ML20091B893

Forwards Response to 840313 Telcon Request for Addl Info Re NUREG-0737,Item II.B.3, Post-Accident Sampling Capability. Procedures for Estimating Extent of Core Damage Based on Methodology in NEDO-22215

ML20091B893
Person / Time
Site:	Hatch
Issue date:	05/22/1984
From:	Gucwa L GEORGIA POWER CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NED-84-235, TAC-44445, TAC-44446, NUDOCS 8405300431
Download: ML20091B893 (5)

Text

G3orcia Power Compary 333 Piec*vont Avenue Atlanta, George 30309 Telephone 404 5264526 Madng Address:

Post Off;ce Box 4545 AHanta, Georgta 30302 Georgia Power L T. Gucwa the southem elecinc system Manager Nuclear Enginee""9 NED-84-235 and Chief Nuclear Eng.neci May 22, 1984 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz, Chief Operating Reactors Branch No. 4 Division of Licensing U. S. Nuclear Regulatory Ccenission Washington, D. C.

20555 NBC DOCKHIS 50-321, 50-366 OPERATING LICENSES DPP-57, NPF-5 EIMIN I. HATCH NUCLEAR PLANT UNITS 1, 2 NUREG-0737 ITEM II.B.3 POST-AOFIDENT SAMPLING CAPABILITY Gentlenen:

Our submittal dated January 26, 1984 prwided information demonstrating the degree to which Plant Hatch cmplied with the criteria of NUREG-0737, Iten II.B.3.

In a March 13, 1984 telephone conference, Georgia Power Cmpany ws raluested to submit additional information regarding certain criteria. In response to that request, Enclosure 1 is submitted.

Please contact this office if there are any questions.

Very truly yours, f f~ Q 1_ =

L. T. Gucwa JH/mb Enclosure xc:

J. T. Beckhan, Jr.

H. C. Nix, Jr.

J. P. O'Reilly (NBC-Region II)

Senior Resident Inspector Oh jo A

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4 ENCIN URE 1 E[EIN I. HNIGI NUCEAR PLANT UNI'IS 1 AND 2 POST-ACCIDENT SAMPLING CAPABILI'IY 1

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INmGXKTIN Post-accident.s apling and analysis capability in accordance with NUREG-0737 Its II.B.3 is.being provided by previously existing in-line contaiment hydrogen and oxygen analyzers in conjunction with the Post-Accident' Smpling Systm (PASS).

Georgia Power Cmpany's January 26,1984 sutwittal provided detailed information regarding cmpliance

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with the criteria of Its II;B.3. _ The Linformation provided herein suppleents the January 26, 1984 sutmittal in respcnse to NRC concerns in the following areas:'

A.

Representative sa pling; B.

Instrment range and accuracy; C..

Personnel radiation exposure; and D.

Core daage estimation.

II. ADDITIONAL INFOWATION.

A.

REPRESENTATIVE SAMPLING

' Reactor coolant smples are obtained frm a jet pump flow-sensing instrment line over the entire range of primary syste pressures.

While scrae BWR licensees have found it necessary to utilize an RHR system tap for. smpling at low. pressure, this was

. c not necessary for Plant Hatch ~ because sufficient motive force for smple flow is provided by the PASS saple pmp and the elevation drop of the smple lines.

Smples taken fra the jet pmp location will be representative of coolant. in the-core because a direct path will always exist from the core region to the smple tap.

For the case of a'small break or no break, reactor water level will be maintained at or' near the normal level. -In the absence of forced reactor recireplation.

flow, operators will be instructed by procedures.to maintain reactor water level high-enough ~ to ensure natural circulation.

She flow of water up through the core' then down past the tap frm which the PASS smple - is taken will ensure that ' the. smple is representative.of coolant in the core.

