. 1, (Generic Letter No. 82-33), at p. 2.

5. NUREG-0737, Supplement 1, further provides, "The proposal to formalize implementation dates in an enforceable document reflects the level of importance which the NRC staff attributes to these requirements," NUREG-0737, Supp. 1, S 1, a t

p. 2, and that procedures are to be established to ensure that the mutually agreed upon schedules are met "without significant delays and extensions." Id. at S 3.5(b), p. 5.

6. With respect to the requirement of a safety Parameter Display System (SPDS), NUREG-0737, Supplement 1, provides specifically that " prompt implementation of an SPDS can provide an important contribution to plant safety," Id. at S 4.l(d),

p. 8; and that " prompt implementation is a design goal and of at S 4.3, p. 9.

primary importance." Id. It further provides that " installation of the SPDS should not be delayed by slower progress on other initiatives." Id. at S 3.1, p. 4 . See .

I also, id. at S 4.3, p. 9.

7. NRC Guidance issued on June 1, 1983, to Project 4 Managers clarifies further that implementation schedules shall be established without further negotiations and shall be put in ,

an enforceable document. The guidance also provides that for operating license applicants an acceptable schedule shall be one in which the applicant commits to implementing all of the requirements in time for the staff to complete its evaluation prior to fuel load. Commonwealth Exhibit 1, Post Tr. 960 at 1 (emphasis added). The guidance provides only that for those whose proposed implementation dates are after fuel load that such commitments are to be made into license conditions. Id.;

R.J. Eckenrode, Staff Dir., Post Tr. 822, at A. 4, p. 3. It finally provides that the negotiation process should be completed by July 1, 1983. Commonwealth Exhibit 1, Post Tr.

960 at 2.

8. Pursuant to a letter f rom Public Service Company of New Hampshire (PSNH), dated April 14, 1983, and designated SBM e n,--- .n --, . n,,~,-->--,,n.,,-,,---w,-o-, -

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499, the applicant,in response to the requirement of NUREG-0737,, Supplement 1, originally committed to implementing its SPDS by December, 1983. Tr. 749; 828; 924-25. The date subsequently, proposed by PSNH for implementation of its SPDS and accepted by the NRC staff was June 30, 1986. Tr. 926. The proposed date for implementation was not after the date scheduled for fuel load. Tr. 926.

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9. NUREG-0737, supplement 1, provides that:

b' , Each operating reactor shall be provided with a safety Parameter Display System that

, is located convenient to the control room operators. This system will continuously display information from which the plant safety status can be readily and reliably assessed by control room personnel who are responsible for the avoidance of degraded

.t. 4 and damaged core events.

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f. The minimum information to be provided shall be sufficient to provide information to plant operators about:

(i) Reactivity control (ii) Reactor core cooling and heat removal from the primary system (iii) Reactor coolant system integrity (iv) Radioactivity control (v) Containment conditions.

NUREG-0737, supp. 1, SS 4.1(b) and (f), at pp. 7-8.

10. The NRC staff has identified a minimum set of approximately 20 plant parameters it believes to be sufficient to provide plant operators with information about the critical safety functions specified in NUREG-0737, Supplement 1. Staff Dir., Post Tr. 822 at 6, A.9.b.

11. Five of the twenty minimum parameters identified by the staff as required to be displayed on SPDS, or one quarter of them, are not displayed on the Seabrook SPDS. The parameters not displayed on Seabrook's SPDS are: (1) ,

containment isolation status; (2) steam line radiation; (3) stack radiation; (4) RHR flow; a nd ( 5 ) containment hydrogen concentration. I_d,.

12. Seabrook's containment isolation status is currently provided on a separate display located approximately 26 feet from the primary SPDS location. Tr. 977. The Staff concluded in SSER Supp. No. 6 S 18.2 at p. 9 that "the containment isolation display is not satisfactorily readable from the prime location to be considered part of the SPDS." See also, Staff Dir., Post Tr. 822 at 9, A.9.g. More specifically, the staff provided:

The conta'.iment isolation display is located a significant distance from the primary SPDS location so that it is difficult to read the legends. Unused cells appear to be randomly located so that pattern recognition is not a viable method of determining containment isolation. Furthermore, the display cells were designed to use two light bulbs each, but heat produced by two bulbs has caused the applicant to remove one bulb per cell. This one bulb condition reduces brightness and readibility and eliminates the redundancy in indication provided by two bulbs.

