ML20248K285

Proposed Tech Specs Re That Such First Type a Test Required by Primary Containment Leakage Rate Testing Program Be Performed During Refueling Outage 18R

ML20248K285
Person / Time
Site:	Oyster Creek
Issue date:	05/28/1998
From:	GENERAL PUBLIC UTILITIES CORP.
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ML20248K276	List: ML20248K272 ML20248K278 ML20248K285
References
NUDOCS 9806100064
Download: ML20248K285 (12)

Text

Attachment 2 Revised Technical Specification Pages i

PDR ADOCK 05000219 P PDR,

4.5 CONTAINMENTSYSTEM Applicability- Applies to containment system leakage rate, continuous leak rate monitor, functional testing of valves, standby gas treatment system operability,inerting surveillance, drywell coating surveillance, instrument line flow check valve surveillance, suppression chamber surveillance,and snubber surveillance.

Objectives- To verify operability of containment systems, and that leakage l from the containment system is maintained within specified values, as outlined in Appendix J of 10 CFR 50. ,

Specificatiort A. PrimaryContainmentLeakageTesting A Primary Containment Leakage Rate Testing Program shall be established to implement 10 CFR -50, Appendix J, Option B, as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance Based Containment Leak Test Program," dated September 1995, as modified by the following exception:

1. The first Type A test required by this program will be performed during refueling outage 18R.

B. Type A Primary ContainmentIntegrated Leak Rate Test (PCILRT).

PCILRT shall be performed in accordance with the Primary Containment Leakage Rate Testing Program.

( C. Type B and Type C local Leak Rate Tests (LLRT)

. 1. LLRT shall be r>erformed in accordance with the Primary Containment Leakage Rate Testing Piogram.

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' 2. The Drywell Airlock, Drywell Airlock electricalpenetration,and  ;

j Drywell Airlock barrel seal shall be local leak rate tested in accordance with the i L Primary Containment Leakage Rate Testing Program.  !

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a. When containment integrity is required, the airlock must be tested at 10 psig within 7 days after each containment access. If the airlock is opened more frequently than once every 7 days, it may be tested at 10 psig once per 30 days during this time period.

- OYSTER CREEK 4.5-1 ' Amendment No.: 132,186

_ = - _ . .-- _ - .- . _ - . - - . . _ ________ _ _ _ _ -

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1 Attachment 3 Risk Evaluation of the Deferral of the Integrated Leak Rate Test (ILRT)

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, ILRT Risk Evaluttion

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1.0 INTRODUCTION

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' he purpose of this evaluatien is to provide continued support of the planning activities that consider the following potential scenarios for Oyster Creek:

1. Continued operation to the end oflicensed life (2009)

2. Sale of Oyster Creek to a third party

3. Early shutdown in September 2000 Specifically, in support of a possible early shutdown, this evaluation addresses the risk impact of deferring the performance of the integrated leak rate test (ILRT) from the refueling outage 17R to refueling outage 18R.

2.0 METHOD The method used in the evaluation of the risk impact of the deferral of the integrated leak rate test is similar to the method used in Attachment 2 of the " Risk Assessment of Dcferred Oyster Creek Projects", Oyster Creek Nuclear Generating Station Docket No. 50-219 (reference 1).

Step 1, Evaluate the Status of the Project within the framework of the various risk analysis studies performed for Oy ster Creck. In this step it is determined whether the risk impact of ILRT deferral can be directly reflected or inferred using the previously developed risk analysis studies.

Resiew the available risk analysis studies (ic., Probabilistic Risk Assessments (PRAs) and External Event (IPEEE) analyses.)

Review the deferred project.

Define whether impact of the deferred project can be directly or indirectly inferred from available risk evaluations.

Step 2 Evaluate the Safety or Risk Impact of the proposed ILRT deferral.

If the risk impact of the ILRT deferral can be directly produced using the available risk studies, perform the evaluation and provide the risk impact.

If the risk impact cannot be directly inferred, however, minor modifications to existing evaluations can be performed, perform modifications and provide the risk impact.

If the risk impact cannot be either directly or indirectly inferred from existing risk evaluations, then perform additional risk evaluations and provide the impact.

Step 3, Categorize the Safety / Risk Impact using categories of high, medium and low. For the ILRT project deferral, this consists of assigning numerical increases in core damage frequency or large early release frequency to pre-defined ranges. Assignment of a risk category allows for the integration of the risk impacts in cases where different figures of merit may be used to evaluate projects or activities.

