ML071020264
| ML071020264 | |
| Person / Time | |
|---|---|
| Site: | Point Beach, 05000001 |
| Issue date: | 02/25/1981 |
| From: | Murley T NRC Region 1 |
| To: | Hanauer S Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2007-0140 | |
| Download: ML071020264 (25) | |
Text
IC;D D FA I of f*UCL FAAr G[ULI..1 t 0I1;Y CUPW.rII&JeON
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MEMORAINDUM FOR: Stephen Hanauer, Director Division of Human Factors Safety (DHFS)
FR(H: Thomas E. Murley, Director Division of Safety Technology COST)
SUBJECT:
DIESEL GENERATOR LOADING PROBLEMS RELATED TO SIS RESET ON LOSS OF OFFSITE POER
REFERENCE:
- Ia-orandum dated 11/5/80: S. Hanauer to F. Schroeder, Acting Director, DST;
Subject:
SAFETY INJECTION SIAL RESET The Division of Safety Technology described in a September 30, 1980, memorandto to you a potential problem at thexPoint beach reactors whereby a Safety Injection System (SIS) reset could result in loss of capability to actuate containment spray automatically. The .erzo also suggested that in other plants other safety functions could be adversely affected by SIS reset and recommended specific action' Later, the Operating Experience Evaluation Branch (OEEB) began'a review of SIS reset from a different aspect due to a recent Preliminary Notification
- (PeO) identifying a problem with diesel generator loading after SIS reset
+
.. and loss of offsite'power. This problem is the same as that raised as Technical. Issue 4/in NUREG-0138 and was considered by the staff tobe resolved on all. operating plants. Therefore, as a result of this PNO and another relatively recent similar report, OEEB performed a review of the history of the problem and the agency's actions to resolve it. In spite of these actions the problem has persisted at least at some plants, indicating a deficiency in the process. Though it is difficult to identify at what; v1Ant the prou_4s t.-ouL du.n, it 6 clear :f:t our suxc.ss was incc;.vlete.
During our review %v determined that the WFS is presently reviewing the problem on new plants and has plans to perform such a review on operating.
plants in the future. Inasmuch as OiFS is addressing this problem (see reference memo).further action on the part of OST is not'planned. However, for your informati~i, we have enclosed our compilation of the history-of the.
problem and the numerous actions taken.in attumpted resolution. The enclosure also identifies the concerns and the actions we believe could bring a final resolution.. - 1'
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S.:Tanaucr 020fB5 ~3 In view of the history of this problec &nd savitng lack of a Ffna1 rosolution, we urgt the {IFS In Its review to consider the need to rcach and document a satisfactory co=pletion of the agency's cotmiAiunnts ralated to Technical Issue Nlo. 4.9 U.1G-01.8o Th.;.:.'*~s F. '::rli-v,.r'f¢,ctor' ovision of S:.-",t.y Tc'.nolo-y Cnclosurcs:
- 1. SIS %set Action as It 8l0rtcs to Dics2l fcnerator Prek1lens
- 2. List of c."lated cr-:nts cc: H. Denton V. Stello T. Murley D. Eisenhut H. Thornburg N'. Moseley E. Jordan W. Garmn.i I1 D. Crutchfield D. Zteimann F. Nolan S. Bryan
- 0. 93pc',fam M. Ernst
.E.: Adensam C. Michelson H. Faulkner D. Ross F. Schroeder R. Baer A. Thadani F. Rosa
/
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$IS RESET ACTIOQN AS IT REL.ATES TO DIESFL. GE PIE RATOR PROB LEMS . ..
In a recent Preliminary Notification, PNO-Y: 80-68A da'ted September 23. 1980.
on San Onofre Unit No. 1, it was reported that the licensee had, during testing$
identified a problem with the design of the diesel generator sequencing circuitry.