SSER Supp. No. 6, S 18.2 at p. 7, Attachment to Staff Direct, Post Tr. 822.

13. The Applicants' witness testified that the Applicants i

have corrected one of the three deficiencies of the containment isolation display noted by the Staff in SSER Supp. No. 6, S 18.2: they have wired the containment isolation display so i

that all unused cells are now set in one location in the light l

box. Tr. 783. The Staf f has not reviewed this correction to determine if an operator located 26 feet away at the Main SPDS display can now readily determine if a containment isolation r

valve, represented by a light cell measuring approximately 2 inches wide by 1 1/2 inches high, has been lef t open. Tr.

783-84; Tr. 856; Tr. 966.

14. The Applicants have not corrected the other two deficiencies relative to containment isolation display: the legends on the display are still not readible from the primary SPDS location, and redundant indication of valve failure is still lacking. Tr. 965.

l 15. Knowledge of containment isolation status is important t

l to safety: if containment isolation valves are left open during an accident, at least with respect to some of the valves, such as the containment purge valves, it would compromise a good degree of the containment's intended f unction and could, at least if two valves in a single line are left

! open, result in the release of radiation. Tr. 784, 863-64, 966-67.

16. - As specified in NUREG-0737, Supplement 1, an SPDS must provide information to plant operators about radioactivity control. NUREG-0737, Supp. 1, S ec t . 4.2.f(iv), at p. 8. The staff considers steam line radiation parameters to be the minimum status indicators of the radiological control function. Tr. 865.

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17. Prompt and accurate knowledge of steam line and stack radiation is important to safety: plant stack radiation data are significant because they indicate whether radiation has been released to the environment and are important for performing population dose projections; steam line radiation is an important variable to know in the event of an accident involving a steam generator tube rupture and can let an operator know when there has been primary or secondary leakage. Tr. 867, 967, 968, 781,

18. Steam line and stack radiation parameters are not currently dispayed on the Seabrook SPDS. Staff Dir., Post Tr.

822 at p. 6 A. 9.b. These two parameters are now displayed on the RDMS display that is located directly behind the STA's position at the main SPDS. Tr. 866.

19. The staff's witness testified that in the event of a problem with steamline or stack radiation the STA's attention would be drawn to the RDMS display in.back of him by auditory alarms. However, the witness did not know which specific parameters on the RDMS were alarmed. Tr. 969.

20. The witness testified that there are approximately one hundred separate hard-wired alarms in the Seabrook control room; that in the event of a multi-failure accident there could be a number of different alarms going off at once; and that, in fact, most accidents of any significance have more than one alarm going off at the same time. One of the problems at the accident at Three Mile Island was that there were so many

different alarms going off at once that the operator did not know where to look. Tr. 1000-1001.

21. In the event of a serious accident at Seabrook there are a number of actions the operator would need to be taking to mitigate the action and he would also need to be paying attention to a number of different parameters and displays.

Tr. 942.

22. Thus there can be no reasonable assurance that in the event of a serious accident where the STA's attention will be quite engaged at the SPDS console in f ront of him and a large number of alarms could be sounding at once, that his attention will actually be drawn in a timely manner to the RDMS display several feet in back of him. Tr. 866, 969, 942, 1000-1001.

23. Applicant's witness did testify that applicants intend to add a radiation screen including steamline and stack radiation parameters to the SPDS prior to proceeding to 5%

power operation. Tr. 774-75. However, that radiation screen will be added only to the SPDS lower level display; the stack radiation parameters will not be included on the top level SPDS display, Tr. 816, and without such parameters included on the top level display there can be no basis for presuming an operator's attention will be drawn to the lower level display when needed Tr. 970.

24. R?sidual Heat Removal ( RHR) flow is another of the critical parameters that the staff deems should be displayed on SPDS to assess critical saf ety f unctions, but which is not

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displayed on the Seabrook SPDS. Staff Dir. Post Tr. 822 at 5, A.8(2); SSER Supp. No. 6 S 18.2 at p. 3; Tr. 777-78, 861; 938-39. RHR flow is displayed on the main control board at two locations to the lef t of the main control board. Tr. 775. It is located approximately 26 feet, 6 inches, to the main control board and approximately 3 feet down that board from the main 4

SPDS location. Tr. 776.

