ML17254A294

Responds to NRC Request for Review of Certain Concerns Re Exxon Nuclear Corp ECCS Evaluation Model.Based on Listed Review W/Exxon & Westinghouse,Stated Concerns Inapplicable to Facility

ML17254A294
Person / Time
Site:	Ginna
Issue date:	03/26/1985
From:	Kober R ROCHESTER GAS & ELECTRIC CORP.
To:	Zwolinski J Office of Nuclear Reactor Regulation
References
NUDOCS 8504030076
Download: ML17254A294 (5)

Text

REGULATORY FORMATION DISTRIBUTION S M (RIDS)

SUBJECT:

Responds to NRC r'equest for r eview of certain concer ns re Exxon nuclear Cor'p ECCS evaluation models Based on listed review w/Exxon 8 'riestinghouse<stated concerns inapplicable to facility, SIZE:

D ISTRIBUTION CODE+

A001D COPIES RECEIVED:LTR ENCL TITLE:

OR Submittal:

General Distribution ACCESSION NBR;8504030076 DOC ~ DATE: 85/03/26 NOTARIZED; NO DOCKET FACIL:50 244 Robert Emmet Ginna Nuclear Plantg Unit ii Rochester G

05000244 AUTH INANE'UTHOR AFFILIATION KOBERgR,HE Rochester Gas 8 Electric Corp.

RECIP ~ NAME>>

RECIPIENT AFFILIATION ZNOLINSKI'iJ.A Oper ating Reactor s Branch 5

NOTES;NRR/DL/SEP

1cy, OL: 09/19/69 05000244 RECIPIENT ID CODE/NAME'RR ORB5 BC 01 INTERNAL: ACRS

'9 ELD/HDS4 NRR/DL DIR NRR/DL/TSRG NRR/DSI/RAB RGN1 COPIES LTTR ENCL 7

6 1

1 1

1 1

RECIPIENT ID CODE/NAME ADM/LFMB NRR/DE/MTEB NRR/Dl /ORAB ETB RE FILE 04 COPIES LTTR ENCL EXTERNAL:

EGSG BRUSKE,S NRC PDR 02 NOTES:

LPDR NSIC 03 05 TOTAL NUMBER OF COPIES REQUIRED:

LTTR 28 ENCL

H (I

III I ~

~

'I lg

ll I

P t ~

t

//////////////

/////////////

II/El)7/IIII

////// 8'vis//

ROCHESTER GAS AND I

stAtC ELECTRIC CORPORATION o 69 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 ROGERtN. KOBER VICe PRESIDSNT ELECTRIC 6 STEAM PttODVCTION March 26>

1985 TSI.CPHONC ARCA CODE Tld 546-2700 Director of Nuclear Reactor Regulation Attention:

Mr. John A. Zwolinski> Chief Operating Reactors Branch No.

5 U.S. Nuclear Regulatory Commission Washington<

D.C.

20555

Subject:

Loss of Coolant Accident Analysis R.

E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Mr. Zwolinski:

This letter is in response to a request from members of the NRC staff to address certain concerns regarding Exxon Nuclear Corporation's (ENC) emergency core cooling system evaluation model.

We have reviewed these concerns with Exxon Nuclear and with our current fuel vendor and core designer Westinghouse Electric Corporation and>

as described below<

have reached the conclusion that the stated concerns are not applicable to Ginna From 1978 through 1983>

Ginna reload fuel was supplied by ENC.

The most recent ENC loss of coolant accident analysis for Ginna was performed in Spring 1982 and was submitted to the NRC with our letter dated August 9< 1982.

That analysis resulted in a calculated peak clad temperature (PCT) of 1928 F.

ENC has confirmed to us that the specific models used for the Ginna analysis did not contain any of the three errors which the NRC Staff has identified to us regarding heat transfer correlations and augmentation factors and mixing assumptions.

Regarding the fourth concern raised by the NRC Staff> the following background information and conclusions are provided.

In 1984<

RG6E began a transition to Westinghouse fuel.