Step 4, Evaluate the Integrated Safety / Risk Impact. Using the categcries established in step three, proside a final integrated risk assessment. This step allows for the risk impacts to be Ilttdef4 1 05/1298

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l considered in an integrated manner and as part of an overall risk management approach. In the case of the ILRT deferral, reference I is resiewed to determine if there are any changes to the integrate assessment performed in that assessment.

l The figures of merit used in the evaluation of the quantitative risk impact are core damage frequency and large early release frequency (LERF). These figurcs of merit are chosen since most presiously perfonned risk studies evaluate the frequency of core damage or large early release frequency. Other qualitative factors such as, consi'.lcration of alternative endstates, (e.g., significant transients) are documented in the individual evaluations. 'Itcse qualitative factors can affect the allocation of a project to a given risk category.

In oversiew, the above methodology agrees closely with the methods for the use of PRA methods in risk informed decision making outlined in the NRC draft Standard Review Plan, "Use of Probabilistic Risk Assessment in Plant-Specific, Risk Informed Decisiomnaking: General Guidance" (reference 1). For comparison purposes:

Steps I and 2 are equivalent to the first element in the draft SRP, Define the Proposed Change.

Steps 2 through 4, correspond to Element 2 of the draft SRP, Conduct Engineering Evaluations.

The third element of the draft SRP, Develop Implementation and Monitoring Strategies is also addressed in steps 2 through 4 on an individual project basis. Each project is evaluated for the potential for risk reduction, including compensatory measures. For example, fire watches have been posted in fire zones which corlain thermolag fire barriers. The evaluation of implementation and Monitoring strategics is performed on an activity or project basis depending on risk impact of the project deferral and the risk reduction achievable with potential compensatory measurcs. Also, performing parts or portions of projects are considered potential compensatory measures. For example, the most risk significant portions of a project may proceed as planned while less risk significant portions are deferred for a single cycle.

The fourth element in the draft SRP is represented in the submittal of the integrated schedule to the NRC. The submittal and supporting documents contain sufficient information to support the conclusions of the acceptability of the deferrals and are available for staff review.

3.0 EVALUATION OF DEFERRAL OF TIIE ILRT 10 CFR 50 Appendix J defines the requirements for Primary Containment Leakage Testing which is intended to ensure Primary Containment Integrity. In September 1995 the NRC approved an alternative to the existing Appendix J requirements. This alternative is known as Option B to Appendix J. In the interest ofimplementing Option B to Appendix J at Oyster Creek, Technical Specification Change Request (TSCR) 242, Revision 2 was submitted to the NRC (reference 2). The NRC accepted the TSCR approving the implementation of Option B to Appendix J at Oyster Creek on September 3,1996 (reference 3).

The Technical Specification Change Request No. 242, revision 2 (reference 2), among other changes, j included a commitment to perform the Type A Primary Containment Leak Rate Test during refueling j outage 17R. i l

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3.1 Step 1 - Evaluation of the Status of the ILRT Project j i

i The goal of this step is to evaluate whether the risk impact of the deferral of the ILRT can be estimated l using existing risk analyses done for Oyster Creek or other available risk analyses performed in the nuclear j power industry.

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In the case of plant specific risk analyses, the impact of the deferral of the ILRT can be estimated using the j existing Level 1 and Level 2 Oyster Creek Probabilistic Risk Assessments (references 4 and 5).

In addition, many evaluations performed in the nuclear power industry are also appropriate for the evaluation of the risk impact of the deferral of the ILRT. These studies include those performed in support of the Option B to Appendix J, such as NUREG-1493, " Performance-Based Containment Leak-Test i Program" (reference 6).  ;

i' These studies include detailed analysis which could not be cost-beneficially produced on a plant specific basis for the deferral of the ILRT. Therefore, the analysis of the risk impact of t' deferral of the IRLT is based heavily on this industry analysis with plant specific features and inwhts incorporated where  ;

appropriate.

l NUREG-1493, " Performance-Based Containment Leak-Test Program" (reference 6) presents summaries of many other risk analyscs done in determining the cost benefit of the extension in Appendix J testing intervals including Type A testing (i.e., ILRT). The following is a list of the significant references used in NUREG-1493- .

1 e NUREG/CR-4330, " Review of Light Water Reactor Regulatory Requirements", which examined, in part, the risk impact associated with increasing the allowable containment leakage rate by usmg l two different methods. {

r r e NUREG 1150, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, Final Report", published in December of 1990.

• NUREG/CR-4550, " Analysis of Core Damage Frequency: Internal Events Methodology",  ;

published in January of 1990. l These studies included evaluations of the Peach Bottom Nuclear Power Station. The Peach Bottom Nuclear Power Station is a General Electric Boiling Water Reactor (BWR-4) with a Mark I type containment.

