NUREG-0410, Speech Entitled Shutdown Decay Heat Removal, for Presentation at 890418-20 NRC Regulatory Info Conference in Washington,Dc
| ML20113G528 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 04/18/1989 |
| From: | Rosalyn Jones Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20092F288 | List:
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| References | |
| CON-IIT05-238-000A-90, CON-IIT5-238-A-90, RTR-NUREG-0410, RTR-NUREG-410 NUDOCS 9202210439 | |
| Download: ML20113G528 (7) | |
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l SHU100WI4 DECAY HEAT REMOVAL BY ROLERT C. J0!JES, JR., SECTION CHIEF REACTOR SYSTEMS BRAllCH DIVISION OF EllCINEERiflG AtlD SYSTEMS TECHNOLOGY OFFICE OF NUCLEAR REACTOR REGULAT!0fl U.S. NUCLEAR REGULATORY C0ftMISS10N FOR PRESENTATI0il AT THE NRC REGULATORY INFORMATION CONFERENCE THE ftAYFLOWER HOTEL WASHINGTON, DC APRIL 18-20, 1989 i
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l SHUTDOWN DECAY HEAT REMOVAL INTRODUCTION Loss of decay heat renoval (DHR) cvents during pressurized-water reactor (PWR) shutdown operations, partitularly when the reactor coolant system (RCS) is in a reduced inventory condition for ir.aintenance operations, are an inoortant safety issue requiring industry attention.
Numerous loss of DHR events 1 ave occurred.
Several industry and NRC publications (for example, References I through 5) have been issued as a result of these events. Nevertheless, operating reattors continued to experience these events.
On April 10, 1987, Diablo Canyon Unit 2 lost DHR for approximately li hours and bulk boiling of the RCS occurred. This event occurred while the unit was shut down with the RCS level near the centerl%e of the hot leg piping. The flRC dispatched an Augmented Inspection Team G1T) to the site to investigate the circumstances associated with the event.
The All found deficiencies in procedures, hardware, and training that contributed to the event and that appeared generic to all PWRs (Reference 6).
To determine whether these issues were generic to all PWRs, the NRC issued Generic Letter 87-12 (Reference 7). This letter requested each licensee with a PWR to provide a description of plant operation dJring the approach to a partially filled RCS condition and during operation while in that mode. Review of licensee responses confirmed that the concerns were generic, as a result, the NRC issued Generic Letter 88-17 (Reference 8) providing reconrnended expeditious actions and long-term progranned enhancements for preventing a loss of DHR event, mitigating a loss of DHR event should one occur, and controlling the release of radioactive material if a core damage event should occur.
This paper discusses the safety significance of shutdown decay heat removal and the NRC reconnendations provided in Generic Letter 88-17 to address the issues.
The inethod being used by the NRC to review utility actions in response to the letter is described. The paper also highlights those areas where NRC review of industry responses indicates that further review may be required.
SAFEiY SIGHlflCANCE As describcd above, loss of DHR events have been of concern for some t itre.
Probabilistic risk assessment (PRit) results (References 9 and 10) indicate that the risk of a core damage accident during nonpower operation is on the order of 10 E-05 with 85 percent of the risk being associated with mid-loop operation.
The PRAs show that operator and other personnel interactions in event mitigation are greater than those associated with power operation.
The uncertainty associated with human interactions is high.
Review of the Diablo Canyon event identified several phenomena that had previously been unrecognized. One concern was the difficulty to control RCS level, and therefore assure continued DHR pump operation, while in a mid-loop condition. This concern was related to inadequate l
instrutnentation, inadequate procedures, and f ailure to understand plant behavior when in this mode.
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i Scenarios sere diso identifiev for Westinghouse and Combustion Engineering-designed plants wherein a loss of CHR durino operation while in a reduced inventory condition could result in core damage in approximatel) i hour; significantly less time than the previously estiinated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> available before the top of the core was uncovered.
In adoition, plant response to a loss of DHR had not been analyzed, and procedural inadequacies were found. Thus, there was concern that effective mitigation strategies for a loss of DHR event may not exist.
Furthermore, thc Diablo Canyon event demonstrated that the containment building could be open to facilitate maintenance operations during shutdown operations and while the RCS was in a mid-loop condition.
Should a loss of DHR event progress to a core damage situation, an unmitigated offsite release could occur.
g Review of industiy responses to Generic Letter 87-12 indicated that many of the problens identified from the review of the Diablo Canyon event existed at mast PWRs.
On the basis of the above information, tne NRC deternined that loss of DHR events during shutdown operation are an important !afety issue requ1 ring industry attention.
