ML030700177
| ML030700177 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 03/13/2003 |
| From: | Spaulding D NRC/NRR/DLPM/LPD3 |
| To: | Cayia F Nuclear Management Co |
| Spaulding D | |
| References | |
| Download: ML030700177 (3) | |
Text
1 GUIDANCE FOR LICENSEE REVIEW OF PRELIMINARY ASP ANALYSIS
Background
The preliminary precursor analysis of an event or condition that occurred at your plant has been provided for your review. This analysis was performed as a part of the NRCs Accident Sequence Precursor (ASP) Program. The ASP Program uses probabilistic risk assessment techniques to provide estimates of operating event significance in terms of the potential for core damage.
The types of events evaluated include actual initiating events, such as a loss of off-site power or loss-of-coolant accident, degradation of plant conditions, and safety equipment failures or unavailabilities that could increase the probability of core damage from postulated accident sequences.
This preliminary analysis was conducted using the information contained in the plant-specific final safety analysis report (FSAR), individual plant examination (IPE), and other pertinent reports, such as the licensee event report (LER) and/or NRC inspection reports.
Modeling Techniques The models used for the analysis of events were developed by the Idaho National Engineering and Environmental Laboratory.
The models were developed using the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) software.
The developed models are called Standardized Plant Analysis Risk (SPAR) models. The SPAR models are based on linked fault trees. Fault trees were developed for each top event on the event trees to a super component level of detail.
Two revisions of the SPAR models are currently being used in the ASP analysis: SPAR Rev. 2 and SPAR Rev. 3.
- SPAR Rev. 2 models have four types of initiating events:
- transients,
- small loss-of-coolant accidents (LOCAs),
- steam generator tube rupture (PWR only),
and
- loss of offsite power (LOSP).
The only support system modeled in Rev. 2 is the electric power system.
- SPAR Rev. 3 models are currently being developed to replace Rev. 2 models. The newer revision models have 11 types of initiating events:
- transients,
- small LOCAs,
- medium LOCA,
- large LOCA,
- interfacing system LOCA,
- steam generator tube rupture (PWR only),
- LOSP,
- loss of component cooling water (PWRs only),
- loss of service water, and
- loss of DC power.
Both revisions have transfer events trees for station blackout and anticipated transient without scram.
The models may be modified to include additional detail for the systems/components of interest for a particular event. This may include additional equipment or mitigation strategies as outlined in the FSAR or IPE. Probabilities are modified to reflect the particular circumstances of the event being analyzed.
Guidance for Peer Review Comments regarding the analysis should address:
- Does the "Event Summary" section:
- accurately describe the event as it occurred; and
- provide accurate additional information concerning the configuration of the plant and the operation of and procedures associated with relevant systems?
2
- Does the "Modeling Assumptions" section:
- accurately describe the modeling done for the event;
- accurately describe the modeling of the event appropriate for the events that occurred or that had the potential to occur under the event conditions; and
- include assumptions regarding the likelihood of equipment recovery?
Appendix G of Reference 1 provides examples of comments and responses for previous ASP analyses.
Criteria for Evaluating Comments Modifications to the event analysis may be made based on the comments that you provide.
Specific documentation will be required to consider modifications to the event analysis.
References should be made to portions of the LER or other event documentation concerning the sequence of events. System and component capabilities should be supported by references to the FSAR, IPE, plant procedures, or analyses. Comments related to operator response times and capabilities should reference plant procedures, the FSAR, the IPE, or applicable operator response models.
Assumptions used in determining failure probabilities should be clearly stated.
Criteria for Evaluating Additional Recovery Measures Additional systems, equipment, or specific recovery actions may be considered for incorporation into the analysis. However, to assess the viability and effectiveness of the equipment and methods, the appropriate documentation must be included in your response. This includes:
- normal or emergency operating procedures,
- piping and instrumentation diagrams (P&IDs),
- electrical one-line diagrams,
- results of thermal-hydraulic analyses, and
- operator training (both procedures and simulation).
This documentation must be current at the time of the event occurrence. Systems, equipment, or specific recovery actions that were not in place at the time of the event will not be considered. Also, the documentation should address the impact (both positive and negative) of the use of the specific recovery measure on:
- the sequence of events,
- the timing of events,
- the probability of operator error in using the system or equipment, and
- other systems/processes already modeled in the analysis (including operator actions).
An Example of a Recovery Measure Evaluation A pressurized-water reactor plant experiences a reactor trip. During the subsequent recovery, it is discovered that one train of the auxiliary feedwater (AFW) system is unavailable. Absent any further information regrading this event, the ASP Program would analyze it as a reactor trip with one train of AFW unavailable. The AFW modeling would be patterned after information gathered either from the plant FSAR or the IPE.
However, if information is received about the use of an additional system (such as a standby steam generator feedwater system) in recovering from this event, the transient would be modeled as a reactor trip with one train of AFW unavailable, but this unavailability would be mitigated by the use of the standby feedwater system.
The mitigation effect for the standby feedwater system would be credited in the analysis provided that the following material was available:
- standby feedwater system characteristics are documented in the FSAR or accounted for in the IPE,
- procedures for using the system during recovery existed at the time of the event,
- the plant operators had been trained in the use of the system prior to the event,
- a clear diagram of the system is available (either in the FSAR, IPE, or supplied by the
3 licensee),
- previous analyses have indicated that there would be sufficient time available to implement the procedure successfully under the circumstances of the event under analysis, and
- the effects of using the standby feedwater system on the operation and recovery of systems or procedures that are already included in the event modeling. In this case, use of the standby feedwater system may reduce the likelihood of recovering failed AFW equipment or initiating feed-and-bleed due to time and personnel constraints.
Schedule Please refer to the transmittal letter for schedules and procedures for submitting your comments.
Reference 1.
R. J. Belles, et al., Precursors to Potential Severe Core Damage Accidents: 1997, A Status Report, USNRC Report NUREG/CR-4674 (ORNL/NOAC-232)
Volume 26, Lockheed Martin Energy Research Corp., Oak Ridge National Laboratory, and Science Applications International Corp., Oak Ridge, Tennessee, November 1998.
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