25. RHR flow is an important indication because it lets plant operators know if the RHR pump is functioning properly.

Tr. 777. It allows the evaluation of the status of heat removal from the primary system when post accident cool dc a has progressed to the point where the RHR system provides the primary heat removal path. Lawrence Livermore Audit Reoort

'1 (Law Liv Rept.) at p. 10, Attachment to Staff Dir., Post Tr.

822.

26. NRC Staff's witness testified that Applicants could defer the installation of RHR flow on the Seabrook SPDS until first fuel reloading because that parameter is displayed elsewhere in the control room. Staff Dir. Post Tr. 822 at ,

p. 7; Tr. 861. However, the Staff witness did not know where in the control room the information on RHR flow was available i

or where it was in relation to the primary SPDS location. Tr.

862; 941. The Staff witness also did not know how the information on the RHR flow was provided on the main control i board, whether it was provided as a direct indication of the parameter or whether is was subject to inference or calculation

for its derivation. Tr. 862. Moreover, the Staff witness did not know how the STA's attention would be drawn to the RHR display elsewhere in the control room, and could testify'only that the operator's attention should be drawn to the RHR flow if that was the area in which he was involved and expecting something to happen. Tr. 942.

. 27. The NRC Staff witness was unable to testify based on personal knowledge as to the basis for his testimony that allowing the installation of the RHR flow parameter to be deferred until first fuel reloading will not pose a threat to the public health and safety. Tr. 940, 941. He testified that he did not understand the safety significance of requiring RHR flow parameter on the SPDS. Tr. 940. The witness testified that it was beyond his technical competence to know whether the RHR system is used when it is no longer practical to use the steam generator for decay heat removal because the reactor coolant system pressure has dropped, and could not answer the question posed to him whether an indication of RHR flow when the RHR system is supposed to be isolated f rom the reactor could indicate that the RHR system has not been successfully isolated. Tr. 1002; 939.

28. Containment hydrogen concentration is another parameter considered by the Staff to be part of the minimum information required to assess critical safety f unctions and that is not displayed on the Seabrook SPDS. Staff Dir. Post Tr. 822 at 5, Tr. at 861.

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29. It is critical for safety to know containment hydrogen concentration levels because high hydrogen concentration in containment poses a challenge to containment integrity. Law.

Liv. Rept. at 10, Attachment to Staf f Dir. , Post Tr. 822. High hydrogen concentration in the containment is one factor that could lead to a hydrogen explosion. Tr. 945. Containment hyrdogen concentration would also provide an important indication that fuel cladding is being damaged. Tr. 778. One of the problems during the accident at Three Mile Island was that operators were not aware of the significant hydrogen concentration in the containment. Tr. 944.

30. The Applicant's witness justified not adding the containment hydrogen concentration parameter to the Seabrook SPDS based on the fact that the parameter is presently indicated on the main control board. Agg. Dir., Post Tr. 739 at p. 3. The witness did not testify as to where on the main control board the parameter was located, where that parameter was situated wth respect to the main SPDS location, or how an

operator's atttation would be drawn to the parameter, except tc state that the parameter would be considered when using emergency procedures to respond to plant upset conditions.

Id. The witness testified that in his opinion there was no need for a containment hydrogen concentration parameter to be ever added to the Seabrook SPDS. Tr. 770.

I 31. The Staff witness testified that deferring the addition of containment hydrogen concentration to Seabrook's

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SPDS is justified because the information on containment hydrogen concentration is available elsewhere in the control room. Staff Dir. Post Tr. 822 at pp. 6-7, A.9.b.

32. The staff witness did not know where in the control room the information on hydrogen concentration was available, but he assumed that such information could not be seen from the STA position at the primary SPDS display. Tr. 944. The witness also did not know whether the information on hydrogen concentration that was available elsewhere in the control room was provided as a direct indication of the parameter or whether it was provided through inference or calculation, and he did not know how long it would take to obtain the information on hydrogen concentration from the main control bcard. Tr. 862.

33. The staff witness was unable to provide a technical judgment as to whether Seabrook's not having a hydrogen concentration parameter in place on its SPDS prior to the first refueling outage would present a safety hazard. Tr. 949-50.

He did not know the basis for the requirement that a hydrogen concentration parameter be displayed on SPDS. Tr. 942-42. And he did not know the significance of a high level of hydrogen concentration in the containment or whether there could be a potential for serious off-site cer;7quences if the operator was not aware of containment hydrogen concentration levels. Tr.

