ML043410089
| ML043410089 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/02/2004 |
| From: | Bezilla M FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC4480 | |
| Download: ML043410089 (7) | |
Text
FENOC FENOC 5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor Ohio 43449 Mark B. Bezilla 493177 Vice President - Nuclear Fax: 419-321-7582 Docket Number 50-346 License Number NPF-3 Serial Number 3110 December 2, 2004 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Davis-Besse Nuclear Power Station Review of Preliminary Accident Sequence Precursor Analysis of February 2002 Operational Conditions (TAC No. MC4480)
Ladies and Gentlemen:
By letter dated October 1, 2004', the NRC submitted for review and comment the preliminary Accident Sequence Precursor (ASP) Program analysis of operational conditions that existed at DBNPS from February 2001 until the plant was shutdown in February 2002.
FirstEnergy Nuclear Operating Company (FENOC) has reviewed the preliminary ASP analysis and finds that the descriptions of plant conditions, plant configuration, and operating procedures in the "Condition Summary" section appear to be accurate. In addition, the "Modeling Assumptions" section accurately describes the manner in which the relevant conditions were modeled, describes a model that in general appropriately characterizes plant conditions, and reasonably accounts for the likelihood of recovery actions that might have been taken if needed. The analysis, however, incorporates several hypothetical considerations that are more conservative than the as-found conditions presented in the applicable reference documents. In accordance with the NRC's guidance for the ASP program, we believe that the NRC should modify the assumptions used in the ASP analysis for Davis-Besse to make them more realistic.
Specific comments regarding these factors, along with additional clarifications, are provided in Attachment 1.
'Davis-Besse Nuclear Powver Station (DBNPS) Letter Log 6243, referred to here as "ASP Letter"
Docket Number 50-346 License Number NPF-3 Serial Number 3110 Page 2 If you have any questions or require further information, please contact Mr. Henry L.
Hegrat, Supervisor - Licensing, at (330) 315-6944.
Very truly yours, MSH : :
Specific Comments Commitment List cc: J. L. Caldwell, Regional Administrator, NRC Region III J. B. Hopkins, DB-1 Senior NRC/NRR Project Manager C. S. Thomas, DB-1 NRC Senior Resident Inspector Utility Radiological Safety Board
Docket Number 50-346 License Number NPF-3 Serial Number 3110 Page 1 Specific Comments
- 1. The method used to determine the Loss of Coolant Accident (LOCA) probability is not consistent with NRC published ASP and Probabilistic Risk Assessment (PRA) principles and guidance. The NRC transmittal letter states the following goal of the ASP analysis program: "In assessing operational events, the Nuclear Regulatory Commission (NRC) staff strives to make the ASP models as realistic as possible regarding the specific features and response of a given plant to various accident sequence initiators." The most realistic assessment would have used as-found LOCA probabilities. Instead, the ASP study uses LOCA probabilities for a hypothetical condition.
One of the factors that was most important to the NRC's calculated Conditional Core-Damage Probability (CCDP) is the assessed increase in the frequency of a LOCA, and particularly of a large LOCA. These frequencies were calculated by the Oak Ridge National Laboratory (ORNL) in Reference 112, taking into account a reduction in the safety margin for the pressure boundary. However, as discussed below, the ASP analysis for Davis-Besse did not use the best-estimate values in the ORNL report, but instead used conservative numbers.
- The "frequency of occurrence" associated with each pressure range in Table B-1 (ASP Letter) reported by ORNL could be reasonably interpreted as the 5%/95%
bounds of a normal distribution. This distribution would account for all causes of higher than nominal RCS pressure. However, the calculated probability of a transient-induced LOCA is arbitrarily increased by a factor of ten (ASP Letter, page B-2) without justification, other than to account for uncertainty in the choice of the distribution. Then the ASP Letter takes this frequency, applied for transients alone, and scales it up for Loss of Offsite Power (LOOP) and Station Blackout (SBO). Since the ASP probability analysis should be a best estimate analysis, the normal distribution is appropriate without an increase. The sensitivity analysis is the appropriate process to examine uncertainties in the results.
2 Williams, P.T., Yin, S., and Bass, B. R., Probabilistic Structural Mechanics Analysis of the Degraded Davis-Besse RPV Head, ORNL/NRCALTR-04/15, Oak Ridge National Laboratory, September 2004
Docket Number 50-346 License Number NPF-3 Serial Number 3110 Page 2
- Oak Ridge National Laboratory estimated the failure pressure for the as-found cavity to be between 2700 and 3300 psig. For a normal RCS pressure of 2185 psig, the resulting probability of a LOCA is 3.95 E-6. The total LOCA probability for all RCS pressure conditions is estimated to be 2.08E-4. We believe that this number is a reasonable estimate of the probability of a LOCA.
However, the LOCA probabilities reported in Table C-1 are on the order of 2E-1, approximately 1000 times higher than that reported in Table B-1.
