ML20149M649
| ML20149M649 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/29/1987 |
| From: | Thadani A NRC |
| To: | Hebdon F, Shao L, Starostecki R NRC |
| Shared Package | |
| ML20149B764 | List: |
| References | |
| FOIA-87-714 NUDOCS 8802260220 | |
| Download: ML20149M649 (17) | |
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NUCLEAR REGULATORY COMMISSION W ASHINGTON. D.C. 29666 1.,
May 29, 19
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R. Starostecki F. Miraglia Gq FROM:
A. Thadani Enclosed is the proposed 50.54(f) letter. We were assisted by the Offices of Nuclear Regulatory Research and Analysis and Evaluation of Operational Data in the formulation of this proposal. The letter is currently being reviewed for transmittal to CRGR on Tuesday June 2, 1987.
I am also sending a copy to J. Zerbe.
Please provide any comments you may have by Monday June 1.
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MEMORANDUM FOR:
Edward Jordan, Director Office cf Analysis anc' Evaluation of Operational Data FROM:
Thomas E. Murley, Director Office of Nuclear Reactor Regulation
SUBJECT:
NRR PLANS FOR RESPONSE TO LESSONS LEARNED FROM DIABLO CANYON LOSS OF RHR EVENT OF APRIL 10, 1987 AND RELATED EVENTS Diablo, Canyon Unit 2, a Westinghouse four loop FWR, experienced a loss of RHR on April 10, 1987 that continued for 85 minutes. The core heated to boiling within 30 to 45 minutes, and for the remainder of the event was cooled by reflux condensation in the steam generators.
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An Augmented Inspection Team (AIT) was dispatched to the site, and spent more than a week conducting the onsite investigation. The AIT concluded that lessons learned from the event are of significance to safety and many of the lessons appear applicable to all PWRs.
The staff considered these lessons to be sufficiently important that an infonnatWn Notice was DrCDared (87-23),
and individual plants have been contacted 4 they were in mid-oop or anticipating mid-loop operation.
Two licensees reacted to the lessons learned by voluntarily closino containment'and restricting other oDerations while in mid-loop operation. We have talked with other plant personnel who are ruith.ing theiLgrocedures anQgdware irfTfght of the lessons learned inf o6Eation.
Industry contacts regarding this issue are ongoing.
The NRR staff has prepared a 50.54(f) letter (attached) which provides additional inforation, and which requires infomation from all PWR licensees pertaining to this issue. This letter has been coordinated with and r9 viewed by AEOD, RES, OGC, and the Augmented Inspection Team leader. OGC has reviewed this package and has no legal objection.
Contact:
W. Lyon, SRXB x27605 l
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Edward Jordan 2
We request a CRGR review of this 50.54(f) letter be scheduled at the earliest i
opportunity.
Thomas E. Nurley, Director Office of Nuclear Reactor Regulation
Enclosures:
As stated cc:
Y. Stello J. Grace A. Davis R. Martin J. Martin DISTRIBUTION Central Files 1
SRXB R/r T. Murley J. Sniezek
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R. Starostecki L. Shao A. Thadani M. W. Hodges R. Jones W. Lyon W. Lyon R/F
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- RSTAROSTECKI:JSNIEZEK DATE 25/ /87
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DFFICIAL RECORD COPY
ATTACWENT 1 RESPONSE TO RECUIREMENTS FOR CONTENT OF PACKAGE SUBMITTED FOR CRGR REVIEW (1)
The proposed generic requirement or staff position as it is proposed 3
to be sent out to licensees.
See enclosed 10 CFR 50.54(f) letter (Attachment 2).
(ii)
Draft staff papers or other underlying staff documents supporting the requirements or staff positions.
Draft Augmented Inspection Team (AIT) Inspection Report, Report Number 50-323/87-18, dated May 19., 1987.
Memo from A. C. Thadant to B. A. Boger, G. C. Lainas, G. M.
Cr d Holahan, and F. Schroeder, "Imediate Notification to PWR Licensees Concerning loss of RHR and Associated Concerns," dated
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May 13, 1987.
Memo from E. Jordan to T. Murley and E. Beckjord, "Loss of Decay Heat Removal Function at Pressurized Water Reactors with Partially Orained Reactor Coolant Systees," dated May 18, 1987.