943. With respect to hydrogen concentration in the containment, the witness did not know what the source of that hydrogen would be or f rom what part of a potential accident i

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sequence or sequences it might come. Tr. 984.

34. The staff witness testified that the noncompliance of the Seabrook SPDS with Item I.D 2 of NUREG-0737, Supplement 1, should not present a serious safety problem at Seabrook in part based on the conclusion that:

The SPDS in its current design should not provide erroneous or misleading information to plant operators and therefore will not increase the potential for operator error in the event of an abnormal occurrence at Seabrook. . . .; [and]

the current Seabrook SPDS does provide useful information to the plant operators (albeit not all the information called for in NUREG-0737, Supplement 1). . . .

Staff. Dir., Post Tr. 322 at p. 4, A.7. Nevertheless, the witness testified that the data validation algorithms on Seabrook's SPDS may not be sophisticated enough to ensure that the data displayed to the operator are valid, and that in some cases incorrect and misleading data would be displayed. Staff.

Dir., Post Tr. 822, at p. 7; SSER Supp. No. 6 S 18.2 at p. 4; Tr. 970-71.

35. With respect to the data validation algorithms the Lawrence Livermore audit repor* provides:

The Seabrook SPDS does not completely satisfy the provisions of Supplement 1 to NUREG-0737 i regarding rapid and reliable display because the l data validation techniques used are insufficient l to provide a highly reliable synthesized value of

! SPDS parameters. . . . The use of high or low values provided by redundant instrumentation may result in a conservative estimation of the status l of Critical Safety Functions but it also ensures that the operator will be misled about safety function status in the event of large instrument errors or on-scale instrument failures. Use of average vilues without additional validation checks does not guarantee the operator will be consistently misled in the conservative direction. PSNH must implement validation methodology that maken more effective use of redundant information available via the MPC.

Law Liv. Rept. at p. 13, S 4.3.2, Attachment to Staff Dir.,

Post Tr. 822,

36. The Lawrence Livermore Audit Report also calls into question the usefulness of some of the data displayed to aid the operator in rapidly and reliably determining safety status when it states:

PSNH could also improve the usefulness of the existing validity screening of input data by tightening the reasonableness band applied to some parameters. For example, at the time of the audit, PSNH was using 0 F as the lower limit for reasonableness check of temperature inputs and 200 percent as upper limit for the reasonableness check of reactor power. The audit team believes more meaningful bounds could be established in both cases.

Id-

37. Applicants' witness did not agree with the Staff that more sophisticated methodology is needed to ensure that valid data are displayed on the SPDS, because operators are required to validate any SPDS conclusions prior to implementing any corrective actions. Tr. 806. However, since the data validation methodology does not guarantee that the operator will be consistently misled in the conservative direction, the operator will not always have an indication on the SPDS of an 4 abnormal condition when one exists, and in that case the

operator will have no cause to validate the improper SPDS i

indication. Tr. 806-807.

38. The NRC staff witness agreed with the Lawrence Livermore Report that the data validation methodology could affect the operator's ability to rapidly and reliably assess plant safety status. Tr. 970-71.

39. The NRC staff also testified that the usefulness of the Seabrook SPDS lower level display formats is in question.

Staf f Di r. , Post Tr. 822 at p. 5, A.8.(9); Tr. 971.

40. The lower level display formats are diagrams of the logic trees for the Critical Safety Functions. Their purpose is to indicate to the person monitoring the SPDS the condition of that Critical Safety Function. Tr. 971. Operators are instructed to use the lower level displays when the SPDS

indicates something that they should react to. Tr. 815.

41. Both the Staff and the Applicants' witnesses agreed that one of the best measures of the usefulness of a piece of equipment is whether or not those by whom it is intended to be used actually use i t. Tr. 844, 760.

42. On May 20 and 21, 1986, the NRC audit team observed a simulator drill conducted by PSNH to demonstrate the use of the SPDS under plant upset conditions. The audit team noted that during the entire course of the drill Critical Safety Functions status was monitored by the Shift Technical Advisor (STA) using hardwired instrumentation and hard copies of the CSF status trees. At no time during the drill did any of the 36 operators involved select for display an SPDS CSF status tree (lower

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level display). Law Liv. Rept, at p. 14, S 4.4.1, Attachment to Staff Dir., Post Tr. 822; Tr. 760-761; 815; 843; 972.

