ML20069G460
| ML20069G460 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 03/23/1983 |
| From: | Goldsmith M, Hubbard R, George Minor SUFFOLK COUNTY, NY |
| To: | |
| References | |
| REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR ISSUANCES-OL, NUDOCS 8303250157 | |
| Download: ML20069G460 (50) | |
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M L'%D 2enc 3 UNITED STATES OF AMERICA NDCLEAR REGULATORY COMMISggO g g BEFORE THE ATOMIC SAFETY AND LICENSING BbARD o
)*
In the Matter of )
)
LONG ISLAND LIGHTING COMPANY )
) Docket No. 50-322 (0.L.)
(Shoreham Nuclear Power ) .
Station, Unit 1) )
)
---) .
SUFFOLK COUNTY SUPPLEMENTAL TESTIMONY ON CONTENTION 7B BY MARC W. GOLDSMITH, RICHARD B. HUBBARD AND GREGORY C. MINOR March 23, 1983 l
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8303250157 830323 PDR ADOCK 05000322
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i TABLE OF CONTENTS ,
Introauction . . . . . . . . . . . . . . . . . . . . . . . 1 I. UNRESOLVED SAFETY ISSUE A-17 . . . . . . . . . . . . 2 O
A. POINT 1: Comments on Conran's Baseline Considerations . . . . . . . . . 2 B. POINT 2: Lack of Progress in Resolving USI A-17 and
. Licensing Implications . . . . . . . . . . 9
- 1. Lack of Progress Toward Resolution of USI A-17 . . . . . . . . . . 10
- 2. The Staff's Supplemental Testimony Provides No Basis for Making the North Anna Findings . . . . . . 16 C. POINT 3: Systems Interaction Study for Shoreham . . . . . '
.:. . . . . . . . . 20 -
II. SAFETY CLASSIFICATION . . . . . . . . . . . . . . . . 22 A. Conflicts in Terminology Unresolved . . . . . . . . . . . . . . . . . . . 24 B. LILCO's FSAR Amendment Insufficient . . . . . . . . . . . . . . . . . . 26 C. Undocumented Safety Classification Program . . . . . . . . . . . . . 30 D. The Staff Reliance on the Standard Review Plan and Regulatory Guides is Not Justified . . . . . . . . . . . . . . . . . 32 E. Correctness of Safety Significance Assumed Rather Than Demonstrated . . . . . . . . . . . 37 1
UNITED STATES OF AMERICA
' NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of )
)
LONG ISLAND LIGET7NG COMPANY )
) Docket No. 50-322 (0.L.)
)
(Shoreham Nuclear Power )
Station, Unit 1) )
)
-)
SUFFOLK COUNTY SUPPLEMENTAL TESTIMONY ON CONTENTION 7B BY MARC W. GOLDSNITH, RICHARD B. HUBBARD AND GREGORY C. MINOR . . .
I am Marc W. Gtidsmith and I am employed by Energy Re-search Group, Inc. located at 400-1 Totten Pond Road, Waltham,
~
MA 02154. A copy of my professional qualifications appears in the record of this proceeding following Tr. 1113.
I am Richard B. Hubbard and I am employed by MHB Technical Associates, located at 1723 Hamilton Avenue, San Jose, CA 95125. A copy of my pr'ofessional qualifications appears in the record of this proceeding following Tr. 1113.
I am Gregory C. Minor and I am employed by MHB Technical San Jose, CA 95125.
Associates located at 1723 Hamilton Avenue, A copy of my professional qualifications appears in the record of this proceeding following Tr. 11'13.
This testimony addresses matters discussed in the February 9, 1983 Affidavit of NRC Staff Member James H. Conran
(" Affidavit"), and the NRC Staff Supplemental Testimony on Contention 7B by Roger J. Mattson, Richard H. Vollmer, Charles E. Rossi, Ashok C. Thadani and Franklin D. Coffman, Jr., dated March 10, 1983 (" Staff Supplemental Testimony").
Our testimony is divided into two parts, consistent with Mr. Conran's Affidavit. The first part, sponsored by Messrs.
Goldsmith, Minor and Hubbard, addresses points raised by Mr.
Conran and the' Staff relating to Unresolved Safety Issue
("USI") A-17 -- Systems Interaction. The second part, spon-
.sored by Messrs. Minor and Hubbard, addresses points raised by Mr. Conran and the Staff relating to safety classification.
I. UNRESOLVED SAFETY ISSUE A-17
~
In his Affidavit, Mr. Conran indicafes his agreement with some of the opinions expressed in Suffolk County's original Contention 7B testimony concerning USI A-17, particularly regarding the lack of progress toward timely resolution of that USI and the difficulty in making the North Anna findings given its current status. In the testimony which follows, we set forth comments on particular points made by Mr. Conran in his l Affidavit and by the Staff in its Supplemental Testimony.
A. POINT 1 -- Comments on Conran's Baseline Considerations (Affidavit, paces 5-9)
Mr. Conran sets forth five " baseline considerations" in characterizing the decisions and actions taken by the NRC Staff and the Commission relating to the systems interaction issue.
2-
Thus Mr. Conran states that: Staff management and the Commission have judged that the systemh interaction issue is "a legitimate safety concern, serious enough to warrant designa-tion as an Unresolved Safety Concern" (Affidavit at 3); that judgment "was reconfirmed and reinforced in the aftermath of the TMI-2 accident" (Affidavit at 3); " staff management and the Commission intended timely resolution of this important issue" (Affidavit at '5); the time originally allowed for resolution of the issue "necessarily implied and, indeed, required assignment of high priority and strong commitment to the USI A-17 program by staff management and the Commission" (Affidavit at 5-6); the post-TMI systems interaction program has a "high priority as-signment and timely resolution objectives (Affidavit at 8); and
~~
" decisions and actions [on A-17] were based broadly on widely-shared cualitative judgments regarding the importance of the issue involved and the necessity for prompt action and timely resolution . . . . The decisions involved were evolved through a highly-visible and open consensus forming process . . . ."
l (Affidavit at 9).
For the reasons set'forth be?'ow, we agree with Mr. Conran:
USI A-17 has been accorded a high degree of safety significance which requires a timely resolution, whether the USI A-17 pro-L gram is viewed as " confirmatory" or is characterized with some other descriptive term. The data and testimony which follow
~
l further confirm the high priority accorded to this USI.
l l
A Since 1977, the NRC has repeatedly confirmed the high pri-ority given to USI A-17. The Task Action Plan for Task A-17, Systems Interaction in Nuclear Power Plants, was first de-scribed in NUREG-0371, Revision 1, Approved Task Action Plan for Category A Generic Activities (December 1977). The original A-17 Task Action Plan stated:
.The problem to be resolved by this task is to establish a systematic process to review plant systems to' determine their impact on various other plant systems.1/
The plan for resolution of USI A-17 was:
to develop and implement, to the extent that a .
study indicates the need, a method of review that will extend the present review techniques in suf-ficient. breadth and depth to assure a systematic and comprehensive review of systems interaction. _,
The plan will also include the development of criteria and procedures to assure that applicants incorporate appropriate systems interaction con-siderations into their design and review process.2/
-1/
Task Action Plan, Task No. A-17, Rev. 0 (November 15, 1977), at 2. This is contained in NUREG-0371, Rev. 1, re-printed as Appendix F to NUREG-0410, NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants (January 1978).
2/ Id. The original USI A-17 Task Action Plan identified the following as the major tasks to be performed:
(a) Establish a uniform designation of plant systems and their associated functional inputs and outputs, and deter-mine the interface points or boundaries where interactions can occur, inc3uding identification of the types of inter-actions.
(Footnote cont'd next page)
_4_
i In 1977, the Staff's Office of Nuclear Reactor Regulation
("NRR") developed criteria for grouping generic technical activities into categories indicative of their priority. Four categories (A through D) were identified. The NRC's program for resolution of generic issues then focused primary attention on the highest priority activities--those in Category A.3/
Category A activities were defined as:
Those gen'eric technical activities judged by the Staff to warrant priority attention in terms of manpower and/or funds to attain early resolution.
These matters include those the resolution of which could (1) provide a significant increase in assurance of the health and safety of'the public, or (2) have significant impact upon the reactor licensing process.4/
/
(Footnote cont'd from previous page)
(b) Compare the Standard Review Plan (SRP) against item (a)'above to determine the extent to which the SRP already adequately addresses interdisciplinary review areas and
' systems interactions. Also, determine the extent to which the SRP includes consideration of systems interac-tions. . . .
