ML20063E155

LOCA ECCS Limiting Break & Exposure Sensitivity Analysis for ENC XN-1 & XN-2 Reloads at Prairie Island Unit 1 w/5% Steam Generator Tubes Plugged Using ENC Wrem Iia PWR Model

ML20063E155
Person / Time
Site:	Prairie Island
Issue date:	06/29/1982
From:	Kayser W, Owsley G, Valentine P SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML20063E147	List: ML20063E144 ML20063E155
References
XN-NF-81-06-S01, XN-NF-81-6-S1, NUDOCS 8207130314
Download: ML20063E155 (13)

Text

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XN-NF-81-06 lI Supplement 1 Issue Date: 06/29/82 I

I LOCA ECCS LIMITING BREAK AND EXPOSURE SENSITIVITY ANALYSIS FOR ENC XN-1 AND XN-2 RELOADS AT PRAIRIE ISLAND UNIT 1

.WITH FIVE PERCENT STEAM GENERATOR TUBES PLUGGED USING ENC'S WREM IIA PWR MODEL l

Supplement 1 Prepared by :

P.J Valentine Fue Response Analysis Prepared by : Mh W.V. Kayser, Manager Fuel Response Analysis j

Concur :

G.F.Owsley, Man'agery Reload Fuel Licensing Approve :

S g}tAl AA-v J.. M6rgan, Mandser Li ensing & Safety Engineering Approve :

Og G.A.SofsefManap&TechnicalServices r'

Fuel Engineering I

/ (ON NUCLEAR COMPANY,Inc.

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pna8ramn

I Il NUCLEAR REGULATORY COMMISSION DISCLAIMER l

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IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxo.i Nuclear-fabricated reloart fuel or other technical services provided by Exxon Nuclear for liaht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by bcensees or applicants before the USNRC which are customers of Exxon Nuclear in their demone ration of comoliance with the USNRC's regulations.

Without derogating from the foregoing, neither Exxon Nuclear nor any person acting on its behalf:

A.

Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not inf nnge privately owned rights; NE or B.

Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, a p-paratus, method, or process disclosed in thrs document.

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i XN-NF-81-06 Supplement 1 I

TABLE OF CONTENTS I

Section Page 1.0

SUMMARY

1 2.0 MODELS IN ASSUMPTIONS.............................

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3.0 ANALYSIS RESULTS..................................

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4.0 REFERENCES

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LIST OF TABLES Table No.

Page 1

Prairie Island Unit 1 Fuel Design Parameters......

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Prairie Island Unit 1 Computer Codes used in Current Analysis.........................

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Prairie Island Unit 1 ECCS Exposure Sensitivity Results for Reloads XN-1 and XN-2 at End-of-Life.

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Prairie Island Unit 1 ECCS Analysis Parameters....

8 LIST OF FIGURES I

Figure No.

Page PrairieIslandUnit1FhversusPeakPellet I

1 Burnup for Reload Fuels XN-1 and XN-2............

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1 Xii-NF-81-06 Supplement 1 1.0

SUMMARY

This document provides the results of LOCA-ECCS analysis for the Prairie Island Unit I reactor to support reactor operation with ENC fuel reloads XN-1 and XN-2 for extended burnup. The analysis assumed 5% steam generator tube plugging in Prairie Island Unit 1.

Previous ENC analysis (l) had supported operating reloads XN-1 and XN-2 to a peak pellet exposure of 47 GWD/MTM. The current analysis extends the burnup dependent total peakinglimit(Fh)topeakpelletburnupsof50GWD/MTMforreloadXN-1and 51 GWD/MTM for reload XN-2.

The corresponding peak rod burnups used in the analyses are 45.5 GWD/MTM for reload XN-1, and 46.4 GWD/MTM for reload XN-2.

Figure 1providestheallowabletotalpeaking(Fh)vs.peakpellet burnup determined by previous analysis, and extended by the current I

1 analysis. The allowable Fg vs. exposure is constant at 2.21 (14.03 kw/ft total, 13.66 kw/ft heat release in the fuel) to a peak pellet burnup of 36.7 and 39 GWD/MTM for reloads XN-1 and XN-2 respectively.

