ML20138K056
| ML20138K056 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/05/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20138F088 | List: |
| References | |
| NUDOCS 9705120100 | |
| Download: ML20138K056 (6) | |
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i ATTACHMENT 1 ARKANSAS NUCLEAR ONE, UNIT 2, NUCLEAR POWER PLANT INDIVIDUAL PLANT EXAMINATION i
STAFF EVALUATION REPORT r
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ENCLOSURE 3 l
9705120100 970505 PDR ADOCK 05000313
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INTRODUCTION On August 28, 1992, Entergy Operations, Inc. submitted the Arkansas Nuclear One, Unit 2 (ANO-2), nuclear power plant IPE submittal in response to Generic Letter 88-20 and associated supplements.
On May 8, 1995, the staff sent questions to the licensee requesting additional information.
The licensee responded in a letter dated October 5, 1995.
A " Step 1" review of the ANO-2 IPE submittal was performed and involved the efforts of Science & Engineering Associates, Inc., Scientech, Inc./ Energy Research, Inc., and Concord Associates in the front-end, back-end, and human reliability analysis (HRA), respectively. The Step 1 review focused on whether the licensee's method was capable of identifying vulnerabilities.
Therefore, the review considered:
(1) the completeness of the information, and (2) the reasonableness of the results given the ANO-2 design, operation, and history.
A more detailed review, a " Step 2" review, was not performed for this IPE submittal. A summary of contractors' findings is provided below.
Details of the contractors' findings are in the attached technical evaluation reports (Appendices A, B, and C) of this staff evaluation report (SER).
In accordance with Generic Letter 88-20, ANO-2 proposed to resolve Unresolved Safety Issue (USI) A-45, "Shutriown Decay Heat Removal Requirements." They also proposed to resolve two other generic safety issues (GSIs) namely, GSI 23, " Reactor Coolant Pump Seal Failures," and GSI 105, " Interfacing Systems LOCA at LWRs." With regard to GSI 23 and GSI 105, the Commission has determined that no further action is required by licensees.
The submittal states that the licensee intends to maintain a "living" PRA.
II.
EVALUATION AN0-2 is a two-loop Combustion Engineering PWR with a steel lined, prestressed, large dry containment. The ANO-2 IPE has estimated a core damage frequency (CDF) of 3.4E-05 per reactor-year from internally initiated events, exclusive of the contribution from internal floods. The licensee estimated the CDF contribution from internal floods at less than IE-06 per reactor-year and did not combine it with other initiating events. The ANO-2 CDF compares reasonably with that of other Combustion Engineering PWR plants. Transients contribute 84 percent, loss of coolant accidents (LOCA) 8.1 percent, station blackout 3.5 percent, anticipated transients without scram 3 percent, and steam generator tube rupture (SGTR) 0.15 percent. (Internal flooding was screened out.)
The import' ant system / equipment contributors to the estimated CDF that appear in the top sequences are:
i) Loss of a DC bus, which represents a relatively large contribution to CDF because it:
(1) can lead to loss of all main feedwater, (2) causes a partial loss of emergency feedwater, and (3) completely fails feed and bleed since power from both DC buses is required to open the ECCS vent valves or LTOP valves..
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i l-ii) Passive failures of the emergency feedwater turbine driven pump A.
l The licensee's Level 1 analysis appears to have examined the significant I
initiating events and dominant accident sequences. The common-cause failure i
i analysis, however, while including many important components, did not include some component groups typically included in IPE/PRAs, such as, circuit i
breakers, electrical swithchgear, certain valve types, air compressors and j
fans. The staff does not believe that this omission masks a vulnerability to l
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severe accidents, but it does represent a weakness in the analysis.
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Based on the licensee's IPE process used to search for decay heat removal (DHR) vulnerabilities, and review of ANO-2 plant-specific features, the staff j
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finds the licensee's DHR evaluation consistent with the intent of the USI A-45 l
(DHR Reliability) resolution and is, therefore, acceptable.
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The licensee performed a HRA to document and quantify potential failures in j
human-system interactions and to quantify human-initiated recovery of failure i
events. The licensee identified the following operator actions as important j
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in the estimate of the CDF: Operator failure to align offsite power to 4160 i
volt buses after failed automatic realignment following plant trip; operator failure to trip tha reactor coolant pumps (RCPs) after loss of component l
cooling wattr (CCW); and operator failure to realign DC buses to the swing l
l battery charger, In general, operator actions were found to be relatively 1
important to piant risk in the ANO-2.
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Weaknesses, however, were identified in the " post-initiator" HRA analysis.
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One weakness is in the area of credit allowed for recovering operator errors.