Oyster Creek is a General Electric Boiling Water Reactor (BWR-2) with a Mark I type containment. A full comparison of plant features is provided in Attachment 1. Given the similarity in containment type, the analysis contained in NUREG 1493 is judged to be appropriate in addressing the deferral of the ILRT for Oyster Creek.  ;

3.2 Step 2 - Evaluate the Safety or Risk Img.act The risk impact on the public of increasing the allowable leakage rate for the Peach Bottom Nuclear Power Plant is small. From NUREG-1493:

" Increasing the containment leakage rate from the nominal 0.5 percent per day to 5 percent per day leads to a barely perceptible increase in the total population exposure. "

In addition, NUREG-1493 investigated alternatives to Appendix J testing frequencies. These alternatives included the extension of the Type A test (ILRT) from the then three per ten year period to once per 20

. years. From NUREG-1493, Summag of Technical Findings:

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" Reducing thefrequency of1)pe A tests (ILRT) from the current three per 10 } ears to once per 20 years wasfound to lead to an imperceptible increase in risk. The estimated

, increase in risk is wry small because IIRTs identify only a few potential containment

\ leakagepaths than cannot be identyled by Type B and C testing, and the leaks that have i beenfound by Type A tests have only been marginally above existing requirements.

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. Resinion 0 Given the insensitivity of risk to containment leakage rate (Chapter 5) and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integratedleakage-rate tests ispossible with minimalimpact on public risk.

The impact of relating the 1LRTfrequency beyond one in 20 years has not been evaluated Beyond testing the performance ofcontainment penetrations, ILRTs also test the integrity ofthe containment structure. "

Other considerations which impact the risk associated with the deferral of the Oyster Creek ILRT include (references 2 and 8):

1. The drywell corrosion issue at Oyster Creek has resulted in extensive inspections of the containment structure. Ultrasonic thickness inspections have been performed to determine wall thickness of the drywell shcIl and an extended inspection plan has been undertaken. These inspections provide a degree of assurance that the drywell structure has sufTicient wall thickness to withstand the peak accident pressure. Furthermore, some assurance is provided that there is no structural degradation occurring which would affect the containment's ability to withstand accident pressure.

2. Oyster Crcck has two means of detecting gross containment leakage. The first is by monitoring the use of nitrogen. If excessive nitrogen consumption is discovered, the cause of the use is investigated. A gross leakage path will be identified due to the use of nitrogen, and will trigger an investigation into the cause. The second way of determining gross leakage is by performance of the torus to drywcl! vacuum breaker leak test. This periodic test monitors drywc!! and torus pressure for a one hour period to calculate bypass leakage across the vacuum breakers. The test

, results would provide an indication if an external leakage path from either the torus or drywell exists. This too will trigger an investigation into the cause of the excessive leakage. While not testing the structure at accident pressure, these methods prove that there has been no gross structural degradation since the last ILRT.

Based on the risk impact information provided in NUREG-1493 as well as the plant specific considerations, the risk impact of the deferral of the ILRT from the current scheduled 17R to the 18R refueling outage is considered imperceptible.

3.3 Step 3 - Categorize Safety / Risk Impacts This report section provides the categorization of the safety / risk impact for the deferral of the ILRT. The risk impacts are categorized into either high, medium or low categories according to the increase in large early release frequency as defimed on Table 1. In the case where non-quantitative results have been used to assess the risk impact, the assignment of the risk category is based on judgement.

Table I- Categorization of Large Early Release Frequency Increases Risk Range Category (Percent Increase)

High > 100%

Medium 10 % - 100 %

Low <10%

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, Revision 0 Table 1, - Categorization of Large Early Release Frequency increases, is derived, in part, from the EPRI PRA Applications Guide (reference 9) which indicates that increases in large early release frequency for Oyster Creek of greater than 36.4% are significant and require additional analysis. The risk associated with the deferral of the ILRT project, based on the analysis in NUREG-1493, is judged to be imperceptible and therefore assigned to the Low risk category.

3.4 Step 4 - Evaluate the Integrated Safety /R sk Impact The purpose of step 4 is to integrate the risk impact of all project deferrals for Oyster Creek including those listed in reference 1, " Supplemental Update of the Integrated Schedule" Given the very small risk increase for ILRT testing frequencies of up to 20 years, as indicated in NUREG-1493, the deferral of the Oyster Creek ILRT isjudged not to affect the integrated risk impact assessment provided in reference 1.

4.0 CONCLUSION

S The deferral of the Oyster Creek ILRT from the 17R refueling outage to the 18R refueling outage results in an imperceptible increase in risk. This is based on the conclusions contained in NUREG-1493 which states that increases in the testing frequency of Type A (ILRT) tests from three in ten to one in 20 years results in an imperceptible increase in risk. Other factors, such as the monitoring of nitrogen usage and the drywell to torus vacuum breaker testing further ameliorate this small increase by allowing for the detection of gross leakage. In addition, drywell thickness inspections provide some assurance that there is no structural degradation u hich would affect the containment's ability to withstand accident pressures.

Based on the above, the risk impact of the deferral of the Oyster Creek ILRT from the 17R refueling outage to the 18R refueling outage is assigned to the Low risk impact categosy and does not affect the integrated assessment performed in reference 1.

