ML20071Q036

to Dresden Unit 2 LOCA Analysis Using Exem/Bwr Evaluation Model MAPLHGR Results

ML20071Q036
Person / Time
Site:	Dresden
Issue date:	12/02/1982
From:	Braun D, Jensen S, Kayer W SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17194B411	List: ML20071Q036 ML20071Q059 ML20071Q087
References
XN-NF-82-88, XN-NF-82-88-R01, XN-NF-82-88-R1, NUDOCS 8212290199
Download: ML20071Q036 (20)

Text

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/sz G.El Jppsen, fj6 nager Dath NSSS System Ahalysis (ECCS)

Reviewed by : '7 #M  % ///23/F1 L.V. Kayser, Manager Date Fuel Response Analysis Reviewed by :

//[23//R R.E. Collingham, ager D'a t e '

Systems Model D opm nt (ECCS)

/7 Approved by: I f! '

7.2 New ps R.B. Stout, Manager Date Licensing & Safety Engineering Approved by:

UV U ' ' ' ' I' '- l G. A. Sofer, Ranager/ Date Fuel Engideering &' Technical Services

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E(ON NUCLEAR COMPANY,Inc.

8212290199 821221 DR 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 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 liconeses of the USNRC which utilize Exxon Nucleer' fabricated reload fuel or other technical services provided by Exxon Nuclear for licht water power reactors and it is true and correct to the best of Exxon Nucleer's knowledge, information, and belief. The information contained herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USNRC's regulations.

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

l A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-motion contained in this docurrent, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for derregos resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.

XN- NF- F00, 766

i XN-NF-82-88 Revision 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 CONCLUSION

S .................................................. 13

5.0 REFERENCES

................................................... 14

l 11 XN-NF-82-88 Revision 1 LIST OF TABLES l

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

l iii XN-NF-82-88 Revision 1 l

CL LIST OF FIGURES 1

FIGURE PAGE 1.1 Dresden Unit 2 MAPLHGR vs. Assembly Average Burnup ........... ......................... .................. 3 l

l l

3.1 Blowdown Hot Channel Cgnter Slab Heat Transfer _

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

9 3.2 Blowdown Hot Channel Center Volume Quality .................... 10 6

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 i

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1 XN-NF-82-88 Revision 1

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 (C0 =1.0). This limiting break formed the basis of the MAPLHGR heatup analyses 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.

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f 2 XN-NF-82-88 Reyision 1 Table 1.1 Dresden Unit 2 MAPLHGR Sumary Assembly Average Burnup MAPLHGR GWD/MTM kw/ft

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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

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[ modified to EXEM/BWR(2) with the NRC approval (3) of the cladding swelling and r 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. The E

} 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 the i

core reflood process. The HUXY/BULGEX code performs the hot assembly heatup, clad swelling andg rupture calculations.

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

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s the reactor depressurization, lower plenum refill, core reflood, and the time E

- at whichzthe Jeflooding liquid is entrained to the maximum power plane in the core (time of hot node reflood). The time of hot node reflood is an input parameter for the heatup calculation.

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\ n i 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 f hot node reflood) to determine the peak clad temperature (PCT) and the percent l

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.

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6 XN-NF-82-88 Revision 1 Table 2.1 Dresden Units 2 aad 3 Reactor System Data l

i Primary Heat Output, MW 2577.5*

1 Total Reactor Sytem Volume, ft3 20160.

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

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 Steam Flow Rate Ib/hr 9.95 x 106

• 3 Feedwater Flow Rate, Ib/nr 9.95 x 106

• Rated Recirculation Pump Head, ft. 570.

Rated Recirculation Pump Speed, rpm 1670.

Moment of Inertia, Ibm-ft 2/ 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-82-88 Revision 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

8 XN-NF-82-88 Revision 1 Table 3.1 Dresden Unit 2 LOCA Analysis Results for ENC XN-1 8x8 Reload Fuel Assembly MAPLHGR Local PCT Average (kw/ft) MWR (OF)

Burnup (%)

(GWO/MTM)

0. 13.0 .8 1900 f

12. 13.0 .7 1856 1

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4.0 CONCLUSION

S 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.

14 XN-NF-82-88 Revision 1

5.0 REFERENCES

(1) Exxon Nuclear Company, " Exxon Nuclear Methodology for Boiling Water Reactors, 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 and Rupture Model", XN-NF-82-07(A), Revision 1, dated November 1982.

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

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

(5) 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, 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 and Reflood Analysis", XN-NF-80-19( A1, Volume 28, Revisici.1, dated June 1981.

(9) Exxon Nuclear Company, "HUXY: A Generalized Multirod Hertup Code with 10 CFR 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 the Deformation and the Onset of Bulging of Zircaloy Fuel Rod Cladding", XN-74-27, Revision 2, dated December 31, 1974.

l i

i 15 XN-NF-82-88 Revision 1 Issue 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)

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hTT' f XN NF 82 84 NP REVISION 1 l

PLANT TRANSIENT ANALYSIS FOR DRESDEN UNIT 2, CYCLE 9 DECEMBER 1982 RICHLAND, WA 99352 Ep NUCLEAR COMPANY,Inc.

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