ML20236T338
| ML20236T338 | |
| Person / Time | |
|---|---|
| Issue date: | 11/20/1987 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR ACRS-2532, NUDOCS 8712010154 | |
| Download: ML20236T338 (11) | |
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..a DATE ISSUED:Nov.20, 1987 PROPOSED MEETING
SUMMARY
.FOR THE INSTRUMENTATION AND CONTROL SYSTEMS SUBCOMMITTEE MEETING OCTOBER 29, 1987 WASHINGTON, D. C.
Purpose:
The Subcommittee on Instrumentation and Control Systems met on October 29, 1987, in Washington, D.C. to review the NRC Staff's proposed resolu-tion of USI A-47, " Safety Implications of Control Systems in LWR Nuclear l
Power Plants."
In addition, the Subcommittee discussed and considered 1
Mr. Basdekas's comments regarding the resolution of this USI. The Subconsnittee meeting was fully open to public attendance, i
Attendees ACRS NRC Staff J. Ebersole, Chairman N. Anderson C. Michelson, Member R. Baer G. Reed, Member D. Basdekas C. Wylie, Member A. Szukiewicz E. Patterson, invited Consultant M. El-Zeftawy, Staff Others S. Bruske INEL W. McCaughey, BG&E C. English, INEL R. Borsum, B&W B. Collins, INEL K. Arn, Serch Lic.
Y. Kim, NUS L. Connor, DSA P. Believeau. NUMARC M. Patterson, Self i
Meeting Highlights, Agreements, and Requests 1
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Mr. J. Ebersole, Subcommittee Chairman, introduced the other ACRS members present and stated the purpose of the meeting. Mr.
Ebersole gave a brief summary of the USI A-47 and highlighted some p
of the NRC Staff's resolutions to this USI. He indicated that-NUREG-1217, " Evaluation of Safety Implications of Control Systems in LWR Nuclear Power Plants", which was prepared by the NRC Staff to document all the technical findings related to this USI, implies-(at least from its title) much more broader scope than it actually is, and the scope of the study is extremely narrow.
II. Mr. A'.
Szukiewicz, Task Manager / Engineering Issues Branch, RES, presented the overview program for USI A-47.
He stated that instrumentation and control systems utilized by nuclear plants are composed of safety grade protection systems and non-safety grade control' systems.
Safety grade systems are used to:
(1) trip the reactorwhenspecifiedparametersexceedallowablelimits;and(2) protect the core from overheating by initiating ECCS systems.
Non-safety grade control systems are used to maintain the plant within prescribed parameters during shutdown, startup and normal load varying power operation. Non-safety grade systems are not relied on to perform any safety functions during or following postulated accidents, but are used to control plant processes.
Although non-safety grade control system failures are not likely to result in accidents or transients that could lead to serious events or result in conditions that safety systems are not able to cope with, in-depth studies have been identified in which a failure or malfunction of the non-safety grade control systems can:
(1)
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potentially cause steam generator or reactor vessel ovarfill; or (2) can lead to a transient that could cause severe vessel over-m C
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-cooling.
In addition, there is the potential for control system failures to result,in plant conditions which may result in unac-1 ceptable M sk.
o Mrf Stukiewicz indicated that the purpose of USi'I-A7 is to perfonn
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a tvstenatic evaluation of non-safety grade control systems that are typically used during nordal ple.nt operations and to identify contrni systems whose failure could:'
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(1) cause transients or accidents identified in the FSAR analysis to be potentially more severe'than previously analyzed; l
. (?)'bdversely affect any assumed or anticipated operator ac' tion during the course of an event; s
f (3) cause technical specification safety limits to be exceeded;'
f or (4) cause transients or accidents to occur at a frequency in
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excess of those established for abnormal operational transients and design basis accidents.
