ML20205F384

Partial Initial Decision.* Partial Initial Decision LBP-87-10 on Onsite Emergency Planning & Safety Issues. Served on 870326

ML20205F384
Person / Time
Site:	Seabrook
Issue date:	03/25/1987
From:	Harbour J, Luebke E, Wolfe S Atomic Safety and Licensing Board Panel
To:
References
CON-#187-2914 82-471-02-OL, 82-471-2-OL, LBP-87-10, OL-1, NUDOCS 8703310152
Download: ML20205F384 (71)

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DOCMETED LBfh-10 UNITED STATES OF AMERICA -. . . ..

NUCLEAR REGULATORY COMMISSION '87 MAR 25 P4:15 ATOMIC SAFETY AND LICENSING BOARD OFFICE Cr 2EcrCTAE.y Before Administrative Judges: DON g m Sheldon J. Wolfe, Chairman Emmeth A. Luebke Jerry Harbour SEWED MAR 261987

) Docket Nos. 30-443-0L-1 In the Matter of ) 50-444-0L-1

)

PUBLIC SERVICE COMPANY ) (On-Site Emergency Planning 0F NEW HAMPSHIRE, et al. ) and Safety Issues)

)

) (ASLBP No. 82-471-02-OL)

(Seabrook Station, Units 1 and 2) )

) March 25, 1987 PARTIAL INITIAL DECISION Appearances Robert A. Backus, Esq., Backus, Meyer & Solomon, Manchester, New Hampshire, for Intervenor Seacost Anti-Pollution League Diane Curran, Esq., Harmon & Weiss, Washington, D. C.

for Intervenor New England Coalition on Nuclear Pollution Carol Sneider, Esq., Office of the Attorney General, Bostnn, Massachusetts, for the interested Conrionwealth of Massachusetts George D. Bisbee, Esq., Office of the Attorney General, Concord, New Hampshire, for the interested State of New Hampshire Thomas G. Dignan, Jr. , Esq. and R. K. Gad, III, Esq. , Ropes and Gray, Boston, Massachusetts for Applicants the Public Service Company of New Hampshire Robert G. Perlis, Esq., Office of the r,eneral Counsel, Washington, D. C., for Staff U. S. Nuclear Regulatory Commission s

8703310132 DR 870325 ADOCK 05000443 PDR ..

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9 d ,' . i. C di, " T5' TABLE OF CONTENTS PAGE I. INTRODUCTION 1 A .' Background 1 B. Content of the Opinion and Findings 6 II. CONTENTIONS 6 A. Contentions NECNP III.1 and NH Classificiation Scheme and Emergency Action Levels ,

6 B. SAPL Supplemental Contention 6 - Compliance of the -

Seabrook Safety Parameter Display System (SPDS)

With NUREG-0737, Item I.D.2 7 C. Electrical Equipment Environmental Qualification Time Duration 14 III. CONCLUSIONS 19 l

l l

IV. FINDINGS OF FACT 20 l 20

! Emergency Classification and Action Level Scheme l Safety Parameter Display System 27 l Electrical Equipment Environmental Qualification Time Duration . 47 l

CONCLUSIONS OF LAW 66 l

ORDER 67 t-i-

o-LBP-87-10 UNITED STATES OF AMERICA

. NUCLEAR REGULATORY COMMISSION ATOMIC' SAFETY AND LICENSING BOARD Before Administrative Judges:

Sheldon J. Wolfe, Chairman Emmeth A. Loebke Jerry Harbour

) Docket Nos. 50-443-0L-1 In the Matter of ) 50-444-0L-1

)

PUBLIC SERVICE COMPANY ) (On-SiteEmergencyPlanning 0F NEW HAMPSHIRE, et_ al. ) and Safety Issues)

)

) (ASLBP No.- 82-471-02-OL)

(Seabrook Station, Units 1 and 2) )

) March 25, 1987 PARTIAL INITIAL DECISION (Operating License)-

OPINION I. INTRODUCTION A. Background On July 9, 1973, the Public Service Company of New Hampshire, et al. (Applicants) had filed with the then U. S. Atomic Energy Commission an application for licenses to construct Seabrook Station, Units 1 and 2. Each of the units is a Westinghouse pressurized water reactor and each is designed to operate at a thermal power of 3411 megawatts. The site of the nuclear generating facility is located on the western side of Hampton Harbor, in the township of Seabrook, Rockingham County, New Hampshire, and is approximately eleven miles

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south of Portsmouth, New Hampshire and forty miles north of Boston, Massachusetts. After a public hearing before an Atomic Safety and Licensing Board, the construction permits were issued on July 7,1976.

The application for operating licenses was docketed by the Nuclear Regulatory Commission on October 5, 1981. Notice of the opportunity for requesting a public hearing was published in the Federal Register on October 19, 1981. (46 Fed. Reg. 51330.) On November 30, 1981, an Atomic Safety and Licensing Board was constituted and the following Administrative Judges were appointed: Helen Hoyt, Chairman, Emeth Luebke and Oscar Paris. On August 25, 1982, the Licensing Board was reconstituted with Administrctive Judge Jerry Harbour being appointed to serve in lieu of Administrative Judge Paris.

Ultimately, pursuant to 10 C.F.R. 92.714(a), the Licensing Board admitted various individuals and organizations as intervening parties (Intervenors), and, pursuant to 10 C.F.R. 92.715(c), it permitted representatives of various interested States and municipalities to participate in the proceedings. Amongst those admitted as intervening parties were: New England Coalition en Nuclear Pollution (NECNP) and Seacoast Anti-Pollution League (SAPL). Amongst those permitted to participate as representatives of interested States or municipalities were the Attorney General of the State of New Hampshire (NH) and the Attorney General of the Commonwealth of Massachusetts (Mass.).

During a hearing held in August 1983, the then presiding (Hoyt)

Board heard the Applicants' and the Staff's evidence upon three on-site

E3 Cb emergency planning and safety issues.1 That Board also heard evidence presented by Apolicants, the Staff and Mass, upon an off-site emergency planning issue. After the closing of the record, the Applicants, the NRC Staff, and NECNP filed proposed findings and conclusions of law with respect to NECNP I.B.2 and with respect to NECNP III.1 and NH 20. NH filed submissions only with respect to NH 20. Applicants, the NRC Staff, Mass., SAPL and NECNP filed proposed findings of fact and conclusions of law with respect to NECNP III.12-III.13. On November 23, 1983, Applicants filed a reply to the various proposed ~ findings.

The Hoyt Board did not issue a partial initial decision with respect to the contentions referred to above. On September 9, 1985, the Board was reconstituted and this Board (the Wolfe Board), consisting of 1 NECNP Contention I.B.2 asserted that Applicants had not satisfied the requirements of GDC 4 that all equipment important to safety be environmentally qualified because Applicants had failed to specify the time duration over which the equipment was qualified.

Similar Contentions NECNP III.1 and NH 20 asserted, in substance that the emergency plans did not contain an adequate emergency classification scheme as required by 10 C.F.R. $50.47 and Appendix E, and by NUREG-0654, 2

As reworded by the Board, NECNP III.12-III.13 read as follows:

Evacuation Time Estimate "The evacuation time estimates provided by the Applicants in Appendix C of the Radiological Emergency Plan are deficient in failing to include an estimate of: 1) the times for evacuation during adverse weather conditions developing on a busy summer weekend; and 2)the times for simultaneous evacuation of beach areas lying NE to SSE of the Seabrook site."

I r

t Administrative Judges Sheldon Wolfe, Chairman, Emeth Luebke and Jerry Harbour, was appointed to preside over all on-site safety and emergency planning issues. (50 Fed. Reg. 37608.) The Hoyt Board retained jurisdiction over all off-site emergency planning issues.3 In an Order issued on November 4, 1985 (unpublished), this Board stated as follows:

We have reviewed the record and have concluded that the record needs to be reopened for the limited purpose of supplementation. It is not our intention, and we will not permit the retrying of issues heard before the closing of the record on August 23, 1983. After a prehearing conference, and af ter discovery, if any, a supplementary hearing will be ordered to take evidence on the above-identified matters pertaining to Contentions NECNP I.B.2, NECNP 111.1 and NH 20, which involve significant health and safety issues, and which were not previously ripe for hearing.

Footnote 2 stated that, if NH Contention 10 (Detailed Control Room Design Review) was not informally resolved, evidence would be taken on that contention as well during the supplementary hearing.

In a Memorandum and Order of July 21, 1986, LBP-86-22, 24 NRC 103, the Board granted NH's motion to withdraw its Contention 10, but, because SAPL had preserved its rights as a joint intervenor with respect i to that contention, it converted NH 10 to and replaced it with SAPL i

! Supplemental Contention 6.

4 3

Subsequently, on March 25, 1986, Judge Hoyt ruled that that Board had jurisdiction over the evacuation time estimate contention in its entirety, both as to the prior litigation and as to any further litigation on that issue before that Board. In a Memorandum and Order of August 14, 1986 (unpublished), this Board ruled that NECNP Contention III.12-III.13 did not present an on-site emergency planning issue.

y _-

l*

In a motion which had been filed on June 17, 1986, Applicants l

requested, inter alia, that the Board's Partial Initial Decision when issued should authorize issuance of an operating license for operation not in excess of 5% of rated power. The Memorandum and Order of July 25,1986 LBP-86-24, 24 NRC 132, reflected that, after considering the evidence presented during the supplementary hearing, the Board would decide in its Partial Initial Decision whether or not to authorize issuance of an operating license for operation of Seabrook Unit I up to and including 5% of rated power.

On September 15, 1986, the Board partially granted Applicants' motion for summary disposition of SAPL Supplemental Contention 6.

(LBP-86-30, 24 NRC _.)

The reopened hearing began on September 29, and proceeded on

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September 30, October 1 and 0ctober 3, 1986.4 The same parties and interested States, which had attended and participated in the 1983 hearing, also attended and participated in the 1986 reopened hearing.

Proposed findings of fact, conclusions of law, proposed forms of decision, and briefs were filed on the following dates: Applicants -

October 30; SAPL - November 7; NECNP - November 12; Mass. - November 12, 4 Limited appearance statements were received during the initial August 1983 hearing and during the reopened 1986 hearing.

i i

1986; NRC Staff - November 26, 1986.5 Applicants filed a reply on December 1, 1986.

On November 25, 1986, Applicants advised that Unit 2 had been officially cancelled.

B. Content of the Opinion and Findings

! Part II of this Opinion discusses and resolves the contentions.

Part III reflects our conclusions. The Board's underlying Findings of Fact and Conclusions of Law are appended and incorporated by reference.

An Order is also appended.

It should be noted that all of the proposed findings of fact and conclusions of law submitted by the parties that are not incorporated directly or inferentially in this Partial Initial Decision are rejected I

as unsupported in law or fact or as unnecessary to the rendering of this l

Partial Initial Decision.

6 II. CONTENTIONS A. Contentions NECNP III.1 and NH Classification Scheme and Emergency Action Levels (Fdgs. 1-13) l 5

NH did not file these submissions. NECNP's submissions were limited to addressing NECNP Contention I.B.2, Mass's and SAPL's submissions were limited to addressing SAPL Supplemental Contention

6. Only the Applicants' and the Staff's submissions addressed all of the on-site safety and emergency planning contentions.

6 These contentions constitute the only remaining issues in controversy with respect to en-site safety and emergency planning matters.

O In substance these contentions assert that, contrary to the requirements of 10 C.F.R. 50.47 and Appendix E and of NU9EG-0654, the emergency plan ~does not contain an adequate emergency classification and action level scheme. It should be noted that, pursuant to a stipulation, the written direct testimonies of the Applicants' and the Staff's witnesses were admitted into evidence and incorporated into the 1986 record as if read. There was no cross-examination, and only the Applicants and the Staff filed proposed findings of fact, conclusions of law and briefs with respect to these contentions. Thus', although these two contentions are no longer controverted issues, we decided to prepare factual findings and to set forth our conclusion.

