ML20092H831

Forwards Response to NRC 831207 Request for Addl Info Re NUREG-0737,Item II.B.2, Design Review of Plant Shielding, Per IE Insp Rept 50-213/83-13.Addl Shielding Studies Will Be Completed by Sept 1984

ML20092H831
Person / Time
Site:	Oyster Creek
Issue date:	06/21/1984
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.2, TASK-TM NUDOCS 8406260287
Download: ML20092H831 (2)

Text

, e GPU Nuclear Corporation Nuclear =g;;;a88 Forked River New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:

Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 NUREG-0737, Item II.B.2 Design Review of Plant Shielding Your letter of December 7,1983 requested additional information relative to NUREG-0737, item II.B.2 " Design Review of Plant Shielding." The questions were generated as a result of Region I Inspection Report No.

50-219/83-13 which reviewed our previous submittals dated January 4,1980 and April 10,1980.

Responses to questions 1 and 5 are provided as an enclosure to this letter. The remaining questions require that additional shielding studies be performed. A contractor has been retained to perform the necessary studies and completion is expected in September 1984. The results and responses to the remaining questions will be forwarded to you shortly thereafter.

Should you require any further infomation on this subject, please contact Mr. Michael Laggart, BWR Licensing Manager at (201)299-2341.

Very truly yours, M.2 P M g #["O ter . Fiedler Vice President and Director Oyster Creek l PDF/ dam Enclosures cc: Dr. Thomas E. Murley, Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, PA 19406 l NRC Resident Inspector s l Oyster Creek Nuclear Generating Station Forked River, NJ 08731

, 8406260287 840621 i

PDR ADOCK 05000219 U PDR l GPU Nuclear Corporation is a subsidiary of the General Pubhc Utilities Corporation

r ENCLOSURE [

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Question 1:

Source Terms were identified as being derived from GE data and documented in l Computer Run SNUMB-7007S dated November 9,1979. Who perfomed this

calculation, and what. type of code was used (for example, ORIGEN)? [

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Response:  ;

Radiation Source Terw Infomation for NilREG-0578 Implementation, General Electric Company, November 1979, Computer Code RIB-D was used to derive the source terms.

4 Question 5:

Explain the reasoning contained in the April 15, 1983 GPU letter to NRC f

! (P.B. Fiedler to D. Eisenhut) entitled " Cycle 10 Refueling Outage Workload", t for cancellation of the commitment for a SGTS Filter Tie-in. Specifically,  !

what analyses have been performed to conclude that: (1) a single SGTS filter i train is capable of handling (without changeout) effluent loading associated j with an excessive MSIY leakage accident, and (2) whether radiological source 1 contribution need be considered for any vital areas (such as the Security l Building) from such a source? l t

Response: l i

By letter dated April 15, 1983, GPU Nuclear provided justification for  !

j cancelling a proposed modification to the Standby Gas Treatment System [

(SGTS). The justification, as stated in the letter, was based upon the NRC's i i

evaluation of SEP Topic XV-19, " Radiological Consequences of a loss of Coolant  !

i Accident." GPU Nuclear was provided with the results of that evaluation by i i letter from Mr. Dennis Crutchfield to P. B. Fiedler dated September 2,1983. .

! A copy of this letter is attached for your convenience. The evaluation  !

i demonstrates that a single SGTS is capable of handling effluent loading.

> Please note in the evaluation that Main Steam Isolation Valve through valve ,

L leakage is not routed to the SGTS. I i  :

! In reponse to the second part of Question No. 5, the filters for the SGTS are located below grade in a concrete pipe tunnel. Because of the obvious f j shielding between this location and other vital areas we did not consider it necessary to consider these components as radiation sources for other vital  !

areas. }

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[gmc ,o UNITED STATES

!N . [j NUCLEAR REGULATORY COMMISSION 2~, c WASWNGTON, D. C. 20555 t ,. / September 2,1982 Docket No. 50-219 -

LS05-82-09-011 Mr. P. B. Fiedler, Vice President and Director Oyster Creek Nuclear Generating Station ,

Post Office Box 388  ;

Forked River, New Jersey 03731

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION, SAFETY EVALUATION OF SEP TOPIC XV-19, RADIOLOGICAL CONSEQUENCES OF A LOSS OF COOLANT ACCIDENT Enclosed is the staff's final evaluation of SEP Topic XV-19 for the Oyster Creek Plant. The evaluation has been revised from the draft evaluation sent to you on June 29, 1982, based on information supplied by you. The revised portions have been marked with a line in the right hand column to designate the changes. The staff now estimates that the 30 day low population zone (LPZ) doses could exceed the allowable specified in 10 CFR 100 by approximately 14% (341 vs. 300 rem) instead of 20% as in the draft evaluation. Since the activity leakage pathway that contributes over 95% (334 rem) of the estimated dose is still from the main steam isolation valve (MSIV) leakage the recommendations outlined in the draft evaluation are still valid.

These are as follow:

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 MSIV leakage then should be lower than assumed by the staff.

2. Evaluate 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 pathways to less than 300 rem.

This evaluation will be a basic input to the integrated safety assess-ment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This assessment may be

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P. B. Fiedler revised in the future if your facility design is changed or if HRC criteria relating to this subject is modified before the integrated assessment is cortpleted.

Sincerel:

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• Dennis M.Reactors Operating Crutchfield, Bran,c;hief h No. 5 Division of Licensing f

Enclosure:

i As stated cc w/ enclosure:

See next page 1

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Mr. P. B. Fiedler ,

i 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 o' Nuclear Regulatory Commission, Region 1 631 Park Avenue King of Prussia, Pennsy1.vania 19406 J..Xnubel .