The report indicated the problem would occur when a safety Injection signal Is blocked in accordance with the LOCA procedure following safety injection initia-tion. Under these circumstances a subsequent loss of off-site power would not produce automatic resequencing of safety injection loads onto the diesel generator supplied buses.
This same concern was presented November 3. 1976. as Technical Issue No. 4
- Loss of Offsite Power SLbsequent to Manual Safety Injection Reset Following a LOCA" in NUREG-0138 and is quoted as follows:
"The ECCS designs are such that beginning about two minutes after occurrence of a LOCA the operator Is required to reset the Safety Injection System (SIS) signal. If the operator does reset the SIS signal (a few minutes after LOCA) and if a loss of offsite power should then occur, prompt operator action would be required to restart the LOCA loads. The logic for startup of the emergency diesel generators would cause automatic sequencing to pick up the normal shutdown cooling loads In some designs and in others no loads would be sequenced (since there would be no accident signal present) rather than the LOCA loads, which would be the case if SIS had not been reset. Tne staff has accepted this design, in some plants provided that procedures for use of the SIS reset, and for actions required in the event of loss of offsite power after SIS reset to sequence on the proper loads without overloading the diesels. are carefully reviewed for adequacy.
"The staff also should consider the effects on the core of loss of power to engineered safety features following a LOCA and after safety injection signal reset.-
"The resolution of this issue should be implemented on all plants."
The staff, in response to these concerns, committed to reexamine the plant opera-ting and emergency procedures, which may direct or permit operator action concerning
$51 reset to be taken as early as 10 minutes after a LOCA signal, and to require such procedures be revised to prohibit SIS reset by operator action earlier than 10 minutes after a LOCA Signal unless it can be shown such action is required In the interest of safety. This commitment was to be implemented by the Office of Inspection and Enforcement (OE). In addition, for all operating PWRs, the OilE was to verify that all safe shutdown loads are automatically loaded following an operator action to reset SIS including those necessary.to assure continued cooling of the diesel generators.
Subsequently. OIS. at the request of URR, issued Temporary Instruction TI 2515/5 January 14. 1977. for the purpose of surveying operating PWRs'regarding diesel generator load sequencing in event of loss of offsite power following a LOCA and after SIS reset. Having reviewed the survey results, NRR concluded that at least 14 operating PWRs did not have written emergency procedures covering necessary corrective operator actions after reset. Also, NRR was unable to,
determine from the survey reports which PWRs had control system designs where operator action would be required prior to SIS reset to Orevent emergency equip-ment from changing its operating position.
As a result of these findings. NRR sent a memo to OIE June 28, 1978. (Stello toThornburg) requesting as follows:
S. Therefore,.'we request that the Office of Inspection and Enforcement initi-ate.a programto assure that the.necessary:procedures are prepared where they do not already exist. (The regulatory authority for requiring su:h procedures is contained in 10 CFR 50.36, Technical Specifications, and 10 CFR 50 Appendix B. Quality.Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.) Specifically, the procedures should set forth:
- 1. The specific operator actions necessary to manually restart the required engineered safety features if a loss of offsite power or an accident occurs after the SIS has been reset and before the engineered safety feature equipment is'returned to the automatic starting sequence mode by clearing the signals which initiated the SIS or resetting the RPS trip breakers.
- 2. The specific operator actions required prior to SIS reset to prevent any equipment from moving out of its emergency mode when the SIS is reset."
In response to the NRR request, OIE issued TI 2515/14 instructing that an inspec-tion to verify items 1 and 2 above be performed prior to October 16, 1978.
Upon completion of these inspections, a memorandum was issued 'from OIE, March 19, 1979, (Moseley to Stello) summarizing the results as follows:
"During these inspections-, a total of 30 facilities were found to have either design features which cannot be blocked by the reset-acttoror- -
adequate procedures available. Procedures were also found at the remaining 16 facilities; however, procedure revision was required based on information discussed in the TI. All deficient procedures were corrected by December 31, 1978.u This action was thought to have completed the resolution of this concern until Licensee Event Report (LER) 79-141/O0T-O North Anna Units I and 2, dated November 20, 1979, which indicated that the situation was'not adequately resolved.
