IR 05000219/1985002

Forwards Insp Rept 50-219/85-02,documenting 850108 Meeting W/Util in Forked River,Nj Re Unacceptably High Integrated Whole Body Dose of 62 Rems Predicted at Main Security Bldg

ML20205A296
Person / Time
Site:	Oyster Creek
Issue date:	04/11/1985
From:	Donohew J Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737 LSO5-85-04-012, LSO5-85-4-12, NUDOCS 8504250472
Download: ML20205A296 (2)

Text

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April 11, 1985

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Docket No. 50-219 LS05-85-04-012 LICENSEES: GPU Nuclear Corporation Jersey Central Power and Light Company FACILITY: Oyster Creek Nuclear Generating Station SUBJECT: MEETING WITH GPU NUCLEAR ON NUREG-0737, II.B.2, PLANT SHIELDING STUDY On Tuesday, January 8, 1985, a meeting was held at GPU Nuclear's (the licensee's) contractor's place of business. Enclosed is the inspection report which documents what happened at this meetin LL h \

ack N. Donohew, Jr. , roject Manager Operating Peactors Bra ch #5 Division of Licensing

Enclosure:

NRC Inspection Report 50-219/85-03

REGION I==

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631 PARK AVENUE g '%#

. . . .,o. 'f KING oF PRUSSIA, PENNSYLVANIA 19406 Docket No. 50-219/DPR-16 GPU Nuclear Corporation ATTN: Mr. P. B. Fiedler Vice President and Director Oyster Creek Nuclear Generating Station-P. O. Box 388 Forked River, NJ 08731

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Gentlemen:

_ Subject: Inspection 50.-219/85-03 .

This letter refNs to the special safety inspection by Mr. E. Kelly on January 8,-1985, at the United Engineers and Constructors, Inc. Philadelphia, Pennsyl-vania office. The inspection consisted of a review of calculations and the results of your shielding study in response to NUREG-0737, Item II.B.2 for Oyster.. Creek.. The findings were discussed with Mrr+1. Laggart of your staff at the conclus. ion of the inspectio . .

,' With the exception of one item, your. shielding studr was found to miet the

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requirements' of TMI Item II.B.2. Therefore, all open items associated with your study, previously documented in Inspection Repont-83-13, have been cloTe However, the results of your study predict a whole body dose of 62 Rem inside of the mairi security'. building. This will be carried as an open item until you propose an ace'ptable e resolutio No reply is requ' ired, and your cooperation with us in this matter is appre- -

ciate .,

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Sincerely,

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, j arry ister, Cn(ief Projects ranch No. I Division of Reactor Projects

Enclosure:

NRC Region I Report 50-219/85-03

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REGION I==

Report No. 50-219/85-03 Docket No. 50-219 License No. DPR-16 . Priority --

Category C Licensee: GPU Nuclear Corporation 100 Interpace Parkway -

Parsippany, New Jersey 07054 Fa;111ty NA.me: Oyster Creek Nuclear Generating Station

Inspection At: United Engineers and Constructors, Inc. Philadelphia, Pa. Office Inspection Conducted: January 8, 1985 Inspecto.r: . .  ! - -a --

E. M. Kegly, Project Enginger, RPS IL .. Date -

Division'of Reactor Projects

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Approved By:' . .

bW Walter Baunack, Acting Chief

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Reactor Projects Section No. lA,

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' Division of Reactor Projects .

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Inspection Summary:

This'Ipec'falinspectionbyoneregion-basedinspector(fivehours)reviewedthe re-analyses of shielding requirements at Oyster Creek in response to NUREG-0737, Item II.B.2, Desig'n Review of Plant Shielding. Four previous open items, from the initial review of the required analyses during Inspection 50-219/

83-13, were close One open item was identified: an unacceptably high integrated whole body dose of 62 Rem predicted at the main security building. The licensee has committed to have their contractor re-analyze this location, using more realistic t.ssumptions regarding occupancy and drywell leakag G

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.' DETAILS Persons Contacted M. Laggart, GPU Licensing B. Hohman, GPU Licensing J. Boorboor, UE&C Nuclear Analysis Group Manager A. Friedman, UE&C'L,1c'ensing Manager-R. Siu, UE&C Senior-Engineer Background and Scope - - - *

Item II.B.2 of NUREG-0737 required a radiation and shielding design review

_. of spaces near~ sysstems that may become contaminated during the course of an accident... That review was intended to identify vital areas - those which require occupancy for recovery or mitigation - in order to assure

- adequate access by means of design changes, shielding or procedural controls. The predicted integrated dose to an individual in those areas identified as vital was required to be within the 5 Rem limit of General Des.ign Criterion 19 for the duration of the ecfteen .