The basis for the LOCA probabilities of 2E-1 in Reference 11 (pages 118, 119) was examined. The method of obtaining these probabilities was not based on the vessel head as-found conditions, but represents the LOCA probabilities for a hypothetical condition of the vessel head that did not exist. Basing the ASP analysis on LOCA probabilities from a hypothetical situation that did not exist is contrary to the stated goal of providing as realistic an assessment as is available, since LOCA probabilities from as-found conditions of the RPV head have been determined and are available in Reference 11 to the ASP analysis.
- The probability of a rod ejection of 1E-2 (page 5) seems too high based on the as-found conditions. The control rod drive mechanism cracking was in its initial stage of development and was not through wall,3 resulting in a significantly lower as-found rod ejection frequency.
The qualitative characterization of the importance of this event is generally appropriate; however, using realistic LOCA probabilities computed from as-found data in Reference 11 would require re-evaluation of the event with respect to other ASPs computed by the NRC. If used, more conservative numbers should be annotated as such and should not be described as best estimate numbers. In addition, comparison of this event with other industry events should include clarification regarding the use of conservative input data.
3 Davis-Besse Letter Serial Number 1-1311, "Response to the Nuclear Regulatory Commission (NRC)
Preliminary Significance Assessment for the Control Rod Drive Mechanism Cracking and Reactor Pressure Vessel Degradation Identified for the Davis-Besse Nuclear Power Station," dated April 24, 2003
Docket Number 50-346 License Number NPF-3 Serial Number 31 10 Page 3
- 2. The event tree for medium LOCAs (Figure 2 on page 18) includes top events for auxiliary feedwater (AFW) and for cooldown of the RCS. The implications of success for these events in the analysis include the following:
- If high pressure injection is unavailable, the cooldown allows earlier inventory control by low pressure injection (LPI).
- Successful cooldown is modeled as a requirement to allow long-term cooling to be accomplished using low pressure recirculation (LPR). Otherwise, the event tree indicates that only high pressure recirculation (HPR) would be an option for long-term cooling.
This event-structure is different from that used in the Davis-Besse Probabilistic Safety Assessment4'5 (PSA) or in the NRC's model for the Significance Determination Process (SDP)6. Neither the Davis-Besse PSA nor the SDP analysis models the need for AFW or active measures to cool down the RCS following a medium LOCA.
There may be conservatism in the PSA and SDP with respect to the ability to avoid core damage in the event of a medium LOCA without HPI available. For the precursor assessment, however, a more significant impact is the assumption that active cooldown would be needed to effect LPR in the long term. The expected response following a medium LOCA would be for the RCS to depressurize sufficiently such that, at the time of depletion of the borated water storage tank (BWST), LPR would be established.
If this change were to be made to the precursor analysis, medium LOCA sequence 4 would no longer be a core-damage sequence. This sequence comprises the second largest contribution to the CCDP, according to Table 1 and Table 3B of the preliminary precursor report. Removing this sequence would change the CCDP from 6.2E-3 to approximately 5.3E-3, and would reduce the importance from 6.IE-3 to approximately 4.4E-3.
4 Individual Plant Examination for the Davis-Besse Nuclear Power Station. The Toledo Edison Company, February 1993.
5 "Davis-Besse Probabilistic Safety Assessment Sequence Analysis Notebook." Davis-Besse Nuclear Power Station, Rev. 2, May 2001.
6 Azarm, M.A., et al., "Risk-Informed Inspection Notebook for Davis-Besse Nuclear Power Station Unit 1". Brookhaven National Laboratory for U.S. Nuclear Regulatory Commission, Rev. 1, November 2002.
Docket Number 50-346 License Number NPF-3 Serial Number 3110 Page 4
- 3. No event tree is provided for small LOCA in the precursor analysis. The event tree could help to define sequences 5 and 3 in Tables 3D and 3E, respectively.
- 4. The title for Table 3E is incorrect. The reference in the body of the table to "SLOCA Sequence 3" is correct, rather than "MLOCA Sequence 2" as indicated in the title.
- 5. One of the cut sets in Table 3E is comprised of the single basic event "DHR-MOV-CF-BWST." This event is missing from the list of basic events in Table 4.
- 6. Several events are identified that are characterized as "operator fails to recover sump in LLOCA (or MLOCA, SLOCA or transient)." These descriptions are misleading, since no credit is given to operator action to recover from sump failures (see the assumptions on page 4). The values for the corresponding events account for the conditional probabilities of failure of recirculation given pump operation at reduced Net Positive Suction Head. The descriptions should be changed to be more relevant to the actual treatment for these events, and used consistently throughout the analysis.
- 7. At the top of page A-4, it is stated that "... the operator will not hesitate to turn off containment spray if the water inventory is needed for decay heat removal.
Procedures allow turning off containment spray if containment pressure is less than 19.5 psig." This statement is correct, except that the current value used in plant procedures is 18.7 psia.
Docket Number 50-346 License Number NPF-3 Serial Number 3110 COMMITMENT LIST The following list identifies those actions committed to by the Davis-Besse Nuclear Power Station, Unit Number 1, (DBNPS) in this document. Any other actions discussed in the submittal represent intended or planned actions by the DBNPS. They are described only for information and are not regulatory commitments. Please notify the Supervisor - Licensing (330-315-6944) of any questions regarding this document or associated regulatory commitments.
COMMITMENTS DUE DATE None NA