A memorandum was then issued from Division of Operating Reactors to OIE, December 20, 1979, (Gammill to Jordan). This memorandum referred to the March 19. 1979, OIE (Moseley) memorandum which stated that as of December 31, 1978, all of the deficient procedures had been corrected. The (Gammill) memorandum also contained the following:
"Since that time we have received a preliminary notification (PNO-iI-79-25) and subsequent licensee event report (LER 79-141/CIT-O) describing design deficiencies that have recently been discovered at the North Anna Powee Station 'Inits 1 and 2, which could cause equipment important to safety to operate less conservatively than assumed in the safety analysis. Specifically,
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it has been found that ccrtain equipment would return to its non-safety mode following the reset of an ESF'signal; thus, protective actions of the affected systems could be compromised once the associated actuation signal is res:t.
"The drawing review during the previous 1E investigation into this matter was carried out at the logic diagram level. As evidenced above, this review did not, In all cases, adequately reflect the actual system design.
When the schematic diagrams were subsequently examined, the abnormalities were found. A review atthe full detail level is needed to determine whether or not the above problem exists at other operating facilities.
-In light of the above, we believe the problem of equipment prematurely moving out of its emergency mode upon an SIS reset could still exist at some operating facilities.- Since this problem, which we once thought to be corrected, has recurred and since we are riot convinced that the licensees
- themselves know wether or not this problem exists at their facilities. we suggest the Issuance of the enclosed proposed IE Bulletin."
Following this, OIE issued Bulletin 80-06 Which covered the concern of equipment changing position after an SIS reset. The responses to this bulletin are to be reviewed by the l&C:Branch. ,However, the bulletin does not.cover the original concern of NUPEG-0138 regarding the:loading of the diesel generators. The North Anna event occurred after the OlE review, indicating that there were still problems wtth SIS reset. There is, therefort,.reason to suspect that the diesel generator loading after SIS reset may still have problems as indicated by the recent PNO-V:
80-68A.on San Onofre-1.
We recognlze that if tVe Sanl, Onofre plant were designed to have no SIS reset.
OIE may not have found it necessary to review the procedures covering manual actions (the response from OIE'implias..this). 1However, in view of the problems identified on North Anna and San Onofre it is obvious that design problems may exist which were neither discovered by the OlE 'review nor by preoperational or normal ongoing testing of these systems. In addition, some licensee's 1ay not know how their design was implemented as indicated by the San Onofre-l report.
TViaf, owin s;' "es.tI'h r::r: Iy wh O .vcvrrin9 pobl'am could be brought to a. final resolution.
Request licensees of operating reactors to:
- 1. ,submit a detailed description of the time sequence-loading (manual and/or automatic) of the diesel generators.under the following conditions:
- 2. describe the te'sts which were performed to verify (1). above.
- 3. describe the procedural limitations on the.SIS reset (in what time frame after a LOCA signal it Is allowed and under what circumstances).
- 4. if reset is not prohibited for at least 10 minutes after a LOCA signal, as, stated by the response to item 4, IUREG-0138, Justify such prior action.
S. describe the manual actions necessary to reestablish the operation of equipment necessary to maintain a safe plant condition for items (I)a and (1)b above and the time frame within which they must be performed.
Identify the plant procedures covering such situations.
- 6. describe any electrically operated equipment required for the continued aperAti.on of the emergency power sources which are not designed to be automattcally loaded onto the emergency buses.
A review of this Information should be performed on each operating PWR to assure that:
- a. dtesel generators are not overloaded because of auto loading without proper sequencing.
- b. diesel generators are not being blocked from loading except as accomp-lished by a sequencer for auto loading.