.c Specific sub-tasks' of this issue were to study steam generator / reactor vessel overfill and overcool transients to dei. ermine the need for prever;tive and/or mitigating design measures. The objective of this USI
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is to evaluate the need for possible control systems changes in v
M. operating reactors and to verify the adequacy of licensing requirements t
,j' lond to propose, if needed; additional criteria and guidelines.
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The NRC. Staff has completed its proposed resolution of USI A-47. The safety. issue is th [, there may be failures initiated or aggravated by
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'non-Safety grade control systems that could lead to plant events that impact the health and safety of the public. An evaluation, within 7,.
prescribed linits, was performed on non-safety grade control systems m
that are typically used during normal plant operation.
Four nuclear f
- g steam system plants we.re evaluated; GE designed BWR, a 3-loop W_ PWR, a e q 3v' N M) once-through 'sMn generator PWR designed by BAW, and a CE designed PWR.
A study was also conductrA'to determine the generic applicability of the
[resultsofhespecificplantsanalyzedtoeachclassofplants.
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The' limitation and assumpiNonn t'nat were used in the study of USI A-47 i
are as follows:
(1) A minimum number of safety-grade protection systems would be L
available <f.oArip the reactor and initiate overpressure protection
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systems or e e rgency/ core cooling (ECC) systems, if needed, during jyl f
transients'in'itiated 5y failures in the control systems.
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(2) Control system failures resulting from common cause events such as earther. takes, floods, fires, and sabotage, or operator errors of 1
omission or cdmission are not addressed in this review. A study I-i of selected multiple control system failures in non-safety grade i-WN__-._--_.-_
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(3) Transients resulting from control system failures during limited conditient of operation (LCOs) or a'.tlicipated transient without e
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if (4) Theplant-specificde's'!gnswereassumedto13avebeenappropriately A
modified to comply with the reqisirements of IE Bulletin 79-27 and H,
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' concluded the following:
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j Results of reference plant analysis can be generally applied i
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Control system design of plants by the same (NSSS) suppliers
'I tre functionally similar,
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,I t Transients resulting from the failure of the same non-safety i '. '
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grade control syst9m on df ffere.nt plants of the same NSSS 5-
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- swp$ lier wil'/Mdiice similar or bounding transients,
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.4 Iap'r'deaint made after the TM1-2 event for the auxiliary
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- feeddter system and idr operator information and training
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?peitly aid in recovery of complex transients, di t
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Plant transients resulting from control system failures can be adequately mitigated by the operators provided that failures do not compromise operation of the minimum number of pro-tection system channels.
(Exceptions noted recomend proce-dure and operator training at CE plants),
Transient or accidents resulting from control system failures are less severe than, and bounded by the transients and accidents analyzed in the FSAR. However, overfeed events were not analyzed in the FSAR.
PWR plant designs having redundant comercial grade (or better) overfill protection systems (from main feedwater overfeed events) that satisfy the single failure criterion were determined to be adequate, BWR plant designs with comercial grade (or better) overfill protection systems (from main feedwater overfeed events) were determined to be adequate.
As part of the resolution of USI A-47, the Staff is proposing four actions for all licensees and applicants of LWRs. These actions are:
1.
A Generic Letter requiring all plants to provide automatic steam generator or reactor vessel overfill protection - most plaints already comply.
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Require all plants to periodically verify operability of overfill l
protection. Most GE and W plants comply.
3.
Require Oconee plants to provide automatic initiation of EFW on low L
steam generator level - all other B&W plants provide this feature.
4 Require all CE plants (with low head safety injection pumps) to reasr.ess emergency procedures and operator training and modify (if necessary) to assure plant shutdown for any SB LOCA.
The selection of the proposed resolution is based on consideration of
'the safety benefits derived from those actions in terms of risk reduction and the cost of implementation.
It has been determined that proposed actions are backfits.
It should be noted that as a result of the Rancho Seco event (December 26,1985) which involved a loss of power to the Integrated Control System (ICS), a reevaluation of all B&W plants is currently underway.
l The NRC Staff is planning to submit the proposed resolution of this USI to the CRGR for review in November 1987, and issuance for public com-ments by March 1988. The schedule to resolve the public comments and prepare the final resolution is approximately September, 1988.