(c) Develop, to the extent necessary, revisions of the SRP based on the results of task (b).
(d) Develop criteria and procedures, including informa-tion requirements, for use by applicant;. in their design and review of plant designs for systems interaction. Id.
at 2-4.
l 3/ NUREG-0410, at A-3.
Id., at Appendix B. Categories B-D included activities 4/
judged as: "important in assuring the continued health and safety of the public but for which early resolution is not (Footnote cont'd next page)
~
a l i
USI A-17 was one of 41 tasks, selected from a total of ap-proximately 133, which the Staff's TecEnical Activity Steering Committee determined to be of the highest priority, and there-fore placed in Category "A."5/ Therefore, the designation of the systems interaction issue as "A"-17 reflects the Staff's view, in 1977, that its resolution could either " provide a sig-nificant increase in assurance of the health and safety of the public, or have significant impact upon the reactor licensing process."
The high priority accorded to- USI A-17 in 1977 was recon-firmed in 1979. In NUREG-0510, the Staff defined an
" Unresolved Safety Issue" as: , ,,
a matter affecting a number of nuclear power plants that poses important questions concerning the adequacy of existing safety requirements for which a final resolution has not yet been .
developed and that involves conditions not likely to be acceptable over the lifetime of the plants affected.17 (Footnote cont'd from previous page) required"; having "little direct or immediate safety, safeguards or environmental significance, but which could lead to improved Staff understanding of particular technical issues or refinements in the licensing process";
and, not " warrant [ing) the expenditure of manpower or funds because little or no importance to the safety, envi-ronmental or safeguards aspects of nuclear reactors or to improving the licensing process can be attributed to the activity." Id.
5/ Id., Appendix A at A-8 and Appendix D.
6/ NUREG-0510, Identification of Unresolved Safety Issues
, Relating to Nuclear Power Plants (January 1979) at 10.
In NOREG-0510, the Staff grouped and categorize 4 issues by activity type (8 categories) and by relative worth based on risk-related criteria (4 categories) to determine which ones were USIs. Task A-17 was identified as involving group 3 type o
activitics (" performing studies to confirm the adequacy of cur-rent Staff safety requirements"); it was placed in risk-related the highest category 1, de. fined as " Potential High Risk Items,"
of the risk-based groupings.1/
After evaluating 86 generic tasks that were in Category A (NUREG-0410 ranking),
the tasks that were in activity groups 1, 2 or 3, an6 the tasks in risk-based categories I or If, as well 4
Z-as reviewing the Abnormal Occurrences reported to Congress dur-
" ing 1977 and 1978, the Staff determined that only 17 generic d issues qualified as USIs.
~ A-17, Systems Interactions, was
]
among them.8/ The significance of the Staff's identification of A-17 as a USI is unmistakeable, in light of the following Staff explanation of how the USI definition was applied:
In applying this definition, matters that pose "important questions concerning the adequacy of existing safety requirements" were judged to be those for which resolution is necessary to (1) compensate for a possible major reduction in the degree of protection of the public health and safety, or (2) provide a potentially significant decrease in the risk to the public health and safety. Quite simply, an " Unresolved Safety 7/ Id., Appendix B at B-4, B-13, and Appendix C at C-1.
_8 /, _I_d. at 13-16.
W Issue" is potentially significant from a public to safety result instandpoint and NRC action onits theresolution is likely/
affected plants.9 (Emphasis added.)
A-17 has never lost its designation as an USI. Rather, as
, Mr Conran notes in his Affidavit (page 3), the safety signifi-cance and importance of USI A-17 was confirmed and reinforced following the TMI-2 accident by the inclusion of a systems in-teraction program in the Commission's TMI Action Plan.10/ The TMI Action Plan provides further evidence of the priority of resolving the systems interaction issue. The Task II.C.3 ef-fort was ranked as having high safety significance and a near term (within 2 years of implementation) benefit.ll/
Recent Staff actions have again doc,umented the continuing _,
high priority given to USI A-17. In the November 10, 1982 draft of NUREG-0933, Rev. O, "A Prioritization of Generic Safety Issues," the Safety Program Evaluation Branch of the NRC's Division of Safety Technology assigned the highest priority to the systems interaction issue.12/ The stated purpose of the priority rankings contained in draft NUREG-0933 is "to assist in the timely and efficient allocation of resources to those 9/ Id. at 10.
10/ NUREG-0660, HRC Action Plan Developed as a Result of the TMI-2 Accident, Item II.C.3 (May 1980).
11/ NUREG-0660, Tables B.2, at B.2-5, and B.3, at B.3-3.
12/ Draft NUREG-0933, Rev. O, at xxxii.
safety issues that have a high potential for reducing risk and in decisions to remove from further consideration issues that have little safety significance and hold little promise of worthwhile safety enhancement."13/ Only 15 of the 74 issues o considered (certain issues were eliminated) were assigned a "high" priority.
Assignment of a BIGH priority means that strong efforts to achieve an earliest practical resolu-tion are appropriate. This is because (a) an im-portant' safety deficiency is involved (though gen-erally the deficiency is not severe enough to require prompt plant shutdown), (b) a substantial safety improvement is li'kely to be attained at_a low enough cost. to make the improvement very worthwhile, or (c) the uncertainty of the safety assessment is unusually large and an upper-bound risk assessment would indicate an important safety deficiency.li/
USI A-17 received a "high" priority rating in draft NUREG-0933.15/
The facts set forth'above document the high priority and safety significance which have been accorded to resolution of USI A-17 since it was first iden;ified. We now turn to the actual progress that has been made toward resolution.
B. POINT 2: Lack of Progress in Resolving USI A-17 and Licensing Imolications In his Affidavit, Mr. Conran disagrees with the Staff's position that the USI A-17 program "provides currently an 13/ Id. at ii-iii.
14/ Id. at iv.
15/ Id. at xxxii.
adequate basis for the ' justification for operation' conclusion required under North Anna . . . .
(Affidavit at 2). Mr.
Conran stresses his view that USI A-17 requires timely resolu-tion but that progress toward such resolution cannot be demon-strated. We agree with Mr. Conran's conclusion that there has been a lack of progress toward resolving USI A-17 which renders the program inadequate to provide the basis for a North Anna finding.
In the testimony which follows, we: (1) support the con-clusion in Mr. Conran's Affidavit concerning the lack of and (2) disagree with progress toward resolution of USI A-17, the Staff's assertions that current licensing requirements provide reasonable assurance that Shoreham can be operated 4
without undue risk to the public health and safety. (Staff Supp. Test. at 3-4;.
- 1. Lack of Procress'Toward Resolution of USI A-17 When the Task Action Plan for systems interaction was ini-tiated in 1977, the date for completion of the task was l
December 30, 1978. 11/ As of January 1979, the NRC still ex-pected to complete " Phase I" of the task (the development of a workable methodology) by September 1979.11/ Phase I was to be 16/ Task Action Plan, Task No. A-17, Revision O (November 15, l
l 1977) at 9. See footnote 1.
)
l 17/ ,NUREG-0510, Appendix A at 12.
I l
l
the ap-done under contract by Sandia Laboratories; " Phase II,"
plication of the methodology to actual plants, was contingent u upon whe~ther Sandia actually identified systems interactions as a serious problem.18/ By September 1979, the target date for M'.
hi" Sandia's Phase I study had been moved back to March 1, 1980.
The reason given for the delay was that Sandia had
- .7
- underestimated the level of effort required for the task.
-6
,.( Phase II was.s'cheduled for completion in March 1981.19/
~
P' In 1980, the NRC revised and rescheduled the plan for res-olution of USI A-17. As of May 1980, the revised plan includ-ed: (1) a plan to develop and demonstrate workable c.J, methodologies for systems interaction analysis; and (2) a g].. series of nuclear power plant systems. interaction reviews.20/
g
% In connection with the development of methodologies, three NRC H3 contractor laboratories (. Lawrence Livermore, Battelle, and Brookhaven) each provided a report to the Staff on the state-of-the-art in systems interactions methodologies.21/ These I. , 18/ Id.