(iii)
Each proposed requirement or staff'p'osition shall contain the sponsoring office's position as to'whether the proposal would increase requirements or staff positions, implement existing requirements or staff positions, or would relax or reduce existing requirerents or staff positions.
Detailed descriptions of related equipment, procedures, conditions, and requirements are prnpnud to be made available to the operators.
Plant technical specifications are oronated to be compared to procedural requirements.
It is proposed that analysis of the transient be examined and possiblv~reneatet incorporating i
conditions into the calculations 'n prevmusty 6nsidered. The l
ac_tivation,of these reconsnendations would increase reovirements.
l (iv)
The proposed method of impinmentation with the concurrence (and any coements) of OGC on the method proposed.
06C has reviewed the proposed requirement of all PWR licensees to provide specific inferination pertaining to the issue. OGC has no t
j legal objection to our ororg31 (v)
Regulatory analyses generally conforining to the directives and d
guidance of NUREG/BR-0058 and NUREG/CR-3568.
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2 This issue deals with an unanalyzed event. Reconenended actions involve mainly procedural change with some urther effort towards anal is and tr Jating.
Losts of implementation would be mJinil in ison to (tee increasTTn sayy which would be significan1.
e (vi)
Identification of the category cf reactor plants to which the generic requirement or staff position is to apply.
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The proposed requirement is applicable to all U.S. Cosmercial pressurized water reactors.
(vii)
For each such category of reactor plants, an evaluation which demonstrates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. The evaluation shall document for consideration infonnation available concerning at,y of the following factors as may be appropriate and ar.y other infonnation relevant and material to the proposed action:
(a) statement of the specific objectives that the proposed action is designed to achieve; The proposed objective is to obtain proposed actions for safely operating in mid-loop conoitions. Actions are to be based on analyset _of conditions which may occur. Response by the licensee ~wou' d be expected before subsecuent mif-7557
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operation at each respective,<p' Tant.
Since many reactors enter mid-loop operation during uopianned shutdowns, this response is considered to be of high priot ity.
(b) general description of the activity that would be required by the licensee or applicant in order to complete the action; As discussed in Section (iii), recoernendations involve providing the operators with detailed descriptions of plant conditions during mid-loop operation, operable and useful equipment available, and procedures used. Technical specif.ications applicable during this conditiWsre tc be evaluated and compared with plant procedural requirements.
Also, a gnalis is to be revgd and possibly repeated incorporating information gained from the Ali inspection.
(c) potential change in the risk to the public from the accidental offsite release of radioactive material; Documentation, analysis, and review of this unanalyzed event would significantly reduce risk of core melt and thervfore greatly reTu&i ri,sk to the public.
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. (d) potential impact on radiological exposure of facility employees and other onsite workers; Implementatian of reconnendations, again, would significantly reduce the risk of core melt and therefore grg reduce risk to facility workers.
(e) installation and continuing costs associated with the action, including the cost of facility downtime or the cost of construction delay; Costs for procedural changes, review, and analysis would be mTWmal. There would be no extension of facility downtime or of construction delay.
(f) the potential safety impact of changes in plant or operational complexity, including the relationship to proposed and existing.
regulatory requirements and staff positions; GOC 34 requires an RHR system which assure the specified acceptable fuel design limits are not exceeded. This 50.54(f) letter (dentifies additional analyses needed to obtain i
assurance.j Inis dcies not preclude further actions under Gl.
(g) the estimated resource burden on the NRC associated with the proposedactionandtheavai)/6tfility of such resources; j
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The NRC resource burden should include the initial monitoring I
of licensee enforcement and the review of possible reanalysis of the loss of RHR event during mid-loop operation. Efforts required by the staff would be moderately low.
(h) the potential impact of differences in facility type, design or age on the relevancy and practicality of the proposed action;
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Impact differences based on facility differences is not.
certain.
It is pro)osed that reconnendations apply to all l
PWRs.
If a vendor ias reason to believe our concern is not applicable to a given facility, then the supporting evidence only need be presented to the staff and will be considered for exeeption. However, we believe this concern to be ap>11 cable to all Westinghouse PWRs and probably to all PWRs wit 1 i
U-tube-designed steam generators.
(1) whether the proposed action is interim or final, and if interim, the justification for imposing the proposed action on an interim basis.