In the analysis performed by Westinghouse for that transition<

they evaluated the plant response to all transients which may be affected by fuel type> including the loss of coolant accident.

Westinghouse was provided with detailed fuel parameters such as dimensions>

densities and enrichments.

Westinghouse performed detailed pressure drop measurements on Westinghouse optimized fuel assembly (OFA) design fuel and ENC fuel assemblies.

Westinghouse concluded that analyzing a complete core of OFA under LOCA conditions was conservative with respect to any potential combination and configuration of Westinghouse and ENC assembly types.

The assumption of modeling a full core of OFA was determined to be conservative for transition cycles for two major reasons:

8504OSOO7S 850S2S l

PDR ADOCK 05000244 P

PDR ADO)

,)O

ROCHESTER GAS AND ELECTRIC CORP.

DATE March 26 >

1985 TO Mr. John A. Zwolinski SHEET NO.

2 The increase in core flow area associated with OFA due to the smaller rod diameter has an important impact on flooding rates during reflood.

A full OFA core configuration decreases core flooding rates which reduces heat transfer coefficients and results in earlier steam cooling.

2.

The OFA has a higher volumetric heat generation rate than the ENC fuel.

The analysis assumes tha't an OFA assembly has the hottest rod and maximum F

H which maximizes the calculated PCT.

Westinghouse also evaluated the impact of any hydraulic resistance difference.

The only portion of the LOCA evaluation model impacted by the small hydraulic resistance difference which exists between the ENC and OFA fuel is the core reflood transient.

Since the hydraulic mismatch is so small> only crossflows due to smaller rod size and grid designs need to be evaluated.

The maximum reflood axial flow reduction for the OFA at any possible peak clad temperature location in the core< resulting from crossflows to adjacent ENC assemblies<

has been conservatively calculated to be one percent.

Analyses were performed which demonstrated that the maximum PCT penalty possible of OFA fuel during the transition period is 4 F.

After this transition>

the 0

Westinghouse ECCS analysis will apply to a full-core OFA without the cross flow penalty.

In addition to the factors identified above<

the burnup levels and/or power levels of the ENC fuel further assure that the limiting fuel is the OFA fuel.

The last region of ENC fuel was loaded in 1983.

Four of these fuel assemblies have burnups of at least 11,000 MWD/MTU< while the other assemblies in this region have burnup levels in excess of 20>000 MWD/MTU.

Eight fresh ENC fuel assemblies were loaded during the Spring 1984 refueling out-age and currently have burnups in excess of ll>000 MWD/MTU.

Maxi-mum burnups will remain below the design limits established by ENC.

The ENC assemblies are not located in peak power locations for Cycle 15 but have F

values at least 5% below that in the limiting OFA locations.

The minimum margin to the F limit pre-0 dieted for an ENC assembly occurs at an elevation where the F

limit is 2.29 and the predicted F

is 2.125.

It should be n8ted that this value is predicted b2sed on load follow operation<

while Ginna typically operates base loaded.

As described in our letter dated April 10<

1984<

we have loaded four ENC annular fuel pellet demonstration assemblies during the current Cycle 14-15 refueling outage.

In addition to the fuel temperature and stored energy benefits which derive from the use of an annular fuel

pellet, these assemblies are loaded in the core periphery and have assembly power levels less than 0.8 times core average power.

These assemblies have been explicitly modeled in the most recent Westinghouse reload analysis.

Thus>

these assemblies also are bounded by the OFA fuel analysis.

ROCHESTER GAS ANP ELECTRIC CORP.

pATE Narch 26, l985 To Nr. John A. zwolinski SHEET NO.

For the reasons stated abovei we have concluded that the analyses performed by Westinghouse with the approved Westinghouse evaluation models bound all fuel currently loaded in the Ginna reactor.

This evaluation demonstrates that the limiting fuel type is the Westinghouse OFA fuel.

Thus<

no reliance is currently made by the ENC LOCA models and potential concerns regarding those codes do not apply to Ginna.

y truly yoursi Roger W. Kober