e(on nuclear company,inc

XN-NF-82-88

Revision 1

Issue Date: 12/02/82

i

DRESDEN UNIT 2 LOCA ANALYSIS USING THE ENC EXEM/BWR

EVALUATION MODEL

MAPLHGR RESULTS 1

Prepared by : J . [ht.,) bw h TV H/23[f2,

D.JO Braun/P.J. Valentine fate '' '

Reviewed by : S E 7' #/2.5/szG.El Jppsen, fj6 nager DathNSSS System Ahalysis (ECCS)

Reviewed by : '7 #M % ///23/F1L.V. Kayser, Manager DateFuel Response Analysis

Reviewed by : '

//[23//RR.E. Collingham, ager D'a t e 'Systems Model D opm nt (ECCS)

f!7/

'Approved by: I 7.2 New psR.B. Stout, Manager DateLicensing & Safety Engineering

Approved by: UV U ' ' ' ' I' '-"

|

G. A. Sofer, Ranager/ DateFuel Engideering &' Technical Services

E(ON NUCLEAR COMPANY,Inc.~

8212290199 821221DR ADOCK 05000

_

-

NUCLEAR REGULATORY COMMIS$10N DISCLAIMER

IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT

PLEASE READ CAREFULLY

This technical report was clerived through research and developmentprograms 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 liconeses of the USNRC whichutilize Exxon Nucleer' fabricated reload fuel or other technical servicesprovided by Exxon Nuclear for licht water power reactors and it is trueand correct to the best of Exxon Nucleer's knowledge, information,and belief. The information contained herein may be used by the USNRCin its review of this report, and by licensees or applicants before theUSNRC which are customers of Exxon Nuclear in their demonstrationof compliance with the USNRC's regulations.

Without derogating from the foregoing neither Exxon Nuclear norany person actirs on its behalf:

l

A. Makes any warranty, express or implied, with respect tothe accuracy, completeness, or usefulness of the infor-motion contained in this docurrent, or that the use ofany information, apparatus, method, or process disclosedin this document will not infringe privately owned rights;or

B. Assumes any liabilities with respect to the use of, or forderregos resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.

XN- NF- F00, 766

i XN-NF-82-88Revision 1

- TABLE OF CONTENTS

SECTION PAGE

1.0 INTRODUCTION AND SUMMARY ..................................... l'

2.0 JET PUMP BWR ECCS EVALUATION MODEL ........................... 4

3.0 RESULTS ...................................................... 7

4.0 CONCLUSIONS .................................................. 13

5.0 REFERENCES ................................................... 14

.

.

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11 XN-NF-82-88Revision 1

LIST OF TABLES

|

TABLE PAGE

)

1.1 Dresden Uni t 2 MAPLHGR Sumary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

2.1 Dresden Unit 2 Reactor System Data .......................... 6

3.1 Dresden Unit 2 LOCA Analysis Results .............. ......... 8

!-1.

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iii XN-NF-82-88Revision 1

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!| CL

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LIST OF FIGURES,

.

1

FIGURE PAGE

1.1 Dresden Unit 2 MAPLHGR vs. Assembly Average| Burnup ........... ......................... .................. 3|

| 3.1 Blowdown Hot Channel Cgnter Slab Heat Transfer_

'

Coefficient, BTV/hr-ftc 0F................................. ... 9

3.2 Blowdown Hot Channel Center Volume Quality .................... 106

3.3 Blowdown Hot Channel Center Volume Coolant T-Temperature for 8x8 Fuel, OF ...................... ........... 11

:.

3.4 Hot Assembly Heatup Results for 8x8 Fuel, _2

1.0 DEG/PS Break .. ................................... ....... 12.-

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1.0 INTRODUCTION AND SUPNARY

This document presents the Maximum Average Planar Linear Heat Generation

Rate (MAPLHGR) results from a LOCA analysis performed for the Dresden Unit 2

reactor. These results were obtained using the NRC approved Exxon Nuclear

Company EXEM(1) jet pump BWR ECCS evaluation model, including the EXEM/BWR

model changes (2) approved by the NRC Staff (3) for fuel rod swelling and

rupture. The generic jet pump BWR 3 LOCA break spectrum analysis was described

in XN-NF-81-71(A)(4), and showed the limiting break for a BWR 3 on a generic

basis to be a double-ended guillitone (DEG) configuration in the recirculation

piping on the suction side of the pump (PS) with a discharge coefficient of 1.0

(C =1.0). This limiting break formed the basis of the MAPLHGR heatup analyses0

reported herein.