At the time of the 1983 hearings, the Applicants' emergency classification and action level scheme was not complete. In light of the supplementary evidence presented during the course of the 1986 hearing, we conclude that Applicants' emergency classification and action level scheme fully satisfies the requirements of 10 C.F.R. 550.47 and Appendix E of Part 50 and meets the guidance criteria of NUREG-0654.

B. SAPL Supplemental Contention 6. Compliance of the Seabrook Safety Parameter Display System (SPDS) With NUREG-0737, item I.D.2. (Findings 14-47)

The central issue of this contention is whether there is reasonable assurance that the health and safety of the population in the immediate vicinity of the plant will be protected if corrections to deficiencies in the Seabrook SPDS are deferrod until the first refueling outage.

O Requirements for the SPDS are set forth in NUREG-0737, Supplement 1,7 a Commission-approved document providing certain post-TMI requirements and guidance to be implemented both by applicants for, and holders of, operating licenses for power reactors, in order to upgrade emergency response capability and facilities.

With respect to litigation of TMI-2 issues in operating license proceedings, the Commission specifically endorsed NUREG-0737 requirements as being necessary for responding to the accident at TMI-2, and categorized the NUREG-0737 requirements, like those-in NUREG-0694,8 as falling into two categories in terms of their relationship to existing regulations:

(1) Those that interpret, refine or quantify the general language of existing regulations, and (2) Those that supplement the existing regulations by imposing requirements in addition to specific ones already contained therein.

(Statement of Policy; Further Coninission Guidance for Power Reactor Operating Licenses, 45 Fed. Reg. 85236, 85238 (December 24,1980).) The requirements for implementing the SPDS fall into the second category.

" Requirements for Emergency Response Capability (Generic Letter No.

82-33)", transmitted to licensees and applicants by letter dated December 17, 1982.

0 "TMI-Related Requirements For New Operating Licenses", June 1980.

Specific requirements for an SPOS were not included in NUREG-0694 but were included as Item I.D.2 in NUREG-0737 which superseded it.

The implementation schedule of TMI Action Plan requirements for applicants for an operating license was given in Enclosure 2 of NUREG-0737 (pp; 2-3 to 2-11). Depending upon safety significance and the immediacy of the need for corrective actions, the schedule required implementation of different items at various times, such as prior to fuel load, prior to initial criticality, prior to full power, by some fixed date, or for some requirements, by a schedule to be determined.

The implementation date for the SPDS requirements fell into the last category. While Supplement I to NUREG-0737 indicated that schedules therein superseded previous schedules, the schedule for implementation of the SPDS remained unfixed and to be set by agreement between the Applicants and Staff. The Board notes that the SPDS was never included among those requirementi whose implementation was required prior to fuel ,

load or prior to initiill criticality.

While Supplement 1 to NUPEG-0737 permits implementation of SPDS requirements by a schedule agreed upon between the NRC Staff and Applicants, it also stresses prompt implementation as an important contribution to plant safety. NUREG-0737, Supplement I does not require implementation prior to initial criticality and no evidence was adduced to indicate that it must be.

The principal function of the SPDS is to aid control room operators during abnormal and emergency conditions in determining safety status of the plant and in assessing abnormal conditions that may require corrective actions to avoid a degraded core.

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The Seabrook Station SPDS is not in full compliance with the requirements (and guidance) provided in NUREG-0737, Supplement 1, because of certain deficiencies of disparate severity found by the NRC Staff in its review of the operating license application. (See fdg. 27, l

infra.) The severity of deficiencies ranges from those in" guidance" items, i.e., methods of achieving particular functions or operations (deficiencies 5, 8 and 9), to absence of minimum or critical plant variables specifically required by NUREG-0737 Supplement 1, as part of the SPDS displays, e.g., absence of displays for residual heat removal (RHR) flow and containment hydrogen concentration variables (deficiency

?).

One SPDS deficiency (No. 7), which had been fully resolved before l the hearing, involved proper isolation of nonsafety grade circuits of i

the SPDS from the Class IE systems to protect the safety systems frcm l

possible interference. Existence of this deficiency, which is controlled by requirements for safe interconnection of safety and L nonsafety related systems as well as by SPDS requirements, was one of the main reasons that impelled this Board on September 15, 1986 to order adjudication of the status of the SPDS. LBP-86-30 at 17, 24 NRC .

Evidence of the resolution of this item received at the hearing was i

uncontroverted, and the resolution is described in Appendix 8-A of Supplement 6 of the SER.

l This Board has not attempted to make any independent evaluation of the relative safety significance of individual deficiencies; indeed the record would not completely support such an evaluatinn. Instead, we

have relied upon NUREG-0737 which sets forth certain requirements for the SPDS and describes some of the critical safety function requirements as " minimum information to be provided," for which the Staff has identified a minimum set of 20 plant parameters that it believes to be sufficient to provide plant operators with information about the critical safety functions. The general standard for resolutions that we have applied is that each of these specific SPDS requirements shall be met, or that equivalent alternative means for the control room operating personnel at the prime SPDS station to obtain the information, shall be implemented prior to operation at levels exceeding five percent of rated power (except for deficiency 11 discussed below). We have determined that three of the deficiencies (Nos. 5, 8 and 9) have been largely resolved between the Staff and Applicants as to how best to achieve certain functions or operations. Indeed NUREG-0737 merely provides guidance and does not mandate how to achieve these ends. Similarly, we could find no clear requirement in NUREG-0737 for SPDS availability calculations (deficiency 10). We, nevertheless, present our findings with respect to the deficiencies 5, 8, 9 and 10, infra.

With regard to deficiency 11, the tests that must be conducted to determine SPDS computer response time are required by NUREG 0737, but meaningful tests or statistics from computer response times must await plant operations at power levels when significant total loads are placed on the main plant computer. In the interim, public health and safety will not be adversely affected by the unknown system response tine under heavy computer loading because the SPDS will be functional and

available, and operating personnel are required to verify any SPDS indications prior to taking any actions on them. Applicants-have made the commitment that, prior to restart following the first refueling outage, load. tests shall have been conducted to determine response times for SPDS indications, sufficient to evaluate SPDS priority requirements on the main plant computer.

.With respect to two of the minimum plant variables identified by ,

the Staff as essential to the provision of information on critical safety functicns as required by NUREG-0737 Supplement.1; viz., residual

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heat removal (RHR) flow and containment hydrogen concentration variables (deficiency 2), we find that Applicants have not met their burden of showing that public health and safety will be protected if addition of these variables is deferred until restart following the first refueling 1

outage. Accordingly, we impose a condition on the operating license to require addition of these indications to the continuous SPDS displays l prior to operation above five percent of rated power.

The requirements addressed in deficiencies 3 (readability of the ,

containment isolation display) and 4 (location of monitors that display  ;

steam line radiation and vent stack radiation parameters) constitute l

three other minimum plant parameters identified as essential to the provision of critical safety function variables. The Applicants aver i

that improvements, already made to the arrangement of lights that i indicate containment isolation valve status, provide to the operator at <

the prime SPDS location the information on containment isolation that is I

required by NUREG-0737, Supplement 1. With regard to the location cf ,

I-

u the two essential radiation monitors, Applicants have committed to establish a radiological control screen on the SPDS prior to plant operation above five percent of rated power. While this screen will require a selection button to call up the radiation monitors, the same information is displayed an arm's length behind the SPDS station on the Radiation Data Management System (RDMS) displays, which have auditory alarms that sound if radiation levels exceed a designated set point. We find that the nearby auditory alarms adequately compensate for the lack of continuous display of the radiological control screen on the SPDS.

Thus we find that, the correcticns already made and the commitment to implement corrections described by the Applicants, when verified by the Staff, will provide reasonable assurance with respect to these three SPDS essential requirements that the health and safety of the public will be protected during operations above five percent of rated power.

We impose a condition on the operating license that the radiological control screen on the SPDS be implemented, as committed to, prior to operation at power levels above five percent. We consider Staff verification of these corrections, and others described below, to be ministerial tasks.

Applicants have committed to correction of another deficiency (No.

1), the lack of continuous display of SPDS variables, as required by

- NUREG-0737 Supplement 1. Two alternative approaches to meeting this requirement were described by the Applicants. (See Fdg. 30, infra) We find, and so condition the license, that either alterrative, if implemented prior to operation above five percent of rated power, and

l subject to Staff verification, will provide reasonable assurance that the requirement that the SPDS display be continuous will be met and that public health and safety will be protected. Also, see deficiency 4, supra.

Two SPDS displays, the subcriticality and core cooling status trees had been found to be capable of providing erroneous indications of the status of these critical safety functions at normal operating power levels. The Applicants have corrected this deficiency (No. 6) so that the status trees will function properly ~ at all power levels or requisite operational modes. Subject to verification of the corrections by the Staff, we find that this deficiency is resolved.

We are imposing license conditions with respect to three of the SPDS deficiencies (Nos.1, 2, and 4) cited by the Staff in SSER-6 that must be corrected prior to plant operation above 5% of rated power.

- With regard to the other cited deficiencies we find that Applicants have demonstrated that any needed correction of certain of them may be deferred until the first refueling outage wihout adverse impact on public health and safety, and that the remainder already have been corrected in a manner that we find will protect public health and safety. Our findings on these corrections already made by Applicants is contingent upon verification by the Staff.

C. Electrical Equipment Environmental Qualification Time Duration (Findings 48-90) l As set forth in our Findings of Fact, infra, assisted by its contractor, the NRC Staff made a preaudit review of the Seabrook l

Environmental Qualification program. Approximately 112 equipment qualification files (EQFs) were examined. The contractor's report, showing many deficiencies, was sent to the Staff in a memorandum dated February 21, 1986. The Staff's reviewer had received a copy of this preaudit report some time prior to February 21. Prior to conducting the environmental qualification audit, the Staft's and its consultant's review team members met with the Applicants and discussed each of the deficiencies found during the preaudit review. Applicants agreed to correct these deficiencies. Between February 24 and February 27, 1986, the review team conducted an audit -- some of the 12 EQFs audited were chosen to determine if Applicants had corrected the deficiencies as they had previously agreed to do. The results of the audit, recorded by the Staff in a Meeting Summary dated April 11, 1986, reflected that specific deficiencies were found in six of the twelve 12 files audited. In Supplement 5 to the Safety Evaluation Report issued July 1986, the Staff noted that the Applicants had " proposed acceptable corrective measures in the form of additional information and file revision to eliminate the deficiencies cited." The Staff concluded in Supplement 5 that "on the basis of the results of its review and subject to confirmation that all audit deficiencies have been corrected, the Staff concludes that the Applicant has demonstrated compliance with the requirements for environmental qualification as outlined in 10 C.F.R. 50.49, the relevant parts of GDC 1 and 4, and $9III, XI, and XVII of Appendix B to 10 C.F.R. 50, and with the criteria as specified in NUREG-0588." The Staft has received a letter from the Applicants notifying it that all deficiencies

have been corrected and that the EQFs have been changed to reflect these corrections.

NECNP has abandoned one aspect of this contention - viz. that Applicants' environmental qualification of electrical equipment program is-deficient in failing to specify the time duration over which the equipment is qualified. As to this aspect of the contention, upon our i review of the record, we conclude that the postaccident qualification time duration for electrical equipment important to safety at Seabrook, g which ?fs required to be environmentally qualified under General Design  !

Criterion 4 and 10 C.F.R. 550.49, has been specified for a period of one 7 year following a postulated accident, or, in the alternative, for the I

time required to perform its safety function plus a margin, as specified in Position C.4 of Regulatory Guide 1.89, Revision 1.