43. Applicants' witness testified, with respect to the issue of the questionable usefulness to the operators of the Seabrook SPDS lower level displays, that the Staff's comments indicate a disagreement on specific usage of the SPDS lower level displays; that it is possible to satisfactorily monitor plant status without SPDS lower-level displays being available on the main computer. Apg. Dir., Post Tr. 739 at p. 6. The Applicants' witness clarified during cross-examination that one could not satisf actorily monitor plant status by using only the

) SPDS top level displays. Tr. 759. Rather, operators could i monitor plant status by patrolling the main control board with hard copies of the lower-level displays. Tr. 759. In the Applicants' view, the enti re " computerized portion" of SPDS

! could be eliminated without defeating the goal of safe and efficient power generation. Tr. 789-90.

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44. The NRC staff witness testified, however, that one of the lessons learned from Three Mile Island is that a control i

room is rather large and that the information on safety-related ,

equipment is spread throughout the control room. Therefore,

the purpose of an SPDS is to bring together a lot of dispersed information into a single picture of safety. Tr. 931, 840.

The SPDS is intended to provide operators with a first indication of things that might be occurring, which they can then verify with their class lE instrumentation. Tr. 840.

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45. The staff witness further testified that the hard copy of the Critical Safety Function Status trees is used only to maintain a dynamic value. Tr. 847. The hard copy is not part f

of the SPDS. Tr. 972. The hard copy does not i ndicate, as do the lower level displays on SPDS, whether a CSF status tree is under challenge or extreme challenge but requires that the operator patrol the main control board to obtain the relevant data on the CSPs. Tr. 845-847; 790.

46. The NRC considers lower level SPDS displays to be an aid to operators in assessing abnormal conditions. Tr. 971.

The NRC witness, along with the audit team, was not able to conclude that the Seabrook lower level displays provide the required operator aid in the determination of plant safety status. Tr. 847.

47. The NRC Staff has not evaluated the reason why the lower level displays on the Seabrook SPDS are not useful to the operators and has asked the utility to evaluate the problem.

Tr. 972. The Staff does not know yet whether it i s a problem with the format, the type of information available, or some

other reason that makes it more difficult to use the lower level displays than the hard copy. Tr. 972, 980.

48. The method for accessing the lower level display on the Seabrook SPDS i s awkward and more difficult than the NRC staff thinks it should be. Staff Dir. 822 at 7; Tr. 855.

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There are several ways of doing it better. Tr. 855. The staff recommends that there be a single operator action for callup of y - + . - - - -' -.v.-ww--y - , -+--w w-*g r- ,.-er-,w m - - -

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each of the second-level displays. Staff Dir. Post Tr. 822 at

9. The audit report states that the dif fi culty in accessing the lower level SPDS displays should be evaluated as a potential source of the operators' reluctance to'use the status tree displays. Law Liv. Reot, at p 17, Attachment to Staff Dir., Post Tr. 822.

49. Another potential deficiency identified by the Staff that could affect the speed and/or reliability of the Seabrook SPDS is the system response time. Staff Dir., Post Tr. 822 at

p. 7, A.9.c.

50. NUREG-0737, Supplement 1, calls for continuous, concise and rapid displays. NUREG-0737, Supp. 1 S 4.l(b) at

p. 7; Tr. 859. The NRC witness testified that a 5 second response time would satisfy the requirement of a rapid

, display. Tr. 859,

51. The Seabrook SPDS response time has only been tested during a lightly loaded sequence. Id.; Tr. 974. The Applicants have not demonstrated that the SPDS update and response times will not be unacceptably affected by the high main plant computer loading conditions expected to occur during response to a severe plant upset. Tr. 857, i

52. There is a very good chance that in the event of a severe accident at Seabrook that a large number of nearly simultaneous processing demands will be made on the main plant computer. Tr. 974. Such high demand could significantly slow down the update rate of the Seabrook SPDS, depending on what i

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priority the SPDS has on the main plant computer operation.

Tr. 974-75. The NRC staff does not know what priority the Seabrook SPDS has on the main plant computer operation. Tr.

975. If the update rate were slowed down the SPDS would provide the operator with out-of-date data upon which he would be giving advice. Tr. 975. A long response time, of more than 8-10. seconds, would also cause the operator to make mistakes because, according to the NRC witness, he would think he asked for something wrong and try it again when nothing appears on the screen. Tr. 859.