Heatup analyses were performed for Dresden Unit 2 Cycle 9 of the ENC XN-1

8x8 reload fuel in Dresden Unit 2 using blowdown boundary conditions from the

limiting 1.0 DEG/PS break. The f1APLHGR results for the heatup analysis are

shown in Table 1.1 and Figure 1.1. The calculations were performed according

to the requirements of 10 CFR 50 Appendix K. The MAPLHGR limits defined in

Table 1.1 satisfy the ECCS criteria specified by 10 CFR 50.46(5),

The analysis encompasses Dresden Unit 2 Cycle 9 only, since the revised

ENC fuel performance model, RODEX2, is expected to be approved by the NRC in

time for application to future Dresden Unit 2 cycles with ENC fuel.

p , . , _ .- - - -

f 2 XN-NF-82-88Reyision 1

Table 1.1 Dresden Unit 2 MAPLHGR Sumary

Assembly AverageBurnup MAPLHGR

GWD/MTM kw/ft

0. 13.0

.

12. 13.0

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[ 4 XN-NF-82-88,

7 Revision 1

h 2.0 J_ET PUMP BWR ECCS EVALUATION MODEL,

s' \

} The evaluation model used for the Dresden Unit 2 LOCA analysis is the ENC

EXEM(1) code package for jet pump BWR plants. EXEM is made up of the GAPEX(6),

; RELAX (7), FLEX (8), and HUXY/BULGEX(9,10) codes. The EXEM code package was--

s

[ '(' modified to EXEM/BWR(2) with the NRC approval (3) of the cladding swelling andr b-

'

rupture model based on NUREG-0630. The fuel initial stored energy and exposed-

,.

$ il fissipn gas release calculations are performed with the GAPEX code. TheE

} FLEX code perfgrms the reactor system refill /reflood calculation from the time'

'of rated core spray until liquid is entrained in the core midplane during thei

_

core reflood process. The HUXY/BULGEX code performs the hot assembly heatup,

clad swelling and rupture calculations.g

3 The RELAX blowdown calculation determines the reactor system behavior

f during the initial portion of the reactor system depressurization transient. A-

separate RELAX / HOT CHANNEL computation is used to calculate cladding-to--

I coolant heat transfer coefficierts and coolant thermodynamic properties for

the m imum power assembly. For this calculation, time-dependent boundary-

conditions are derived from the RELAX / BLOWDOWN analysis. This blowdown

g calculation also supplies reactor system conditions at the time of rated lower

. pressure core spray flow to initialize the system refill /reflood transient

calculation.

I The FLEX system refill /reflood analysis predicts the latter segment of\'

the reactor depressurization, lower plenum refill, core reflood, and the times

E

at which the Jeflooding liquid is entrained to the maximum power plane in the-z

core (time of hot node reflood). The time of hot node reflood is an input

parameter for the heatup calculation.

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The HUXY/BULGEX heatup calculation uses calculated parameter-s from GAPEX'

-,

(f uel stored energy and fission gas release), RELAX (time of rated spray, decay

' power,heattransfercoefficientsandcoolanttemperatures)?andFLEX(timeof

hot node reflood) to determine the peak clad temperature (PCT) and the percentf

|>

oxidation of cladding, A symmetric center peaked axial ower profile was used.

Through a series qf heatup calculations at different burnups, the plant MA)LHGR

, limits are determined. ; . ;

Dresden Upits 2 and 3 r[ actor system data appropriate for this analysis-

are given in Table 2.1. > ENC ref oad fuel is compatible hydraulically and.s,, o

neutronically with't.he NSSS vendor fuel. The FLEX refill /reflood calculation

was performed with the leakage holes and leakage paths minimized to conserva-

tively bound the Dresden Unit 2' Cycle 9 mixed core as well as future cores.