[ However, as to a second aspect of the contention, NECNP proceeds to urge that Applic' ants' EQFs do not contain either complete or accurate documentatioh{ demonstrating that each safety component is capable of performing its safety function for the duration in which it is required to be functional during an accident. It argues thus that Applicants have failed to provide reasonable assurance that Seabrook's safety a

^

equipment can survive an accident for the requisite duration. In

$ support of its position, NECNP alleges first that five of Applicants' environmental c.ualification files reflected deficiencies and that the l

" systemic and pervasive nature of Applicants' noncompliance with the NRC's environmental qualifichti_on requirements is ccnfirmed by the NRC audit," io.which six of the twelve equipment qualification files

c.

audited by the-NRC Staff showed deficiencies. It also alleges that the Staff's sampling technique was flawed in examining only twelve equipment qualification files. (NECNP Brief at 9,10.) Finally, it alleges that, given the large number-of deficiencies found during the preaudit, it was.

premature to begin the audit until the extensive corrections found at the preaudit stage had been effected. (NECNP Prop. Fdg. 84.)

NECNP's first allegation is without merit since its basis is faulty. Except for a missing maintenance requirement document in one of the five files alleged by NECNP to be deficient, there is no evidence that the equipment listed in the five challenged files was not properly qualified or that _the files failed to meet the recordkeeping requirements of 10 C.F.R. 550.49. We have directed in our Order, infra, that the missing maintenance requirement document be supplied. As to the six file deficiencies found during the audit of the Applicants' EQFs, four merely called for addition of clarifying or supporting information already in Applicants' possession, and two called for corrections to two equipment items observed during a walkdown inspection. In a letter to the Staff, Applicants have confirmed that all file deficiencies have been corrected. We do not find that the l

audit deficiencies suggest, much less confirm, a " systemic and pervasive" noncompliance with environmental qualification requirements, and there is no evidence to support such an allegation. Instead, the record shows that Applicants have responded to the audit findings by correcting the deficiencies.

l l

NECNP's second allegation is also without merit since it lacks evidentiary support in the record. It did not present an expert witness to testify that the Staff's sampling technique was flawed and it did not cite the , testimony of any witness called by the Staff or by Applicants in support of such a tsarren allegation. Moreover,.the record reflects that some of the twelve audited EQFs were selected to determine whether Applicants had corrected the deficiencies which they had agreed to do.

Finally, NECNP's third allegation is without merit. In its proposed finding 61 which we have adopted, NECNP asserted that the Staff generally performs an audit after it has reviewed a license applicant's equipment qualification program and concluded that it is basically adequate. (See Fdg. 57, infra.) As reflected above in our discussion of NECNP's first allegation, the environmental qualification program at the time of the audit was basically adequate, and thus the Staff's audit had not been conducted prematurely.

In light of our discussion, we conclude that the eleven equipment qualification files, which had been challenged by NECNP during the hearing, are complete and accurate and thus show that each safety component is capable of performing its safety function for the duration in which it is required to be functional during an accident. We also conclude that there is no evidentiary bast for the allegation that Applicants systemically and pervasively failed to comply with the Comission's environmental qualification requirements.

L.

t' III. CONCLUSIONS

. The Board concludes that Applicants' emergency classification and

- action _ level scheme fully satisfies.the requirements of 10 C.F.R. 650.47 and Appendix E of Part 50 and meets the guidance criteria of NUREG-0654.

We conclude that, except for three SPDS deficiencies which must be corrected prior to plant operation above 5% of rated power, the Applicants have established that the other SPDS deficiencies contested in this proceeding either will have no adverse impact on the public health and safety if corrections are deferred to the first refueling outage or have been corrected by the Applicants in such a manner so as i - to protect the public health and safety.

We conclude that the postaccident_ qualification time duration of electrical equipment important to safety at Seabrook, which is required l to be environmentally qualified under General Design Criterion 4 and 10

) C.F.R. 550.49, has been specified for a period of one year following a postulated accident, or, in the alternative, for the time required to perform its safety function plus a margin, as specified in Position C.4 of Regulatory Guide l.89, Revision 1. Finally, we conclude that, except f for a document absent from one file which we have ordered to be included, the eleven equipment qualification files, which had been

- challenged by NECNP during the hearing, are complete and accurate and thus show that each safety component is capable of performing its safety 1- function for the duration in which it is required to be functional F

during an accident. There is no evidentiary support for the allegation 1

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that Applicants systemically and pervasively failed to comply with the envirormental qualification requirements.

9 FINDINGS OF FACT Emergency Classification and Action Level Scheme

1. NECNP Contention III.1 asserts:

The emergency plan does not contain an adequate emergency classification and action level scheme, as required by 10 C.F.R. 950.47(b)(4) and NUREG-0654, in that ,

(a) No justification is given for the classification of various system failures as unusual events, alerts, site area emergencies, or general emergencies.

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

(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.

9 The factual background is set forth in the introduction to our opinion, supra. At the close of the reopened 1986 hearing the Board directed the parties to file, and stated a party would be deemed to be in default if it did not file, proposed findings of fact and conclusions of law and briefs and a proposed form of order or decision (Tr. 1024). Further, inter alia, the Board instructed that proposed findings should be integrated and based upon the original (1983) record and upon the instant (1986) record.

Finally, the Board instructed that the August 1983 transcript should be cited as 1-Tr. followed by the page number in order to distinguish it from the September-October 1986 transcript (Tr.

1025).

6 (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 implemented 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.

NH Contention 20 asserts:

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 take 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.

$50.47 and Appendix E and NUREG-0654.

2. 10 C.F.R. 950.47(b)(4) requires that emergency plans meet the following criteria:

(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:

IVB. Assessment Actions The means to be used for determining the magnitude of and for continually assessing the impact of the release of radio-active 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 emergency 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.

3. During the 1983 hearing, only Applicants' panel (Messrs.

Anderson, Thomas and MacDonald) testified (ff.1-Tr.1483) and the Staff's witness (Mr. Sears) testified (ff.1-Tr.1691)._10 Relying upon cross-examination, New England Coalition Nuclear Pollution (NECNP) and New Hampshire (NH) did not present any witnesses. In the 1986 reopened hearing, additional testimony was presented by Applicants' panel (Messrs. MacDonald and Thomas) (ff. Tr. 487) and by the Staff's panel (Messrs. Perrotti and Bryan) (ff. Tr. 489). The testimonies of the two panels were incorporated into the record by stipulation (Tr. 485-87; Tr.

489)andnocross-examinationwasconducted. Only the Applicants and the Staff filed proposed findings of fact, conclusions of law, and briefs with respect to NECNP Contention III.1 and NH Contention 20.

4. An emergency classification and action level scheme is designed to enable responsible personnel in the control room to 10 During the 1983 hearing the following exhibits were admitted into evidence: Staff Ex. 1 - Safety Evaluation Report dated March 1983; Staff Ex. lA - Suppl.1 to the SER dated April 1983; Staff Ex. IB -

Suppl. 2 to the SER dated June 1983; Staff Ex. 2 - Final Environmental Statement dated December 1982.

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

4 recognize and declare an emergency of a particular category or severity so that onsite and offsite emergency response organizations can be contacted and so that corrective actions can be taken to restore the reactor to normal (or stable) conditions. (MacDonald, 1-Tr. 1495-97; Sears, 1-Tr. 1700-03).

5. Tne emergency classification and action level scheme for Seabrook set forth in Applicants' Radiological Emergency Plan was first transmitted to the NRC and to all the parties to the proceeding on June 27, 1983. The scheme utilizes a symptomatic approach to emergency recognition and classification. (App. Ex. 1, ff. 1-Tr. 1483 at 5-1; MacDonald, 1-Tr. 1486-87). Subsequent amendments were made to the scheme (App. Exs. I and 2, ff. Tr. 487).

6. 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 persennel on-site or the public off-site. An alert indicates a substanti31 degradation of station safety marnins which could affect on-site personnel safety, ceuld 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 loss of containment integrity. (App. Ex. 1, ff.

p 1-Tr. 1483 at 5-1 and 5-2). The four classes of events included in the Seabrook scheme are consistent with the classes identified in NUREG-0654, Appendix 1.11

7. The symptomatic approach 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 recognition of various degrees of challenge to said functions. (App.

test. , ff.1-Tr.1483 at 15-16). The five critical safety functions are: subcriticality, core cooling, heat sink, reactor' coolant system integrity, and containment integrity. Color-coded status trees, based on plant events 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 - imediate action required). (App. Ex. 1, ff.

Tr. 487 Figure 5.6). The classification scheme at Seabrook relates the 11 The Board takes official notice of pertinent Commission's NUREGs and Regulatory Guides.

f- .

r ,

s status of the critical safety functions to the four emergency action classifications. (M., Figures 5.1 through 5.5).

8. In addition to the status of the five critical safety functions, Applicants' scheme takes into account thirteen miscellaneous emergency conditions (Id. , Figure 5.6). Each of these conditions is related to at least one of the four emergency classifications. (M. )

9. 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. (Sea rs',

1-Tr. 1699-1700). The framework as described in findings 6-8 above fully meets the requirements of 10 C.F.R. 950.47(b)(4) and Part 50, Appendix E.

10. 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. test. , ff. Tr. 487 at 3; Staff test. , 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. (StaffEx.4). Subsequent Staff inspections verified that the corrective actions, identified in Section 13.3.2.3 of Supplement 4, have been completed. (Staff test., 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 i

I

r standard of 10 C.F.R. 950.47(b)(4) and the guidance criteria of NUREG-0654. (Staff test., ff. Tr. 489 at 4; Staff Ex. 4,913.3.2.3).

11. Based on the evidence adduced in the 1986 hearing, the Board concludes that the open items that were discussed in the hearing in 1983 have now been satisfactorily resolved. In particular, the Board finds:

a. All the Seabrook-specific set points for the critical safety function status trees have now been selected. (App.

test., ff. Tr. 487 at 4; Staff Ex. 4 at 13-10; cf.

MacDonald, Tr.-1 1489-91,1511-13,1544-45; Thomas, Tr.-1 1 1516-23,1545).

b. Applicants have now incorporated indications and alarms from six different condition monitors as emergency action levels. (App, test., ff. Tr. 487 at 4; Staff Ex. 4 at 13-10 and 13-11; cf. Sears, Tr.-1 1717-20).

c. Applicants have now performed an acceptable comparison between their emergency action levels and NUREG-0654. (App.

test., ff. Tr. 487 at 5; App. Ex. 2, ff. Tr. 487; Staff test., ff.

Tr. 489 at 5; cf. Sears, Tr -1 1717-20),

d. Applicants have now completed the training of operators in the use of the emergency action levels. (Staff test., ff.

Tr. 489 at 5; cf. MacDonald, Tr.-1 1506-08 and Sears, Tr.-1 1711-13).

12. Applicants have also revised their treatment of fire and control room evacuation events so that the treatment is now consistent C r_ ___m

with the guidance contained in NUREG-0654. (Staff test. , ff. Tr. 489 at 5-7).

13. Finally, training on the Seabrook Station emergency classification system has also been provided to representatives of the State of New Hampshire Civil Defense Agency and Department of Public Health Services. Both agencies have indicated their agreement with the procedure used to classify emergency conditions. (App. test., ff.

Tr. 487 at 4). ,

Safety Parameter Display System

14. As originally admitted, SAPL Supplemental Contention 6 (formerly New Hampshire Contention 10) challenged the adequacy of two aspects of the Applicants' control room design, i.e., the Detailed Control Room Design Review (DCRDR) and the Safety Parameter Display System (SPDS). Following this Board's partial granting of sumary disposition with respect to the DCRDR issues, the surviving portion of this contention with respect to the SPDS asserted:

The Seabrook Station control room design does not comply with NUREG-0737, item 1.D.2.

