ML20149L874

Safety Evaluation for Ies Utilities of IPE in Response to GL-88-20 & Associated Suppls for License DPR-49

ML20149L874
Person / Time
Site:	Duane Arnold
Issue date:	11/12/1996
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20149L652	List: ML20149L649 ML20149L654 ML20149L874
References
GL-88-20, NUDOCS 9611190205
Download: ML20149L874 (4)

Text

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, pnurg a t UNITED STATES g ,g NUCLEAR HEGULATORY COMMISSION 4

o 2 WASHINGTON, D.C. 20666-0001

'4 9 . . . . . ,o STAFF EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

IES UTILITIES INC.

CENTRAL IOWA POWER COOPERATIVE CORN BELT POWER COOPERATIVE DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331 1

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1.0 INTRODUCTION

On November 30, 1992, the Iowa Electric Light and Power Co. (subsequently renamed IES Utilities Inc. (IES)) submitted the Duane Arnold Energy Center

(DAEC) Individual Plant Examination (IPE) in response to Generic Letter (GL) 88-20 and associated supplements. On January 6, 1995, the staff sent the

licensee a request for additional information (RAI). The licensee responded in a letter dated June 26, 1995.

i A " Step 1" review of the DAEC IPE submittal was performed and involved the efforts of Science and Engineering Associates, Inc., Scientech, Inc., and Concord Associates in the front-end, back-end, and human reliability analyses, respectively. The Step 1 review focused on whether the licensee's method was capable of identifying vulnerabilities. The review considered (1) the completeness of the information and (2) the reasonableness of the results, given the DAEC design, operation, and history. A more detailed review, a i'

" Step 2" review, was not performed for this IPE submittal. A summary of the contractors' findings is provided below. Details of the contractors' findings are in the attached technical evaluation reports (Enclosures 2, 3, and 4). l In response to the RAI, the licensee stated that the original IPE analysis, as described in the IPE submittal, has been updated. The licensee designated the updated version as the Revision 3 PSA (Probabilistic Safety Assessment). The technical evaluation reports, therefore, are based on information contained in the DAEC IPE submittal and Revision 3 PSA along with the licensee's responses to the RAI.

In accordance with GL 88-20, the licensee proposed to address Unresolved Safety Issue (USI) A-45, " Shutdown Decay Heat Removal Requirements." No other specific USIs or generic safety issues were proposed for resolution as part of the DAEC IPE.

Enclosure 1 9611190205 961112 PDR ADOCK 05000331 P PDR

2.0 EVALUATION DAEC is a single-unit General Electric BWR-4 with a Mark I containment. The DAEC Revision 3 PSA has estimated a total core damage frequency (CDF) of 1.5E-5 per reactor year. Internal flooding is a negligible contributor to CDF. Loss of offsite power transients, including station blackout, contribute 42%, anticipated transients without scram 22%, transients 17%, loss of support systems 16%, and loss of coolant accidents 3%. The important system / equipment contributors to the estimated CDF that appear in the top sequences are common cause failure of both diesel generators, failure of alternate low pressure injection (fire pump), and failure of reactivity control (scram). The licensee's Level 1 analysis appears to have examined the significant initiating events and dominant accident sequences.

Based on the licensee's IPE process used to search for decay heat removal (DHR) vulnerabilities and review of DAEC plant-specific features, the staff finds the licensee's DHR evaluation consistent with the intent of the JSI A-45 (Decay Heat Removal Reliability) resolution.

The licensee performed an human reliability analysis (HRA) to document and quantify potential failures in human-system interactions and to quantify human-initiated recovery of failure events. The licensee identified the following operator actions as important in the estimate of the CDF: failure to manually depressurize the primary system, failure to initiate torus cooling, failure to vent containment via torus vent, failure to recover offsite power within 30 minutes, and failure to recover emergency AC power within 30 l minutes.

The licensee evaluated and quantified the results of the severe accident progression through the use of a containment event tree and considered uncertainties in containment response through the use of sensitivity analyses.

The licensee's back-end analysis appeared to have considered important severe accident phenomena. Among the DAEC conditional containment failure j probabilities, early containment failure is 41% with overpressure being the primary contributor; late containment failures are 29% with failure of )

containment venting being the primary contributor; and bypass is less than 1%.  !

The containment remains intact 29% of the time. The licensee's response to '

containment performance improvement program recommendations is consistent with the intent of Supplement 3 to the Generic Letter.

Some insights and unique plant safety features identified at DAEC are as follows ,

1. Installation of hardened wetwell vent. l

2. Diverse means for establishing alternate vessel injection.

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3. Diverse dual ring bus arrangement to increase the reliability of the l offsite power supply system. i

4. Large air accumulators for feedwater regulating valves. l l

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5. Ability of equipment to operate without HVAC for extended periods. I

The licensee used the criteria in Nuclear Management and Resources Council (NUMARC) 91-04, " Severe Accident Issue Closure Guidelines," to screen for i

plant-specific vulnerabilities, and none were identified. Plant improvements, l 4

however, were identified and have been implemented. These improvements were  !

to change a portion of the control building fire protection system from a

" wet" pipe system to a " dry" pipe system to reduce the potential for flooding-related accidents in the control building.

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3.0 CONCLUSION

i Based on the above findings, the staff notes that (1) the licensee's IPE is complete with regards to the information requested by GL 88-20 (and associated 4 guidance NUREG-1335), and (2) the IPE results are reasonable given the DAEC design, operation, and history. As a result, the staff concludes that th:s licensee's IPE process is capable of identifying the most likely severe accidents and severe accident vulnerabilities and, therefore, that the DAEC IPE has met the intent of GL 88-20.

1 It should be noted that the staff's review primarily focused on the licensee's '

!' ability to examine DAEC for severe accident vulnerabilities. Although certain i 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 findings (or quantification estimates) that stemmed from the examination. Therefore, this staff evaluation does not constitute NRC approval of endorsement of any IPE 1 material for purposes other than those associated with meeting the intent of i GL 88-20.

l Principal Contributors : Nelson Su Glenn Kelly 1

Date: November 12, 1996 a

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• DUANE ARNOLD NUCLEAR POWER PLANT INDIVIDUAL PLANT EXAMINATION TECHNICAL EVALUATION REPORT (FRONT-END)

Enclosure 2 ,

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