- c. the design accommodates automatic loading of equipment necessary for continued operation of the diesel generators.
- d. if manual actions are required to reload the diesel generators sufficient time .isavailable to perform such actions and that the plant has operating procedures covering such actions.
Safetyrevaluation reports to document the. findings of these reviews showing resolution of these concerns would provide a documented basis of the staff commitment in NUREG-0138.
Attachment:
Related Documents List
- ' ll - .~ll 4eil
- eJ~nNI
- .. a
-Related Documents:
- 1. .NUREG-0138
- 2. Memorandum: V. Stello, Director, DOR/NRR to H. Thornburg, Director, DRO/IE, dated June 28, 1978,
Subject:
Emergency Operating Procedures Governing SIS Reset At Operating PWRs.
- 3. Inspection and Enforcement Temporary Instruction TI 2515/5i January 14. 1977.
(Survey of Operating PWRs Regarding D/G Load Sequencing in Event of Loss of Offsite Power Following a LOCA and Subsequent to SIS Reset).
- 4. Inspection and Enforcement Temporary Instruction TI 2515/14, July 14. 1978, (Survey of Operating PWRs to Assure that Facilities Having an SIS Reset Feature Have Written Procedures to Cover All Necessary Operator Actions Prior to and After SIS Reset).
- 5. Memorandum: N. Moseley, Director, DRO/IE to V. Stello, Director, DOR/INRR dated March 19, 1979,
Subject:
Emergency Operating Procedures Governing SIS Reset At Operating PWRs - Memorandum dated June 28, 1978.
- 6. North Anna Unit 1 Licensee Event Report (LER) 79-141/OlT-0 dated November 20, 1979.
- 7. Memorandum: December 20, 1979, W. Gammill, Acting Director DORP/DOR to E. Jordan, Assistant Director, DROI/IE,
Subject:
Engineered Safety Feature Rcset Design Deficiences.
- 8. Memorandum: January 14, 1980, D. Crutchfield, Chief, SEPB/OOR, to D. Ziemann, Chief, ORB2/DOR, Subhct: Resolution of SEP Topic VII-2, ESF System Control and Uesign.
- 9. Memorandum: D. Ziemann, Chief, ORB-2/DOR to E.L. Jordan, Assistant Director, OROI/IE,
Subject:
Request for Reinspection of SIS Reset Procedures.:
- 10. Mfuorandum: February 20. 1980, W. Gammill, Acting Director, DORP/DOR to E. Jordan, Assistant Director, DROI/IE,
Subject:
Comments on IE Draft Bulletin.
Engineered Safety Feature Reset Design Deficiencies.
- 12. Region Report, Page 203 of Volume II, Part 2.
- 13. Memorandum: May 23, 1980, Frank Nolan to S.E. Bryan,
Subject:
Comments -
Rogovin Report Allegation RE: IE Error in Concluding That Adequate Procedures Were 1n Place in All Operating Reactors Including TMI-2, to Cover All Necessary Qperator Actions Before and After SIS Reset.
- 14. Memorandum: June 2, 1980, S.E. Bryan, Assistant Director for Field Coordination, ROI/IE to N.C. Moseley, Director, ROI/IE,
Subject:
Rogovin Report Statement on Inadequacy of SIS Reset.
- 15. San Onofre Unit 1 Preliminary Notification PNO-V-80-68A September 23, 1980,
Subject:
Potential Failure of Emergency Diesel Generator Load Sequences to.
Perform as Described in FSAR. .%
- 16. Memorandum: September 30, 1980, Frank Schroeder to Robert Bernero,
Subject:
Safety Injection Signal Reset.
17: Memorandum: September 30, 1980, Malcolm Ernst to Gus C. Lainas, Subject; Containment Spray Pump Actuation Logic.