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 West State Street - CN 112 Trenton, New Jersey 08625 Mayor Lacey Township 818 Lacey Road Farked 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 Post Office Box 388 Forked River, New Jersey 08731 e e e

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. i XV-19 LOSS-OF-COOLANT ACCIDENTS RESULTING FROM A SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR' COOLANT PRESSURE BOUNDARY (RADIO-LOGICAL CONSEQUENCES) - OYSTER CREEK 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,i.em. LOCA's

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result in excessive fuel damage or melt unless coolant is .rehenished.

Excessive fuel damage can result in significant radiological consequences to the environment via leckage from*the containment. SEP Topix XV-19 is

' 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. .

11. 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 froh 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.

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. 1 In addition, 10 CFR Part 100.11 provides dose guideline values for reactor l

. l siting assessments.

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i III. R' ELATED SAFETY TOPICS ,,

fi Topic II-2.C. " Atmospheric Tvansport and Diffusion Characteristics for Acci- i 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-

down or to mitigate the consequences of an accident are maintained. Various other topics examine such areas as containment integrity and solation, post accident chemistry, ESF systems, combustible gas control and control room I

habitability. .

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i IV. REVIEW GUIDELINES' I

. 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 citigating features are acceptable only if the resulting doses at the exclusion area and low population zone bounda' ries are within the guideline values of l

10 CFR Part 100.  :

V. EVALUATION

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

sizes. The submittal, however, did not provide sufficient detail to permit - l an. independent analysis and questions were sent to the licensee on April 7, 1982 by teletype. Based on the licensee's response to the questions dated i April 28,1982 (in a letter from Drew G. Holland of GPU Nuclear to Robert Fell e

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S 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 s ources:

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 conclusions reached in this evaluation, the licensee should confirm this statement by submitting the details on how each leakage path was considered in arriving at the conclusion that no containment leakage bypasses the area processed by the SGTS.) The calculated dose f rom contai nt leakage is derived solely from the 0.5% per day Technical Specification leakage limit from the primary containment, complete mixing in the secondary containment and then processing by the SGTS prior to release '.o the environment.

Based on information provided by the licensee on filter efficiencies, the I

! staff has determined that an appropriate value for the filter eff,1ciencies f is 90%.

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 i

l was determined from the acceptance criteria of the plant's test program for i

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these valves. 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 i for 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

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

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3. Post-LOCA Leakage from ESF Systems Outside Primary Containment: Because the l 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 total leakage in the calculation of the ESF component leakage contribution to the LOCA doses.

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

5 seconds. The licensee should submit confirmation of these plans and a schedule for their installation. The staff has evaluated the potential l

I contribut' ion to the LOCA dose from operation of the purge system during the e

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onset of an accident and has determined that the contribution is much less than 0.1 rem and, therefore, is negligible.

VI. CONCLUSION The calculated doses and assumptions used to arrive at these doses are presented in Table XV-1 and XV-2, respectively. The evaluation indicates that the 0-30 day LPZ thyroid dose guideline is exceeded by approximately 14%. The staff notes that a major portion of this dose is attributed to i MSIV leakage. As noted earlier, the licensee needs to provide information to support the statement (in the April 28, 1982 letter) that no containment leakage bypasses the area served by the SGTS.

The staff concludes that because of the uncertainties in the calculation of l

the doses and because the estimated thyroid doses exceed the 0-30 day 10 CFR 100 thyroid dose LPZ guideline value by only approximately 14%, any l

! plant backfit considerations can be appropriately 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

g. Hrs. Hrs. Rem Rem Rem Rem

!! 0.0 2.0 3.8 0.1 1.4 0.1

zW gjll LJ J 2.0 4.0 - -

3.8 0.1 4.0 8.0 - - 0.3 0.1 8.0 24.0 - -

0.2 0.1 24.0 96.0 - - 1.0 0.2 96.0 720.0 - - 0.4 0.2 w

+ 37.5 96.0 - - 170 0.1 i!'; MR W2 "i 96.0 720.0 - - 164 0.1 uJ 0.0 2. 0 < 0.1 < 0.1 - .-

g;ll kJ J 0.0 720.0 - - 0.01 <0.01 Total LOCA doses 3.8 0.2 341 1.0 *

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

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

1. Reacter stretch power (Mwt) 1934

2. Fission product release fractions (percent)

a. Iodines 25

b. Noble gases 100

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

4. Primary containment leak rate (%/ day) 0. 5

5. SGTS filter efficiency (percent) 90 (all forms of iodine)

6. MSIV leak rate (scfh) 11.5

7. SGTS bypass leakage 0

8. ESF leakage into reactor building (gpm) 1. 0

9. Purge system flow rate (cfm) 1000

10. Time required for purge system isolation (sec) 5 11 X/Q's (sec/ cubic meters) t -Ground level release for MSIV leakage I-0-2 hour EAB* (414 m) 7.6 E-4***

• 0-8 hour LPZ** (1208 m) 6.5 E-5 I 8-24 hour " 4.3 E-5 l 1-4 day 1.7 E-5 l

4-30 day 4.8 E-6 Elevated release used for containment leakage (fumigation conditions) l 0-2 hour EA8 1.1 E-4 l 0-4 hour LPZ 4.2 E-5 l

[ (non-fumigation conditions) 4-8 hour LPZ 9.1 E-7 8-24 hour LPZ 2.5 E-7 1-4 day LPZ 1.7 E-7 4-30 day LPZ 2.5 E-7

• Exclusion Area Boundary (10 CFR 100)

• • 0 uter boundary of Population Zone (10 CFR 100) l ***7.6 E-4 = 7.6 x 1 = .00076

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