ML20138H667

Safety Evaluation Accepting Util IPE Submittal in Response to GL 88-20

ML20138H667
Person / Time
Site:	Three Mile Island
Issue date:	12/19/1996
From:	NRC
To:
Shared Package
ML20138G356	List: ML20138G346 ML20138H589 ML20138H667
References
GL-88-20, NUDOCS 9701060166
Download: ML20138H667 (5)

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P c-ENCLOSURE

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THREE MILE ISLAND, UNIT 1, NUCLEAR POWER PLANT INDIVIDUAL PLANT EXAMINATION STAFF EVALUATION REPORT i

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I 9701060166 961219 PDR ADOCK 05000289 P

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I.

INTRODUCTION On May 23, 1993, the General Public Utility Co. (GPU) submitted the Three Mile Island, Unit 1, (TMI) Nuclear Power Plant Individual Plant Evaluation (IPE) submittal in response to Generic Letter 88-20 and associated supplements. On August 3,1995, the staff sent questions to the licensee requesting additional information.

The licensee responded in a letter dated December 6, 1995.

A " Step 1" review of the TMI IPE submittal was performed and involved the efforts of Brookhaven National Laboratory in the front-end, back-end, and j

human reliability analysis (HRA). The Step 1 review focused on whether the

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licensee's method was capable of identifying vulnerabilities.

Therefore, the review considered:

(1) the completeness of the information, and (2) the j

reasonableness of the results given the TMI, design, operation, and history. A more detailed review, a " Step 2" review, was not performed for this IPE submittal. Details of the contractors' findings are in the attached technical evaluation report of this staff evaluation report (SER).

In accordance with Generic Letter 88-20, TMI proposed to resolve Unresolved Safety Issue (USI) A-45, " Shutdown Decay Heat Removal Requirements."

(See a discussion of this isSection II). The-licensee also proposed, and the staff agrees to consider resolved, USI A-17, " Systems Interactions in Nuclear Power Plants," with the submission of the internal flood portion of the submittal.

In addition, the licensee listed GSI-23, " Reactor Coolant Pump (RCP) Seal J

Failures," in the IPE. The Commission has decided not to take additional rulemaking action at this time and plans to issue a generic letter on this issue at a later date.

The submittal does not state whether the licensee intends to maintain a "living" probabilistic risk assessment, although the staff strongly recommends they do so, particularly in light of the fact that the licensee indicated they plan to use the IPE models for on-going risk management activities.

II.

EVALUATION TMI is

%cox bd Wilcox PWR with a large, dry containment. The TMI IPE has of 4.2E-05 per reactor-year from estimated a core damage frequency (CDF) floods are not included in the above internally initiated events.

Internal total and they add another 3E-6 per reactor year to the CDF. The TMI CDF compares reasonably with that of other Babcock & Wilcox PWR plants.

Transients contributed 60 percent to the total CDF, loss of coolant accidents, 38 percent, interfacing system LOCA, 2 percent, and steam generator tube rupture (SGTR), 0.4 percent. The important system / equipment contributors to the estimated CDF that appear in the top sequences are:

1.

Failure of the High Pressure Injection (HPI) system, 2.

Failure of the Decay Heat River Water and Closed Cooling Water system, and 3.

Failure of recirculation from the reactor building sump. )

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i-i The licensee's Level 1 analysis appears to have examined the significant initiating events and dominant accident sequences.

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Based on the licensee's IPE process used to search for decay heat removal l

(DHR) vulnerabilities, and review of TMI plant-specific features, the staff j

finds the licensee's DHR evaluation ' consistent with the intent of the USI A-45 i

l (DHR Reliability) resolution and is, therefore, acceptable.

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The licensee performed an HRA to document and quantify-potential failures in human-system interactions and to quantify human-initiated recovery of failure 4

events. The licensee identified the following operator actions as important l

in the estimate of the CDF:

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Failure to switch over to reactor building sump recirculation following l

a large loss of coolant accident.

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Failure to refill the borated water storage tank (BWST) given a SGTR or

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very small LOCA.

3.

Failure to throttle HPI flow after engineering safeguards actuation.