53. There can be no assurance that during an accident situation when it is most needed that the SPDS update rate and response time will be within an acceptable amount of time. Tr.

857; 860.

54. NUREG-0737, supplement 1, requires that the SPDS continuously display information from the plant safety status that can be readily and reliably assessed. NUREG-0737, supp.

1, S 4.1.b at p. 7. The SPDS is supposed to be used in addition to other control room instrumentation used to

, determine safety status and serves to aid and augment those other components. NUREF-0737, Supp. 1, S 4.1.c at p. 7.

55. At Seabrook, all control room displays could be selected such that no SPDS is provided in the control room.

Staff Dir. Post Tr. 822 at 2. The Applicant's witness testified at one point in the hearing that the Applicants did not know when the software would be developed and input into i

the computer to allow for a continuous display. Tr. 764.

Later in the hearing they testified that the software has been

developed and that it should be in place prior to proceeding to 5% power operation. Tr. 805. Until such time as a continuous 4

display is in place, the Applicants propose to have the SPDS Shift Technical Advisor CRT continuously display the top level SPDS screen. Tr. 764. However, Applicants have not adopted any administrative procedures to ensure that the SPDS will be continuously displayed as stated. Tr. 805.

56. The NRC does not yet have accurate availability calculations for Seabrook's SPDS. Tr 973.

57. The NRC finds system availability of note than 99% to be acceptable. Tr. 973; Staff Dir., Post Tr. 822 at p. 7, A.9.c.(3). One of the problems during the incident at the Davi s-Besse plant last year was that the SPDS was not available. Tr. 990.

58. The availability calculations for Seabrook's SPDS do not yet include the availability of Reactor Vessel Level l

Instrumentation Systems (RVLIS) and the Radiation Data Management System (RDMS). Staff Dir., Post Tr. 822 at 7 7, A.9.c.(3).

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59. The availability calculations which are available for the Seabrook SPDS are based on records maintained by PSNH on the availability since December, 1985, of the Main Plant l

Computer (MPC) system, and on PSNH's stated intention to maintain a complete set of MPC spare parts on site and to have

_, . ~ . _ _ _ _ _ __- _ _. __..._ _ _ __ _ ___ _. . _ _ _ _ _ .

qualified maintenance staff available on all shifts. Law Liv.

Rep't. at p. 13, Attachment to Staff Dir., Post Tr. 822.

However, Monacorp., the company that is the manufacturer of the Main Plant Computer, is no longer in business. Tr. 787

60. The NRC staff witness testified that he was the person at NRC principally responsible for the recommendation that modifications to the Seabrook SPDS could be handled as license conditions and deferred until first fuel reloading. Tr.

867-68. The witness testified that he based his recommendation to allow deferral of the SPDS on the onsite audit of May 20-23, I

1986 and on a review of a draft of the Lawrence Livermore audit report. Id. In fact, the witness had already concluded by April, 1986, a month before the audit, when the Seabrook Statior. Safety Evaluation Status Report was issued, that any deficiencies in t.he SPDS identified by the audit would not need to be corrected until the first ref ueling outage. Tr. 930; i

Comm. Ex. No. 2, Post Tr. 964.

61. The Seabrook SPDS is not yet acceptable. Tr. 934; Staff Dir., Post Tr. 822. According to the Staff, Applicants are still obviously having problems putting in place an SPDS that the Staf f thinks will do the job required by the lessons learned from Three Mile Island. Tr. 982.

62. NRC Staff has proposed that conditions be imposed on the Seabrook operating license which would require that all

, deficiencies in the Seabrook SPDS be corrected prior to restart after the first refueling outage. Staff Dir., Post Tr. 822.

i l

A period of time of 12 to 18 months typically elapses between initial start up and the first refueling for a pressurized water reactor such as Seabrook. Tr. 834. The period af plant operation f rom initial start-up to fi rst refueling is not any safer than any other subsequent period of plant operation. Tr.

768; 837; 975.

63. Staff witness testified that a prime basis for his decision to allow the deferral of corrections to the Seabrook SPDS until the first refueling outage is that the information not available on the Seabrook SPDS is available elsewhere in the control room. Staff Dir. Post Tr. 822 at p. 4, A.7,; Tr.