- 18. Memorandum: September 30, 1980, Frank Schroeder to Stephen S. Hanauer,
Subject:
Safety Injection Reset.
u"ITED STATES NUCLEAR REGULATORY COMY.1SSICr, -
I' - WASIHINGTON. 0. C. 20s5"
,E'ORAtIDUT FOR: 0. G. Eisenhut, Director Division of Licensing FROM: R. J. Mattson, Direýtor Division of System Integration
SUBJECT:
PROTECTI`ON AGAINST NAVERTENT BORON DILUTION During our review of the accident analyses submitted for recent PWR fuel re-load aoplications, we have concluded'that puny of these PWRs are not adequately protected against inadvertent boron dilution during shutdown and refueling.
A detailed description of the inadequacies and the basis for the concern are orovided in Enclosure 1. We have recuired u;grading of the protection system for boron dilution events on a case-by-case basis during our reload reviews.
However, there are plants operating cycle after cycle without a special reason to submit a revised safety analysis report. Thus, Inadequacies in protection against boron dilution events remain uncorrected. In order to make sure that all ooerating PWRs are properly protected against inadvertent boron dilution, we have concluded that a generic review of inadvertent boron dilution events and related protective measures be carried out by each licensee.
- n~losui 2 *is a draft. of a letter. we request you Issue to all licensees-of ooerating PWR nuclear power plants. It requests they review their protection systems for Inadvertent boron dilution accidents against the staff's accept-bility criteria (provided) and comult to upgrade their protection system as necessary. This requirement has been reviewed by OST and they have concurred in this action.
Roger J. Mattson, Director Division of System Integtation Office of Reactor Regulatory
Enclosures:
As stated cc: T. Murley P. Check T. Soees B. Sheron
Contact:
J. Laaksonen, X28507 B. Sheron, XZ9453
...................... ~
ENCLOSUR" I.
- --7Y ASSE-S-':."ENT OF BOPOr DILUTION EV':ES
- . BACKGRCU',O Reactivity control in'P.(Rs is accomplished using control rods and boron.
During power operation, the control rods are usually removed from the core, and the boron concentration of the primary coolant is adjusted to control reactivity.
In the event that the boron concentration is inadvertently di'.qtH d,,'ing power operation, the resulting reactivity.insertion *i11 increat, wo reactor power and automatic safety systems will act to shut down the rvi:',r dnd maintain the plant within safety limits.
When the plant is shutdown and the control rods are inserted, boron is still required to be dissolved in the primary coolant in order to maintain the necessary shutdown margin. In the event of an inadvertent dilution of the primary coolant boron concentration while at shutdown, there are no automatic safety systems which will automatically mitigate the event. Mitigation is accomplished by the operator, who must be alerted to the event, diagnose the cause, and take the proper corrective action before loss of shutdown margin occurs.
The present criteria for assessing the acceptability of plants to accomodate boron dilution events is stated in section 15.4.6, Part II of the Standard Review Plan. The key elements of these criteria are that:
(1) in the absence of a single failure, the consequences of the event should be within the limits prescribed for anticipated operational occurrences.
'V -
(2) the time available for the operator to tate corrective action between actuation of alarm and loss of snutcown margin s.uld ze 30 minutes during refueling and 15 minutes during all other modes.
moreover,Section III of 15.4.6 of the SRP states that "... the redunaancy of alarms that alert the operator to be unplanned dilution is confirmed.'
1*. STATEMENT OF PROBLEM
- a. Recent Problems Recent reviews of reload applications and operating license applications have shown that significant deficiencies exist in protection against boron dilution events.
o Case review of San Onofre II and III, Waterford, St. Lucie II, and Palo Verde have indicated the detection and alarms available during shutdown and refueling did not meet the single failure criteria.
o Calvert Cliffs reload application did not have redundant detection and alarm systems. A boronometer which was claimed to be available during shutdown was found not be be capable of measuring boron concen-tration during certain modes of operation (measurement range needed not within instrument capability). The plant computer alarm cited was not audible and would require a technical specification on t.,e operator to check the alarm readout every 8.8 minutes.
o V. C. Summer's initial submittal measured operator action time from the time of the dilution event starting rather than the time of the alarm,
a Detectlon and Alarm cirmuits are generally not safety grade but control grade.
o No technical specifications generally exist which prohibit detection and alarm systes from being taken out of service during ooera=::na*
modes in which they would ba required.
o ANO-Z had no audibl* alar.s for the operator to respond to. The audible count-rate meter was ;,the proposed prot*tction.