Table 1 provides differences in fuel design parameters for reloads XN-1 and XN-2.

Athigherexposures,theFhlimitdecreaseslinearlyto2.009at47GWD/MTM I

peak pellet burnup for both reloads XN-1 and XN-2.

Above 47 GWD/MTM peak pelletburnupFhdecreaseslinearlyto1.94at50GW9/MTMforreloadXN-1 and 1.91 at 51 GWD/MTM for reload XN-2.

AreductioninFhathighexposures is necessary to offset the adverse effects of fission gas release on predicted clad rupture and flow blockage in the postulated LOCA. TheFhreduction I

occurs at sufficiently high burnup that it is not anticipated to be restrictive for the projected core operation.

Operation of the Prairie Island Unit 1 totheFhlimitdetailedinFigure1willassurethattheplantoperatesin conformance with 10 CFR 50.46 criteria (2),

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2 XN-NF-81-06

!g Supplement 1 l3 lI Table 1 Prairie Is'and Unit 1 Fuel Design Parameters I

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Reload Reload XN-1 XN-2 Pellet Diameter (in)

.3565

.3555 m

Diametric Gap (in)

.0075

.0075

)I Clad Inner Diameter (in)

.364

.364 Clad Outer Diameter (in)

.424

.426 Fuel Length (in) 144.

144.

Plenum Volume (in3)

.575

.575 Prepressurization (psig) 290 305 Dish fraction

.01

.01 I

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2.30 (2.21,0)

(2.21,39) 2.20

( 2.21,36.7 )\\

\\s Reload XN-2

\\

\\

-e Reload XN-1 \\

\\

5 2.10

\\

e

\\

e

\\

5

\\

l l

l w

1 2.00

\\(2.009,47)

\\

\\

(l.94,5C )

(l.91,51) '

1.90 0

10 20 30 40 50

{M u :-

PEAK PELLET BURNUP, GWD/MTM

{[

o-Figure 1 PrairieIslandUnit1FfversusPeakPelletBurnup Sh for Reload Fuels XN-1 and XN-2

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XN-NF-81-06 Supplement 1 2.0 N0DELS IN ASSUMPTIONS The analysis used the ENC WREMll flow blockage methodology (3) and not the NRC flow blockage model as described in NUREG-0630(4).

The NUREG-0630 model as implemented by ENC (5) is currently under review by the NRC Staff, as well as PWR ECCS model updates (6). The currently approved ENC flow biockage model, along with the WREM-IIA (7) ECCS models and the NRC enhanced fission gas release patch to GAPEX(8), give conservatively high peak cladding temperatures (PCTs) compared to the results with the models cLerently under NRC review. PCTs were also computed in accordance with the NRC interim upper plenum injection model(9).

The computer codes used in these analyses are listed in Table 2.

These codes have been approved by the NRC for use in licensing analysis and are I

part of the ENC WREM-IIA model. The fission gas release predicted by GAPEX at the desired exposure is adjusted by the NRC enhancement formula for exposures greatr.r than 20 GWD/MTM. The effect of uncertainties in internal rod pressure on rupture pressure have also been considered in the analysis.

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Supplement 1 Table 2 Prairie Island Unit 1 Computer Codes used in Current Analysis I

Calculation Computer Code Blowdown

)

I

)

Hot Channel

)

)

Normalized Power

)

RELAP4/ ENC 28FB

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Accumulator

)

Injection

)

Containment Pressure CONTEMPT /LT22 ENC Reflood REFLEX Heatup T00DEE2/UJAN82 Stored Energy

)

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GAPEX Fission Gas Release

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APCT for UPI NRC Model with Westinghouse modifications I

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6 XN-NF-81-06 P

Supplement 1 3.0 ANALYSIS RESULTS Calculations were made for the previously ide..tified limiting i

break (10), 0.4 DECLG, with 5% of the steam generator tubes plugged.

The results of the analysis are given in Table 3.

Table 3 shows peak cladding temperatures less than 22000F, the local maximum metal-water reaction less than 17% and core wide metal-water reactiot, less than the 1% limit. The resultssatisfiedthe10CFR50.46criteriaattheFhlimitslisted.

