ML20054K851

Forwards Safety Evaluation of SEP Topic XV-19, Radiological Consequences of Loca. Thirty-day Low Population Zone Doses Could Exceed Allowable Specified in 10CFR100

ML20054K851
Person / Time
Site:	Oyster Creek
Issue date:	06/29/1982
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Fiedler P CAROLINA POWER & LIGHT CO.
References
TASK-15-19, TASK-RR LSO5-82-06-123, LSO5-82-6-123, NUDOCS 8207060220
Download: ML20054K851 (10)

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June 29, 1982 Docket No. 50-219 LS05-82-06-123 Mr. P. B. Fiedler Vice President and Director - Oyster Creek Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION, SAFETY EVALUATION OF SEP TOPIC XV-19, RADIOLOGICAL CONSEQUENCES OF A LOSS OF COOLANT ACCIDENT General Pubite Utilities letter dated November 4,1981, transmitted for our review your Safety Analysis Report for SEP Topic XV-19, Radiological Consequences of a Loss of Coolant Accident.

Enclosed is our evaluation of this topic. The staff conservatively estimated that the 30 day low population zone (LPZ) doses could exceed the allowable specified in 10 CFR 100 by approximately 20% (357 rem vs. 300 rem). The activity leakage pathway that contribute over 90% (334 rem) of the estimated dose is from main steam isolation valve (MSIV) leakage, assuming that the main steam system integrity is lost downstream of the isolation valves at the turbine stop valves.

In light of the significant MSIV dose pathway it is recommended that GPU evaluate the following items:

1.

Perform a more realistic analysis for MSIV doses factoring in the effects of drywell pressure vs. MSIV leakage rate as a function of time. The total itSIV leakage then should be lower than assumed by the staff.

2.

Evalur.te the merits of directing the turbine building ventilation exhaust through a charcoal filter system.

3.

Evaluate the merits of installing MSIV leakage prevention systems.

4.

Any other procedure or system modifications that will limit the total LOCA doses from all pathwaysto less than 300 rem.

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This topic evaluation will be a basic input to the integrated assessment.

The evaluation may change if your facility is changed or if NRC criteria are modified before completion of the integrated assessment.

Sincertly, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing

Enclosure:

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Mr. P. B. Fiedler CC G. F. Trowbridge, Esquire Resident Inspector Shaw, Pittman, Potts and Trowbridge c/o U. S. NRC 1800 M Street, N. W.

Post Office Box 445 Washington, D. C.

20036 Forked River, New Jersey 08731 J. B. Lieberman, Esquire Commissioner Berlack, Israels & Lieberman New Jersey Department of Energy 26 Broadway 101 Commerce Street New York, New York 10004 Newark, New Jersey 07102

, Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I

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631 Park Avenue King of Prussia, Pennsylvania 19406 J..Knubel BWR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, New Jersey 07054 Deputy Attorney General State of New Jersey Department of Law and Public Safety 36 W.st State Street - CN 112 Trenton, New Jersey 08625 Mayor Lacey Township 818 Lacey Road Forked, River, New Jersey 08731 U. S. Environmental Protection Agency Region II Office ATTN:

Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Licensing Supervisor Oyster Creek Nuclear Generating Station

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Post Office Box 388 Forked River, New Jersey 08731 I

XV-19 LOSS-0F-COOLANT ACCIDENTS RESULTING FROM A SPECTRUM OF POSTULATED PIPING BREAKS WITHlN THE REACTOR' COOLANT PRESSURE B0UNDARY (RADIO-4 LOGICAL CONSEQUENCES)

I. INTRODUCTION

' Loss-of-coolant accidents (LOCA's) are postulated breaks in the reactor coolant pressure boundary resulting in a loss of reactor coolant at a rate, in excess of the capability of the reactor coolant makeup sys' em.

LOCA's t

result in excessive fuel damage or melt unless coolant is te lenished.

Excessive fuel damage can result in significant radiological consequences to the environment via leakage f~ om~the containment.

SEP Topix XV-19 is r

' intended to assure that the radiological consequences of a design basis LOCA from containment leakage,,ESF leakage, containment purge and leakage through the main steam isolation valves (MSIV's) are within the exposure guideline values of 10 CFR Part 100.

I II.

REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility. The LOCA is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety.

In addition,10 CFR Part 100.11 provides dose guideline values for reactor siting assessments.

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R' ELATED SAFETY TOPICS Topic II-2.C, " Atmospheric Transport and Diffusion Characteristics for Acci-dent Analysis" provides the meteorological data used to evaluate the offsite doses.

Topic III-5. A, " Effects of Pipe Breaks on Structures, Systems and ~

. Components Inside Containment" ensures that the ability to achieve safe shut-

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down or to mitigate the consequences of an accident are maintained.

Various other topics examine such areas as containment integrity and isolation, post accident chemistry, ESF systems, combustible gas control and control room habi tabili ty.

IV. REVIEW GUIDELINES The review of the radiological consequences of a LOCA was conducted in accord-ance with Appendices A, B, and D to Standard Review Plan 15.6.5 and Regulatory Guide 1.3.

The plant is adequately designed against a LOCA and the dose mitigating features are acceptable only if the resulting doses at the exclusion area and low population zone bounda' ries are within the guideline values of 10 CFR Part 100.

V.

EVALUATION In the licensee submittal to NRC, the licensee provided a full spectrum of loss-of-coolant accidents as a result-of various primary system pipe break ~

sizes.

The submittal, however, did not provide sufficient detail to pemit an independent analysis and questions were sent to the licensee on April 7, 1982 by teletype.