19/ NUREG-0606, " Unresolved Safety Issues" Summary (September l 4, 1979) at 1.
20/ NUREG-0660, Item II.C.3, at II.C-7.
State-of-the-Art 21/ NUREG/CR-1859, " Systems Interaction:
Review and Methods Evaluation" (Lawrence Livermore)
(January 1981); NUREG/CR-1896, " Review of Systems Interac-
[
tion Methodologies" (Battelle Columbus) (January 1981);
NUREG/CR-1901, " Review and Evaluation of System Interac-tion Methods" (Brookhaven) (January 1981).
l b l
reports, published in January 1981, all suggested that a combi-nation of existing methodologies could be used to provide a systematic approach to systems interaction anal 3;es. The NRC incorporated this guidance into the Staff Summary Letter Report, "The Approach to Systems Interaction in LWRs" and in the " Initial Guidance for the Performance of Systems Interac .
tion Analyses at Selected LWRs."22/ Although the guidance document is in. circulation within the Commission, to our knowledge the recommended methodologies have not been systemat-ically implemented at any plant.
~
As of May 1980, the Staff's USI A-17 program also included review of three plant-specific systems interaction studies: the Diablo Canyon Seismic Systems Interaction Walkdown; the San Onofre Seismic Systems Interaction Walkdown; and the Indian Point 3 Systems Interaction Review. The Diablo Canyon and San Onofre' studies were each limited to spatially coupled interac-tions initiated by seismic events.13/ The Indian Point study includes a walkdown and, in addition, a dependency analysis.24/
22/ Systems Interaction Section, Staff Summary Letter Report, "The Approach to Systems Interactions in LWRs" (June 1981); Reliability and Risk Assessment Branch, " Initial Guidance for the Performance of Systems Interaction Analyses at Selected LWRs (Guidance for Interim Use and Comment)" (January 7, 1962).
23/ Memorandum from Richard Savio to ACRS Members, " Status of
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the Review of the Diablo Canyon Systems Interaction Study (October 30, 1980); Memorandum from L.S. Rubenstein to R.L. Tedesco, " Evaluation of San Onofre Units 2 and 3 Seismic Interaction Program" (April 7, 1981) at 4.
l 24/ 'ACRS Subcommittee on Safety Philosophy, Technology, and Criteria, Meeting Transcript, (February 26, 1982) 27, 28, 34, 35, 65-69.
12 -
As of the end of September 1981, the NRC Staff had com-pleted its initial evaluation of the methods to be used in the Diablo Canyon and San Onofre systems interaction studies.25/
In a meeting on July 24, 1981 with the Indian Point licensees,
" the Staff commented upon its review of a preliminary submittal In October 1981, a of the proposed Indian Point 3 study.21/
Staff proposal for the review of four plants using NRC-developed methodologies was submitted to the NRR Director.
That proposal estimated that a systems interaction analysis of a nuclear plant and its review would take about eighteen months.22/ Thus, as of October 1981, the Staff intended to gather data from seven plants in the course of its USI A-17 program. Two of the plant reviews (Diabio Canyon and San Onofre) were to be limited in scope; the other five (Indian Point and the four reviews using NRC-developed methodology) were to be more complete.
The situation today is very different. As Mr. Conran notes at page 11 of his Affidavit, the evaluation of 25/ Diablo Canyon SER, Supplement 11 (October 1980); San Onofre SER, Supplement 2 (May 1981).
26/ Meeting Summary and Status Report for Meeting with PASNY/Ebasco on Proposed Indian Point 3 System Interaction Program, July 24, 1981 (undated).
Memorandum from Thomas Murley to Harold Denton, "Imple-27/ mentation of Systems Interaction Interim Guidance" (October 30, 1981).
interactions identified in the Diablo Canyon walkdown has not been submitted to the NRC, nor is there an agreed-upon date for submittal. Similarly, the unevaluated search results of the Indian Point 3 systems interaction study will not be submitted to the Staff until late 1983. (Conran Affidavit at 11).
The status of the four plant reviews has also changed. As Mr. Conran states in his Affidavit (pages 19-20), following the submission of the October, 1981 Staff proposal recommending the initiation of the four plant reviews, no authorization was received from the NRR. In February, 1982, however, the NRC stated (in a letter to the ACRS) that "[The Staff] proposes to begin soon with reviess of four near-term operating license plants using two different methodologies #for two plants each."28/ It now appears, however, that the Staff USI A-17 (See program has discarded the four plant reviews altogether.
Staff Supp. Test. at 6-7).
The foregoing " progress" or lack thereof toward resolution of USI A-17 has been criticized by the ACRS. In January, 1982, the ACRS stated:
The Committee believes that it is already l
j past the time when its recommendation for a systems interactions study on Indian Point 3 l
should have been completed. The ACRS is also disappointed with the absence of even a lim-ited review of systems interactions by eitner the NRC Staff or the Applicant in some of the i
recent operating license reviews.
28/ Memorandum from William J. Dircks, NRC, to Paul Shevmon,
,ACRS, " Systems Interactions" (February 12, 1982).
1
- The Committee believes that the matter of systems interactions has been delayed for too long. Rather than delay another several years for most plants while the Staff decides on the ultimate approach, it appears that a e
- staged approach may be preferable, with lati-4.
tude to the licensee to do as good a job as he now can.29/
! Despite this view expressed in January 1982, today, about 14 months later, there has still been no substantial progress to-e, ward resolution.
The Staff has stated in its Supplemental Testimony (page
- 7) that it now believes a basis for new licensing requirements
.5 g could result from the A-17 program- in October 1984. .However, based on past history, there certainly is no guarantee that this new schedule will be met. Moreover, the results of the
- g. y .
A-17 program as now described in the Staff's Supplemental Tes-e .,
timony-(page 7) will yield data from only one plant, Indian g-i Point 3, which is a PWR.30/ Although the methodology to be used for the Indian Point study now planned by the Staff may be applicable to BWR studies, the Staff apparently does not plan to test those methods on BWRs as part of the USI A-17 program.
Therefore, there is no indication that either the current Staff 29/ Letter from J.J. Ray, Acting ACRS Chairman, to William J.
Dircks (January 8, 1982).
30/ The two other plant specific systems interactions studies (Diablo Canyon and San Onofre), both of PWRs, are not mentioned in the Staff's latest proposed A-17 plan.
(Staff Supp. Test. at 7) f e
plans for A-17, or the ultimate result if those plans are implemented, will have applicability to the Shoreham plant.
4 Accordingly, not only has progress been exceedingly slow toward USI A-17 resolution, but the " resolution" which the Staff now projects for October 1984 has not been demonstrated to be ap-plicable to Shoreham. Based on the foregoing, we conclude that the NRC's progress toward resolving USI A-17 has been'inade-quate.
- 2. The Staff's Supplemental Testimony Provides No Basis for Making the North Anna Findings The Staff asser.ts in its Supp.lemental Testimony (page 4) that USI A-17 "is confirmatory in nature." This statement ruggests that the Staff believes the issue has less safety sig-nificance than other USIs and that its timely resolution is not an important or necessary consideration in the licensing process. .
We disagree. First, even'if USI A-17 is " confirmatory,"
it has been accorded extremely high priority. (See discussion in Point 1). A USI of such high priority--even if
" confirmatory"--must have timely resolution or material progress toward resolution such that there is a factual basis for the Staf f's opinion that existing lf censing review processes are adequate. Further, despite the Staff's state-ments that "A-17 was initiated to confirm that present review pr.ocedures and safety criteria provide an acceptable level of r
independence for systems required for safety by evaluating the potential for the more important undesirable interactions between and among systems" (Staff Supp. Test. at 5) (emphasis in original),.31/ the fact remains that such confirmation has
- no't yet occurred. (See discussion of A-17 progress.) More-over, every NRC priority ranking of generic issues has placed resolution of USI A-17 in the highest possible category.
Therefore, ev;en if the adjective " confirmatory" were applicable to the Staff's anticipated results of an eventual A-17 resolu-tion, that fact does not mean' that achieving the resolution is
" unimportant" or " insignificant." ~
The Staff states in its Supplemental Testimony (page 5) that: /
Progress in this program [ Task A-17] to date has provided no indication that present review proce-dures and criteria do not provide reasonable as-surance that the effects of potential systems in-teractions on plant safety will be within the effects on plant safety previously evaluated (i.e.
within the design-basis envelop).