This is a g l pr go reog*_**gt.
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. 1 (viii) for each evaluation conducted pursuant to 10 CFR 50.109, the proposing Office Director's determination, together with the rationale for the determination based on the consideration of paragraph (i) through (vii) above, that:
Ta) there is a substantial increase in the overall protection of uA'l public health and safety or the connon defense and security to b,I g [y be derived from the proposal; and b
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(b) the dinct and indirect costs of implementation, for the facilities affected, are justified in view of this increased protection.
The memorandum from E. Jordan to T. Murley provides the rationale to require further analysis by the licensees addressing unexamined conditions during mid-loop operation and also to provide corresponding information to the operators.
Currently, inaccurate instrumentation readings and unanalyzed air entraprent phenocena during this mode of operation have led to the condition where water level in the RCS is allowed to fall too low. Yortexing causes a loss of the RHR pumps and could evolute into overpressurization of the RCS.
With further gialysis and operator awareness, the risk of overpressurization is grg reduced.
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(ix) for each evaluation conducted for prpposed relaxations er decreases in current requirements or staff positions, the proposing Office Director's detereination, togeth# with the rationale for the determination based on the considerations of section (i) through (vii) above, that:
(a) The public health and safety and the connon defense and security would be adequately protected if the proposed reduction in requirements or positions were implemented, and (b) The cost savings attributed to the action would be substantial enough to justify taking the action.
Kapplicable.III l
ATTACHMENT 2 TO:
All Licensees of operating PWRs and holders of construction pennits for PWRs Gentlemen:
SUBJECT:
LOSS OF RESIDUAL HEAT REMOVAL (RHR) DURING MID-LOOP OPERATION
- pursuant to 10 CFR 50.54(f), the NRC is requesting infonnation to assess safe operation of Pressurized Water Reactors (PWRs) when the Reactor Coolant System (RCS) water level is below the top of the Reactor Yessel (RV). The principal concerns are (1) failure of the RHR system to meet the design basis of the Technical Specifications (Tgn Criterion 34 (10 CFR Part 50 Appendix. A), and.), in plant, such as General Desi
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lyzed impact upon safety.
Our concerns regarding this issue have increased over the past several years, and lessons learned from the April 10, 1987 Diablo Canyon loss of RHR event require an assessment of operatiens and planned operations at all PWR facili-ties.
Study of the Diablo Canyon event has led to identification of unana-lyzed conditions which are of si never came close to core damage,gnificance to safety. Although Diablo Canyon and could have withstood the loss of RHR 1
condition for over a day with no operator action, slightly different condi-tions could have led to a core damage accident' within several hours. One unanalyzed condition involves boiling with/) the RCS in the presence of air, leading to RCS pressurization with the potential for ejecting RCS water via cold leg openings, such as could exist during Reactor Coolant Pump (RCP) or loop isolation valve repairs. The lost water would no longer be available to j
cool the core, and this could significantly decrease the time to core damage if makeup were unavailable. The pressurization could also affect the capabil-ity to provide r.akeup water to the core.
Other unanalyzed situations are also possible, and occurred at Diablo Canyon.
The seriousness of this situation is exacerbated by the practice of conducting operations with the equipment hatch removed, and by the lack of procedures which address prompt containment isola-tion should the need arise.
Loss of RHR and related topics are not a new concern to the NRC staff. This topic has been addressed in numerous consnunications and investigations. Yet, events continue to occur at a rate of several per year. This condition needs to be fully considered in order to ensure compliance with Consission require-rents. Therefore, we request you to provide a safety assessment of operation of your plant during the approach to mid-loop condition and while in that operating condition. Your safety assessment is to include the following:
- Mid-loop operation as used here is the condition where water level in the reactor coolant system is below the level of the top of the reactor vessel.
2 1.
A detailed description of the circumstances and conditions under which your piant will be entered into and brought through a draindown process and operated at mid loop conditions. This is to include the information l
bases used for decisions, procedures generation, and training.
Examples of the type of infonnation required are the time between full power operation and reaching a mid-loop condition, requirements for minimum SG a
levels, restrictions regarding removal of equipment for maintenance and testing while in mid-loop, restrictions regarding testing and maintenance that could perturb the NSSS, requirements pertaining to isolation of containment, and the time required to replace the equipment hatch should replacement be necessary.