The actions requ'ested are given in Generic Letter 88-17 and are discussed below. To reinforce the importance of this issue, the Director of Huclear Reactor Regulation sent letters to the Chief Executive Officer of cach utility requesting that they assure the issues are properly addressed.
Furthermore, in recognition of the important role of the operators in prevanting and mitigating loss of DHR events, individual letters were sent to each licensed operator.
GENERIC LETTER 88-17 l
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The NRC recognized that any actions to be placed on PWR licensees must recognize the unique aspects of nonpower operation, including the need for maintenance activities. Thus the staff sought a flexible solution that would provide reasonable protection of the public health and safety.
To achieve this objective, mcetings were held with representatives of the various pWR Owners Groups.
These were successful in assisting the NRC in meeting its goal.
Generic Letter 88-17 provides two sets of recommendations, expeditious actions and programmea enhancements, which are to be addressed by licensees. These are discussed below.
Recommended Expeditiras A m ont The expeditious actiors are primarily directed toward mitigating offsite release :,hould severe core damage occur following a loss of DHR ever t.
Practical actions to reduce the litelihood of a loss of DHR event and to provide improved response to a loss of DHR event are also reccomended. The eight expeditious actions specified can be summarized as:
Discuss the Dicbio Canyon event and other industry events with plant personnel.
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Implement procedures and administrative conti015 to assure containment closure tfore possible uncovering of the core.
l Provide at least two independent, continuous temperature indications representative of core exit conditions.
Provide at least tuo independent, continuous RCS water level indications.
Implement controls to avoid nperations that perturb the RCS, Provide two means, and associated procedures, of adding inventory I
to the RCS to mitigate a loss of DHR event.
.implent procedures to assure that all hot legs are not blocked simultaneously by nozzle dams unless an adequate vent path is 1
provided.
Implement procedures to assure that all h9t legs are not blocked j --
simultaneously by closed loop stop valves unless an adequate vent
- J path is provided.
The generic letter specifies that these actions should be implemented before operation in a reduced inventory condition.
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Recomended P,rogrammed,Enhancemeny Piograreed enhancement recomendations emphasize improved understandino, better proced."es, better utilization of existinu equipment, and better instrumentation.
The ::- + % recommendations are:
t% >ide reliable indication of parameters that describe the state c'.he RCS and performance of systems used to cool the RCS for both tw mal and accident conditions.
W,sement procedures during reduced inventory operation, including nurmal ard emergency operation, that cover the NSSS, the containment and supporting systems.
4 Assure that adequate equipment, of Mgh reliability, is available for cooling the RCS and woiding a loss of DHR. Maintain sufficient backup equipment for mitigating a loss of DHR cvent.
Conduct analyses to develop a basis for procedures, instrumentation installation and response, and equipment /NSSS interaction and response.
E Review phnt Technical Specifications (TSs) and propose modifications for those TSs that restrict or limit the safety benefit 1
of the generic letter.
These actions were developed to deal with the root cause of loss of DHR events and to provide improved operator response to a loss of DHP event if one occurs.
Containment closure capability is to be retained. Thus, these actions l
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l represent a defense in depth pr.ilosophy of accident prevention, accider.t mitigation, and containment. Recognizing the lead time needed for analysis, i
design, and installation activities, an isnplementation schedule of appruximately 18 to 24 months is given in the generic letter.
l RMIEW METHOD The review method being employed by the staff consists of three basic elements. First, the licensee's responses to the generic letter will be -
given an audit review to ensure that the licensee appears to be meeting the intent of the generic letter. Observations will be provided in those areas where the licensee responses indicate that additional action may be needed.
An implementation audit will be performed by regional personnel or resident inspectors to assure that cocmitted actiens are properly implemented. A l-
-Temporary Instruction (Re#crence 11) will help Suide this effort for the expediticus actions. Review of 1:nplementation of the progranmed enhancenent reenmmendations will be performed in a similar manner.
Finally, for those responses that indicate t. hat the generic letter recommendations inay not be properly addressed, inspections will be perfortned to assure that appropriate actions are being pursued. Further plant-specific actions may be required to assure that licensees address this important safety issue.
The staff cecognizes that there are likely to be plant-spec'fic issues that will arise when licensees address the reconnendations given in the genet ic i
letter.
The staff review method is flexible and can address these issues when they arise as has been desconstrated by numerous discussions we have had with v
j licensees. We will continue to work with licensees to assure that the fundanental objectives of the generic letter are satisfied while considering l
plant-specific issues.
