ML19250B739
| ML19250B739 | |
| Person / Time | |
|---|---|
| Issue date: | 10/22/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Ahearne J NRC COMMISSION (OCM) |
| References | |
| NUDOCS 7911050048 | |
| Download: ML19250B739 (8) | |
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fiD10RA'iDUM FOR: Comissioner John F. Aheame Lee V. Go:M:k THRU:
Lee V. Gossick, Executive Director for Operations
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Harold R. Denton, Director, Office of riuclear Reactor Regulation
SUBJECT:
SAFETY I!1PLICATIO"S OF CONTP,0L SYSTEI15 AND PLA'IT DYRN1ICS In_tmduc_ti_on and_Sumary, By nemorandum to you dated September 4,1979,11r. Denetrios Basdekas identified a nunber of concerns related to control system design and plant dynanics. This memorandum addresses those concerns and discusses related work that lRR has either planned or is undenray.
fir. Basdekas maintains that, because design criteria are inadequate and there is no detailed staff review of plant control systens, it cannot be concluded that the staff safety reviews are adequate to ensure that plant designs are acceptable.
In addition, he contends that control system nalfunctions should be considered as initiators of anticipated operational occurrences
- or postulated accidents.
Further, these nalfunctions, together with the effects of other nomally functioning control systems, should be considered during and subsequent to A00s or accidents.
In assessing the inpacts of these mlfunctions on the conser,utnces of both transients and accidents, ifr. Basdekas believes that the analytic modeling must accurately describe the various dynamic processes. Uithout such as assessment, he concludes that there may be sequences of events not nou considered in the safety analyses for which inadequate nitigating featurcs Ive been provided.
He cites TMI-2 as an example.
Mr. Basdekas nakes a number of recomendations for addrassing the concerns he has raised. These include:
1.
Failure Mode and Effects Analyses (FI'.EA) of control systems for each plant;
- 2. -Establisinent of design criteria for control systeas; 3.
Establidaent of requirenents for ccntrol system design and installation; 4.
Revision of the Standard Review Plan (SRP) to include the detailed review of control systems; 5.
Training and/or hiring of suitably trained staff to perfom the control sy3 tem revices; and, 6.
Dereting of operating plents until a. preliminary review of control systens has been conpleted for each plant.
FAETHpated operational occurrences (A00s) are those events which are expected to occur at least once during the life of the plant.
3, f, 887 172 7911050 ]0 g
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i Commissioner John F. Ahearne..
In the discussion which follows, we describe the review process presently used to judge the 6dequacy, from a safety standpoint, of plant protection systems, our treatment of control systems in that process and efforts that are planned or underway to provide added assurance that this process is adequate or identify changes necessary to satisfy Commission safety requirements. As this discussion indicates, we share some of the same concerns that Mr. Basdekas raises and we believe that the work we have initiated addressed those concerns. We agree with the need to investigate control system failures and design inadequacies. How-ever, we do not assign the same importance to the review of plant dynamic and control system performance, including stability, as does Mr. Basdekas. We do plan to investigate the possibility of simulating the dynamics of control systems in a representative B&W plant but we do not believe there is sufficient justifi-cation for an imediate detailed review of control system dynamics at all operating plants.
Finally, while we agree with the need to investigate the effects of control system failures and design inadequacies, we do mt believe there is sufficient evidence to suggest that conclusions drawn from safety analyses are not valid. Therefore, we do not believe there is adequate justification for the recontiendation to reduce power at operating plants pending a preliminary review of control systems.
Discussion As Mr. Basdekas notes in his memorandum, the staff has not reviewed control systems in detail. The staff requires that all applicants for an operating license demonstrate by analysis that the plant is designed to mitigate the effects of a defined set of anticipated operational occurrences and postulated accidents.
In assessing the effects of anticipated events, it is assumed that the events can be initiated by single control system malfunctions. These mal-functions are non-mechanistic in that no cause for the malfunction is identified nor are other associated malfunctions consideced.
For example, the loss of all main feedwater is considered an anticipated event, but, in analyzing this event, it has not been necessary to identify, for example, that a power supply failure caused the loss of feedwater and the coincident malfunction of other equipment powered by that same supply. The staff followed this approach, reasoning that the event would not be substantially changed because of the specific component which was assumed to have failed. This simplified the staff review since it would not be required to identify all single failures which could cause the event regardless of the probability of its failure. Further, the analysis assumed that all control systems respond as designed (unless the equipment mal-function is associated with a particular control system). All plant neutronic and thennohydraulic parameters are assumed to be at their worst-case values at the time the event is initiated.