- b. Operational Histosy Li cmnsee Event-Reports batwen 7969 and the present were examined to determine all events in which the boron concan:ration of the primary coolant was Inadvertently changed (both increased and decreased). These are provided in Tablea.
This review indicated that 25 such events have occurred. Using the total
,/ number of PWR reactor years as approximately 300, the -paobab444ty of an eve~ ci Intration is Inadvertently changed Is.,25/300
.8f0.1 per reactor y~et 1f I /iA rtv&&wftt 10I lesAt%.i.i
"-~ soci vd-t'*S Of significance also is that a boron dilution path presently not considered In the safety analysis reports has beem' identified. This Is the back leakage of secondary coolant Into the primary through steam generator tube leaks when the systw it shutdown and the primary pressure Is below the secondary pressure.
For all of the events repo.ted, mitigating action was taken by the operator before the shutdown margin was lost. No information was found on the Instrumentation that alerted the operator to the event, or the rate of
boron concentration change (it Is reasonable to assume that the rate was less than that assumed in FSAR analyses).
Safety Sqnfflcance The consequences of uMitigated boron dilution event are typically not analyzed irn Safety Analysis Reports. Rather, the analyses focus on dionstratng that sufficient time (per SRP 15.4.6) texsts ftr the operator to Identify the event and take the proper corrective action priOr-to loss of shutdow margin..
A boron dilution event which occurs during either shutdown or refuelibg Is no me-ed by any automatic protection system. The only means of protecting against. a loss of shutdown margin and return to criticality is the operator being alerted to the event and acting properly and timely.
lip Whfl e an ar a/ent could be made that the detection and alam system neds be safety grade, the present control grade systems are considered suffi-cient. However, because the event Is mitigated solely by operator action, It Is Imperative that the detection and alarm systams be sufficiently reli-able and- available during all times in which they would be needed.
An unmitigated boron dilution event would probably result in the most severe consequences during refueling with the reactor head off and the primary system partially draIned.
ftadiation associated with the power-burst would have a direct leakage path to the containment with associated contamination of any workers who were present in containment at the time of the event.
- ,' * - ,:o *, *- ,. *., , ... ,-. .., , . . . . .. . : - ,
.- ,W~t*.acor vessel head in pl aceA system overIreSsure and fuel..ar/ge
.ptentital-exist,,: Considerationof ts 'event, froma Atne l £hoCk..and-.
poin may also 1. .arranted....
Based on the present SRPcatagorlzation ofa4boron dilution event as an ant .'
C pat*e Operational occurrence (AOO) wch c Is' supprted, by .the operational,.?
i ?.::.
. . t.. r able -t& 1., t ,s e"pacted that an unmiti gat-d boron diluton *u .
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"..., . ..- .. exceed the criteri afor AO's (No fuel failure. r1 press , f an operator eror ratils 0.1 to 0.01 per, demand', -the peobabitty of an u*mitiated bor dilutio event. s on theorder of 10 -to 1 Prooosed.Criterla Based on the above asSteSsment int is concluded that all opeating PWRs-sould meet theacceptance criteria set forth'in SRP section 15.4.5 in order to*.assure, that they are adequately protected against. boron'dilution, events._---
The foIowi ng guidance Is provided for demonstrating that the SRP cre .rla are met
- 1. -When analyzingp.boron dilution during col d shutdown and refuel ing'modes'.
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