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Inspection 50-219/83-l'3 assessed the General Public Utilities Nuclear j' (GPUN)' corporation shielding design review for dyster Creek originally

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outlined in a January 4,1980 let.ter to the NRC. 'That inspection con-cluded that additional information was required, cecessitating a-re-anaTy-sis. A . number of quastions were identified in that report (0 pen Items 83-13-01 throughiO4)-which involved: .

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orig.in,of core source term calculations;,

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consideration of airborne concentration in the reactor building; -

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specification of vital areas and associated doses;

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cancellation of Standby Gas Treatment System (SGTS) fil.ter tie-in

..t..- modificatio '

These questions wer.e presented to GPU in a December 7, 1983 letter from NRR, and the itcensee's response dated June 21, 1984, answered, in part, and committed to additional shielding studies. United Engineers was then retained to perfctm a re-analysi This inspection reviewed those result l

The cancellation of the originally proposed SGTS tie-in was assessed i during an inspection at UE&C offices on September 11, 1984, and documented i in Detail 9 of Inspection Report No. 50-219/84-28. That inspection found the bases. used by GPUN to cancel the modification to be justifie Those <

bases were summarized in a September 18, 1984 GPUN letter to NRR and approved.by NRC in a safety evaluation issued to GPUN by NRR on October 2, 1984.

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. Calculations UE&C utilized an in-house version of computer code QAD-CG, with a revised numerical integration option, to model post-accident sources. The prin-cipal sources were the isolation condenser and the reactor. building upper space. The latter was the major contributor to vital area dose rate UE&C code CCC-448/QAD-UE was benchmarked and accepted as a documented code as reported in the February 1984 Radiation Shielding Information Center (RSIC) newslette ~ '

The results of UE&C calculation set Nos. 7450-111-51 (December 1984) and 54 (January 1985) were reviewed with the cognizant engineer, R. Siu. The

_ . formen calculation, addressed doses due to, airborne activity inside the reactor building, while the latter detai. led the overall assessment of access evalua tions for vital areas, including dose map: and integrated exposures. The selected vital areas were found to have maximum dose rates and integrated 30-day exposures as follows:

. _ , . . .Ma'ximum Dose Integrated.Shele Body -- .

Vital Area Rate (mR/hr) ExposureICRem) . Occupancy -

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Control Room Security Bld , Continuous

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(Main Gate) 960 . 62 ._ Continuous -

Diesel Bld .1 negligible Intermittent <

PASS Room' .. 30 - within five Intermittent '

Hot Chemistry Lab 3,100 within five (Note 1) Intermittent TSC 13 negligible Continuous Stack RAGEMS' 1,300 0.7 (Note 2) One-Time .

(inside)

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Turbine Bld .8 (Note 2) One-Time RAGEMS (outside)

' Alternate Hot Lab 60 within five Intermittent Notes: Short duration access (less than h hour) for first 24 hr . Assuming 10 minute round trip transit and 20 minute service time to change nitrogen bottles for Radiation Analyzer and Gaseous Effluent Monitoring System (RAGEMS).

With the exception of the Main Security Building, all vital areas were appropriately identified, and all were predicted to meet the 5 Rem limit for occupancy for the duration of the accident or expected intermittent sta The Security Building, with current analytical assumptions, does not meet the exposure criterio Subsequent phone conversations with GPUN licens-ing representatives on January 9 and March 11, 1985, co.cluded that more realistic assumptions will be employed to ascertain if predicted doses

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could.'be lowered to within the 5 Rem limit. These would include: (1)

pressure-dependent, pr.imary to secondary leakage rates; and (2) use of occupancy factors in the building, similar to control room habitability studies. The resolution of the post-accident radiological habitability of the Main Security Building will be followed as an unresolved item (50-219/85-03-01). Conclusions -

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All assumptions employed were found to be conservative, reasonable and l accurately reflective of plant desig It is concluded that the results of GPUN's shielding study satisfactorily meet the requirements of NUREG -

0737, Item II.B.2, contingent upon resolution of the habitability of the  !

_ .. main security. building. Therefore, unresolved Items 50-219/83-13-01 through 04 are considered close .. Exit Interview

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The results of thi.s in.spection were discussed with M. Laggart, GPUN Licensing, at the conclusion of this inspecttaq"ind-again (for the-- -

proposed approach to reducing the Main Securit9LBuilding exposure pre '

, diction) in phone conversation on January 9 and March 11, 1985. _

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