Similarly, in analyzing postulated accidents, plant control systems are assumed
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to respond normally except that no credit is taken for such a response that would be of bentfit in mitigating the effects of the accident.
It has been assumed that the consequences of design basis accidents (e.g., LOCA, steamline break) would not 887 173 i
Comissioner John K. Ahearne be significantly Miected by control system mg1 functions because of the rapid change in plant parameters during such accidents.
l We believe that the review appmach followed by the staff has been an effective use of resources for evaluating the adequacy of plant designs. The analytical demonstration that the plant safety systems can successfully mitigate the effects of the defined set of anticipated operational occurrences and postulated accidents provided the staff with adequate basis to conclude that the designs of these protection systems were adequate and that the consequences of these design basis accidents would not be significantly affected by malfunctions in plant contml systemss The staff has recognized that there am drawbacks in the approach discussed above in that the events considered in the analysis do not bound all events which can be postulated. For example, recently in a letter from Westinghouse Electric Corporatior to one of their operating plant costomers (Attachment 1), a number of control systens could potentially malfunction if impacted by adverse environments due to a high energy line break inside or outside containment. Westinghouse indicated that the effects of such failures could lead to high energy ifne break consequences more severe than those presented in the safety analysis reports. The staff responded by issuing a letter to all operating light water reactors (Attachment 2) requesting that each licensee review their plant design in light of this concern and respond within (20) days with regard to whether operation of their plant should be modified, suspended, or revoked.
It is expected that evaluations will be perfomed to evaluate the consequences of these and other rotential control system failures which can be postulated to ensure that while this safety concern ray exist, the overall conclusions regarding the adequacy of plant protection features and operator actions necessary to mitigate these events are adequate to creet all saTety criteria necessary to permit continued plant operation.
The staff has raised questions mgarding the acceptability of multiple challenges to the reactor protection system due to problems related to control system actions at several B&W plants (Attachment 3). The Crystal River events mentioned by Mr. Basdekas are discussed in Attachnent 1 The events were either initiated by equipment malfunction or operator induced. While none of these events led to significant consequences, the frequency with which these events have occurred has highlighted the need to give greater regulatory attention to the control systems involved.
In a very related way the " Lessons' Learned' Task Force Status Report and Short-Tem Reconnendations, NUREG-0578" required in Section 2.1.9 that analysis of design 4
and off-nomal transients and accidents scenarios be perfomed including operator i
actions not previously analyzed. This position requires that, in addition to the nomal single failum assurnption, consequential failures shall also be considered.
The staff also required the perator errors that could cause the complete loss of safety function shall also be considered. Thus it is expected that through these efforts a variety of event trees will be investigated for their probability 887 1/4
a Comissioner John F. Ahearne e r.
of occurrence as well as possible consequences.
In response to this requirement the B&W Owner's Group (THI Effects Subcomittee) has' discussed with the staff a program they intend to follow to be responsive to this requirement.
- Brbfly, the program has the following objectives:
Investigate a wide range of reactor plant transients, including failures not normally considered in Safety Analysis Reports.
Provide appropriate information sto the plant operators to enable them to deal effectively with abnormal transients.
Promote a better understanding of system fundamentals and abnormal transient operation.
The B&W owners have stated that the engineering support to accomplish these ob.iectives are estimated at 30,000 man-hours, independent of the efforts that will be provided at each licensee plant. The staff is currently reviewing tne program to better understand how responsive this program is to the requirement stated in NUREG-0578 and the time necessary to implement the program.
Recognizing the importance of control systems and the role those systems can play in both the initiation and mitigation of off-normal events, the staff has a number of other initiatives either in the planning stage or presently underway to enhance our knowledge of these systems. These initiatives are aimed at impmving our understanding of possible control system failure mechanisms and their frequency of occurrence, and establishing the effects of these failures.
As a followup to the TMI-2 events, the Comission issued orders to the B&W operating plants. As part of these orders, B&W was required to submit to the NRC staff a failure modes and effects analysis of the Integrated Control System. This analysis has been completed and the results are included in a B&W report entitled "Integratei Control System Reliability Analysis," BAW-1564, August 1979.