With all due respect, we believe this statement is not entirely correct. The only study in the A-17 program that compared potential systems interaction events with specific Standard Review Plan ("SRP") requirements--the Sandia study--identified deficiencies in the SRP. For example, in its Phase I study 31/ See also Shoreham SER, at B-10, for an almost identical statement.
modelling limited portions of the Watts Bar PWR, Sandia found specific examples of failures in the SRP to cover particular adverse systems interactions.32/ Examples of these
.- deficiencies in the SRP are set forth in the Attachment to this testimony which reproduces Table 7.1 from the Sandia study.
The Sandia study methodology was found to be inadequate for n
e; broad systems interaction detection; however, the results of
," the limited Sandia review are important because inadequacies in the SRP were detected. Thus, the existing review procedures, i.e., the SRP, have not be'en shown to be adequate.33/
- Weaknesses in the SRP also have been discovered as a result of actual events that involved adverse systems interac-
~~
tions. For example, the Quad Cities -flo6 ding incident led to
' changes in the SRP with respect to internal flooding and high energy line breaks; similarly, the Browns Ferry. fire led to SRP change's with respect to Pire protection and cable separation.
Therefore, in the past, certain potential adverse systems 32/ See NUREG/CR-1321, Final Report-Phase I Systems Interac-tion Methodology Applications Program, (Sandia National Laboratories) (April 1980).
33/ The Sandia study did not conclude that the soft spots in the SRP--areas when adverse interactions were not accounted for--would have resultedHowever, in effects outside the the fact that
' so-called " design-basis envelop."
Sandia did identify that the SRP is not complete in ac-counting for systems interactions, together with the over-all lack of A-17 progress, in our opi.nion, undercut the confidence that can be placed in the existing review pro-l cedures.
1 interactions have not been evaluated in the SRP until after the l cvents had actually occurred. Thus, the SRP has addressed certain systems interactions on a reactive basis, and has been found not to address at all certain other interactions.
The initial results of the Diablo Canyon study, which was limited to the identification of seismic interactions, identi-fled. some 677 potential interactions. Of these, 228 resulted in plant modifications.31/ Because the results of the study have not been fully evaluated, it is difficult to determine how
.many of these modifications were critical to the plant's safe operation. However, the sheer number of changes suggests that systems interactions are indeed slipping-through the review process.
Thus, the few studies on systems interactions done for the Staff and by utilities do not corroborate the adequacy of the existing review process. Rather, these studies have led to changes in plant design. Thus, we agree with Mr. Conran:
Although the results of these efforts have not yet been fully-evaluated by the utilities involved and reviewed by the staff, in several instances on the basis of licensees' own prudent judgment, modifications to facil-ity designs have already resulted as a result of system interaction reviews. (Affidavit at 13).
1 31/ Diablo Canyon SER, Supplement 11 (October 1980), at 6-1.
1
Finally, we disagree with the Staff's " reasonable assur-ance" conclusion (Staff.Supp. Test. at 5, quoted earlier) as applied to Shoreham because, to our knowledge, there have been no generic or specific BWR systems interaction studies performed or reviewed by or on behalf of the Staff. Therefore, to our knowledge, the Staff has no study data on which to base a BWR- or Shoreham-specific conelusion. In addition, as discussed above, the " progress" in the A-17 program has not yielded any results upon which such a conclusion could be based.
C. PpINT 3: Systems Interaction Study for Shoreham Mr. Conran urges that all licensees and operating license
~~
applicants should be required to commenc6 limited systems in-teraction reviews. (Affidavit at 12-13). He believes such studies would produce beneficial findings for the specific
~
plants involved and would also provide data needed by the Staff in its USI A-17 program. (13. at 14). We agree on the need for systems interactions reviews at Shoreham.35/ Set forth below i
I are our views on the outline of such reviews.
l 35/ The matters discussed earlier, including the deficiencies in the SRP, the lack of focus of the Staff's USI A-17 ef-fort on BWR systems interactions, the potential adverse interactions found where SI studies have been conducted, and the lack of progress by NRC on generically resolving the USI A-17 issue, all indicate the need for systen s in-teraction studies at Shoreham.
l 1
The NRC-sponsored studies done by the national laboratories indicate that a multiple ~ step process using several methodologies is necessary to assure a systematic eval-uation of a plant for systems interactions. These studies also indicate that there are several established methods to accomplish the task and that no,particular method is better than another. However, all the methods seem to follow the same grouping of major actions to review and identify systems inter-actions. These methods were circulated in draft form within
.theNRCalmosttwoyearsago.)f/ In January 1982, a draft
" Initial Guidance for the Performance of Systems Interaction Analyses at Selected LWRs (Guidance for Interim Use and Comment)"31/ was published. This guidan6e should be considered in preparation of the program for Shoreham systems interaction analyses.
The scope of a study at Shoreham should account for func-tionally and spatially coupled systems interactions. Human in-teractions are more complex, involving a combination of By plant-specific and generic industry-wide considerations.
concentrating on the first two interactions (spatial and 31/ Staff Summary Letter Report, "The Approach to Systems Div.
In-teractions in LWRs" Systems Interaction Section RRAB, of Safety of Technology (June 1981).
37/ RRAB, Div. of Safety Technology, Office of Nuclear Reactor Regulation (January 7, 1982).
functional), the impact of human interactions con ba significantly reduced. The three step process recommended by the NRC38/ should be used with one or more methods used iteratively between and within each step. The identification of spatial couplings will require that a systematic visual in-spection be made by a multidisciplinary team to provide joint Further, it judgment on whether systems interactions exist.
will be important to do this detailed walkdown with checklists for the interaction types (e.g. , Diablo-type targets and sources) with dif ferent types of engineers evaluating items with respect to their individual specialities.
II. SAFETY CLASSIFICATION The purpose of this section is to sit forth our evaluation of the adequacy of the proposed "fix" to the Shoreham safety classification matters presented in the Staff Supplemental Tes-timony. 'Our testimony also addresses the related matters presented in the Conran Affidavit.
The Staff proposes a two step "fix" to ensure that LILCO will comply with General Design Criterion 1 ("GDC-1") of 10 CFR Part 50, Appendix A during the operation of Shoreham. (Staff l
38/ The process involves: (1) "Model the plant to select the combinations of systems for detailed evaluation"; (2)
" search'the selected combinations of systems"; and (3)
- " evaluate the discovered systems interactions against criteria for corrective action." Id. at 2-3.
Supplemental Testimony at 11) . First, the Staff has proposed for the operational phase that the ASLB find that:
[T]he Staff's definition of 'important to safety' is correct and binding on the Applicant. . . .
(Staff Supplemental Testimony at 11).
Second, the Staff requested LILCO to commit in the FSAR that Shoreham will comply with GDC-1 during operations as follows:
Amend the.FSAR to commit for non-safety related s truc tur.e's , systems, and components, to include in the preventive and corrective maintenance program, the design change control program, the procedures for procurement of equipment, the procedures for modifications and removal of equipment from service, and the OA program, a provision that,.as a minimum, the equipment and associated software shall be accorded the safety significance given to it in the FSAR, the technical specifications and the emergency operating procedures.. The charters _
and decisions of the Review of Oper~ations Commit-tee, the Offsite Nuclear Review Board, and the
- Man.ager of Quality Assurance shall also reflect
~
these considerations. (February 18, 1983 Letter, Eisenhut of NRC to Pollock of LILCO, attached to Staff Supplemental Testimony).
LILCO in a March 2 letter to the Staff agreed to amend the
' Shoreham FSAR and further, in a March 8 letter, LILCO submitted examples of the language it intends to incorporate in the pro- ,
posed FSAR amendment (these LILCO letters are included as at-tachments to the Staff Supplemental Testimony). The Staff appears satisfied by LILCO's FSAR commitment. (Staff Supple-mental Testimony at 10-12).
In our opinion, the Staff's proposed two step "fix" and LILCO's proposed FSAR amendments ar,e deficient. Our testimony
. ~
below discusses the deficiencies by covering the following matters:
o Conflicts in Terminology Uniesolved o LILCO's FSAR Amendment Insufficient
, o Undocumented Safety Classification Program o The Staff Reliance on the Standard Review Plan and Regulatory Guides is not Justified r o Correctness of Safety Significance Assumed Rather Than Demonstrated A. Conflicts in Terminology Unresolved Mr. Conran expressed concern that the terminology differ-t ences between LILCO and the Staff are more-than just a language problem. He' states: ..