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i The licensee evaluated and quantified the results of the severe accident progression through the use of a containment event tree but did not examine uncertainties in containment response through the use of. sensitivity analyses, j

The licensee's back-end analysis appeared to have considered important severe accident phenomena. The licensee identified the following conditional containment failure probabilities:

Early containment failure is 3 percent with hydrogen overpressure (with both high and low vessel pressure sequences) l the primary contributor; late containment failure is 63 percent with basemat meltthrough and overpressurization due to steam and non-condensible gas being the primary contributors; and bypass is 4 percent with interfacing systems LOCA and SGTR the primary contributors. According to the licensee, the containment remains intact 30 percent of the time. Early radiological releases are dominated by bypass and most late releases are due to benign, rather than catastrophic containment failure. The licensee's response to containment performance improvement program recommendations is consistent with the intent of Generic Letter 88-20 and associated Supplement 3.

The licensee based the majority of their containment response analysis on work done for the Oconee nuclear plant co..'ainment. While the staff finds this approach acceptable as a search for vulnerabilities, based on the similarities i

of these large, dry containments, the submittal makes it difficult to determine what new TMI-specific information regarding accident progression the licensee learned. We also note that the submittal did not include a containment phenomena sensitivity analysis as requested by Generic Letter 88-20. This represents a weakness in the IPE submittal, especially in light of the fact that the " template" approach was used. Although the template approach is an acceptable method for Generic Letter 88-20 compliance (which relies on existing analyses of similar containments in other nuclear plant *:e the staff believes that the licensee could have gained more of an understanding of the range of accident progression outcomes, and therefore the TM1 containment performance, had sensitivity analysis been performed. M

D Some insights and unique plant safety features identified by the licensee at TMI are:

1.

The turbine driven main feedwater pumps will continue to run for most transients, with puiap flow output automatically matched to the decay heat level.

2.

The turbine driven emergency feedwater pump has a mechanical linkage for control and is, therefore, not dependent on DC power for long term control in station blackout scenarios.

(The IPE did not take credit for this feature.)

3.

The czergency feedwater system pumps (two motor driven and one turbine driven) have been shown by licensee testing to not need bearing cooling for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.

One pressurizer power operated relief valve (PORV) and two safety valves can be used for feed and bleed. The PORV block is usually open and it depends only on DC power, not on nitrogen supply or instrument air.

5.

The RCP seals use new high temperature 0-rings which, according to the licensee, show a reduced leakage potential following a loss of all seal cooling.

6.

Recirculation switchover is accomplished manually.

The licensee defined.a front-end vulnerability as any core damage sequence i

exceeding IE-04 per reactor year. They defined a back-end vulnerability as any containment bypass or large early containment failure that exceeds IE-06 per reactor year.

Based on these definitions the licensee did not identify any vulnerabilities.

Plant improvements, however, were identified. These procedural improvements (none hardware-based) are listed below. They have not yet been implemented, however, their contribution to the reduction to CDF was assessed by the licensee to be smail:

1.

Operator action to isolate the steam generator experiencing a tube rupture and cooling down the reactor coolant system via the intact steam generator in cases where HPI is lost (CDF reduction approximately 1 percent).

2.

Operator action to refill the borated water storage tank during SGTR and very small break LOCA sequences (CDF reduction 4 percent).

Procedures are in place for refilling the BWST for small break LOCAs and the BWST level is monitored by numerous groups during emergency drills.

3.

Operator action to verify the closure of stop-check valve MUV-14 when aligning for piggyback recirculation from the reactor building sump (CDF reduction " negligible" but the risk achievement worth is high due to the potential bypass path to the environment.) _.

I Even though the licensee assessed the CDF reductions from these plant im.irovements as being small, we recommend that the licensee be encouraged to make these procedural changes and advise the staff upon their completion.

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 associated guidance in NUREG-1335), and (2) the IPE results are reasonable given the TMI 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 TMI IPE has met the intent of Generic Letter 88-20.

i It should be noted that the staff's review primarily focused on the licensee's ability to examine TMI for severe accident vulnerabilities. Although certain aspects of the IPE were explored in more detail than others, the review is not intended to validate ',he accuracy of the licensee's detailed findings (or i

quantification estimates) that stemmed from the examination. Therefore, this SER does not constitute NRC approval or endorsement of any IPE material for purposes other than those associated with meeting the intent of Generic Letter 88-20.

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