The report includes a number of recomendations by B&W regarding improvements in the performance of the ICS and related systems. The staff is presently reviewing this report with the assistance of Oak Ridge National Laboratory. Recomendations regarding possible system improvements will be developed and future work will be defined.
As part of this effort, ORNL is investigating the possibility of producincl a computer simulation of a representative B&W plant which would include plant control systems. Such a simulation, if it proves feasible, would allow us to evaluate a variety of different kinds of control system failures including the effects of plant dynamics.
The staff has for some time recognized the need for criteria for equipment and systems important to safe plant operation but which need not be designed in compliance with safety system requirements.
In 1977, 887 175
ccamissioner.lohn F. Aheame,
the Office of Standards Development was requested to begin the development of such criteria but no work was done because of unavailability of manpower in both OSD and NRR. We have recently held discussions with OSD regarding the need to begin the development of.these criteria and they agree with the need to proceed. Further work is being delayed until the Lessons Learned Task Force decides on the scope of equipnent to be cove ad by the criteria.
Prior to the TMI-2 event, the staff had began to investigate the interaction of the various plant systems. This activity, defined in Task Action Plan TAP-A17 " Systems Interaction in Nuclear Power Plants,' involves the application of fault tree methodology as a means of systematically reviewing plant systems for susceptibility to systems interactions. Particular emphasis is being placed on the presuned redundancy and independence of safety systems. As Mr. Basdekas notes in his memorandum, this analysis does not treat the dynamic aspects of control-protection system interactions.
We believe that this detailed analysis of control system malfunctions is unnecessary at this time.
Westinghouse also has a study underway that is closely related to A-17.
As a part of our review of the Westinghouse Integrat,:1 Protection System (IPS), we requested that an analysis be made of possille interactions between the IPS and the plant contml systems and/or the engineered safety features (seeHUREG-0493). The objective of this analysis is to assess the degree to which these interconnected systems are suscept.ible to common mode failure. The methodology which is currently being developed by Westinghouse for this purpose makes use of fault tree analysis. The Westinghouse study will not only give us additional insight into the interaction of complex control and protection systems, but it should also provide us with additional guidance on methodology for assessing the impact of control systbem failures for other plant designs.
Finally, we are planning to devote more manpower to the analysis of operating experience. Events have occurmd in the past which have mceived in-sufficient eeview effort. Such events can indicate the existence of control system problems and possible problems associated with operator ermrs eThisin knowledge should be fed back into the review process. It will also be useful input to a technical assistance effort to be initiated shortly on contml room design improvements.
We believe each of these initiatives will add to our understanding of the importance of control system malfunctions and operator action and help us confirm the adequacy of our current review process. Our approach emphasizes only those concerns that we believe deserve innediate attention, thereby ensuring that limited staff msources are used wisely. We have not concladed that these concems are of sufficient 887 176 h
Comissioner John F. Aheame -
significance to warrant either the plant-by-plant control system analysis or the temporary reduction in power that Mr. Basdekas suggests would be prudent.
I hope this memo has been responsive to thle concem highlighted by Mr. Basdekas.
If you have any questions, I will be glad to discuss them with you at your con-venience.
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Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
As stated cc: Chaiman Hendrie Comissioner Gilinsky Comissioner Bradford 88/
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Comissioner Kennedy OGC OPE SECY
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,L ATTACHMENT 1 REPRINT Westinghouse Electric Corporation Water Reactor Division Nuclear Service Division Box 2728 Pittsburgh, Pennsylvania 15230 August 30, 1979 PSE-79-21 Mk. F. P. Librizzi, General Manager Electric Production Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101
Dear Mr. Librizzi:
Public Service Electric and Gas Co.
Salem Unit No. l OUALIFICATION OF CONTROL SYSTEMS As part of a continuing review of the environmental qualifications of Westinghouse supplied NSSS equipment, Westinghouse ha,s'also found it
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necessary to consider the interaction with non-safety grade systems.
This investigation has been conducted to determine if the perfomance of non-safety grade systems which may not be protected from an adverse environment could impact the protective functions performed by NSSS safety grade equipment. The NSSS control and protection systems were included in this review to assess the adequacy of the present environ-mental qualification requirements.