1 III. Mr. Demetrios Basdekss, Reactor and Plant Safety Issue Branch /RES, expressed some concern regarding the resolution of USI A-47 and the special significance impacting the B&W plant safety reassessments.
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Control Syst:ms Mtg. 0ct. 29,1987 l-3 Some of Mr. Basdekas's concerns are related to the impact of instrumentation and control systems on various aspects of B&W plant performance (e.g., the effects of external events, operator errors of omission and comission, comon mode - comon cause and cascade failures, steam generator overfill leading to a main steam line waterhammer and break, steam generator tube rupture, related BOP, and systems interactions).
Mr. Basdekas also expressed some concern regarding the completeness and adequacy of implementation of all the requirements of IE Bulletin 79-27 and NUREG-0737 issued in the wake of TMI-2 accident for all B&W plants. He also indicated that the question of recriticality during overcooling condition has not been evaluated.
Mr. Basdekas comented that the ACRS should get more aggressive in its review of this USI and be more forceful in its letters to the Comission, instead of writing the " pussy-cat" kind of letters.
IV. As a result of the Subcommittee's discussion, the Subcommittee members raised some concerns regarding the following:
Mr. Michelson expressed some concern regarding the set of limitations and assumptions that were developed to confine the USI A-47 study.
For instance, the internal floods, inadver-tent actuation of control systems, wetting down due to fire, harsh environmental conditions, etc. all are not considered in the scope of the study and review.
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Mr. Ebersole expressed some concern regardino the question of steam generator cverfill leading to a main steam line break.
Also, steam generator tube rupture with more than one steam generator blowing down was not assessed.
Mr. Ebersole expressed some concern regarding the lack of protection ageinst temperature rise and its effect on control-systems. Also, the lack of protection against excessive initiating failure frequency.
Mr. Michelson expressed some concern regarding the safety implications of the design of control systems and related BOP systems and the need to coordinate its resolution with USI A-17 " System interactions."
Mr. Michelson questioned the validity of the study and review of USI A 47 and commented that the NRC Staff did not perforn any investigation related to an event such as fire in the control room, to determine the consequences and implications of such an event.
Mr. Michelson commented that the Staff should state in their reports (NUREG-1217 and NUREG-1218) the limitations regarding theanalysisofreturn-to-powertransient(e.g.,thespecifics of thermal-hydraulic model).
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s Mr. Ebersole commented that as a result of the Rancho Seco and l
Davis Besse events, it appears that " transfer-of-faults" was permitted. Mr. Ebersole requested additional infomation and more clarification from the Staff regarding their inves-a tigation.
Mr. Wylie questioned the assumption that all requirements of IE Bulletin 79-27 and NUREG-0737 were adequately implemented and requested more clarification. The NRC Staff responded by stating that 97% of the cases were adequately implemented and agreed to supply more clarification.
Mr. Reed expressed some concern regarding the question of recriticality during overcooling conditions. The Staff agreed to supply more information.
Mr. Michelson questioned the justification and validity of the j
l staff's requirement to upgrade the existing main feedwater
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overfill protection systems for different PWR designs (e.g.: W and CE).
1 Mr. Michelson commented that control systems failures result-ing from seismic events should be considered in the scope of the study.
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Future Activities I
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l l t The Subcommittee Chairman of'the Instrumentation and Control Systems will brief the full Committee in November 6, 1987 regarding the Subcom-mittee activities related to USI A-47.
Also, the NRC-Staff will brief the Committee on the same subject. A letter commenting on the rate of progress is anticipated.
NOTE:
Additional meeting details can be obtained from a transcript of this meeting available in the NRC Public Document Room, 1717 H Street, NW., Washington, D.C. or can be purchased from Heritage Reporting Corporation, 1220 L Street, NW, Washington, D.C. 20005 (202) 628-4888.
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