932. Applicants rely almost entirely on the fact, that even without an SPDS operators can obtain all the necessary information on plant safety status elsewhere in the control room, to justify their conclusion that completion of the Seabrook Station SPDS may be deferred until first ref ueli ng .

64. NUREG-0737, Supplement 1, states that an SPDS is supposed to be in addition to other safety monitoring equipment in the control room, that the purpose of an SPDS is to aid and augment those components; and that prompt implementation of SPDS provides an important concibution to plant safety and is a design goal and of primary importance. NUREG-0737; Supplement 1, at SS 4.1.c 4.1.d, and 4.3, pp. 7-9.

65. An SPDS aids operators in rapidly and reliably assessing the safety status of the plant. Tr. 935; NUREG-0737, supp. 1, S 4.1.b at p. 7. The purpose of SPDS is to give i

j operators a first indication of things that might be occurring which they can then verify with their class IE instrumentation Tr. 840.

66. It is the purpose of an SPDS to pull together a number of parameter values so that the STA or other operator can handle a large number of off-normal conditions or at least recognize these conditions and recognize some pattern to their occurrence. Tr. 984-85.

67. At Three Mile Island, Unit 2, most critical parameters were also on display somewhere in the control room. Tr. 936.

68. One of the lessons learned from Three Mile Island is

, that a control room is rather large and that the information on safety-related equipment is spread throughout the control room. The purpose of SPOS is to bring the major things together to show a general picture of safety. That is the basis for the NUREG-0737, Supplement 1, requirement that in addition to the control room equipment there be one concise

. display in the control room where all the critical safety i

parameters are displayed. Tr. 931, i

(b) Conclusions of Law

69. Attorney General Bellotti adopts and incorporates by

reference SAPL's proposed conclusions of law Nos. 56-61.

70. The Board finds, based on the amount of time that Applicants have had to implement an SPDS, the Applicants' i

1

I previous commitments to the Staff to implement a complete SPDS prior to fuel loe. ding, and the level of importance placed on prompt implementation of an SPDS by NUREG-0737, supplement 1, that deferring the implementation of a complete SPDS for Seabrook until the first refueling outage does not meet the requirements of NUREG-0737, Supplement 1, for " prompt i mpleme nt at i on . " - (See Findings Nos. 2-8.)

71. The Board finds that the Applicants have not sustained their burden of demonstrating that the health and safety of the public near the Seabrook plant can be protected when the Seabrook SPDS display does not contain five of the twenty minimum parameters necessary to assess Critical Safety Functions. (See Findings Nos. 9-12, 16, 18, 24, 27, 28, 33, 60-62, 63, 67.) The Board finds that knowledge of the missing parameters is important to safety (see Findings Nos. 15-17, 25, 27, 29) and that there has been an inadequate showing by Applicants that operators will in all cases be appropriately apprised of such parameters in the event of a serious accident at Seabrook. (See Findings Nos. 13, 14, 19-22, 23, 26, 30, 32, 38.) Furthermore, the Board finds that the purpose of having one concise display of all critical parameters so that the STA can get an overall picture of the safety status of the plant is defeated when some of those parameters are set in various locations throughout the large control room. (See Findings Nos. 44, 45, 54, 64, 66-68.)

72. The Board finds that the Applicants have not met their E

e O'

burden of establishing reasonable assurance that the health and safety of the public near Seabrook will still be protected if deficiencies in Seabrook's SPDS are not corrected until the first reloading outage, because the existing SPDS could lead to operator error in assessing accident conditions (See Findings Nos. 34, 35, 37, 38,-52) and may not be available or useful to operators when most needed. (See Findings Nos. 36, 39, 40-43, 46-59.)

Proposed Form of Decision For the reasons stated above, Applicants have failed to meet their burden of proof pursuant to 10 C.F.R. S 2.732, and no operating license for Seabrook Station may issue until or unless the Applicants have corrected all deficiencies of SPDS set forth in Section 18.2 of the SSER Supplement No. 6 and have performed sufficient validation testing to ensure that the usefulness of the SPDS is thoroughly established.

Respectfully submitted, FRANCIS X. BELLOTTI, t- s- '

By: - 4 Carol S. Sneider Assistant Attorney General Environmental Protection Division Department of the Attorney General One Ashburton Place, Room 1902 Boston, MA 02108-1698 (617) 727-2265 DATED: November 12, 1986