Baseri on the licensee's response to the questions dated April 28,1982 (in a letter from Drew G. Holland of GPU Nuclear to Robert Fell

, of NRC)', the staff performed an analysis of the radiological consequences according to the current NRC criteria.

The radiological consequences of this accident result from the'following sou rces:

1. Containment Leakage:

The licensee in his April 28, 1982 letter indicates that there is no containment leakage which bypasses the SGTS filters.

(Because any bypass leakage paths can alter the conclusiony reached in this evaluation, the licensee should confirm this statement by hubmitting the details on how each leakage path was considered in arriv.ing at the conclusion that no containment leakage bypasses the area process,ed by the SGTS.) The calculated dose from containment leakage is derived solely from the 0.5% per day Technical Specification leakage limit from the primary containment, r

complete mixing in the secondary containment and then processing by the SGTS prior to release to the enviroiiment.

Pending receipt of the following infor-mation (requested via telephone, C Nichols of NRC to Yosh Nagi of GPU Nuclear):

a. the SGTS charcoal filter thickness;

b. the SGTS charcoal filter material; and l

C. the residence time of tn'e air stream in the charcoal filter.

1 The staff has used the SGTS charcoal filter efficiency assumed by the licensee.

This value (99%) is the highest allowed by NRC.

Should NRC disagree with this value after review of the abqve requested information, the calculated doses would rise.

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2. Main Steam Isolation Valve Leakage:

Oyster Creek does not have a main steam isolation valve leakage control system (MSIV-LCS).

In our analysis, we have assumed that the MSIV's leak at a rate of 11.5 scfh.

The value of 11.5 scfh was determined from the acceptance ' criteria.of the plant's test program for i

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The staff has estimated that a holdup of fission products will occur in the 100 foot section of main steam piping between the outboard isolation valve and the turbine stop valves. Leakage is assumed to occur at ground level.

The resulting 0-30 day LPZ doses based on the 11.5 scfh per MSIV is 334 rem f or the thyroid and 0.2 rem whole body. The length of the. main' steam pipe section between the outboard main steam isolation valve and the turbine stop valves is critical to this conclusion. The estimated length of pipe (100

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feet) was supplied by the licensee, and because of its importanca to the calculation, should be verified by the licensee.

3. Post-LOCA Leakage from,ESF Systems Outside Primary Containment:

Because the ECCS leakage will be to the reactor building and the SGTS includes an ESF grade filtration system which filters the reactor building exhaust, we have not calculated the doses from passive failures (according to Appendix B to Standard Review Plan Section 15.'4.5).

We have calculated the doses resulting from anticipated operational leakage. No Technical Specification limit on the l

leakage from ESF systems outside containment exists.

We have assumed one gpm i

l total leakage in the calculation of the ESF component leakage contribution to l

the-LOCA doses.

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4.

Containment Purge:

The existing purge valves will close in about one minute from an initiating signal.

The licensee in his April 28, 1982 letter indicates plans to replace these valves with ones that will close within 1

5 seconds.

The licensee should submit confirmation of these plans and a schedule for their installation.

The staff concludes that with the above modifications, the purge dose contribution will be small. However, because l

the licensee has not responded to all the questions regarding the purge

t system, the staff is unable to reach any ' conclusions as the effect of this contribution in terms of comparison with the exposure guidelines of 10 CFR Part 100.

VI. CONCLUSION The calculated doses and assumptions used to arrive at these doses are presented in Table XV-1 and XV-2, respectively.

The evaluati,on indicates that the 0-30 day LPZ thyroid dose guideline 10 exceeded by' ahproximately 20%. The staff notes that a major portion of this dose can,be attributed to MSIV leakage.

As noted earlier, the licensee needs to provide information to confirm the assumed SGTS filter efficiency, to complete the purge dose contribution and to support the statement (in the April 28, 1982 letter) t' at no containment leakage bypasses the area served by the SGTS.

The staff concludes that because of the uncertainties in the calculation of the doses and because the doses exceed the 0-30 day LPZ guideline value by only approximately 20%, any plant backfit considerations should be pursued during the integrated assessment.

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TABLE XV.-1 RADIOLOGICAL CONSEQUENCES OF A LOCA AT OYSTER CREEK Duration Exclusion Area Boundary Low Population Zone From To Thyroid Whole Body Thyroid Whole Body Hrs.

Hrs.

Rem Rem Rem Rem

+5 0.0 2.0 2.6 0.8 0.2 0.1 iE m 2E 2.0 8.0 2.6 0.4

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O.4 8"!

8.0 24.0 24.0 96.0 9.4 0.2

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7.0 0.1 96.0 720.0 w

aE 37.5 96.0 170 0.1 3M E"i 96.0 720.0 164 0.1 w

E 0.0 2.0 (0.1

<0.1 u.M 0.01

<0.01 0"i 0.0 720.0 Total LOCA doses 2.7 0.9 357 1.4 i

• The leakage from this source is assumed to start 37.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the accident l

and,' therefore, there is no contribution to the EAB dose.

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TABLE XV-2 ASSUMPTIONS USED IN THE ANALYSIS OF THE RADIOLOGICAL CONSEQUENCES OF A LOCA AT OYSTER CREEK

• 1. Reactor stretch power (Mwt) 1934

2. Fission product release fractions (percent)

a. lodines 25

b. Noble gases 100, ".'

3. Primary containment volume (cubic feet) 180,000

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4. Primary containment leak rate (%/ day) 0.5

5. SGTS filter efficiency (percent) 99' ' '

(all forms of iodine)

6. MSly leak rate (scfh) r 11.5

7. SGTS bypass leakage 0

8. ESF leakage into reactor building (gpm) 1.0 1

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