We further focussed the issue in controversy as:

[S]ince the SPDS is not currently at an optimum, i.e.,

incomplete, in light of the deficiencies which are listed in Draf t License No. NPF-56 at C-9 and in light of five additional deficiencies which will be listed in Supplement 6 to the SER, [is there] reasonable assurance that, in deferring improvements to the SPDS until the first refueling outage, the safety of the population in the imediate vicinity of the plant will be protected?

(Board Memorandum and Order LBP-06-30, 24 NRC (September 15,1986).)

15. NUREG-0737, dated November 1980, is a letter to licensees of operating power reactors and applicants for operating licenses forwarding post-TMI requirements which have been approved for implementation (NUREG-0737,atiii). Requirements for implementation of the SPDS are included under Item I.D.2 in NUREG-0737. The implementation schedule for the SPOS is shown as "TBD" (to be determined) rather than as required before operation at some specified

. power level, or prior to a fixed date, as is shown for other requirements. Supplement 1 to NUREG-0737, dated Decemb'er 17, 1982 provides additional clarification on requirements for emergency response capabilities, including those for the SPDS. The requirements set forth in NUREG-0737, Supplement I have been reviewed and approved by the Comission (on July 16,1982). The document notes that the requirements therein "are to be accorded the status of approved NUREG-0737 items as set forth in the Commission's Statement of Policy: Further Commission Guidance for Power Reactor Operating Licenses (45 FR 85236),

December 24,1980)." While NUREG-0737 Supplement I also indicates that any schedules for implementation of requirements therein supersede previously set schedules for those items, the SPDS implementation schedule remained indefinite. (NUREG-0737, Supplement 1, at 2, 5; see Fdg. 18, infra).

16. The purpose and function of the SPDS is described as:

l The SPOS should provide a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determining the safety status of the plant.

Although the SPDS will be operated during normal operatiuns as well as during abnormal conditions, the principal purpose and

e_

! function of the SPDS is to aid the control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnomal conditions warrant corrective action by operators to avoid a t degraded core. This can be particularly important during anticipated transients and the initial phase of an accident.

(Emphasis added)

NUREG-0737, Supplement 1 at 7.

17. The minimum information required to be provided to the plant operators by the SPDS shall include information about five designated criticalsafetyfunctions(CSFs):

(i) Reactivity control (ii) Reactor core cooling and heat removal from the primary system (iii) Reactor coolant system (RCS) integrity (iv) Radioactivity control (v) Containment conditions The specific parameters to be displayed shall be determined by the licensee. (Id.at8.)

18. NUREG-0737, Supplement 1 addresses implementation schedules for the post-TMI emergency response requirements (including the SPDS) at several places. The general scheduling instructions state:

You will note that the enclosure does not specify a schedule for completing the requirements. It has become apparent, through discussions with owners' groups and individual licensees, that our previous schedules did not adequately consider the intenration of these related activities. In recognition of this and the difficulty in implementing generic deadlines, the Commission has adopted a plan to establish realistic plant-specific schedules that take into account the unique aspects of the work at each plan't. By this plan, each licensee is to develop and submit its own plant-specific schedule which will be reviewed by the assigned NRC Project '.

Manager. The NRC project Manager and licensee will reach an

agreement on the final schedule and in this manner provide for prompt implementation of these important improvements while optimizing the use of utility and NRC resources.

.... For holders of construction permits and applicants for operating licenses, plant-specific schedules for the implementation of these requirements will be developed in a manner similar to that being used for operating reactors, taking into consideration the degree of completion of the power plant.

(M., transmittal letter at 2), and:

Specific implementation plans and reasonable, achievable schedules for improvements that will satisfy the requirements will be established by agreement between the.NRC Project Manager and each individual licensee. -

(!_d.at5.)

19. While the above findings do not show that NUREG-0737, Supplement I requires a fully-complying SPDS by any fixed date, or prior to issuance of an operating license, the importance and safety significance of prompt implementation of an SPDS is emphasized elsewhere in the document, viz:

Prompt implementation of an SPDS can provide an important contribution to plant safety. The selection of specific information that should be provided for a particular plant shall be based on engineering judgment of individual plant licensees, taking into account the importance of prompt implementation.

(!_d.at8),and Prompt implementation of an SPDS is a design goal and of primary importance. The schedule for implementing SPDS should not be impacted by schedules for the control room design review and development of function-oriented emergency operating procedures. For this reason, licensees should develop and propose on integrated schedule for implenentation in which the SPDS design is an input to the other initiatives.

If reasonable, this schedule will be accepted by NRC.

(M. at 9.)

I

20. The Board heard evidence on SAPL Supplemental Contention 6 in Portsmouth, New Hampshire on October 1 and 3, 1986. The Applicants presented direct testimony from Messrs. Lawrence A. Walsh and George S.

[ Thomas (ff. Tr. 739); the NRC Staff presented the direct testimony of Mr. Richard J. Eckenrode (ff. Tr. 822). Seacoast Anti-Pollution League (SAPL) and Massachusetts (Mass.) 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, and no evidence had been presented during the 1983 hearing.

21. The SPDS is designed to provide 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. (Staff test. , ff. Tr. 822, at 2.) The SPDS primarily serves to accumulate important safety information in one centralized location. (Eckenrode, Tr.985-86,995-96,998,1001.)

22. 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. (Staff test., ff. Tr. 822, at 2; App. test., ff.

Tr. 739, at 1-2; Eckenrode Tr. 978-79.) The SPDS is used to refer operators to various other displays and controls in the control room where corrective actions are to be taken if needed. (Stafftest.,ff.

Tr. 822, at 2; Walsh, Tr. 808; Eckenrode, Tr. 839,979.)

23. Operators are trained to respond to emergencies both with and without the SPDS. (App. test., ff. Tr. 739, at 2; Walsh, Tr. 812, 817; seeNUREG-0737, Supplement 1,14.1.c,at7.)

24 The Seabrook SPDS is incorporated as a function within the main plant computer. The displays are presented on cathode ray tubes (CRTs) that are an integral part of the control room displays. The i designated primary SPDS CRT is located near the center of the control room at the shift technical advisor (STA) station. The SPDS displays tray be selected and presented at any of six other CRTs on the main control board. Operator access is through the existing keyboards used for accessing all plant programs and displays. (Staff test., ff.

Tr. 822; Section 18 of the Seabrook Safety Evaluation Report (SSER-6),

(Staff Ex. 6 at 1, ff. Tr. 822.)

25. The top-level SPDS display format consists of six color- and position-coded bars representing the summary status of the six critical safety functions (CSFs). Each CSF status tree is displayed on the second-level fornat, which includes parameter values and a color- and shape-coded status circle for each tree branch. The color-coded summary bar for the six functions appears in the lower left corner of each CSF status tree. (Section 18 of SSER 6, Staff Ex. 6 at 2, ff. Tr. 822.)

26. Applicants submitted their SPDS report to the NRC Staff by letter dated January 6,1986 (SBN-920). (Staff test. , ff. Tr. 822, at3.) Additional information was submitted to the Staff by letter datedApril2,1986(SBN-987). The Staff and its consultants reviewed the infornation submitted by Applicants and conducted an onsite audit of the $POS in May of 1986. (Ld. at 4; Staff Ex. 6 (Audit peport), ff.

Tr.822.) The results of the Staff's review are set out in Section 18 of SSER-6 (Staff Ex. 6 (SSER-6 and Appendix 18A), ff. Tr. 822).

27. On the basis of its documentation review and information gathered at the onsite audit, the Staff concluded that the Seabrook SPOS does not fully meet the applicable requirements of Supplement No. I to NUREG-0737. Eleven deficiencies, including the six listed in Draft License No. NPF-56, at C-9 (see Fdg.14, supra), were set out in Section 18 of SSER-6 (Staff test., ff. Tr. 822, at 5; Staff Ex. 6 (SSER-6, at 6-10), ff. Tr. 822.) These are listed here, and findings applicable to each are presented below:

(1) The SPDS display is not continuous.

(2) RHR (Residual Heat Removal) flow and containment hydrogen concentration variables are considered by the Staff to be part of the minimum information required to assess the j CSFs and are not displayed on the SPDS.

(3) The containment isolation display is not satisfactorily ,

- readable from the prime SPDS location.

(4) The SPOS does not display sufficient radiation variables.

(5) Several human engineering discrepancies have been ,

1 identified, i.e., awkwardness of calling up the lower level displays and inconsistency of heat sink display geometry with other displays, in addition to items (1),

(3),(6),and(9),

(6) Two CSF status trees (subcriticality and core cooling status) are not mode dependent and have the potential for ,

misleading the operator.

{-

(7) The Westinghouse RVLIS (Reactor Vessel Level Instrument System) isolators, used to protect RVLIS from SPDS, have I not yet been approved by the Staff (but see Fdg. .39, infra).

(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 SPDS display formats to the operator is in question.

(10)RVLISandRDMSavailabilityhasnotyetbeenfactored 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 of loading.

I j

28. Based on reasoning chiefly addressed in Staff Prop. Fdgs. 57, 59-61 and App. Prop. Fdgs. 56-57 and Response Fdgs. at 1-3, including the fact that the Seabrook SPDS, while incomplete, is functional and i

useful, both the Staff and the Applicants take the position that correction of a_nyn incomplete SPDS requirements can be deferred until the i end of the first refueling outage without adversely affecting public health and safety. (See Staff Test., ff. Tr. 822, at 4-5,101 App.

test.,ff.Tr.739,at1-2,7.) The Board rejects this position, f because it runs counter to the thrust of the contention as restated by  !

us(Fdg.14 supra). Intervenors SApL and Mass. take the opposite  ;

position that NUREG-0737 and its Supplement 1 provide requirements for a (

1 ,

o

6 complete SPDS and that all deficiencies must be cured prior to operation of the plant. We reject this position as not supported by our opinion l- or our findings. We now address the specific deficiencies seriatim in j findings below.

29. SPDS display is not continuous (deficiency 1). The Staff l

found that because the Shift Technical Adviser (STA) at the SPDS has the i i

capability to call up displays other than the SPDS at the SPDS terminal, the Seabrook SPDS is not a continuous display as required by NUREG-0737, Supplement 1. The Staff requirement for resolving this discrepancy is that either the CSF (critical safety function or " top level") sumary display must be added to all CRT (cathode ray tube) formats accessible on the STA's CRT, or a dedicated CSF summary display needs to be added l

to the STA station. (Staff test., ff. Tr. 822, at 8; Staff Ex. 6 l (SSER-6),ff.Tr.822,at5-6.)

l 30. The Applicants have comitted to dedicate the SPDS terminal so i

that a continuous display of the CSFs will be achieved or,

alternatively, through a test function and test computer, Applicants will have an SPDS display on every CRT format in the control room and l regardless of what display is called up this CSF monitor display will be j shown. The Applicants indicated that at least the separate dedicated ,

CSF display at the SPDS terminal could be achieved prior to full power ,

operation. (App. test., ff. Tr. 739, at 2-3; Walsh, Tr. 764-65,  !

l 804-05.) The Board finds that implementatinn of either alternative [

prior to operation above 5 percent power to provide a continuous SPDS l

l l

display of CSFs at the STA station provides reasonable assurance with respect to this matter that public health and safety will be protected.

31. RHR flow and containment hydrogen concentration indications t (deficiency 2). Indications of these parameters are not specifically required by NUREG-0737, Supplement 1 to be included as part of the SPDS.