Specific requests regarding selected topics are included in the following items.
2.
A detailed description of the instrumentation and alarms provided to the t
operators for control of thermal and hydraulic aspects of the NSSS during operation under the conditions of concern. You should describe temporary piping used for instrumentation and'the quality control process to assure proper functioning of such connections, piping, and instrumentation, including assurance that they do not contribute to loss of RCS inventory i
or otherwise lead to perturbation of the NSSS during mid-loop operation.
3.
Identification of all pumps which can be used for control of NSSS inven-J.
tory, include:
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a.
Pumps you require be operable or captble of operation. Where such purps may be temporarily removed ce service for testing or mainte-nance, such information is to be neluded, b.
Other purps.
c.
An evaluation of the above with respect to applicable Technical i
Specification (TS) requirements.
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A description of the co_ntainment closure condition you require for the conduct of operations under the' conditions of concern. Examples of areas i
j of consideration are the equipment hatch, personnel hatches, containment i
J purge valves. SG secondary side condition upstream of the isolation valves (including the valves), piping penetrations, and electrical i
penetrations. You are to compare your requirements to applicable TS i
requirements.
I 5.
Reference to and a sumary description of procedures prgt_1y in the j
control room of your plant which describe operation for Sie conditions of concern. You response should include:
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a.
Procedural guidance pertinent to timing of operations, required i
instrumentation, cautions and warnings, and critical parameters.
b.
Operations control and comunications requirements regarding opera-tions which may perturb the NSSS, including restrictions upon testing and maintenance operations which could upset the condition j
of the NS$5.
c.
Differences between mid loop operation and other Mode 5 operations, i
such as operation with a reduced RHR flow rate to minimize the Itkelihood of vortexing and air ingestion.
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Restrictions involving RHR system maintenance and testing.
e.
Response to loss of RHR, including regaining control of RCS heat removal, operations involving the NSSS if RHR cannot be restored, control of efflueni from the containment if containment was not in an isolated condition at the time of loss of RHR, and operations to provide containment isolation if containment was not isolated at the i
time of loss of RHR.
Guidance pertinent to timing of operations, cautions and warnings, critical parameters, and notifications is to be clearly described.
f.
The analysis base used for procedures development, including a i
description of the analyses which illustrate Nuclear Steam Supply System (NSSS) response to nomal operation and to mitigative ac-tions. We are particularly interested in your treatment of draindown to mid-loop, analysis of minor variations from expected behavior such as due to air entrainment and de-entrainment, boiling in the core with and without RCS pressure boundary integrity, calculations of approximate time to core damage, vortexing, level differences in the RCS and the effect upon instrumentation indica-tions, and treatment of air in *,he RCS/RHR system, including the impact of air upon NSSS and instrumentation response.
6.
A brief deJgption of trt1ning provided to operators and other affected l
pert 6nnel uat is specific to the issue of mid-loop operation. We are particularly interested in such areas as maintenance pg_rionnel training i
regarding avoidance of perturbing the NSS$ and fesponse to loss of decay heat removal during mid-loop operatiop'y 7.
Identificatinn of additional resources provided to the operators during mid-loop operation, such as assignment of additional personnel with specialized knowledge involving the phenomena and instrumentation.
When you address the above itees, you should inclu4 consideration of the clarifications and background information provided ii fnclosure 1. Additienal information will be contained in the NRC Augmented Inspection Team report, NUREG 1269, ' Loss of Residual Heat Removal System, Diablo Canyon Unit @, April 10, 1987', a draft copy of which will be foreworded to you in the near future.
Your response is to be signed under oath or affirmation, as specified in 10 CFR 50.54(f), and will be used to determine whether or not your license should be modified, suspended, or revoked. We request your response within 60 days of receipt of this letter. This information is reau_ ired to assess conformance h' a \\,Mf PWRs with their licensing basis and is therefore exempt from backTit i
hi requirmnents. Our review of your submittal of infonnation is not subfict to
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d fees under the provision of 10 CFR 170. We suggest you consider providing a i
d f portion of your response in association with your respective owners group since much of this issue is of generic origin,
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This request for information was approved by the Office of Management and
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Budget under clearance number 3150-0011 which expires September 30, 1989.