I no longer believe that our { Staff and LILCO) 7 differences involve only a language problem to be sorted out mechanically . . . . [U)nderstanding of e
the fundamental safety concepts underlying the in the usage of the term 'Important to Safety' regulations cannot be imposed, (as for example Understanding must beby a condition to license). (Affidavit at
~
developed, and demonstrated . . . .
31-32)-
The Staff proposes that the Board "fix' LILCO's improper usage of the term "important to safety" by the Board's issuance of a post-facto edict and thus to require LILCO to use the Staff's definition of "important to safety" during the opera-tional phase. Such an edict does not address or remedy the potential deficiencies in classification and QA/QC imple-me'ntation resulting from LILCO's improper usage of the
classification terminology Juring the design and construction P h ase.
Absent issuance of an edict, LILC proposes to continue during operations, as it did during design and construction of
- (lis h,:j.
Shoreham, to apply its own terminology rather than that uti-
.M ;
lized by the NRC Staff. For example, at the February 18 jg5
-[ '
meeting, Mr. Mattson of the Staff inquired:
a
[:Bjas LILCO committed to use in operation the ter-if*f
'. minology'8important to safety', even though you
), didn't necessarily design with that terminology?
i,! Or have you not committed? Tr. 66.
Both Mr. Pollock and Mr. McCa'ffrey of LILCO responded that
.[.hh
"[wl e have not committed." Id. (Emphasis added).
- ,& Mr. Pollock
_9 :
ne .
- '*J also added: "I think we have it in testimony and what have you that our approach to importance to safeti is that it is safety "i
4 u
related, in our judgment." Id. (emphasis added).
.g ,
. ~1, Thu's, the efforts of the Staff to persuade LILCO to use
[
the Staff's definition of impor, tant to safety have been unsuccessful. The differences are further demonstrated by the exchange of letters between LILCO and the Staff (beginning with LILCO's letter of December 16, 1982) and the comments by Mr.
Mattson in the Februa'ry 18, 1983 meeting on classification l
issues:
I think in your response to Novak's letter of January 10th, if you had said we will accept theand Staff's definition as we move into operations, if you had built into that procedures and a quali-ty assur ance program and what have you -- that is, you would have attempted to determine the impor-tance to safety of equipment as you han.dled it in operations -- we would not havo had today's meeting, or we could have had a very short meeting. . . . If you had agreed in response to that [ January 10, 1983) offer for you to agree to accept our definition of 'important to safety'.
You obviously did not do that, and sent a letter back in reiterating your December 16th offer. So we were (sic) at a standoff, if you will, todayT-Tr. 143-144 (Emphasis added). ,
The agreement to disagree between the Staff and LILCO, and the Staff's proposed "fix" by a. Board edict, in our opinion, do not constitute an acceptable resolution of the conflicting use of safety classification terminology. Rather, as suggested by Mr. Conran, and as set forth in greater detail in the remainder of our testimony, we.believe understanding of safety classifi-cation terminology must be developed and demonstrated, not im-posed. -
B. LILCO'S FSAR Amendment Insufficient A fundamental flaw of the FSAR "fix" proposed. by the Staff and ad' opted by LILCO is that it.does not resolve the root cause of the problem (i.e., the difference in terminology and its effect in classifying and establishing standards of quality for important to safety but nonsafety-related components). In effect, the Staff "fix" has added no new requirements for Shoreham except inclusion of the associated software and adding the ROC, ONRB and Manager of QA to those to be bound by the resolution. Thus, the FSAR inserts do not ensure that LILCO has paid proper attention to all important to safety equipment and that the proper level of QA has been applied.
Evidence of the ineffectiveness of this resolution is seem in the proposed inserts to the FSAR which LILCO, in its March 8 letter, says are indicative of the types of changes that will be incorporated.39/ Insert A simply reiterates the words of Insert Ef,';f the Staff directive and calls it a " corporate policy".
I"' '. B specifies that maintenance people on nonsafety-related
. 'j SSEC's, shall, in exercising th'eir judgment on the appropriate
.e> '
measures to be applied, maintain the safety significance
....k Finally Insert C
'+
accorded to them in FSAR, TS's and EOP's.
says the QA manager shall also consider the safety significance *
' accorded to nonsafety-related SS&C's and computer software
~
given to them in the FSAR, TS's and EOP's.
j Because there is no systematic, cleprly documented, and 5([3
,. .- c.
well defined " safety significance" of nonsafety-related SS&C's
- jjf, s .s .
in the referenced documents (FSAR, TS's, and EOP's), these
.sJ .
L ' # 5, {
inserts provide no meaningful direction to Shoreham personnel Indeed,
- regarding how to implement the FSAR commitments.
. is adding the computer sof tware to the LILCO commitment meaningless since, to our knowledge, it is not mentioned in the TS's or the EOP's and is given only a functional description in
' the FSAR; in no case is it assigned a " safety significance."
In our opinion, the proposed LILCO commitments provided in the FSAR amendment inserts will not result in an effective 39/ March 8, 1983 letter, Pollock to Eisenhut, SNRC-854.
solution to tha cofoty cicssificotjon issus. Rathar, th9y perpetuate the existing difference (as addressed by Conran at 4
pp. 28-33). There will likely be littl.e or no substantive im-pact from these statements since the Shoreham documents, par-ticularly the FSAR, will continue to be permeated with the o.
fundamental difference in terminology and approach to classifi-cation which the resolution is intended to correct.
The Staff's proposed "fix" does not require LILCO to review the FSAR' and determine the true meaning of the term "im-portant to safety" at every place that it appears in the FSAR.
'The "fix" also does not require LILCO to review the'FSAR to de-termine whether the term "important to safety" should be used in some instances where the term " safety-related" now is used.
Such reviews and FSAR correction, where ppropriate, should be part of any "fix."
Incorrect usage of terms can influence the Staff's review of FSAR commitments. For example, the Staff acknowledged dur-ing prior testimony on Contention 7B that it had previously er-roneously approved LILCO's commitments in section 3.1 of the FSkR related to GDC-1 due to misunderstanding of the terminolo-gy of the LILCO commitment. Thus, in our opinion, LILCO should
~
to safety" pI?mptly correct all usages of the term "important and " safety-related" in the FSAR in accordance with the Staff's definitions. The Staff should then evaluate the changes in LILCO's commitments to confirm that all the Shoreham design,
construction, and operation commitments aro in cecordanco with NRC safety requirements.
In addition, Staff reviewers involved in plant assessments during the 40 year life of Shoreham should not be faced with the potential confusion resulting from LILCO's incorrect termi-nology.- Both Mr. Baass and Mr. Mattson of the Staff expressed concern at the February 18 meeting that LILCO's informal safety classification,could result in problems. Mr. Baass observed:
^
I think the question here is, are you really ad-dressing the safety aspects. I think that's the question, and we are not hearing an assurance that your system does address that. Tr. 124. ,
Mr. Mattson twice questioned whether LILCO's informality in safety classification might result in a serious problem at some time in the future. First, Mr. Matison followed Mr.
Haass's previous statement with his query:
The" point Walt was making is the answer we got was they stay, kicking the tires day in and day out, they see the plant, they know its operation, but Walt Haass was making the point ah, but there are Chapter 15 events, for example, or other accident situations that don' t happen, God willing , never, but they certainly don' t happen day by day.
Will, over a period of time, cognizance of the im ~
portance of a piece.of equipment, maybe a tertiary system to the functioning of safety equipment, be lost because the FSAR relevance of the equipment
' is not by procedure, continually brought before the person making the judgment about what to do?
Tr. 125.
Earlier.in the February 18 meeting, Mr. Mattson had expressed a similar question about future activities:
9 That's a wonderful philosophy, Mr. Pollock. Your philosophy is the kind of philosophy we want to hear from people at your level. I don' t quarrel with that a bit. Your philosophy of wanting availability and safety to go hand in hand and have an excellence of operation at all levels, that is super stuff. I wish everybody had that philosophy.
' But what about 30 years from now? Tr. 78 TEmphasis added). .