As a result of this review, several systems were identified which, if subjected to an adverse environment, could potentially lead to control system operation which may impact protective functions. These systems are:
Steam generator power operated relief valve control system Pressurizer power operated relief valve control system Main feedv:ater control system Automatic rod control system 887 178
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P, Page 2 PSE-79-21 Each of the above mentioned systems could potentially malfunction if impacted by adverse environments due to a high energy line break inside or outside containment.
In each case, a limited set of breaks, coupled with possible consequential control malfunction in an adverse direction, of the above events could yield results which are more limiting than those presented in the plant Safety Analysis Reports.
In all cases, however, the severity of the results can be limited by operator actions together with operating characteristics of the safety systems.
We believe these systems identified do not consti:ute a substantial safety hazard.
However, Westinghouse recommends you re. iew them to determine if any unreviewed safety questions or significant deficiencies exist in your pl ant (s).
To assist you ia understanding these concerns, Westinghouse will hold a seminar in Pittsaurgh on Thursday, September 6 at Westinghouse R&0 Center, Building 701, with all. our operating plant customers. The seminar will address the potential impact of these concerns for various plant designs and various licensing bases.
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Please contact your WNSD Regional Service office to confirm your attendance at the seminar. We will provide additional details concerning the agenda and other meeting arrangements as they become available.
Very truly yours, ORIGINAL SIGNED BY
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F. Noon, Manager Eastern Regional & WNI Support SR4/CC13&l4 cc:
H. J. Midura H. J. Heller R. D. Rippe T. N. Tay1or R. A. Uderitz bO[ }[9 C. F. Barcl ay W_
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ATTACHMENT 2 ga are%
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September 17, 1979
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ALL OPERATING LIGHT WATER REACTORS
SUBJECT:
POTENTIAL UNREVIEWED SAFETY QUESTION ON INTERACTION BETWEEN NON-SAFETY GRADE SYSTEMS AND SAFETY GRADE SYSTEMS You were infomed by IE Infomation Notice 79-22, issued September 14, 1979, that certain non-safety grade equipment, if subjected to an adverse environ-ment, could impact the safety analyses and the adecuacy of the protective functions perfomed by safety grade equipment. Enclosed is a copy of IE Infomation Notice 79-22, and reprinted copies of an August 30, 1979 Westinghouse letter and a September 10, 1979 Public Service Electric and Gas Company letter which address this matter.
The NRC is concerned that a similar potential may exist at other operating light water reactor facilities, including yours, for an unreviewed safety matter related to the effects of the environment on control systems resulting from high energy line breaks inside or outside containment and the results of these effects on the required safety systems.
In accordance with 10 CFR 50.54(f) you are requested to provide within twenty (20) days of the date of this letter, written statements, signed under oath or affirmation, which will enable the staff :o determine, in light of the concerns discussed above, whether or not ynur NRC license (s) to operate your nuclear power generating facility (fes) should be modified, suspended, or revoked.
Sincerely,
- W/Wk Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
1.
IE Infomation Notice 79-22 2.
Westinghouse's August 30, 1979 Letter
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3.
Du ice Electric and Gas T
Ocapany's September 10, 1979 Letter (Reprint) 9 v{
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.t ENCLOSURE 1 s
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AHO ENF0F. CEMENT WASHINGTON, O.C.
2055'5 September 14, 1979 IE Information Notice No. 79-22 QUALIFICATION OF CONTROL SYSTEMS Public Service Electric ar j Gas Company notified the NRC of a potential unrevis,,e:
safety question' at their Salem Unit 1 facility.
This notification was based on a continuing review by Westinghouse of the environmental qualifications of equipaer-that they supply for nuclear steam supply systems.
Based on the present status of this effort, Westinghouse has informed their customars that the performance of non-safety grade equipment subjected to an adverse' environment could impact the protective functions performed by safety grade equipmant.
These non-safetf grade systems include:
Steam generator power operated relief valve control system Pressurizer power operated relief valve control system Main feedwater control system
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Automatic rod control system These systems could potentially malfunction due to a high energy line break 4 side or outside of containment.
NRC is also concerned that the adverse environment could also give erroneous information to the plant operators.
destinghouse states that the consequences of such an event could possibly be more limiting than.results presented in Safety Analysis Reports, however, Wstinghouse also states that the severity of the results can be limited by operator actions together with operating characterisitics of the safety systems.