5.0 REFERENCES

1. GPU Nuclear Corporation, " Oyster Creek Nuclear Generating Station, Long Range Plarming Program (LRPP), Supplemental Update of Integrated Schedule",6700-97-3042, October 1,1997.

2. GPU Nuclear Corporation, " Technical Specification Change Request No. 242, Revision 2:

Implementation of 10 CFR 50 Appendix J, Option B",6730-96-2228, July 17,1996.

3. Nuclear Regulatory Commission, " Oyster Creek - Issuance of Amendment RE: Implementation of 10 CFR Part 50, Appendix J, Option B (TAC No. M94855), September 3,1996.

4. GPU Nuclear Corporation, " Oyster Creek Probabilistic Risk Assessment (Level 1)", Volumes 1 through 6, November 1991.

5. GPU Nuclear Corporation, " Oyster Creek Probabilistic Risk Assessment (Level 2)", June 1992.

6. Nuclear Regulatory Commission, " Performance-Based Containment Leak-Test Program",

NUREG-1493, Final Report, September 1995.

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7. Pacific Northwest Laboratories, " Review of Light Water Reactor Regulatory Requirements",

NUREG/CR-4330, PNL-5809, Volume 2, June 1986.

8. GPU Nuclear Corporation, " Implementation of Appendix J Option B for Primary Containment Leakage Testing", Safety Evaluation, SE-000240-002, Revision 0, February 13,1996.

9. Electric Power Research Institute (EPRI), "Probabilistic Safety Assessment Applications Guide",

EPRI TR-10';96, Final Report, August 1995.

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'l ENCLOSURE 1:

COMPARISON OF PEACH BOTTOM AND OYSTER CREEK BASIC RCS AND CONTAINMENT FEATURES l

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I Table 2: Basic RCS and Containment Comparison Table i' , Plant Name Peach Bottom Oyster Creek Type of Reactor BWR/4 BWR/2 Type of Containment Mark I Mark I ~

l Reactor Core l

l Thermal Power (MWl) 3,293 1,930 l NumberofFuct Assemblics 764 560 i Number of Control Rods 185 137 Reactor Vessel l Inside Diameter (inches) 251 213 l Inside Height (feet) 72.92 64 l Design Pressure (psig) 1,250 1,250 l Number of Safety Valves 2 9 l Lowest Safety Valve Sctpoint (psig) 1,230 1,224 l Safety Valve Capacity (kib/hr) 925 654 l Number of Relief Valves 11 5 l Lowest Relief Valve Setpoint (psig) 1,105 1,070 l Relief Valves Capacity (kib/hr) 889 558 i Reactor Coolant Recirculation Number of Loops 2 5 Number of Pumps 2 5 Inlet Pressure (psig) 1,148 1,025 l Outlet Pressure (psig) 1,326 1,065 i

Number of Jet Pumps 20 0 l l Flow Rate per Pump (gpm) 45,200 32,000  !

l RHR System l

Number of Loops 2 3

Number of Pumps 4 3 l Flow Rate per Pump (gpm)

. 10,000 3,000 Number of Heat Exchangers 4 3 Maximum Capacity of Heat l Exchanger (Btu /hr) /Two Heat Exchanger Set 70,000,000 11,000,000 per Heat Exchanger ]

RHR Service Water System Number of Pumps 3 2 l

! Flow Rate per Pump (gpm at psid) 14,000 6,000 Co e Isolation Cooling System Type RCIC None Capacity (gpm at psid) 616 at 1,120 ,

Emergency injection Systems Number of HPI Pumps 1 Flow Rate per Pump (gpm at psid) 5,000 at 1,120  ;

Number of LPI Pumps 4 0 Flow Rate per LPI Pump (gpm at psid) 10,000 at 20 i Number of Core Spray Pumps 4 4 Flow Rate per Core Spray Pump (gpm at psid) 3,125 at 122 3,400 at 110 lirtdef4 C-1 05/12/98

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t Table 2: Basic RCS and Containment Comparison Table i

, Plant Name Peach Bottom Oyster Creek -

Type of Reactor BWR/4 BWR/2 Type of Containment Mark I Mark I Containment Drywell Material and Construction Steel Steel Drywell Free Volume (ft2) 175,800 180,000 Drywell Design Temperature (oF) 281 281 Wetwell Materialand Construction Steel with Steel Steel Liner Wetwell Minimum Free Volume (A3) 127,000 126,000 Wetwell Minimum Water Volume (R3) 123,000 82,000 Wetwell Design Temperature (oF) 281 150 Design Pressure (psig) 56 35 Vent Configuration Diagonal large- Diagonallarge-diameter pipes ending diameter pipes ending in ram's head in ram's head distribution manifold distribution manifold and vertical piping and vertical piping venting below the water venting below the level of the pool. water level of the pool.

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