However, Staff review of the Applicants' SPOS paraneters found that the fiveCSFsspecifiedinAUREG-0737(see Fdg. 17, supra) are not fully covered by the parameters to support the somewhat different critical safety functions selected by the Applicants in the Seabrook SPOS design (correspondence between the two sets of CSFs is presented in Staff Ex. 6 (Audit Report at 10), ff. Tr. 822). RHR flow and hydrogen concentration parameters are among those minimum or critical plant variables found missing from the SPDS by the Staff (also, see containment isolation and radiationparameters, infra). (Staff test., ff. Tr. 822, at 6-7; Staff Ex.6(SSER-6),ff.Tr.822.)

32. The Applicants' position is that they are still negotiating with the Staff as to whether RHR flow and hydrogen concentration parameters should be displayed on the SPDS; their belief is that indications of these parameters on the main control panel are sufficient from the safety standpoint and their inclusion on the SPOS display is not necessary. (App. test. , (f. Tr. 739, at 3; Walsh, Tr. 768-70.) The Staff continues to require that RHR flow and hydrogen concentration parameters be added to the SPOS, but its position is that addition of these to the SPDS may be deferred without undue public health and safety impact until the first refueling outage. (Staff test., ff. Tr. 822, at

l r

..O 4-7,10-11.) On Board and cross-examination, however, the Staff .

witness, a human factors engineer, couched his response with respect to deferral in terms of reliance upon the staff review practices set forth in N'J REG-0737, Supplement 1 (14.2b, at 8; also see Mass. Ex. 2, ff.

Tr. 964), and credibility of the Staff's position was undermined by its witness' apparently poor understanding of the underlying operational systems, the challenge to which is required to be shown by the SPDS.

(Eckenrade,Tr. 834-37,940-44,978,and984.)

33. The Board finds that the Applicants have notimet their burden f of proof 1. demonstrating that there is reasonable assurance that the  !

public healt5 and safety will be protected if addition of RHR flow and hydrogen concantration parameters to the SPDS is deferred until the first refuelir,; outage.

34. Conttinment isolation display is not readable from the prine SPDSlocatior(deficiency 3). Containnent isolation indications are ,

also among the minimum or critical plant variables required by the Staff i as part of the SPDS. While the containment isolation status indicators are not displayed at the SPDS console, a bank of valve pcsition indicator lights showing containment isolation status on the main control panel is visible from the prime SPDS location. The discrepancy cited by the Staff is one of pattern recognition and the Applicants aver that it has been resolved. (Staff test., ff. Tr. 822, ttA.9.a. A.9.c.

! A.9.g. at 6, 8, 9; Eckenrode Tr. 863; Walsh, Tr. 771-72,781-84.)

l '

! 35. The bank of valve position indicator lights showing containment isolation status on the main control panel is about 2G feet  !

i I

I._______._________________._______.________________

from the prime SPDS station. The lights are in boxes with windows, in a matrix (orgrid) arrangement. Previously, some of the boxes that were not used were blank and the blanks were randomly placed in the matrix.

The bank of lights has been rewired so that light boxes for components are grouped in a systematic order and the blanks are all in one location and off to one side. (Walsh Tr. 771-72,781-83.) The Staff witness, a human factors engineer familiar with the position and arrangement of this bank of lights, testified that if containment isolation has been called for in the plant, the corrections described by Applicants'

- witness would enable an operator at the prime SPDS location to determine containment isolation status from the bank of indicator lights on the '

main panel. (Eckenrode,Tr. 965-66, 986; also see Staff Ex. 6 (Audit i

Reportat8(14.1.2))). Staff review of Applicants' corrections to the installation, however, has not yet taken place. (Walsh.Tr.782-84; Eckenrode Tr. 656.) Based on the foregoing evidence, the Board finds that, subject to Staff verification of the described corrections already implemented, there is reasonable assurance that public health and safety will not be adversely affected by deferral of addition of containment isolation indicators to the SPDS console until restart following the first refueling outage.

36. The SPDS does not display sufficient radiation variables l

(deficiency 4). This item specifically refers to two radiation l parameters, steam line radiation and stack radiation, that are also l minimum or critical plant variables that are not displayed on the SPDS -

console. The Applicants have committed to establish a radiological l

(_ - _ _ __ _ _

\ .  ;

control CSF screen on the SPDS, which is a requirement of NUREG-0737, Supplement 1, prior to plant operation above 5 percent of rated power.

l There will be a selection button to enable picking up of the screen that will show all radiation monitors, but radiation parameters will not be added to the top level SPDS display. (Staff Ex. 6 (SSER-6, Audit Report at 9), ff. Tr. 822; Walsh, Tr. 774-75, 806, 816. Also see, supra, Fdgs.17,30.) Also, these radiation variables are continuously displayedontheRDMS(RadiationDataManagement System)whichis located on a panel just behind the prime SPDS location *("about an arm's length"away). The RDMS has auditory alarms to inform operators when radiation exceeds a designated set point. (Walsh,Tr. 774-75, 780, C05 06; Eckenrode. Tr. 866,969,986.) Based on the foregoing evidence and subject to the App 1tcants' connitment and Staff verification thereof, the Board finds that there is reasonable assurance that public health and safety will not be adversely affected by deferral of addition [

of steam line radiation and stack radiation monitor continuous displays to the SPDS censole until restart following the first refueling outage, i

37. Humanengineeringdiscrepancies(deficiency 5), in addition to human factors aspects of other deficiencies addressed separately herein(v3., deficiencies 1,3,6and9),theStafffoundthatthe format of the heat sink indicators of the SPDS displayed the flow data value above the decision block instead of below the block as do all the other formats, and that the SPDS display callup r.ethnd for the first two CSF status trees is awkward. (Staff test., ff. Tr. 022, at 9; Staff Ex.6(AuditReportat17),ff.Tr.022.) The heat sink screen forriat  ;

L has been changed and is now consistent in its labeling with the other formats on the SPDS display. (App. test., ff. Tr. 739, at 4; Walsh, Tr.777.) Thus, subject to Staff verification of this improvement, this 1

deficiency is resolved. As to the SPDS callup method, operators currently are required to position a cursor and press two buttons simultaneously. The Staff reconsnends that a single callup action be implemented, but finds that the curr[rit callup method, while it could be improved, is adequate in that the requested improvement would mean a difference between about one-half second and one and one-half to two seconds in time. (App. test., ff. Tr. 739, at 5; Staff test., ff.

Tr. 822, at 9; Eckenrode, fr. 855,968.) The Board agrees.

38. Subcriticality and core cooling status trees are not mode dependent (deficiency 6). The problem with these displays was that they would indicate that these CSFs are being challenged during normal operations which would have misled the operators. The subcriticality statusdisplaywouldhaveindicatedred(underextremechallenge) whenever reactor power exceeded 5 percent. Similarly, because the reactor coolant system (RCS) subcooling criteria used by the status tree might not always have been met during power operation, the status of corecoolingmighthaveerroneouslybeenindicatedasorange(under severe challenge) during normal power operations. (Stafftest.,ff.

Tr. 822, at 5, 7; Staf t Ex. 6 (Audit Report at 12), ff. Tr. 822.)

Corrective changes have been made to the SPOS so that these status trees function properly at all power levels (are now mode dependent). (App.

test., ff. Tr. 739, at 4 (as corrected at Tr. 730); Walsh, Tr.174.)

O Applicants are preparing documentation of these changes for Staff review. (Walsh, Tr. 814.)

39. RVLIS isolators (deficiency 7). The problem cited by the Staff involved the requirement for properly qualified interface devices between the SPDS and the Class IE safety-related instrument systems, the purpose of which is to protect the Class IE systems fram interference.

Prior to the September / October 1986 hearing the RVLIS isolation devices were analyzed and tested by the Applicants, and the Staff in its evaluation concluded that the RVLIS isolators were acce'ptable and that the proposed license condition requiring their installation and approval prior to exceeding 5 percent reactor power had been met. (StaffEx.3 (Appendix 18A), ff. Tr. 822; App, test., ff. Tr. 739, at 4-5; Staff test. , f f. Tr. 822, at 8-9.) No party challenged the resolution of this noncompliance, including the deferral of replacement of GA RM-80 isolator devices used elsewhere in the SPDS with approved non-fused devices until the first refueling outage. (Stafftest.,(f.Tr.822, at 8-9; Staff Ex. 6 (SSER-6, at 8, Appendix 18A, at 18A-3), (f.

Tr.822.)

40. SPDS data validation algorithms may not be sophisticated enough to ensure valid data are displayed to the operator _

(deficiency 8). The issue here is presentation of reliable synthesired data on the SPOS. Concern was raised that a paraneter valire could be within an acceptable range but significant1/ different from other ,

measures of the same parameter, causing the average value displayed to be incorrect and possibly misleading. The source of the concern is the

a .: +

'x ,

s 1 e

',4 e ^' .

. 42 ,

N -

- s  ;

, S. ,% .

SPDS algorithm; it utilizes only Unge che: king, averaging ."md auctioneering (i.e., seleeffon of highdst or lowest values in a set).

Accor, ding to the Staff's conhuli. ants, their audit Oncluded specifically ,

that P3NH $st implement data validation methodology that maies more effective use of, or interchannel comparison of, rddundant information available via the me.in plant computes . (tpp. test., ff. Th. 739, at 6; ,

' Staff test., f t.s[r. 822, at' 70-Staf f Ex. 6 (SSER-6, at 4, Audit Report at 12-13), ff.. Tr. 822; Eckenrode, fr. 839, 842-43; Walsh, Tr. 806,

- p 809.) ,

41. According to tho Applicants, the present algorithm is not

inadequate for the Msk. Although under the circumstance where an

?

average value 'is erroneously offset by a single high (or low) value'in f

theset,andtheSDDSdoesnot.indicatranabnormalsituatjon, alerting q the operator to validate the SPDS parameters is not the only function of the SPDS. ( App, test. , (f. Tr. 739, at. 6; Walsh, Tr. 007,049.)

~

According to the Staff, it is cost likely thet an individual param6ter value in a set would have been picked up by an operator at the taala control board through an alarm by the tine the' operator (3TA) et the SPDS, alerted by the top level displayiwent to the lower Icvel display toseewhaitheindividualparameturqiuewas. (Eckenrode, Tr. M5; see ckmrode., fr. 935.) A_ fnetiori .

the nperator would be quickly e.lerted by an alarm of a singit parameter v3 t de. in a set een .if thu single parameter value did not offset the aurage of the sfl sufficiently to affect the top level SP0t display. The Staff believes that tuore is not itkely to be any cre fusion resulting from use of the m .. A _ . m.

x, f ,N ,

D- , .

5),;

u ,

,. A j 43 -

current algorithm in the SPDS but, because there is a potential for confusion, Staff has asked the Applicants to examine it. (Id.,Tr.988.

42. Thus the Board sees the resolution of this issue, which is but part of the general requirements for overall speed and reliability in determining the safety status of the plant, as one of guidance and

'hegree of reliability (see Staff Ex. 6 (Audit Report Section 4.3), ff..

[ Tr.822). For those instances in which a potential for misleading information may occur in the SPDS top level displays because of the

. vafidation aliorithm, the Board finds that in this case reliance by the plant operators on alarms and displays on the main control board is an

- adequate interim compensating procedure. Accordingly we find that I

deferral of changes to the SPDS algorithm employed in the main plant cdputer until the first refueling outage will not adversely affect the p6blic health and safety.

43. Usefulness of the lower level SPDS display formats (deficiency 9). The question posed by this item is utility of the lower

! level displays y the SPDS. During Staff observation of an accident h simulation at Seabrook, it was observed that the Seabrook operators did not use the SPDS lower level displays on the SPDS terminal, but instead useid har,d copy representations of the lower level display. (Staff test., ff. Tr. 822, at 7; Walsh, Tr. 759-61; Eckenrode, Tr. 972,

1. 979-80.) The Staff did not identify any problem with the contents of the lower level displi;ys, but wanted an evaluation of why the operators used the hard copy representation rather than the SPDS lower level display; i.e., was there a format problem, or the like, that made it

l

. 4 m.