ECCS analysisparametersusedtoestablishtheFhlimitsof2.21arelistedin Athighexposures,theradialpeakingwasreducedfortheFh Table 4.

limits less than 2.21.

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Supplement 1 Table 3 Prairie Island Unit 1 ECCS Exposure Sensitivity Results for Reloads XN-1 and XN-2 at End-of-Life I

Reload Reload XN-1 XN-2 Peak Pellet Exposure (GWD/MTM) 50 51 TotalPeaking,Fh 1.94 1.91 Peak Cladding Temperature (PCT) (OF) 1964 1891 I

Maximum Local Zr/H O Reaction (%)

2.2 1.6 2

at Time (Sec) 297 297 Core Wide Zr/H O Reaction (%)

1.0 1.0 2

Hot Rod Burst Time (sec) 171.9 160.4 Hot Rod Burst Location (ft) 8.88 8.63 Rupture Pressure (psia) 982 837 Subchannel Flow Blockage (%)

47 51 Time PCT (sec) 291.1 290.1 PCT Location (ft) 9.63 9.63 Maximum Zr/H 0 Reaction Location (ft) 9.63 9.63 2

.aPCT for UPI (OF) 1 1

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XN-NF-81-06 S u pp l e...e n t 1

'I Table 4 Prairie Island Unit 1 ECCS Analysis Parameters

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Reactor Power (MWt) 1683.0 (102%)

i Reactor Pressure, psia 2250.

i Heat Release in Fuel 97.4%

Nominal Hot Assembly, Radial Peaking, Fr 1.4904 i

Nominal Hot Rod Locak Peaking, F1 1.04 i

l Nominal Engineering Factor, Fe 1.03 i

l Nominal Axial Peaking, Fa 1.3843 i

Nominal Total Peaking Fg=F f F Fa 2 21 rle

{

Axial Power Peak Lecation, X/L 0.5 J

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XN-NF-81-06 I

9 Supplement 1 4.0 REFERENCE 3 1.

Exxon Nuclear Company, LOCA ECCS Limiting Break and Exposure Sensitivity I

Analysis for ENC XN-1 and TNTTlieloads at PraIFITislariTlT5ftTwith

~

ITv~e percent 5 team Generator Tubes Plugged Using ENC WREM-IIA PWR Evaluation Model, XN-NF-81-06(P), February 1981.

2.

10 CFR 50.46 and Appendix K of 10 CFR 50, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors, Federal Register, Volume 39, No. 3, January 9, 1974.

3.

Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC-WREMII", XN-76-27, July 1976.

4.

D.A. Powers and R.0. Meyer, Cladding Swelling and Rupture Model for LOCA Analysis, NRC report NUREG-0630, Apri1 1980.

5.

Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, XN-Np-82-07(P), March 1982.

6.

Exxon Nuclear Company Evaluation EXEM/PWR ECCS Model Updates, XN-NF 20(P), February 1982, and Supplement 1, March 1982.

7.

Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA, XN-NF-78-30, and Amendments, August 1978.

8.

Letter, D.F. Ross to W.S. Nechodom, " Fission Gas Release for Fuel at I

High Burnups", January 1978.

9.

Safety Evaluation Report on Interim ECCS Evaluation Model for Westinghouse I

2-Loop Plants, Analysis Brancn, Division of Systems Safety, Office of Nuclear Regulations, November 1977.

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10. ECCS Large Break Spectrum Analysis for Prairie Island Unit 1 Using ENC l

WREM-IIA PWR Evaluation Model, XN-NF-78-46, November 1978.

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I-XN-NF-81-06 Supplement 1 Issue Date: 06/29/82 I

Distribution M.J. Ades 0.J. Braun J.C. Chandler R.E. Collingham G.C. Cooke l

S.E. Jensen W.V. Kayser D.C. Kolesar J.E. Krajicek J.N. Morgan L.C. O'Malley G.F. Owsley G.A. Sofer P.J. Valentine H.E. Williamson Document Control (10)

Northern States Power /L.C. O'Malley (3)

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