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Counents on burden and duplication eay be directed to the Of fice of Management j
and Budget. Reports Management Room 3208, New Executive Office Building, Washington D. C. 20503, 4
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Sincerely,
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ENCLOSURE 1 TO ATTACHMENT 2 INFORMATION PERTINENT TO LOSS OF RFS! DUAL HEAT REMOVAL SYSTEMS WHILE IN MID-LOOP OPERATION j
Many maintenance and test activities conducted during an outage require lower-ing the Reactor Coolant System (RCS) water level to below the top of the Reactor Vessel (RV) and many times to the centerline elevation of the RV nozzles. This operating regime is typically known as "mid-loop" operation.
It places unusual detnands upon plant equipment and operators due to narrow control margits and limitations associated with equipment, instrumentation, pr6cedures, training, and the ability to isolate containment. Difficulty in controlling the plant while in this condition often leads to loss of the Residual Heat Removal (RHR) System, as illustrated in Table 1.
l Although this issue hat 4 en the topic of many consnunications and investi-gations, events continue to occur at a rate f, several per year.
If Recent knowledge has provided additional irisight into these events. Although the full isplications of this knowledge remain to be realized, our preliminary assessments have clearly established real and potential inadequacies associat-ed with mid-loop operation. These include not understanding the Nuclear Steam Supply System (NSSS) response to loss of RHR, inadequate instrumentation, lack of analyses which address the issue, lack of applicable procedures and train-ing, and failure to adequately address the safety impact of loss of decay heat removal capability.
The following items are applicable to these conclusions:
i 1.
Plants enter an unanalyzed condition if boiling occurs following loss of I
l RHR. For example:
l a.
Unexpected RCS pressurization can occur.
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l No pressurization would occur with a water / steam filled RCS with l
water on the Steam Generator (5G) secondary side as RCS steam would l
condense in the SG tubes and the condensate would return to the RV.
Air in the RC$ can block the flow of steam through passages, such as j
j the entrance portion of $G tubes, so that steam cannot reach cool j
l surfaces.
Failure to condense the steam causes RCS pressurizaticn l
1 until sufficient compression of the air occurs that steam can reach l
cooled tube surfaces. This pressurization occurred during the f
l April 10, 1987 event at Diablo Canyon. Pressure reached 5 to 10
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j psig, and would have continued to increase if RHR had not been restored. The operators initiated event temination by allowing water to flow from the Refueling Water Storage Tank (RWST) into the l
RCS. Increasing pressure would have eliminated this option, and l
would have jeopardized options involving pumps with suction lines l
l aligned (in part) to the RCS.
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Water that ordinarily would be agilable to cool the core might be j
forced out of the RV, thereby reducing the time between loss of RHR and initiation of core damage.
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This is a potential concern whenever there is an opening in the ccid j
leg, such as may exist for repair of Reactor Coolant Pumps (RCPs) or loop isolation valves. Upper vessel / hot leg pressurization could 1
force the RV water level down with the displaced water lost through
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the cold leg opening. A corresponding level decrease wovid occur in the SG side of the crossover pipes between the SGs and the RCPs.
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j This occurrence could be particularly serious if the cold leg j
opening were large or makeup flow to the RCS smal), as from a I
charging pump. Cold leg injection with elevated pressure in the upper vessel may not provide water to the core. Hot leg injection would probably be effective, l
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2.
RCS water level instrumentation my provide inaccurate infomation, j
There are many facets to this issue.
Instrumentation may be indicating a
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level that differs from level at the RkR suction line, a temporary instrument may be in use with no indication or alams in the control l
l room, and design and installation deficiencies may exist. We have I
observed the following:
i a.
Connections to the RCS actually provide a water level indication up-
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stream of the RCP location. This water level is higher than the
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I water level at the RHR suction connection due to flow from the injection to the suction locations and due to entering water momen-tum, which increases level on the RCP side of the cold leg injection iccation.
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Ingestion of air at the RHR suction connection will result in trans-porting air into the cold legs, which can potentially increase pressure in the air space in the sil'd legs relative to the hot legs.
l level instrumentation may respon[to such a pressure change as though RC$ level were changing.
In addition, such a pressurization would move cold leg water into the hot legs and upper RV (or the reverse if a depressurization occurs).