._ __

__ .

_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _

t

i 6 XN-NF-82-88'

Revision 1

Table 2.1 Dresden Units 2 aad 3 Reactor System Data|,

i Primary Heat Output, MW 2577.5*1

Total Reactor Sytem Volume, ft3 20160.

| Total Reactor Flow Rate, Ib/hr. 98.0 x 106|

Active Core Flow Rate, Ib/hr 87.27 x 106

Nominal Reactor System Pressure,(upper plenum), psia 1017.

Reactor Inlet Enthalpy, BTU /lb 525.3

Recirculation Loop Flow Rate, Ib/hr 17.11 x 106

6Steam Flow Rate Ib/hr 9.95 x 10 * 3

6Feedwater Flow Rate, Ib/nr 9.95 x 10 *

Rated Recirculation Pump Head, ft. 570.'Rated Recirculation Pump Speed, rpm 1670.

2Moment of Inertia, Ibm-ft / rad 10950.

Recirculation Suction Pipe I.D., in. 25.78

Recirculation Discharge Pipe I.D., in. 25.46

Fuel Assembly Rod Diameter, in** 0.484

Fuel Assembly Rod Pitch, in** 0.641

Active Core Height, in** 145.24

______________.

102% of rated power*

** ENC fuel parameters

.

_ . - . _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _ _ _ . . _ _ _

/ XN-NF-82-88Revision 1

3.0 RESULTS

The MAPLHGR results for the Dresden Unit 2 reactor have been calculated

using the break shown to be limiting in the generic BWR- 3 break spectrum

analyses (4): a double-ended guillotine break (DEG) with a discharge coeffi-

cient of 1.0 in the recirculation suction piping. The blowdown and re-

fill /reflood calculations for this break are presented in the BWR 3 break

spectrum analysis report. That analysis used Dresden Units 2 and 3 plant

specific system data.

A bounding hot channel calculation has been performed for this MAPLHGR

analysis in which the fuel stored energy has been maximized over the exposure

range of interest for ENC XN-18x8 fuel. The maximum fuel stored energy occurs

at an assembly average burnup of 0 GWD/MTM. This bounding hot channel

calculation provides heat transfer coefficients, fluid temperature and fluid

quality at the plane of interest for the HUXY/BULGEX calculations. These hot

channel calculated parameters are shown in Figures 3.1 through 3.3. Figure 3.4

is a heatup vs. time plot calculated by the HUXY/BULGEX code at an assembly

average burnup of 12 GWD/MTM for the XN-18x8 fuel design.

The HUXY/BULGEX calculated results and corresponding MAPLHGR limits for

ENC XN-18x8 reload fuel are shown in Table 3.1 and Figure 1.1. These results

conform to the NRC requirements specified by 10 CFR 50.46. Table 3.1 shows the

average burnup of the hot assembly (not planar burnup), MAPLHGR, peak local

metal-water reaction, and peak clad temperature.

.

1_. - _ _-- , _ - . . .

I


Table 3.1 Dresden Unit 2 LOCA Analysis Results forENC XN-1 8x8 Reload Fuel

Assembly MAPLHGR Local PCTAverage (kw/ft) MWR (OF)Burnup (%)(GWO/MTM)__________ ______________ __________ ___________

0. 13.0 .8 1900f

12. 13.0 .7 1856

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2; Coefficient for 8x8 Fuel, 8TU/hr-ft 0F ~$

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DRESDEN 2+ GAD-HUXY+ 1.0 DEG/PS8 + BOL+_

111 4 56 7 8?. 9 10 11 11 13 14 153 to is 17 la is 2a 11.

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40 80 120 1G0 200 240 180 320

TIME ( SEC 1-- - _ _ _ - _ - _ _ _ _


4.0 CONCLUSIONS

A LOCA-ECCS analysis has been performed for the Dresden Unit 2 reactor

using the EXEM/BWR ECCS Evaluation Model in conformance with Appendix K of 10

CFR 50. The limiting break was identified as the 1.0 DEG break in the

recirculation suction piping (4). The limiting Maximum Average Planar Linear

Heat Generation Rates (MAPLHGR) based on this break were developed for ENC fuel

for the exposures given in Tables 1.1 and 3.1, and Figure 1.1. These limits

apply for ENC XN-1 8x8 reload fuel.