\' s_

• 4 \

l'

• more' difficult to use the SPDS display than the hard copy? (Eckenrode, j

i Tr.9Ys-80.) l

44. The Applicants' explanation at the hearing'of the operators' use of the hard copy was that the SPDS lower level screens and the hard copy' version show the same thing, and that when verifying SPDS indications on the main control board, as they are required to do, some operators prefer to pick up the hard copy in case they forget what they are looking for. Other operators simply use the SPDS screen and " walk the board" to verify it. (Walsh, Tr. 815-16; see also-SAPL Ex. 2, ff.

Tr.1016.) ThusthehardcopyreprepentationsoftheSPDSlower-level displays serve a memory-assistar.ce function. While the Applicants

.s continue to be required to furnish the requested evaluation of the

{

f utility of the lower-level displays to the Staff, Uis is6 aid finds that there is reasonable assurance that the lower-level displays on the SPDS,

. and the procedure whereby operators may utilize hard copy N representations of the lower level screens while verifying indications l

, on the main control board pose no threat to public health and safety.

T On its. face l, the procedure described during the hearing for utilizing the hard copy representations to aid operators' memory would appear to improve safety. ,

RVLIS and RfiP.S availibility has not yet been factored into

~

45.

overall availability c'alculations (deficiency 10). According to the Staff, system availability data indicated an acceptable (over 99%)

availability for the SPDS, but the calculations did not include the availability of RVLIS or' RDMS data input. (Staff test., ff. Tr. 822,

'O at 7; Eckenrode, Tr. 973.) The overall availability calculation cannot be made prior to the actual interface of both units (RVLIS and RDMS) with the SPDS. -( App. test. , ff. Tr. 739, at 6-7.) These apparently are separate data processing systems which will input data to the SPDS via the main plant computer. (Staff Ex. 6 (Audit Report at 2-3, 6 (13.4.2),

11), ff. Tr. 822; Eckenrode, Tr. 973.) The additional calculations would involve only the RVLIS and RDMS availability and would not affect availability of other SPDS parameters. (Eckenrode, Tr. 974.) RVLIS and RDMS availability is not expected to have a significant impact on overall SPDS availability, and there is no evidence to suggest that RVLIS and RDMS availability will be signifcantly less than that of other Seabrook plant computer-controlled data processing units. (An unavailability criterion (.001) is given in the guidance document NUREG-0696 at 8.) Thus the Board finds that up to the first refueling outage, the high availability calculated for the SPDS alone, without the RVLIS and RDMS availability calculations, provides reasonable assurance that public health and safety will be protected.

4

46. System response time -- a system load test is needed to verify the worst condition of loading (deficiency 11). Although system response times for the SPDS appear to be satisfactory (most factors are updated every five seconds), the Staff observations were made during a lightly loaded sequence. (Staff test., ff. Tr. 822, at 7; Staff Ex. 6 (Audit Report at 18), ff. Tr. 822.)' There is a very good chance that in the event of a severe accident a large number of nearly simultaneous processing demands will be made on the main plant computer, but whether

the update rate of the SPDS indications would be slowed down depends upon what priority the SPDS has in the main plant computer. (Eckenrode, Tr. 974-75.) During a period of heavy load on the main plant computer, even if update rates were delayed, the SPDS would be available as long i as_the main plant computer is running. (Eckenrode, Tr. 857-58.) From a

- human factors standpoint a delay in updating could lead to a mistake on

.the part of the operator. (Eckenrode, Tr. 859.) However, since no operator actions are taken at the SPDS station and any SPDS indications must be verified on the. main control panel prior to taring actions (Fdg.

22, supra) a delay in updating the SPDS indications is not likely to lead to incorrect actions or operations.

47. Applicants have agreed to perform a system load test under heavier loading conditions, which in order to provide meaningful results, would require some level of plant operation. (App. test., ff.

Tr.-739. at 7; Walsh, Tr. 788; Et.kenrode, Tr. 989.) The Staff witness was ura'le to say whether an adequate system load test was part of-a low power testing program. (Eckenrode,Tr.987.) Subiect to the commitment by the Applicants to perform meaningful systems load tests if power operations are authorized, and because the SPDS would be available even during overloading conditions, and because SPDS indications must be verified prior to taking any actions, the Board finds that there is reasonable assurance that deferral of evaluation of results of a future system load test until the first refueling outage will not adversely affect public health and safety.

Electrical Equipment Environmental Qualification Time Duration

48. -NECNP Contention I.B. asserts that:

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

49. NECNP does not now urge that the time durations of the equipment important to safety have not been specified in Applicants' equipment qualification files. Since NECNP has abandoned this aspect of its contention but has proceeded to contest another aspect, upon our review of the record we decided to render the foliowing ultimate finding upon the abandoned aspect. We find that the postaccident qualification time duration for electrical equipment important to safety at Seabrook, which is required to be environmentally qualified under General Design Criterion 4 and 10 C.F.R. 950.49, has been specified for a period of one year following a postulated accident, or, in the alternative, for the time required to perform its safety function plus a nargin, as specified in Position C.4 of Regulatory Guide 1.89, Revision 1. NECNP instead urges that the specified time durations are unsupportable because of incorrect or incomplete equipment qualification files and thus that Seabrook's safety equipment cannot survive an accident for the requisite duration. (NECNP's brief and proposed findings filed November 12, 1986.)

50. During the 1983 hearings, the Applicants' panel (Messrs.

Maidrand and Anderson) (ff.1-Tr. 970), and the Staff's panel Messrs.

LaGrange and Walker) testified (ff.1-Tr. 990). In the 1986 reopened

D hearing, the Applicants' panel (Messrs. Salvo, Thomas and Woodward) .

testifi.ed (ff. Tr. 357). The Staff called a witness (Mr. Walker) to testify (ff. 494). No other witnesses were offered by any party.

However, NECNP did cross-examine the Applicants' and Staff's witnesses.

Only the Applicants, the Staff and NECNP filed proposed findings of-fact, concl/Jsions of law and briefs with respect to this contention.

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

Cri?.erion 4 - Environmental and missile design bases.

Stricctures, systems, and components important to safety

. shall be designed to acconinodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing 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.

52. 10 C.F.R. 950.49 specifies the requirements that must be met to demonstrate compliance with GDC 4, relating to the environmental qualification of electrical equipment important to safety that is located in a potentially harsh environment. In conformance with

'10 C.F.R. 550.49, electrical equipment may be qualified in accordance with the acceptance criteria specified in Category I of NUREG-0588. In addition, guidance as to the means by which 10 C.F.R. 650.49 may be satisfied is provided in Regulatory Guide 1.89 (Walker, ff. 494, at 2).

Pegulatory Guide 1.89 which endorsed the Standard, IEEE 323-1974, provides that electrical equipment be qualified to withstand an accident environment after having been exposed to pre-accident conditions for the qualified life duration under the normal operating ccnditions. (App.

~_ _. . __ __ . _ _ _ _ _ _ _ . ._ _ _

test. ,- ff.1-Tr. 970 at 9-10.) The focus of testinony in this proceeding was on the post-accident qualification time duration, and the documentation in Applicants' environmental cualification files.

53. Requirements for maintaining records, in auditable form, of Environmental Qualification ("EQ") of electrical equipment important to safety are specified in 10 C.F.R. 950.49(j), which provides:

A record of the qualification, including documentation in paragraph (d) of this section, must be maintained in an auditable form for the entire period during which the covered item is installed in the nuclear-power plant or is stored for future use to permit verification that each item of electric equipment important to safety 4

covered by this section--

(1) Is qualified for its application; and (2) Meets its specified performance requirements when it is subjected to the conditions predicted to be present when it must perform its safety function up tc the end of its qualified life.

Section 50.49(d) specifies:

(d) The applicant or licensee shall prepare a list of electric equipment important to safety covered by this section. In addition, the applicant or licensee shall include the following information for this electric equipment important to safety in a qualification file:

(1) The performance specifications under conditions existing during and following design basis accidents.

(2) The voltage, frequency, load, and other electrical characteristics for which the performance specified in accordance with paragraph (d)(1) of this section can be ensured.

(3) The environmental conditions, including temperature, pressure, humidity, radiation, chemicals, and submergence at the location where the equipment must perform as specified in accordance with para-graphs (d)(1) and (2) of this section.

54. At the time of the 1983 hearing, Applicants had completed approximately 80% of their review of their equipment qualification program to determine whether all electrical equipment important to safety. could be qualified for a harsh environment duration of one year (Maidrand,1-Tr. 978). As of that time, the Staff had not received Applicants' environmental qualification submittal in order that it could perform an audit of Applicants' qualification files to verify that electric equipment important to safety located in a harsh environment was qualified for one year or for the required operativ'e time determined plus a margin. (Staff test., ff. 1-Tr. 990 at 2, 3.)

55. The Staff made a preaudit review of the Seabrook qualification program based on 63.11, Amendment 56 of the Seabrook Station Final Safety Analysis Report and on the Applicants' EQ Submittal, Revision 2.12 It was assisted by a contractor, EG&G, the prime contractor of the Idaho National Engineering Laboratory. The contractor's report, showing many deficiencies, was trensmitted to the Staff by a memorandum dated February 21, 1986. (NECNP Ex. 13.) The contractor's report also stated that the deficiencies, while a cause for concern, did not necessarily mean that the equipment was unqualified and that the Applicants should resolve the deficiencies and document the 12 While the oral testimony did not specifically indicate the number of EQ files reviewed, perusal of Table 2 in the EG&G report (NECNP Ex. 13) indicates that all 112 files were available at the time for examination.

.- _ _ . - = . _ _ _ _ - . _

'6 resolutions in an auditable form. (Id. at 4.) The Staff reviewer had a copy of the EG&G preaudit report some time prior to February 21, 1986.

(Waiker, Tr. 697.)

56. Prior to conducting the EQ audit, the review team members met with the Applicants and discussed each of the deficiencies found during the preaudit review. Applicants agreed to correct them. (Walker, Tr. 700.)

57. During the period of February 24 through February 27,1986, the Staff's reviewer and consultants conducted an audit of 12 equipment qualification files as part of their environmental qualification review.

(NECNP Ex. 11.) The Staff generally performs an audit after it has reviewed the equipment qualification program and concluded that it is basically adequate, and after an applicant has agreed that it has sufficiently completed its environmental qualification program.

Moreover, in choosing files for audit, the Staff attempts to achieve a random selection, except where it believes that there could be problems, or lack of information, or any indication that there are reasons to believe that a file may not be complete. (Walker, Tr. 692-93.) For the purposes of the instant cudit, some of the files were chosen to determine if the Applicants had corrected the deficiencies as agreed (Walker, Tr. 696-97).

58. Results of the audit were recorded first in a report to the Staff from its consultants dated March 31, 1986. (NECNP Ex. 12.) This was followed by an exit interview Meeting Summary dated April 11, 1986 prepared by the Staff to document observations and comments made by the

Staff and its consultants to the Applicants at the end of the February 24-27 audit. (NECNP Ex. 11.) The general comments noted in the Meeting Sumary indicated that, inter alia, the Staff did not agree 13 with the way that the Arrhenius equation was used to calculate postaccident operability time. (fECNP Ex. 11 at 1.)