REVIEW FINDINGS - EXPEDIT10t!S ACTIONS l
Licensee responses to the recomended e peditious actions have been received and are being reviewed. Our-review indicates that most licensees appear to be meeting the intent of the generic letter. Generic observaticns from reviews completed to date are tiost licensee responses ii'.dicate that the Diablo Canyon and other loss of DHR events have been discussed with operations perscarel.
It was the intent of the recommendation that these discussions be held with all personnel whose actions can affect reduced itsentory operation.
- With respect to containment closure, many responses did not indicate what administrative controls will be in place to track open containment penetrations.
In addition, it is unclear from review of the responses whether some licensees are only tracking those paths with direct access from the containtoent atmosphere to the outside atmosphere. The staff expects all containment penetrations to be controlled, including those representing an indirect release path to the environinent (e.g.,
penetrations from the containment to the fuel handling or auxiliary building.)
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D Some licensees state that closure of the equipment hatch is achieved by installation of four bolts, without discussion as to whether any checks would be made to assure that tht, closure requirement, mating of the sealing surfaces, specified in the generic letter is inet.
Cenerally, action to be taken to reconcile differences between water level instruraants was not specified.
Care must be taken to resolve discrepancies ds this otidission has caused loss Cf DHR events.
Where licensees are using tygon tubing to meet the expeditious requirements, the staff reenmmends that daily walldowns inspections of the installation be perfortred when it is in use and an additional walkdown be conducted imediately before to its being placed in use.
Several response, have indicated that the pressurizer be opened for RCS venting whee all nozzle dams are installed.
Supporting analyses for this configuration are not provided.
Analyses are necessary to assure that whatever means are used for venting provide an adequate venting capability.
The pl6nt-specific observations are being transmitted to each licensee and will be included in the implementation.
CONCLUSIONS Since the Diablo Canyon event, substantial work has been performed to understand the causes of loss of DHR events and the response of the plant to such a loss. These investigations have indicated that improvemerts were needed in procedures, hardware, and training to prevent or effectively mitigate such cvents.
Numercus discussions were held with industry during the formulation of the recommendations proposed in Generic Letter 00-17.
These discussions have resulted in a solution that will significantly reduce the risk associated with shutdown decay heat removal while minimizing the impact on plant operations during this mode.
Although the actions specified in the generic letter are directed towards LHE while in a reduced inventory condition, the stcff believes that inany of the l
recommendations reflect good operational practices and control of shutdown activities.
We encourage licensces to consider the appropriateness of thcse recommendations to all modes of shutdown decay heat removal.
REFERENCES 1.
J. A. Haried, " Evaluation of Events invo?ving Decay Heat Removal Systems in Nuclear Power Plants," Oak Ridge National Laboratory, NUREG/CR-2799, Jnly 1982.
2.
G. Vine, et al., " Residual Heat Removal Review and Safety Analysis, Pressurized Water Reactors," Nuclear Safety Analysis Center, NSAC-52, January 1983.
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3.
" Loss or Degradation of Residual Heat Removal Capability in PWRs,"
Significant Operating Experience Report (50ER) 85-4, Irctitute of Nuclear Power Operations. August 28, 1985.
4.
II. Ornstein, " Decay Heat Removal problems at U.S. Pressurized Water Reactors," Case Study Report, NRC Office for Analysis and Evaloation of Operational Data, AE00/0503, December 1985.
5.
E. L. Jordan, " Loss of Decay llent Removal Due to Loss of fluid Levels in Reactor Coolant System,* NRC Office of Inspection and Enforcement Information Notice 86-101, December 12, 1986.
6.
" Loss of Residual Peat Removal System, Diatlo Canyor,, Unit 2 April 10, 1987," U. S. Nuclear Regulatory Commission, NUREG-12C9, June 1987.
7.
F. J. Mira911a, " Loss of Residual Heat Removal (RHR) While the Reactor Coolant System is Partially Filled," NRC O'fice of Nuclear Reactor Re9ulacion, Generic Letter 87-12, July 3, 1987.
8.
D. H Crutchfield, " loss of Decay Heat Removal," NRC Office of Nuclear Reactor Regulation, Gencrir, Letter 88-17, October 17, 1988.
9.
" Zion Nuclear Plant Residual Heat Removal PRA,* NSAC-84, July 1985.
1 10.
T. L. Chu, et al., "lmproved Reliability of Residual Heat Removal il Capabilities in PWRs as Related to Resolution of Generic issue 99,"
i NUREG/CR-5015, BNL-NUREG-52121, Brookhaven National laboratory, May i
1988.
11.
NRC Inspection Manual, Temporary Instruction 2515/101, "Ioss of Decay Heat Removal tGeneric Letter 88-17) 10 CFR 50.54(f)," February 16, 1989.
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