In our opinion, GDC-1 requires documented, systematic con-sideration of all SS&C's important to safety and assignment of OA/0C measures'in order to assure that the items will not be degraded as a result of activities occurring during the 40 year life of the plant. Absent complete correction of FSAR terminology, there will always be concern whether LILCO person-nel and Staf'f reviewers properly underst,and LILCO's FSAR com- ,
I mitments.
C. Undocumented Safety Classification Procram A,further deficiency in the Staff "fix" is that it does not require LILCO :o rely on an objective, systematic, docu-
~
mented program for considering SS&C's and for assigning OA/0C requirements. Rather, LILCO appears to depend upon a highly subjective design and operating philosophy, which will remain essentially unchanged under the Staff "fix." For example, Mr.
Pollock stated the following on February 18 with regard to how Shoreham's management systems will control the operation of the facility:
i t
. 1 It's not a definsd procedure; it 10 through i training, it is through our oporating philosophy and years of operating philosophy and maintaining the integrity of the facility. So it is training.
And now I can get into our training programs, our operator training programs, which are not just
) specifically licensed-required [ sic) training, but
- -" 7, it is balance of plant training and operating phi-
- losophy. Tr. 65.
s..
~
Mr. Pollock acknowledged there was no labeling on a draw-(.40 m r .
ing or on the Shoreham equipment to identify LILCO Category 2 fff g
~
nff; (nonsafety-related) items which are important to safety:
$*rf '
[T]here ,i's not a tag that says 'important to safe-SN3 c .1 ty.' But the training and philosophy of the plant t?'t -- and I can't disassociate operating reliability and operating availabili.ty with safety. They are
" hand in hand . . . .
s'., t, .
F 4.
So our philosophy is, by definition, an interpre-tation of what does it mean to the operating
,i;.?:
reliability of the plant if that equipment is ,
going to be, and the availability of that plant, -
J h.J., and that manifests itself in a. safety issue as
}q well.
- shD,
- n.2
- .9 So I. don't have a terminology of important to J."SE safety in that connotation, but I am trying to
^~' defin.e how we approach it, which we think does address that concept of whpt is important. Tr.
67-68.
Mr. McCaffrey of LILCO also acknowledged LILCO's failure to utilize the NRC safety terms, but he relied on other LILCO progre.ms to compensate for the failure to systematically iden-tify SS&C's important to safety:
You don't need to agree on the terminology. That's where we continue to have the problem. I think Mr. Novak said it's the care and fee [dling. You can achieve the same assurance, I hope, in your mind, from the examples and the thought process and the programs that are in place and the feed-backs and the updates and all of that should give
givo you the sense of -- wa don' t ignore that other-than-safety-related. We dor.' t have to call it important to safety. It has, obviously a certain importance, but I think we achieve the same effect by the programs we have. Tr. 68.
In our opinion, compliance with the regulatory require-ments for SSEC's important to safety cannot be demonstrated by
, ,d references to a philosophy, a thought process, training, or
~
narrow progrars and procedures. These, of course, are impor-
- tant but cannot substitute for documented safety classification measures. Thus, in our opinion, the regulations clearly require that the classification and resulting QA/QC measures
/. for SS&C's important to safety, but not safety-related, be set forth in an objective, documented, controlled mechanism.
i% .. "
b D. The Staff Reliance on the StanIard Review A-7 Plan and Reculatory Guides is Not Justified
'S:
The Staff asserts that LILCO has demonstrated compliance E$
with the substantive regulatory requirements for plant items "important to safety" during'the design and construction phase by virtue of the Applicant's alleged compliance with Regulatory Guide 1.70 and the SRP. (Staff Supplemental Testimony at 10, 12). For design and construction, Mr. Conran states that the SRP and Regulatory Guides can, perhaps, provide a safety net or backstop to mitigate serious misunderstandings in the meaning of the regulatory terms. (Affidavit at 32). We believe, how-ever, that the Staff and Mr. Conran, to the extent he agrees wi'th the Staff witnesses, are not in a position to conclude that LILCO's alleged compliance with these Staff guidance documents ensures proper treatment of SS&C's important to safety during the design and construction phase.
For example, the Staff suggests that it is satisfied that LILCO has applied proper quality standards to all SS&C's impor-tant to- saf ety. Bowever, the Staff's ability to reach such a conclusion is sharply undercut by the Staff's admission in its contention 7B , Findings that the Staff does not review or inspect the adequacy of LILCO's quali_y assurance activities for SS&Cs important to safety but not safety-related.
The Staf f does not review the quality assur ance.
program for items important to safety but not safety-related, nor does it inspect for compliance with such a program. (NRC Staff Proposed Opinion, Findir.gi of Fact, and Conclusions of#Law, Volume "
2, Finding 7B:82, p. 85.)
Similarly, we disagree with the Staff reliance on Regula-tory Guide 1.70 and the SRP for its conclusion that during require-d.e sign and construction IILCO h-as satisfactorily met ments for SS&C's important to safety but not safety-related.
The Staff does not know the degree to which LILCO has complied with this Staff guidance.
The Staff does not identify ~in the SER the extent to which 'Shoreham complies with all applicable or current Regulatory Guides and the SRP. The Shoreham SER also fails to document the justification for any instances of non-compliance with the preceding Staff licensing practices.
Thus, the Staff is simply not in a position to assert that n
LILQO has complied with its regulatory guida.ce.
~
The Staff has recognized that its review practices are in-adequate to provide detailed knowledge concerning an appli- .
Thus, in a cant's compliance with Staf f reg'ulatory guidance.
c h'. ,
June 13, 1980* letter to Mr. Dircks, Executive Director for i Operations, Mr. Denton, NRR Director, while stating his overall fj;}(g'l.; ,
- n. .. . .
":b*J.
belief that the public is adequately protected by the Staff's
. .p_s ._
( pg-(
safety review process (a conclusion which we do not support),
b , L: ; sets forth the following general problems in documenting the yphyg
- .L7.i extent of licensee compliance with the Staff requirements, 3.$. ,.j7 which cast serious doubt upon the Staff's reliance upon the
~.jf-P
- h. N completeness of its review process.for shoreham:
M.%.y; The first problem involves the f act that the
- .I's. .
63*A1 staff's current review procedures are not directed
.',E;;; toward 'providing a detailed and specific account-ing of compliance with each and eve *ry regulatory Rather,
- iP:.1 requirement and related regulatory guide.
vM.: , ; the radiological sr.fety review of operating f-?, . - -
license applications is based on the Standard OL;2,5 Review Plan (SRP) which incorporates by reference es applicable regulatory guides and Conformance all approvedwith in Branch Technical Positions ...
the Standard Review Plan and its references is
??.
generally believed to constitute compliance with applicable NRC regulations, although a systematic analysis to establish this congruence has not been conducted.
A second problem is that the staff's review is of the audit type; that is, not all plant features l are reviewed by the staff for conformance to the Standard Review Plan. Given the nature of an audit review, it is not possible for the staff to demonstrate in detail that an application is in complete compliance with all elements of either the Standard Review Plan or the applicable regula-tions.
I A third problem involves, the f act that there are some regulations for which there is no.
l -
l l i l- - - - - -. - - . - - - -. . - , , _ _ _ _ - - - - . _ _ , _ . - _ _ _ - - - . - . _ -
corresponding guidance to reviewers in the SRP.
One known example is that General Design Criterion 51, " Fracture Prevention of Containment Pressure Boundary," is not explicitly refe' fenced in the relevant SRP sections.
'. , , A fourth problem is that the staff Safety Evalua-
- tion Reports have always been written to summarize
~ f'. ' , the results of the audit reviews and were not in-
'? ." '
tended to document all aspects of the review. .
- These reports tend to highlight those areas in f,
which disagreements occurred between the staff and
> ~(,Y c the applicant and the way in which these areas ir' 1 were resolved. Therefore, it is not always
, yt, possible to find in these reports an accounting of the conformance of these applications to some of i' the NRC regulations or regulatory guidance that
- '- received most of the staff attention in these reviews. (emphasis added).