59. In response to the Staff's comment on improper application of the Arrhenius equation, Applicants recalculated the postaccident operability times for all equipment files using the methodology recommended by the Staff. The results were that equipment in all files, j except eleven, met Applicants' original goal of 40-year normal operating life plus one year postaccident life. Technical justifications were given for the postaccident operability durations of equipment in the

! ' eleven files not meeting the one year postaccident life. (App. test.

ff. Tr. 357 at 4-17; App. Exs. 2 and 7 (13).) The Staff reviewed the qualification information for equipment items in those files and found that they met the requirement of 100 days or the postaccident time margin requirements specified in Position C.4 of Revision 1 of Regulatory Guide 1.89 and were thus acceptable (Staff Ex. 5 at p. 3-24 f

I and Table 3.1.)

13 The Arrhenius equation is a time / temperature relationship which compares the test time and temperature with the time and temperature equivalency in the plant, with a constant in the equation which is representative of the materials of the device.

(Woodward, Tr. 482.)

1

r.

1 o

60. The Meeting Sumary also reflected that six of the twelve EQ files audited contained deficiencies that required correction. (I_d. at 1-2; Walker, Tr. 517.) Of the six files, four called for supporting or clarifying information. They were: (a) one file (#113-01-01) should be updated to include test information that had been provided by Applicants during the audit, (b) a second file (#174-15-01) should be supplemented to include additional information justifying the use of a test sequence different from that specified in IEEE 323-1974, (c) a third file

(#113-06-01) should include a statement specifying that' submergence qualification was not required, (d) a fourth file (#236-11-06) should be supplemented to include clarifying test report data in the equipment summary evaluation. (NECNP Ex. 11 at_2.) Two of the six audit deficiencies addressed two specific equipment items observed during a plant walkdown conducted as part of the audit. They were (e) three internal wires and a terminal block in a Limitorque Motor Operator (EQ File #248-37-01) were not identifiable and must be replaced with qualified components, and (f) an ASCO Solenoid Valve (EQ File

1. NSSS-220-02) had two different equipment identification numbers on it, which situation must be rectified. (Id.)

61. Applicants' responses to the NRC audit observations, contained in letters of April 3 and April 10, 1986, were attached to the prefiled testimony of Messrs. Salvo, Thomas and Woodward. These responses indicate that Applicants have completed, or have committed to complete, actions on the deficiencies and open items noted in the audit report.

(App. test., ff. Tr. 357 at 20-21, Exs. 2, 7.)

62. With respect to the files requiring clarifying or supporting information,' as found during the February 24-27, 1986 audit, the Staff noted in_ Supplement 5 of the Safety Evaluation Report (July 1986) that

"[t]he applicant proposed acceptable corrective measures in the form of additional information and file revision to eliminate the deficiencies cited." With respect to the two deficiencies noted during inspection of the installed equipment during a plant walkdown conducted as part of the audit, the Staff noted that "[t]he applicant proposed acceptable corrective measures for the deficiencies that were fou6d and connitted to correct all deficiencies by fuel load." (Staff Ex. 5, 53.11.4 at p.

3-25.)

63. With respect to the overall Seabrook EQ program..the Staff, in SSER-5 (Id., $3.11.5 at p. 3-25) concluded:

The Staff has reviewed the Seabrook program for the environmental qualification of electrical equipment important to safety and safety-related mechanical equipment. The purpose of the review was to determine the adequacy and scope of the qualification program and to verify that the methods used to demonstrate qualification are in compliance with applicable regulations and standards.

On the basis of the results of its review and subject to confirmation that all audit deficiencies have been corrected, the Staff concludes that the Applicant has demonstrated compliance with the requirements for environmental qualification as outlined in 10 C.F.R. i 50.49, the relevant parts of GDC 1 and 4, and 5%III, XI, and XVII of Appendix B to 10 C.F.R. 50, and with the criteria as specified in NUREG-0588.

64 Typically, the Staff asks an Applicant to notify it by letter when all deficiencies have been corrected and the EQ Files have been i

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

~- - - __ - . . - - - -

J F

changed to reflect those corrections. Here the Staff has received such a letter from the Applicants (Walker, Tr. 688,713).

65. In addition to the six audited EQ Files discussed above, during the 1986 hearing, through cross-examination NECNP challenged the adequacy of several EQ Files.14 (See generally Tr. 358-457.) Our findings with regard to the specific items in the five EQ files challenged in NECNP's proposed findings are set forth below.

66. EQ File #113-01-01, item: Anaconda 5 kV power cable (multiple conductor) -- The multiple-conductor cable was qualified on the basis of comparison with test results from a single conductor cable as tested by Anaconda. The construction of the tested specimen was referred to as being " exactly similar" to that of the individual conductor in the multiple conductor cable. (NECNP Ex.1, Reference 6, 5-page attachment to letter from Anaconda Company to United Engineers and Constructors, dated December 10, 1979, at p. 2.) Applicants' witness felt that the term " exactly similar" meant that both the tested cable and the multiple conductor cable are similar within the bounds of environmental qualification so that the test report adequately represents equipment supplied to Seabrook and that the test is a representative test of that equipment. (Woodward, Tr. 368-69.)

14 With respect to at least one of the files, concerns, which had been raised in NECNP's opposition of July 2,1986 to Applicants' motion for issuance of a partial initial decision authorizing low power operation, were addressed by Applicants' witnesses in their prefiled testimony (Apps. test., ff. Tr. 357 at 18-20).

J 3

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i-

! 56 -

j 67. The Board does not find that the term " exactly similar" in this instance is confusing and notes that the Reference 6 letter supra, explicates the similarities and presents other information as to why the testing of the single conductor cable would be representative of and applicable to the multiple conductor cable. Environmental qualification f by testing of similar items, with a supporting analysis, is acceptable according to 10 C.F.R. 550.49(f)(2). No evidence was adduced to challenge the tests or supporting analysis presented in the EQ File

1. 113-01-01 (NECNP Ex. 1, Reference 6). Thus, the environmental I qualification is acceptable and adequately documented.

68. EQ File #113-19-01, items: ITT Suprenant RG-58 Coaxial cable i 1

and RG-11 triaxial cable -- These two items were qualified by comparison with similar types of cable, RG-11-U and RG-59-U coaxial cables, that were tested. The bases for similarity and qualification by comparison of the two untested cables with those tested was explicated in a letter from the manufacturer. (Woodward, Tr. 378-82; NECNP Ex. 4, Equipment Summary Evaluation at p.1 of I', and Reference 4, letter ITT Suprenant Division to United Engineers and Constructors dated February 11,1983.)

69. The absence of qualification for submergence was justified on the ground that the ITT Suprenant cables are not installed below postulated plant flood levels; hence submergence qualification is not required. The basis for this conclusion was a plant walkdown that is documented in the EQ File by a letter from Impell Corporation to Yankee Atomic Electric Company, dated February 2,1986. (Woodward, Tr. 377-78;

f e

A NECNP Ex. 4, Reference 10.) The Board finds this conclusion to be justified.

70. Justification for similarity of the untested cables, generally, was that all four ITT Suprenant cables were similar in construction details and the materials used to construct them were identical. Further, the dimensions of the untested RG-11 triaxial cable and the tested RG-11 coaxial cable are identical through the first shield, and the triaxial cable has an additional shield and jacket of materials identical to that of the coaxial cable. (NECNP Ex. 4, Reference 4.) While NECNP challenged the similarity between the types of cable as not being documented in the EQ File (NECNP Prop. Fdgs.

15-19), the Board found little difficulty in accepting the manufacturers certification, or for that matter, in locating testing requirements, materials specifications, and dimensions of all four cables in the EQ File provided by NECNP, (NECNP Ex. 4, Reference 1 at 3-8,12-13, Appendices A and B.) Thus, the Board finds that justification for environmental qualification of cables RG-58 coaxial and RG-11 triaxial by comparison with tested coaxial cables RG-11-0 and RG-59-U is adequately documented in the Applicants' EQ Files.

71. EQ File #113-20-01, items: ITT Suprenant 300V instrument cable (and MM-IR-12 instrument rack) -- This cable was not subjected to a submergence test, but was qualified for 30 days submergence by immersion in tap water and conducting a high potential test based on 80V/ mil of insulation thickness after completion of the 100-day SLB/LOCA testing where peak temperatures reached 390*F, peak pressures reached

b 113 psig, 100% humidity was maintained through the 100-day test and cables were exposed intermittently to chemical spray. The cable specimens were energized and electrical leadings were maintained throughout the 100-day test. Since the greatest depth of flooding.that this cable will experience in the plant under accident conditions is three feet, producing a static pressure of about 1.3 psi, the static pressure is regarded as negligible in comparison to the 113 psig pressure during the test. The presence of high temperature, 100%

humidity, and chemical sprays, with the high pressure is considered adequate to account for the submerged condition of this cable for a 30-day duration. (NECNP Ex. 5, Reference 14 at 1-2.) Further basis supplied for acceptance of the judgment that this procedure adequately qualified the cable for 30 days submergence was that the cable had undergone thermal and radiation aging to end of life conditions prior to the test sequence and that actual moisture contact with the cable would not have produced more severe conditions of stress. (Woodward,Tr.

404-06.)

72. The 300 kV ITT Suprenant instrument cable associated with i

three valves located inside the reactor containment building and subject to submergence due to LOCA flooding was found, with other items, not to meet a one year postaccident operability tirie, as shown in Applicants' prefiled testimony. That testimony indicated that the valves served by the cable would close in less than one minute, which time when added to l

the one hour margin required by Regulatory Guide 1.89 results in a required operating life of 61 minutes. Applicants engineering analysis

e c.

further indicated that once the inboard letdown isolation valve has .

closed, it has performed its safety function and is not required again in the near term or for long term recovery operation. The other valves (accumulator tank isolation valves) are normally open during power operation and also receive an SI signal to open. Applicants' engineering review determined that all the valves would perform their safety function within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and that long term failure of the cables does not result in a change in valve position. Thus Applicants concluded that the 30-day postaccident operability qualification of the associated cables has sufficient margin to ensure that the required safety function has been performed. (App. test., ff. Tr. 357 at 9 and

. App. Ex. 4 at 1-2.) We agree.

73. As another matter, NECNP challenged the completeness of the EQ File #113-20-01 with respect to Instrument Rack MM-IR-12 which is supplied by the 300 kV instrument cable. According to NECNP, the EQ File indicates that submergence qualification for the cable is not required because the instrument rack has been downgraded to Operability Code C, and that no explanation or justification for the change (in the instrument rack Operability Code) is provided in the file. (NECNP Prop.

Fdgs. 22-23. See NECNP Ex. 5, at Environmental Qualification Assessment Report, n. 9, p. 11.)

74. Equipment categorized as Operability Code C is that which may see a harsh environment, including submergence, subsequent to design 1~

basis accidents, but which performs no safety function relative to mitigating the accident or putting the plant in a safe condition after

the accident.' Operability Code C equipment. is also evaluated to determine that' failure of the equipment due- to environmental conditions will not affect.the safety of the plant. (Woodward, Tr. 386-87; see .

also Regulatory Guide l'.89, Appendix D, 1 3.c.)

75. - The Impell . Corporation, for the Applicants, reviewed' locations of Class IE electrical equipment in the plant and determined that'some.

l were located below. flood level for the specific equipment locations'.

With regard to instrument rack MM-IR-12, located below flood level in the mechanical penetration area Impell found that the' rack, its accessories and the-transmitters are not qualified for submergence. It

-recommended that the equipment be relocated above flood level unless it can be shown that operability for the MELB (moderate energy line break) is not required. (NECNP Ex. 5, Reference 12 at 4; Woodward, Tr. 387.)

76. United Engineers and Constructors performed a specific review and determined that no piece of equipment in instrument rack MM-IR-12 was required to perform any safety function during an MELB.

Accordingly, the instrument rack was downgraded to operability Code C.