Fur,ther, Westinghouse has recommended to their customers that they review their. systems to determine whether any unreviewed cafety questions exist.
This Information Notice is provided as an early notification of a possibly significant matter.
It is expected that recipients will review the information for pcssible applicability to their facilities, No specific action or response is requested at this. time.
If NRC evaluations so indicate, further licensee
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a:tions may be requested or required.
If you have questiens regarding this matts-f ease contact the Director of the appropriate NRC Regional Office.
N: -ritten response to this Information Notice is re
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ENCLOSURE 2 REPRINT Westinghouse Electric Corporation Water Reactor Division Nuclear Service Division Box 2728 Pittsburgh, Pennsylvania 15230 August 30, 1979 PSE-79-21 Mr. F. P. Librizzi, General Manager Electric Production Public Service Electric and Gas Company 80 Park Place Newark, New Jersey 07101
Dear Mr. Librizzi:
Public Service Electric and Gas Co.
Salem Unit No.1 OUALIFICATION OF CONTROL SYSTEMS As part of a continuing review of the environmental qualifications of Westinghouse supplied NSSS equipment, Westinghouse h4s'also found it necessary to consider the interaction with non-safety grade systems.
This investigation has been conducted to determine if the performance of non-safety grade systems which may not be protected from an adverse environment could impact the protective functions performed by NSSS safety grade equipment. The NSSS control and protection systems were included in this review to assess the adequacy of the present environ-mental qualification requirements.
As a result of this review, several systems were identified which, if subjected to an adverse environment, could potentially lead to control system operation which may impact protective functions. These systems are:
Steam generator power operated relief valve control system Pressurizer power operated relief valve control system Main feedwater control system Automatic red control system 887 i82
Page 2 PSE-79-21 Each of the above mentioned systems could potentially malfunction if impacted by adverse environments due to a high energy line break inside or outside containment.
In each case, a limited set of breaks, coupled with possible consequential control malfunction in an adverse direction, of the above events could yield results which are more limiting than those presented in the plant Safety Analysis Reports.
In all cases, however, the severity of the results can be limited by operator actions together with operating characteristics of the safety systems.
We believe these systems identified do not consti.ute a substantial safety hazard.
However, Westinghouse recommends you review them to determine if any unreviewed safety questions or significant deficiencies exist in your pl ant ( s).
To assist you in understanding these concerns, Westinghouse will hold a seminar in Pittsburgh on Thursday, September 6 at Westinghouse R&0 Center, Building 701, with all our operating plant customers. The seminar will address the' potential impact of these concerns for various plant designs and various licensing bases.
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Please contact your WNSD Regional Service office to confirm your attendance at the seminar. We Iiill provide. additional details concerning the agenda and other meeting arrangements as they become available.
Very truly yours, ORIGINAL SIGNED BY
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F. Noon, Manager Eas+ern 'cgional & WNI Support SR4/CCl3&l4
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H. J. Midura H. J. Heller Y
R. D. Rippe T. N. Taylor R. A. Uderitz C. F. Barcl ay W 9
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ENCLOSURE 3 s
REPRINT PUBLIC SERVICE ELECTRIC AND GAS COMPANY Salem Nuclear Generating Station P. O. Box $6 Hancocks Bridge, New Jersey 08038 Sept, ember 10, 1979 Mr. Boyce H. Grier Director of USNRC Office of Inspection and Enforcement Region I 631 Park Avenue King of Prussia, Pennsylvania 19406
Dear Sir:
REPORTABLE OCCURRENCE 79-58/0lP SALEM NO. 1 UNIT LER This letter will serve to confirm our telephone report to Mr. Gary Schneider of the Regional NRC office on Friday, September 6,1979, advising of a potential reportable occurrence in accordance with Technical Specification 6.9.1.8.
We have been notified by our Engineering Department that a Westing-house conducted review of the environmental qualifications of Westinghouse supplied NSSS equipment has identified that conditions associated with high energy line breaks inside or outside containment and their impact on non-safety control systems may constitute an unreviewed safety question.
The control systems concerned are steam generator power operated relief valve control, pressuri::er power operated relief valve control, main feedwater control and automatic rod control systems.
A detailed report will be submitted in the time period specified by the Technical Specifications.