~
Fur ther, the Shoreham FSoR does not provide assurance that
^
the safety classification treatment for SS&C's important to
. . . 2, ggg;A: safety'is adequate. For example, the SR,P calls for application 3Md of GDC-1 to all important to safety items in Section 7.0 of the
.JSU
,M;;;f FSAE. 'SRP Rev. 2, July, 1981, S 7.0. LILCO's FSAR for si3h]
wc ,
' ' Shoreham does not mention GDC-1 as an applicable criteria but does list IEEE 33610/ and Regulatory Guide 1.3011/ for most items. Those items which are not shown as having either GDC-1, l
l IEEE 336 or Regulatory Guide 1.30 as criteria or guidance would FSAR be assumed to have.little, if any, safety significance.
- l. .. -
40/ IEEE 336 "Insti ilation, Inspection, Testing Requirements for Instrumentation and Electrical Equipment During the Construction of Nuclear Power Generating Stations," 1971.
l 41/ Regulatory Guide 1.30, " Quality Assurance Requirements for l
the Installation, Inspection, and Testing of Instrumen-i l
tation and Electric Equipment," 1972.
l l
Table 7.1.1.2 lists 38 major systems categories, four of which indicate the above standards and regulatory guide are not ap-plicable. Among the systems not receihing these considerations are the Remote Shutdown System and Recirculation Pump Trip, both of which clearly are important to safety. This omission is inconsistent with our view of the importance-to-safety and safety significance of these items. Until LILCO prepares a list'of SS&C's, identifies the safety significance of individual pieces of equipment, and defines the applicable codes, regulations, and QA/QC measures for each, there can be no demonstration that the Shorehan treatment, in fac.t, is con-sistent with the importance-to-safety of the item or with the SRP. . _,
Similarly, the SRP and Section 7.142/ of the Shoreham FSAR are also not consistent and there is no way to verify that-all important to safety equipment is included in the FSAR nor that the equipment included is considered in the manner. called for by the SRP. The review method proposed in SRP Section 7.1 (p.
7.1-14) for determining compliance with GDC-1 QA requirements However, Section 17 refers the reviewer to Section 17 of SRP.
of the SRP only lists criteria for the OA review of safety-related SS&C's. Likewise, Section 17 of the Shoreham
--42/
SRP Rev. 2, July, 1981, Section 7.1, and FSAR Section 7.1, including Figure 7.1.1-2.
l l
\ .
i
FSAR only refers to the QA/QC progrtm for safety-related itcms.
Therefore, it is not possible to draw t,he overall conclusion that the SRP and FSAR assure proper safety classifications for design and construction activities.f3/ Indeed, even the latest revision of the SRP does not insure proper QA treatment for im-portant to safety, but not safety-related, equipment since there is no description of what' equipment is included in this category (Appendix D of Section 3.2.2 of the July, 1981 version of the SRP is still in the course of preparation) . Likewise, neither Section 3.~2.2 nor Section 17 of the latest version of the SRP provides detailed criteria concerning how to. meet the requirements of GDC-1.
In our ' opinion, the Staff's general,. reliance on LILCO's alleged compliance with the SRP and Regulatory Guides is with-out subs;tantiation until the validity of the Staff's assump-tions is properly verified.
E. Correctness of Safety Significance Assumed Rather Than Demonstrated Recognizing the conflict in safety classification termi-nology between the Staff.and LILCO, the Staff at the conclusion 13/ The Staff states that LILCO has identified and properly treated SS&C's important to safety and that this has been documented in the SER. Staff Supplemental Testimony at
- 12. However, the Staf f's SER is totally silent regarding QA applied by LILCO for SS&C's important to safety but not safety-related.
of the February 18 meeting issued a letter to LILCO requ2 sting The FSAR additions pro-that the FSAR commitment be amended.
posed by the Staff call for measures to ensure that:
[E)quipment and associated software shall be accorded the safety'sionificance given to it in the FSAR, the technical specifications and the emergency operating procedures. (Letter Eisenhut to Pollock, February 18, 1983) (Emphasis added).
This "fix" thus assumes that the safety significance
~
accorded to equipment and software in the FSAR, TS's, and EOP's
~
is correct.11/' Given the sharp difference between LILCO and the Staff in interpretation of the key regulatory term impor-tant to safety, we do not agree that such an assumption can or ,
should be made. Rather, LILCO should be required to' demon-strate that equipment and software have been assigned the i
appropriate safety significance and the Iommensurate OA treat'-
ment. Without such a demonstration, the "fix" merely specifies
\
that deficiencies built into the LILCO safety assessment clas-sification and assignment .of OA levels to specific items will Thus, be perpetuated during the operating phase of the plant.
f The term " safety significance" has not been defined by the 11/ Staff either in its letters to LILCO or in the Supplemen-tal Testimony. It must be defined if LILCO is to attempt l to comply with the Staff "fix." In our view, determining I
safety significance should include analysis of factors
' such as the effects of the performance or failure of a component on a safety function, service or on a water),'an'd support function the (e.g. power distribution, BVAC, potential for misleading the operator, contributing to a distracting event, or causing a reduction in safety margins.
e since the future treatment of SS&C's during operation will be it based on the oricinal classification a,nd QA/QC requirements, is imperative that LILCO demonstrate that the initial design, 7,,]
-c fabrication, and construction activities, including classifica-33
.s tion and QA/0C requirements for all SS&C's important to safety, '
.,Eff 7
h 48 were systematically applied and implemented in accordance wit jf? .
2:s the Staff requirements.
'I?]' . < ,
To make the demonstration urged herein, LILCO should be b$
~,.i required to identify components which are important to safety
.f3j identi-5p and for each item'of equipment and associated software,
+UQ LILCO U(,ipgj fy the safety significance and QA' activities applied.
y;; should then assess whether the safety significance and QA 4.jjy .
activities for items identified in th,e first step are correct, b555
$ and if not, amend, as required, the various programs identified fjy@
EdNd in the Staff February 18 letter (i.e., corrective and preven-
~
- y tive maintenance program, OA program, etc.)
to reflect the proper safety significance and QA treatment.
In order to carry out the foregoing steps -- indeed, for the Staf f's proposed "fix" to have any substance -- LILCO must 2 ,' ,
Without such a
: compile a list of SS&C's important to safety.
list, it is impossible for LILCO to demonstrate that the proper safety significance and QA controls are being applied, because .
is to be applied will the precise items as to which the "fix" remain only a generalized concept, rather than a concrete set of items whose safety significance can be clearly assessed.
Mr. Conran also calls for a listing by LILCO of SS&C's in-portant to safety.
[A~) condition for (i.e., prerequisite'to) a license in this case should be development by LILCO of a listing of 'Important to Safety' struc-tures, systems and components for Shoreham . . . .
~
(Affidavit at 32).
We support Mr. Conran's position. It is our belief that such a listing of components is essent?al if LILCO is going to demon-strate that it does know how to classify and treat important to safety SS&C's.
Recent events on February 22 and 25, 1983 at the Salem nu-
. clear plant underscore the importance of properly assessing the .
safety significance of components and providing the commensurate level of QA and preventive maintenance. The scram breakers involved in the Salem events weie supposedly classi~
fled as safety-related but the safety significance of attached trip functions (undervoltage and shunt trip coil) was not prop-
- /
erly assessed. The result was that the attached trip functions were not identified as safety equipment and did not receive required maintenance. Their failure to operate properly was a direct result of a maintenace deficiency 45/ and prevented the
-operation of the automatic reactor trip.
~
I l
~
45/ Letter Rahe (W) to Denton (NRC), Mar. 1, 1983, Attached to BN 83-26, Mar. 3, 1983.
The Salem events are directly relevant to Mr. Conran's concern because they show why LILCO must systematically identi-L',
fy equipment important to safety and demonstrate that the con-h sideration given to such equipment (including such matters as
,d:
3
~s maintenance) is consistent with its safety significance. The 6.f:q,.- ,
{.?h, Salem failures demonstrate also that the Staff must be assured ifj that LILCO has implemented a sy'stematic quality program that k.' provides the n.ecessary care for the equipment accord.ing to its es safety significance.dl/ In our opinion, the only means to en-
, , (*
5 [,)j sure that the safety significance and treatment, including '
', 3 maintenance, of items important .to safety has been carefully cQ 2:M considered is to specifically identify--list--these items as a
'c10
..'yu
- ,;-;j j
first ne'ce'ssary step in rectifying LILCO,.'s classification n%d Accordingly, a retrospective Shoreham safety yl$3 difficulties.
g Ejjj classifipationandqualityprogramreviewshouldbepromptly This review should have, at a minimum, the follow-f initiated.
ing major elements:
Ni n
'3 16/ For example, the Salem plant had a Q list and a Master Equipment List but the scram breaker's shunt relay (which is important to safety) was not included on the safety-related list. Thus, for years the relay was given no maintenance. When it failed the first time (Februhry 22, 1983), it was given ordinary commercial treatment which proved to be inappropriate. The result was that the (SECY-83-98 breakers again failed on February 25, 1983.
and Attachments to BN 83-26). ,
- 1. Review Shoreham SS&C's and their associated software and assess their safety significance.