The report of the change is an Engineering Change Authorization (ECA No. 03/114514A dated 2-21-86) that is found in the EQ File, NECNP Ex. 5, Reference 16). The ECA shows the signoff that indicates that several engineering disciplines reviewed the operability requirements and determined that there is no impact on downgrading the equipment from operability Code A to C. A pencilled-in change reflecting the change on the Class IE equipment list also is present in Reference 16 and this change will be reflected in the next scheduled revision of the harsh

~ _

e

.G equipment list for Seabrook (Woodward,' Tr. 386-91;- NECNP Ex. 5,

' Reference 16.)

77. The Board finds that detailed explanation of the criteria used to downgrade the instrument rack MM-IR-12 in the ITT Suprenant cable EQ File #113-20-01, beyond that contained in the file, is not necessary to satisfy the requirement of 10 C.F.R. 950.49(j) that EQ files be

. maintained in auditable form. No other requirement for inclusion of more detailed explanation in +.ha EQ File was averred by NECNP or any other party, and we know of none.

78. EQ File #NSSS-220-03, items: limit switches RH-ZS-618 and RH-ZS-619 -- NECNP challenged this file on essentially the same basis as the foregoing instrument rack [Fdgs. 72-76, supra). NECNP alleges incompleteness because there is no explanation in the EQ File of the reason for downgrading the limit switches to Operability Code C.

Indeed, the same ECA (No. 03/114514A) is referenced to document the change in operability code. (NECNP Prop. Fdg. 29; NECNP Ex. 9 Reference 12 at 2; Woodward, Tr. 446-48.)

79. The Qualification Evaluation Worksheet for the components in this file indicates that all items are located above flood level.

Applicants' witness believed that this was incorrect and that the entry for the "above flood level" question should be "no" rather than "yes" because some equipment is located below flood level. Note 1 on the same page indicates that the limit switches RH-ZS-618 and RH-ZS-619 are located below postulated flood level but that the Operability Code has i

r been changed to' Code C. (Woodward,Tr. 446-47; NECNP Ex. 9,

. Environmental Qualification Worksheet at 2.)

I 80.~ For the same reasons held in Fdg. 77, supra, for the )

instrument rack, we find that NECNP's allegation of incompleteness of EQ File #NSSS-220-03 for the limit switches lacks merit.

81. EQ File #174-15-01, item: Transamerica Delaval level i

transmitters and silicone oil-filled-conduit riser assembly -- This equipment measures containment water level from a ball on a rod sensor )

- l and transmits the corresponding water-level electrical signals to the i

control room. A thirty-minute submergence test had been conducted on the level transmitters, but in order to qualify the transmitters for a l

one year submergence duration, Applicants designed and installed a riser device of metal conduit with sealed connections through which the interconnecting wires run, and which is filled with silicone oil to prevent moisture intrusion into the transmitters. (Woodward, Tr. 429; NECNP Ex. 7, Qualification Evaluation Worksheet at 1 and Reference 7 at 5.)

82. Each of the conduit riser assemblies is configured in an inverted "U" shape with its downward-pointed legs of unequal length tenninating at the junction box or splice box attached to one of a pair of level transmitters. Only the lower transmitters (ID: CBS-LE-2384-1 and CBS-LE-2385-1) of each riser assembly are below flood level and subject to submergence. A " Tee" fitting with a threaded plug or cap is located at the high point of each inverted U-shaped riser assembly to permit filling both legs with Dow #710 silicone oil. The risers, as 1 r iu-

l e

t constructed, are intended to provide a static head of four feet six inches (4'-6") of oil above postulated flood level to counter the 6 5/8 inch head of water that would cover the lower transmitters. (NECNP Ex.

7, Equipment List at 1, Environmental Qualification Assessment Report at 1 and 11 Reference 5 at 1-4, Reference 7 at 2, 3, 5 and 9; Woodward, Tr. 452 (post line 16)-454.) The oil used in the risers is the same as that used in the equipment boxes and was used in the environmental qualification test configuration. (Woodward, Tr. 436.)

83. NECNP challenged two aspects of the riser assemblies, principally on grounds associated with information provided in Reference 7 of EQ File #174-15-01, the Engineering Change Authorization (ECA) which was prepared to obtain authorization and to provide instruction on installation of the risers. First, the ECA is inconsistent with other parts of the EQ file in that the flood levels differ by 2'-8" and the instructions to fill the riser would result in fill levels that differ by four feet relative to postulated flood levels.15 Second, NECNP asserts that elimination of a pressure test to check for leak tightness of the riser assembly, that was originally called for by the ECA design, compromises the ability of the transmitter 15 In Proposed Finding 40 NECNP mistakenly states that a flood level of (-)18'is 2'-8" lower than flood level (-)20'-8". An elevation of (-)18' is 2'-8" higher than elevation (-)20'-8". Also, the fill levels (" Tee" fittings), as constructed, are in excess of one foot above the minimum (-)17'-6" level specified in the ECA. Thus the as-built fill levels are not inconsistent with those in the ECA installation instructions. (See Fdg. 84.)

l d 6

to function for the duration of an accident in which it might be f

submerged. (NECNP Prop. Fdgs. 35, 37, 40, 45; NECNP Ex. 7 at 2, 3, 5, 9; Environmental Qualification Assessment Report, p.11 at n.11.)

84. The Engineering Change Authorization (ECA) of this EQ file (NECNP Ex. 7. Reference 7) contains the installation instructions for installing the risers. Sheet five of the ECA illustrates the concept of filling one riser to a six-inch-minimum point above the flood level which corresponds to the spreader fill cor:nection (" Tee"). This is not inconsistent with the fill instructions on sheet five of the ECA. Once j the equipment was installed, the ECA is no longer the drawing of record 3 for the plant. The postulated flood level in this location at the time the ECA was prepared was (-)18'. (Salvo, Tr. 451; Woodward, Tr. 452, 454; Thomas 455.)

85. After issuance of the ECA (NECNP Ex. 7 Reference 7) the postulated flood level at this location was changed to (-)20'-8".

(Woodward, Tr. 454.) Also, actual measurements of the riser assembly during a plant walkdown indicate that the elevation of the filler " Tees" are 8" above (-)17'-1 3/8" (equals (-)16'-5 3/8") and 91" above (-)17'-

13/8"(equals (-)16'-37/8"). (NECNP Ex. 7 Reference 5 at 2, 4.)

Filling the risers to the respective levels of the filler " Tees" will provide an oil head of 4'-2)" above flood level at one riser and 4'-

41/8" above flood level at the other riser. (Ld. at 1, 3. ) This would counter the 6 5/8" head of water that would cover the submerged transmitters during the postulated flood. These elevations and static pressure heads are in approximate, but not exact, agreement with the

c:

a-design as specified in the "Special Conditions" for acceptance of the submergence qualification of this equipment. (NECNP Ex. 7, EQ Assessment Report at 1; see supra Fdg. 82.) Thus, we find NECNP's allegations concerning the design of the risers and the alleged discrepancies in flood level, fill level and filling instructions to be without merit.

86. In regard to the elimination of the pressure test of the riser assemblies (Fdg. 83), the original purpose of the 60 pounds per square inch test was to examine the assemblies for leak tightn'ess. Upon later review, Applicants determined that the equipment could be damaged by the pressure test, so the test was eliminated. (Salvo,Tr.450,NECNP Ex. 7, Reference 7 at 2 (Rev. "C"), 3, 5, 9.)

87. According to the Applicants a visual examination was performed to verify that the system was leak tight. (Salvo, Tr. 451.) Applicants also asserted that leakage could be adequately monitored during periodic calibrations that take place at intervals of a year to 15 months and, generally, during entries into the containment. (Thomas,Tr.433-35.)

88. Applicants also asserted that under accident conditions the design of the equipment is such that pressure on both sides of the device would be equalized, and that there would be no differential pressure on the system other than the static head of the liquid which is minimal. (Salvo,Tr.480-81.) No explanation of how pressure equalization would be achieved by the design was given, however.

89. The environmental qualification duration for submerged conditions of the containment water level measuring transmitters depends

n upon an. adequata level of silicone oil remaining in the riser devices to maintain a small static head to counter the 6" static head of water above the lower units under submerged conditions. Absence of a differential pressure head in the system under environmental pressure conditions is also required. (NECNP Ex. 7, Environmental Qualification Assessment Report, at 1.)

90. No maintenance requirements are specified to maintain the qualified life of the level transmitters in EQ File 174-15-01. (Id.at 3.) The Board directs that maintenance requirements for the silicone oil filled riser assemblies be developed and included in the EQ file to insure that an adequate level of oil is continuously present in the riser assemblies to maintain the qualified life of the level transmitters under the required environmental conditions.

CONCLUSIONS OF LAW The Board has considered all of the evidence presented by the parties and the entire record of this proceeding. All issues, arguments, or proposed findings presented by the parties, but not addressed in this partial initial decision, have been found to be without merit or unnecessary to that decision. Having resolved all on-site safety and emergency planning issues in controversy, pursuant to 10 C.F.R. $$50.57(c) and 50.47(d), the Board authorizes issuance of a license to operate Seabrook Station, Unit I up to 5% of rated power, subject to the condition set forth below in paragraph number one of our Order. We find that there is reasonable assurance the Seabrook Station,

o I

o Unit 1, can be operated up to 5% of rated power without endangering the public health and safety, and that the state of on-site emergency preparedness provides reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

Further, if the other Licensing Board, which is considering off-site emergency planning issues, determines to authorize a full-power operating license, prior to the issuance thereof Applicants must have satisfied the trree conditions set forth belos in paragraph number two of our Order.

ORCER WHEREFORE, in accordance with the Atomic Energy Act of 1954, as amended, and the rules of practice of the Commission, and based upon the foregoing Findings of Fact and Conclusions of Law, IT IS ORDERED that:

1. Upon making the applicable findings required under 10 C.F.R. 550.57(a), the Director of Nuclear Reactor Regulation is authorized to issue a license authorizing low power testing and operation limited to 5% of rated power for the Seabrook Station, Unit 1, provided that, prior to the issuance thereof, Applicants shall have developed and placed in the appropriate environmental cualification file, maintenance procedures required to insure that an adequate level of oil is continuously present in the silicone-oil-filled riser assemblies associated with the containment water level transmitters to maintain the qualified life of the transmitters

a under the required postaccident environmental conditions (see fdg.

90, supra);

2. If a full-power operating license is authorized by the other Licensing Board which is considering off-site emergency planning issues, prior to the issuance thereof, Applicants, with respect to the Safety Parameter Display System, shall have:

(a) dedicated the SPDS terminal so that a continuous display of the Critical Safety Functions will be achieved g, by means of a test function and test computer, have an SPDS display on every cathode ray tube format in the control room to continuously display the SPDS top level display (see fdg. 30, supra);

(b) provided for continuous display of residual heat removal and hydrogen concentration critical safety function variables at the prime SPDS station (see fdg. 33, supra); and (c) established a radiological control screen at the prime SPDS station which, at the minimum, can be called up by the operator and will display steam line radiation and stack radiation parameters (see fdg. 36, supra).

Pursuant to 10 C.F.R. 92.760(a) of the Commission's Rules of Practice, this Partial Initial Decision will constitute the final decision of the Comission forty-five (45) days from the date of issuance, unless an appeal is taken in accordance with 10 C.F.R. 92.762 or the Commission directs otherwise. See also 10 C.F.R. 992.764, 2.785 and 2.786.

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6 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. 62.762(c).

THE ATOMIC SAFETY AND LICENSING BOARD 1.U N Sheldon J. Wblfe, Chairman ADMINISTRATIVE JUDGE p

Jerry Harbour ADMINISTRATIVE JUDGE

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r r 6^ ^$^ h.

Emmeth A. Luebke ADMINISTRATIVE JUDGE Dated at Bethesda, Maryland this 25th day of March,1987.