Very truly yours, Original Signed By H. J. Midura Manager - Salem Generating Station AWK:jds CC:
General Manager - Electric Production Manager - Quality As'surance 39II05007 6
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y ATTACHMENT 3
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September 13, 1979 Docket Nos.: 50-269, 50-270, 50-287 50-289, 50-302, 50-312 50-313, 50-346 FACILITIES:
Oconee Nuclear Station, Unit Nos. 1, 2, and 3 (0conee)
Three Mile Island Nuclear Station, Unit No. 1 (TMI-1)
Crystal River Nuclear Generating Station, Unit No. 3 (CR-3)
Rancho Seco Nuclear Generating Station (RS)
Arkansas Nuclear One, Unit No. 1 (ANO-1)
Davis-Besse Nuclear Power Station, Unit No. 1 (DB-1)
LICENSEES:
Duke Power Company (Duke)
Metropolitan Edison Company (Met-Ed)
Florida Power Corporation (FPC)
Sacramento Municipal Utility District (SMUD)
Arkansas Power & Light Company (AP&L)
Toledo Edison Company (TECO)
SUBJECT:
SUMMARY
OF MEETING HELD ON AUGUST 23, 1979, WITH THE BABC0CK &
WILC0X (B&W) OPERATING PLANT LICENSEES TO DISCUSS RECENT (POST TMI-2) FEEDWATER TRANSIENTS On August 23, 1979, members of the NRC staff met with representatives of the B&W operat'ing plant licensees, and the B&W Company, to discuss feedwater transients at B&W operating plants which have occurred subsequent to the Three Mile Island Unit 2 (TMI-2) accident. is a copy of the meeting agenda. A list of
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attendees is provided as Enclosure 2.
BACKGROUND Following the TMI-2 accident, an NRC staff review of the B&W designed plants' response to feedwater transients concluded that they have an unusual sensitivity to these types of transients. The Comission issued Orders during May 1979, which confirmed that the plants would shut down and remain shutdown until several short-term action items were accomplished. These items were required to mitigate the consequences of feedwater transients in these facilities. Subsequent to the accomplishment of these items and the lifting of the Comission Orders, several feedwater transients have taken place at the B&W operating plants. This meeting was called by the NRC staff to review those transients with respect to: (1) plant response, (2) operator action, (3) ICS response, and (4) licensees' actions to preclude similar events.
88i 185 M
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, DISCUSSION The meeting was divided into two parts:
(1) The first part involved presentations by FPC, AP&L, Duke and SMUD.
Each licensee discussed the feedwater transient events which have taken place at its facility.
(2) The second part of the meeting involved a discussion of the consequences of experiencing a loss of pressurizer level indication (LOPLI) during transient events on B&W plants.
RECENT FEEDWATER TRANSIENTS The following is a listing of the feedwater transients which were discussed during the meeting:
FACILITY TIME &'DATE DESCRIPTION OF TRANSIENT CR-3 0259 8/16/79 Reactor trip on high RCS* pressure, 72% power, reactor trip when "C" RCP* was secured.
1125 8/16/79 Reactor trip on high RCS pressure, 45% power, 3 RCPs operating, S/G* underfeed 0706 8/17/79 Reactor trip on high RCS pressure, 48% power, 3 RCPs operating, S/G underfeed 1825 8/17/79 Reactor trip on high RCS pressure, 26% power, 3 RCPs operating, S/G underfeed 0202 8/2/79 Reactor trip on low level in both S/Gs,10% power (automatic. anticipatory reactor trip)
ANO-1 1749 8/13/79 Reactor trip on high RCS pressure, 75% power (Automatic reactor-trip-on-turbine-trip did not work)
Oconee 1 0333 6/11/79 Reactor trip on loss of main feedwater, 99% power, (automatic anticipatory reactor trip) 0752 6/11/79 Reactor trip on loss of main feedwater (manual),
1% power Oconee 2 0344 5/7/79 Reactor trip on high RCS pressure, 28% power, fee 2046 6/3/79 Rea DUPLICATE DOCUMENT Rancho Seco 1714 7/12/79 e
Entire document previously au entered into system u WCS.- reactor coolant system y @ / () d U VA
/ /g q g RCP - reactor coolant pump ANO
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S/G - steam generator No. of pages:
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