2.
Assess whether the safety siinificance accorded in guiding documents (such as the FSAR and QA manuals f, and procedures) is correct.
- 3. Prepare a list of SS&C's which are important to safe-i ty. The Shoreham list should be accomplished.at the equipment component level, rather than at the system level, due to the likelihood of divergent QA/QC
~ measures being' assigned to civil, mechanical, and .
electrical components within a particular. reactor system and since QA measures are generally assigned at"the equipment component level.47/
- 4. Identify the QA/QC program elements applicable to
'. each item and assess whether these elements are
~ commensurate with the identified safety significance.
- 5. Complete the requirements of the "fix" defined by the NRC Staff.
l 47/ A report' prepared for the NRC by EG&G, (EG&G-EA-6109)
" Identification and Ranking of Nuclear Plant Structures, Systems, and Components,November, and Graded Quality Assuranceis one such equip-1982, Guidelines -- Dr af t",
i ment listing and safety classification review effort of The EG&G' Report includes a partial which we are aware.
listing and ranking by QA level of SS&Cs important to safety. It also provides 23 factors or guidelines for ranking components within the systems, as to their QA level (EG&G, attached to BN 83-13, Feb. 17, 1983, at pp.
i 16-17).
~
- 6. Document the results of the above steps in the FSAR.
In our opinion, absent such' a review there can be no as-sur ance that Shoreham complies with GDC-1.
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NUREG/CR-1321 SAND 80-0384 '
AN 00 FINAL REPORT I'3.ASE I W
SYST._.S INTIRACTION METH0bOLOGY APPLICATIONS PROG Rv. ..
CarjJ.Boyd -
Vallis R. Cramond ,
Steven W. Eatch Jack V. Hickman Alan M. Kolaczkowski Dermond W. Stack --
Nuclear Puel Cycle Safety Researd'h Department I
Manuscript Sub=itted: De c e=be r 21, 1979 Date Published: April 1980 Sandia Naticnal Laboratories Albuquerque, Kev Mexzco $7165 operated by Sandia Corporation for tbe
, U.S. Department of Inergy Prepared for Office of Standards Develop ent and Office of Nuclear Reactor Regulation V. S. Nuclear Regulatory Co=:ission Washingten, DC 20555 NRC FIN No. A1113
TADLE 7.1 Generic Analysis Concluelons Commonality Comments Systeme _
Interactions of Host Significance Hot Covered in the SRP Ac t u a t io,n- -- Potential Single Event Pressurizer Power Operated Resulting in Dreach Relief Valves Fall Open Location of Boundary and Relief Isolation Valves Fail'Open Auxiliary Feedwater Fails Cooling - Potential Subtle Inter- :
actions Could Lead to Loss of Decay llcat Removal Auxiliary Feedwater Fails Location - Potential Subtle Inter- .i actione Could Lead to
- S
- Loss o f Decay lle~at Removal Hotive - Lo ss o f Ab solute Pres-Pressucizer lleatere Fail Power sure Control, Reduce's Hargin of Safety Location - Source of Water Neces-RWST Output Valves Fail .sary for Inventory Cl.oecd ,Hakeup Needed to Haintain DilR l
l 9J .
C, o
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFO?.E THE ATOMIC SA'FETY AND LICENSING BOAR .
)
)
In the Matter of )
)
1 LONG ISLAND LIGHTING COMPANY ) Docket No. 50-322 (0.L.)
3.
Q' ' (Shoreham Nuclear Power Station,
)
)
.a
"" Unit 1) _)
- ^
fT- _
CERTIFIC ATE OF SERVICE
(* .
I hereby certify that copies of Suffolk County Goldsmith, Supplemental Richard B.
4.} Testimony on Contention 7B by Marc W.
Hubbard and Gregory C. Minor were served upon 4
!! $),
i
" as otherwise indicated. - ,
4
. . Ralph Shapiro, Esq.
Lawrence J. Brenner, Esq. (*) -
- Cammer and Shapiro
~:! Administrative Judge 9 East 40th Street Atomic Safety and Licensing Board New York, New' York 10016
?s U.S. Nuc) tar Regulatory Commission
, (( Washington, D.C. 20555 Howard L. Blau, Esq'.
,-d 217 Newbridge Road f*) Hicksville, New York 113C
~~' Dr. James'L. Carpenter Administrat'ive Judge Esc Atomic Safety and Licensing Board W. Taylor Reveley III, U.S. Nuclear Regulatory 20555 Commission Hunton & Williams
- Washington, D.C. P.O. Box 1535 707 East Main St.
Richmond, Virginia 23212 E.
Mr. Peter A. Morris Administrativa Judge (*)
Atomic Safety and Licensing Board Mr. Jay Dunkleberger
~"'
U.S. Nuclear Regulatory 20555 Commission New York State Energy Off Washington, D.C. Agency Building 2 Empire State Plaza Edward M. Barrett, Esq. Albany, New Y6rk 12223 General Counsel Long Island Lighting Company i 250 Old Country Road Stephen B. Latham, Esq.
j '
. ineola, New York 11501 M Twomey, Latham & Shea P.O. Box 398 Mr. Brian McCaffrey 33 West Second Street Long Island Lighting Company Riverhead, New York 119 17 5 East Old Country Road Hicksville, New York 11801 1
l l
, .- ,-.,s . _ . _. . -
- . I Mr. Jeff Smith Marc W. Goldsmith Shoreham Nuclear Power Static I.ergy Research Group, Inc. P.O. Box 618 -
400-1 Totten Pcnd Road 02154 Nerth Country Road Writham, Massachusetts - _fWading River, New York 1179 Joel Blau, Esq.
MHB Technical Associates 1723 Hamilton Avenue New York Public Service Commission Suite K a The Governor Nelson A. Rockefeller San Jose, California 95125 Building 4'f' Empire State Plaza
,' Hon. Peter Cohalan Albany, New York 12223 Suf folk County Executive
- Cou6ty Executive /Legislativt David J. Gilmartin, Esq. ,
Building Suffolk County Attorney Veterans Memorial Highway
'*l County Executive / Legislative Bldg. 11788 g Hauppauge, New York
" Veterans Memorial Highway 11788 Hauppauge, New York Ezra I. Bialik, Esq..
- Assistant Attorney General
"' Atomic Safety and Licensing '
Environmental Protection Bu S. Board Panel New York State Department o U.S. Nuclear Regulatory Commission Law a'@
Washington, D.C. 20555 2 World Trade Center 2, New York, New York 10047 2f4 Docketing and Service Section office of the Secretary #Atomic Safety and Licensing C;Z U.S. Nuclear Regulatory Commission 7aj , Appeal Board ali Washington, D.C. 20555 O.S. Nuclear Regulatory e
r' Commission _
Bernard M. Bordenick, Esq. (*) .
Washington, D.C. 20555 c.2 - David A. Repka, Esq.
U.S. Nuclear Regulatory Commission Matthew J. Kelly, Esq.
Washington, D.C. 20555 Staff Counsel, New York State Public Service Comt Stuart Diamond 3 Rockefeller Plaza Environment / Energy Writer Albany, New York 1222,3 NEWSDAY 11747 Long Island, New York Stewart M.. Glass, Esq.
(*) Regional Counsel Daniel F. Brown, Esq. Federal Emergency Manageme Atomic Safety and Agency Licensing Board Panel 26 Federal Plaza l
U.S. Nuclear Regulatory Commission New York, New York 10278 Washington, D.C. 20055 James B. Dougherty, Esquire '
3045 Porter Street, N.W. Karla,J. Letsehg Washington, D. C. 20008 KIRKPATRICK, w _ CKHART, HII CHRISTOPHER & PHILLIPS 1900 M Street, N.W., 8th I DATE: March 23, 1983 Washington, D.C. 20036
(*) Hand Delivered
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