Operation of the Dresden Unit 2 reactor with ENC fuel within the limits

defined by Table 1.1 assures that the Dresden Unit 2 Emergency Core Cooling

System will meet the acceptance criteria as required in 10 CFR 50.46. That is:

1. The calculated peak fuel element clad temperature does not exceed

the 22000F limit.

2. The amount of fuel element cladding that reacts chemically with

water or steam does not exceed 1% of the total amount of zircaloy in

the reactor.

3. Tne cladding temperature transient is terminated at a time when the

core geometry is still amenable to cooling. The hot fuel rod

cladding oxidation limit af 17% is not exceeded during or after

quenching.

4. The system long-term cr oling capabilities provided for previous

cores remains applicable to ENC fuel.

- _ _ _


5.0 REFERENCES,

(1) Exxon Nuclear Company, " Exxon Nuclear Methodology for Boiling WaterReactors, Volume 2, "EXEM: ECCS Evaluation Model Summary Description",XN-NF-80-19(A), Revision 1, Volume 2 dated June 1981.

(2) Exxon Nuclear Company, " Exxon Nuclear Company ECCS Cladding Swelling andRupture Model", XN-NF-82-07(A), Revision 1, dated November 1982.

(3) United States Nucleir Regulatory Comission, " Safety Evaluation Report onXN-NF-82-07, Revision 1: Exxon Nuclear Company ECCS Cladding Swelling andRupture Model", October 14, 1982.

(4) Exxon Nuclear Company, " Generic Jet Pump BWR 3 LOCA Analysis Using the ENCEXEM Evaluation Model", XN-NF-81-71( A1, dated October 1981.

(5) 10 CFR 50.46 and Appendix K of 10 CFR 50, " Acceptance Criteria forEmergency Core Cooling Systems for Light Water Cooled Nuclear PowerReactors", Federal Register, Volume 39, Number 3, dated January 4,1984.

(6) Exxon Nuclear Company, "GAPEX: A Computer Program for Predicting Pellet-to-Cladding Heat Transfer Coefficients", XN-73-25, dated August 1973.

(7) Exxon Nuclear Company, " RELAX: A RELAP4 Based Computer Code for Cal-culating Blowdown Phenomena", XN-NF-80-19( A1, Volum 2A, Revision 1,dated June 1981.

(8) Exxon Nuclear Company, " FLEX: A Computer Code for Jet Pump BWR Refill andReflood Analysis", XN-NF-80-19( A1, Volume 28, Revisici.1, dated June1981.

(9) Exxon Nuclear Company, "HUXY: A Generalized Multirod Hertup Code with 10CFR 50 Appendix K Heatup Option - User's Manual", XN-CC-33(A), Revision 1,dated November 1, 1975.

(10) Exxon Nuclear Company, "BULGEX: A Computer Code to Determine theDeformation and the Onset of Bulging of Zircaloy Fuel Rod Cladding", XN-74-27, Revision 2, dated December 31, 1974.

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15 XN-NF-82-88Revision 1Issue Date: 12/02/82

DRESDEN UNIT 2 LOCA ANALYSIS USING THE

EXEM/BWR EVALUATION MODEL

MAPLHGR RESULTS

Distribution

M.J. Ades

J.C. Chandler

R.E. Collingham

G.C. Cooke

C.J. Braun

L.J. Federico

W.V. Kayser

S.E. Jensen

D.C. Kolesar

J.E. Krajicek

J.L. Maryott

S.L. Schell

P.J. Valentine

L.C. O'Malley/ CECO (60)

Document Control (10)i

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- - - . , , - , . . , . - - -.-n.- . ,e ,r-v - r ,- - - - -

.. . _-

hTT' f XN NF 82 84 NP

REVISION 1

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PLANT TRANSIENT ANALYSIS FOR

DRESDEN UNIT 2, CYCLE 9

DECEMBER 1982

RICHLAND, WA 99352

Ep NUCLEAR COMPANY,Inc.

.

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e(on nuclear company,inc

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