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Use of long, small diameter tubing which can lengthen instrument l
response time and cause perturbations such as RCS pressure changes.
to appear as level changes, tubing elevation changes which can trap air bubbles or water dnplets, and tubing which can become kinked or constricted, c.
Some installations provide no indication in the control room, yet level is important to safety. Some provide one indication. Others provide diversity via different instrumentation, but do not provide independence due to comon connections.
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Tygon tube insultations with faint level r. arks at one foot inter-vals, with ne provision for holding the tube in place, e.
Instrurentation in which critical inspections were not perforsed l
after the installation.
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Instrurentation where no provisions were made to assure a single phase in connection tubing or that tubing was not plugged.
g.
Use of instrumentation without performing an evaluation of indicated RCS level behavior and instrument response.
3.
Vortexing and air ingestion fr6m the RCS into the RHR suction line are l
not always understood, nor is NSSS response understood for this condi-1 tion.
l a.
On April 10,1987, Diablo Canyon opetators reduced indicated RCS l
level to plant elevation 106' 6"$beediately following 56 tube 1
I draining, and observed erratic PHR pump current indications. Resto-j ration of level to 106' 10' was reported to have eliefnated the L
I problem. RHR operation was terminated a few hours later at an indicated level of 107' 4' due to observed erratic RHR pop current indication. The Licensee later reported that vortexing initiated
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under those conditions at 107' S 1/2", and was fully developed at f
107' 3 1/2*. Procedures in place at the time of the event indicated the minimum allowable level to be 107' 0" (the hot and cold leg i
centerline elevation) or 107' 3".
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b.
Additional phenomena sppear to occur under air ingestion conditions, i
These include:
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(1) RHR pumps at Diablo Canyon were reported to handle several percent air with no discernible flow or pump current change I
frce that of single phase operation.
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5 (2) One postulate is that air in the RHR/ Reactor Coolant system can i
migrate or redistribute, and thus cause level charges which are at variance with those one wculd expect. This is a possible explanation for observed behavior where a lowering of RCS water level is followed by a level increase. Water in the RHR appears to be replaced t7 P r.
Similarly, an increase in RCS water level that is followed by a decreasing level my be due to void' in the RHR system being replaced by RCS water.
1 Failuri to c".*rsttod such behavior leads to operator mistrust i
of ley, i trst wntation and to operational errors.
1 c.
Operators typ.caily will start another RHR puep if the operating l
purp is lost. Experience and an understanding of the phenorana clearly show that loss of the second 'sp Should be expected. The l
cause of loss of the first pump shoule b m ly be corrected before attempting to run another RHR pump.j l
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d.
Typical mid-loop cperation provides a high RHR flow rate, which may be required by TS, but which my be unnecessary under the unique conditions associated with mid-loop operation. Air ingestion prob-l 1ess are less at low flow rates.
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Only limited instrumentation my be available to the operator while in mid-loop operation.
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level indication is any times available only in containment via a l
Tygon tube. Some plants provide one or more level indications in the control roce, and additionally provide level alarus, 1
J b.
Typically, RHR system temperature indication is the only teeperature l
provided to the operators. l.oss of RHR leaves the operater with no j
RC3 tosperature indication. This can result in violation of Techni-I cal Specifications, as occurred at Diablo Canyon on April 10 when i
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the plant entered Mode 4, unknown to the operators, with the con-tainment equipment hatch removed.
It also resulted in failure to recognize the seriousness of the heatup rate, or that boiling had initiated, c.
RHR pump mctor current and flow rate may not be alanned and scales may not be suitable for mid-loop operation, d.
RHR suction and discharge pressures may not be alanned and scales may not be suitable for mid-loop operation.
5.
Licensees typically conduct mid-loop operations with the containment equipment hatch removed and with operations in progress which impact the
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ability to isolate containment.
Planning, procedure's, and training do not address containment closure in response to loss of RHR or core damage events. This is inconsistent with the sensitivity associated with mid-loop operation and the history of lo(s of RHR under this operating condition.
M 6.
Licensees typically conduct test and raintenance operations which can perturb the 2CS and RHR syster while in mid-loop operation. The sensi-tivity of mid-loop operation and '5e historical record indicate this is not a prudent activity.
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e ID ENCLOSURE G REGULATORY ANALYSIS 6
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