In the-case of :a ? large break, water -supplied by the Core Spray system will flow ' down through the core then up through the. jet pumps past ithe. tap frm which _ the PASS sample is taken.. Again, flow past the' smple tap.diractly fra the-core.will ensure that

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the saple is representative of core conditions. -

1 Representative saples of drywell atmosphere-will,be assured by heat : tracing ; maintaining' an - enviromentl of approximate 1y' 2500F J

tin the sample lines which will prevent condensation and plateout.-

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INSTRLMENT RANGE AND ACCURACY Tables 1, 2 and 3 provide stmnaries of the. instrtment ranges and accuracies for the analyses required by Iten

.II.B.3.

The accuracies are based on : the results of factory tests.

Analyses which are.potentially sensitive to chemical interference (Boron snd Chloride) were tested using the NRC test matrix. The testing

'did not-account for radiation effects; however, the instrtments in 7

une were designed to withstand a total integrated dose of 10 rads with no loss of accuracy.

TABLE 1 PASS REACTOR COOLANT ANALYSES ANALYSIS RANGE ACCURACY 10-l ci/ml-10 Ci/ml

< Factor of 2 Gross Radioactivity p

Gana Spectrta Isotopic 4 Factor of 2 Boron 100-6500 ppn

+15.4,-40.6 ppn Chloride 0.1-20 ppn 40.5 ppn: +0.4,-0.07 ppn 0.5-20 rpn: +18%,-4.8%

Dissolved Hydrogen 0-100 Vol %

+2% of Full Scale TABLE 2 PASS DRYWELL A1MOSPHERE ANALYSES ANALYSIS RANGE ACCURACY 10-3 ci/ml-105 Ci/ml

< Factor of 2 Gross Radioactivity p

A Gamna Spectrta Isotopic

< Factor of 2 TABLE 3 CONTAIPMENT ATMOSPHERE MONITORING -

ANALYSIS RANGE ACCURACY Hydrogen Content 0-10/0-30 Vol %

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PERSOt@EL - RADIATION ' EXPOSURE

%e majority of post-accident sapling and: analysis ' operations are 1

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renotely ~ ~ controlled from - low: radiation 1 Mareas,.: resulting lin

' negligible radiation exposures., Only cin-the case of failure of

~ in-line.' analysis y capability would - it be necessary, to enter ; the

~ Post-Accidenti Sample < Room < (PASR).for grab sapling and-incur-significant radiation exposure.

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The maximun individual l radlation. exposure for the grab. smpling

operation has been calculated. using ~ the : following conservative assunptions:

1.

Regulatory Guide 1.3 source terms are used.

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2.

Sapling is performed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following ' onset of the accident.

3.

Purging of equipnent and lines prior to saple ' roan entry would reduce dose rates by a factor of ten,

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A six minute entry into the'PASR is required.

%is estimate is pased on testing experience.

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%e' calculation results in a maximum dose rate in the PASR of:15.1 Ren/hr and a ruaximun individual whole body dose of 1.51 Ren..While i

. this exposure estimate is ' within the NUREG-0737 criterion,. it should be noted that in an actual accident ' situation, grab sanpling would ~ be. preceded by extensive '~ surveys ~ and AIARA'. analysis' to-j minimize radiation' exposure.

j-D.

CORE DM1 AGE ESTIMATION b

A procedure for estimating the extent - of ' core, danage based on radionuclide concentrations and other plant-paraneters hasJ been j

implenented ' at Plant Hatch.

%e procedure' is Dased on1the generic l

methodology developed by General. Electric' in NEDO-22215. A copy of the procedure, HNP-4848, was transnitted to ' the' NRC by our lletter

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dated February 10, 1984.

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%e nunerical-factors 'in Section.Diof ' HNP-4848.are the-result of Hatch-specific l modifications ^ to the reference data in ; NEEO-22215.

%ese factors were derived per Appendixt A guidance to' account for differences ' between Plant Hatch and the. " reference' plant". - %e -

guidance in ' Notes; A.and B of the Appendix was: used - in generating '

these factors for ;the Plant Hatch procedure'..

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