The IPE estimates for operator error recovery appear optimistic and could j
significantly underestmate the overall human error probability. Another j
weakness is in the lack of treatment of dependencies among multiple actions i
when these actions are modeled in the fault trees. Consequently, dependencies among multiple human actions appearing in the same cutset may not have been treated. Although the licensee has indicated they believe the effect of these 4
weaknesseses is negligible, the staff believes that they are weaknesses, i
nonetheless.
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The licensee evaluated and quantified the results of the severe accident
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progression tnrough the use of a containment event tree and considered l
l uncertainties in containment response through the use of sensitivity analyses.
i-The licensee's back-end analysis appeared to have considered important severe l
accident phenomena. The licensee identified the conditional containment i
failure probabilities for ANO-2 as follows:
early containment failure is 12.2 j
percent with failure to is'olate containment and high pressure melt i
ejection / direct containment heating the primary contributors; late containment i
failure is 13.9 percent with overpressure from loss of containment heat removal-(non-station blackout) and basemat meltthrough being the primary L
contributors, and bypass is 1.1 percent with SGTR and interfacing systems LOCA 3
the primary contributors. According to the licensee, the containment remains intact 72.8 percent of the time.
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Early radiological releases are dominated by bypass, namely, interfacing j
systems LOCA (failure of an RCP seal cooler tube) and SGTR; late releases are
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- l dominated by containment overpressure failure ruulting from (non-station blackout) loss of containment heat removal sequences.
The licensee's response to containment performance improvement program recommendations is' consistent with the intent of Generic Letter 88-20 and associated Supplement 3.
Some insights and unique plant safety features identified by the licensee at ANO-2 are:
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1.
The ability to perform once through feed-and-bleed cooling even though the plant has no power operated relief valves.
It is accomplished by opening the pressurizer ECCS vent valves or low temperature overpressure protection valves i
and injecting coolant via the high pressure safety injection pumps. Without credit for feed and bleed cooling, the CDF would increase by a factor of more than 5.
2.
The HPSI pump seals do not require cooling in the injection mode, only in the recirculation mode.
3.
The RCP seals are of a special design stated to be highly resistent to leakage in the event external seal cooling water is lost.
This, coupled with an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> battery life and diesel generators that are estimated to be an order of magnitude more reliable in their starting function compared to generic data, results in station blackout being a relatively small contributor to CDF.
4.
The unit auxiliary transformer, which powers plant loads during normal operation, shifts loads off to one of two start up transformers. This design tends to increase CDF because its failure to occur sucessfully will lead to a loss of offsite power condition.
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The transfer from ECCS injection to recirculation is automatic.
The licensee defined a vulnerability as either:
(a) Front-end sequence groups with a valid mean CDF greater than IE-04 per year, or (b) Containment event tree endstate groups involving containment failure / bypass that have a valid mean CDF greater than 1E-05 per year. No vulnerabilities were identified by the licensee.
Plant improvements, however, were identified. These improvements, listed below, have been implemented:
1.
Additional CCW relief capacity to mitigate RCP seal cooler tube rupture.
2.
Addition of a new auxiliary feedwater pump.
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3.
Addition of a new alternate AC power source (in response to the Station Blackout Rule.)
In addition, the following procedural improvements have been made:
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The shutdown cooling system procedure was modified to include an additional check that the shutdown cooling suction line isolation valves are closed. t
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The station blackout procedure was modified to assure that the 3/4 inch containment atmosphere monitoring system line is isolated during station blackout.
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The emergency operating procedures were modified to avoid isolation of the emergency feedwater pump discharge from both steam. generators.
III.
CONCLUSION Based on the above findings, the staff notes that: (1) the licensee's IPE is complete with regard to the information requested by Generic Letter 88-20 (and l
associated guidance in NUREG-1335), and (2) the IPE results are reasonable given the ANO-2 design, operation, and history. As a result, the staff concludes that the licensee's IPE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities, and therefore, that the ANO-2 IPE has met the intent of Generic Letter 88-20.
It should be noted that the staff's review primarily focused on the licensee's ability to examine ANO-2 for severe accident vulnerabilities. Although certain aspects of the IPE were explored in more detail than others, the review is not intended to validate the accuracy of the licensee's detailed i
findings (or quantification estimates) that stemmed from the examination.
Therefore, this SER does not constitute NRC approval or endorsement of any IPE l
material for purposes other than those associated with meeting the intent of Generic Letter 88-20. The staff has identified weaknesses in the front end and HRA portions of the IPE and believes that application of the IPE in support of risk-based regulatory applications, beyond those associated with Generic Letter 88-20, require additional treatment in these areas.
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t APPENDIX A ARKANSAS NUCLEAR ONE, UNIT 2 l
INDIVIDUAL PLANT EXAMINATION i
l TECHNICAL EVALUATION REPORT (FRONT-END) l 1
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