Text

LOCA ECCS Analysis MAPLHGR Results
	ML20112H515
Person / Time
Site:	Susquehanna
Issue date:	12/14/1984
From:	Braun D, Jensen S, Krajicek J; SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:	;
Shared Package
ML17139C813	List: ML20112H484; ML20112H497; ML20112H504; ML20112H515;
References
	XN-NF-84-119, NUDOCS 8501170175
	Download: ML20112H515 (51)
	v • d • e

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XN -NF ~l19 SUSQUEHANNA UNIT 1 LOCA-ECCS ANALYSIS MAPLHGR RESULTS DECEMBER 1984 RICHLAND,WA 99352 ERON NUCLEAR COMPANY, INC.

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XN-NF-84-119 Issue Date: 12/14/ti SUSQUEHANNA UNIT 1 LOCA-ECCS ANALYSIS ,

1 l

MAPLHGR RESULTS Prepared by: , k l _ $3cm,s f2/is/R1

~D. dl Braun BWR # Safety Analysis

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Prepared by: _ / / z//a/a g CTE. Kf ajicek BWR Safety Analysis Concur: /. w,_ /df5d/

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S. E./Jensen, Lead Engineer BWR dafeti' Analysis /

Concur: h/ A R. E. Collinghpei, Manager

/2/o/jp BWR Safety Analysis f

, L. I Concur: I b - i/{ / h M ^

J. N. M6rgan,Managerg Custo{ner Services Engineering M /_

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Approve: i- j-R. B. Stout, Manager Licensing ,& Safety Engineering Yr Approve: 'I MN OdctJY G. A. 56fpr, Manager Fuel Engineering'& Technical Services ERON \ UC _EA R CO V 3A \ Y, \ C.

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1 NUCLEAR REGULATORY CCMYlS$3CN C:SCLAIMER IMPORTANT NOTICE REGARDtNG CONTENTS AND USE OF Tl4tS DOCUMENT PLE ASE PE AD CAREFULLY This technical report was rferived through research and deve'coment programs sponsored by Exxon Nuciear Comoany. Inc. It is be ng sub-mitted by Exxon Nuclear to the USNRC as part o8 a technical contre bution to facdita*e safety anaiyses by licensees of the USNRC whicn utilize Exxon Nucteer fabricated raioad fusi or other techncat services

, provided by Exxon Nuclear for liant water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, informat on, and beCef. The informat.on contained herein may be used by the USN8C

.n its revie* o' this reoort, and by tiensees or appli: ants beve the

/

USNRC whien are customers of Exxon Nuclear in their demonstracen of comoliance *,t9 the USN RC's regulations.

\

Wit %ut de egat'ng from the foregoing neitner Eaxos Nuc. ear not any comn acting nn its benalf; I

A. Menos am, warranty, excress or irnoised, with re!oect to the ac:: racy, como'eteness, or usefulness of tBe in*c'- l motion containec in this documet, or that tae use o8 any indo-stion, apoeratus. method, or process c setosec in tnis cocument wilt not infnnge privatelt onnec ng ts.

or B. Assemes my haoilities witn respect to the use of, or for carrages remit:ng from the use of, any information, ao-caratus. method, or process dractosed in this document.

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i XN-NF-84-119 TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

................................................ 1 2.0

SUMMARY

..................................................... 2 3.0 JET-PUMP BWR ECCS EVALUATION M0 DEL.......................... 5 3.1 LOCA DESCRIPTION....................................... 5 3.2 EXEM/BWR APPLICATION TO SUSQUEHANNA bNIT 1............. 5 4.0 ANALYSIS RESULTS........................................... 13 4.1 LOCA ANALYSIS......................... ............... 13 4.2 MAPLHGR RESULTS....................................... 13

5.0 REFERENCES

.................................................43 l

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11 XN-NF-84-119 l

' List of Tables Table Page 3.1 Susquehanna Unit 1 Reactor System Data...................... 9 4.1 Susquehanna Unit 1 Limiting Break Event Times.............. 15

' 4.2 'Susquehanna MAPLHGR Results For ENC 8x8 Reload Fuel............................................ 16 i

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iii XN-NF-84-119 List of Figures Figure Page

-2.1- Susquehanna Unit 1 MAPLHGR vs. Hot Assembly Burnup Results for ENC 8x8 Reload Fuel...................... 4

3. l' System Blowdown Nodalization for Susquehanna Unit 1......................................... 10 3.2 Hot Channel Nodalization for Susquehanna Unit 1...... .. .. .. 11 3.3 -System Refill /Reflood Nodalization for Susquehanna Unit 1......................................... 12 4.1 Blowdown System Pressure, 0.4 DEG/RD Break................. 17 4.2 Blowdown Total Break Flow, 0.4 DEG/RD Break............... 18 4.3 Average Core Inlet Flow, 0.4 DEG/RD 8reak.................. 19 4.4 Blowdown Average Core Outlet Flow, 0.4 DEG/RD Break........................................... 20

-4.5 Blowdown HOT CHANNEL Inlet Flow, 0.4 DEG/RD Break........................................... 21 4.6 Blowdown HOT CHANNEL Outlet Flow, 0.4 DEG/RD Break........................................... 22 4.7 . Blowdown Intact Loop Jet Pump Drive Flow, 0.4 DEG/RD Break........................................... 23 4.8 Blowdown Intact Loop Jet Pump Suction Flow, 0.4 DEG/RD Break........................................... 24 4.9 Blowdown Intact Loop Jet Pump Exit Flow,

0. 4 D EG /R D B r e ak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 5 L 4.10 Blowdown Broken Loop Jet Pump Drive Flow, 0.4 DEG/RD Break........................................... 26 4.11 Blowdown Broken Loop Jet Pump Suction Flow, 0.4 DEG/RD Break........................................... 27 4.12-Blowdown Broken Loop Jet Pump Exit Flow,

0. 4 D E G /R D B r e ak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 r

4.13 Blowdown Upper Downcomer Mixture Level,

0. 4 D EG /R D B re a k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9

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l iv XN-NF-84-119 l

List of Figures (Cont.)-

l' Figure Page 4.14 Blowdown Middle Downcomer Mixture Level, ,

0.4 DEG/RD Break........................................... 30 l 4.15 Blowdown Lower Downcomer Mixture Level, 0.4 DEG/RD Break........................................... 31 4.16 Blowdown Lower Downcomer Liquid Mass, 0.4 DEG/RD Break........................................... 32 4.17 Blowdown Upper Plenum Liquid Mass, 0.4 DEG/RD Break........................................... 33 4.18 Blowdown Upper Downcomer Liquid Mass, 0.4 DEG/RD Break........................................... 34 4.19 Blowdown Lower Plenum Liquid Mass, 0.4 DEG/nD Break........................................... 35 4.20 Refill /Reflood System Pressure, 0.4 DEG/RD Break........................................... 36 4.21 Refill /Reflood Lower Plenum Mixture Level, 0.4 DEG/RD Break........................................... 37 4.22 Refill /Reflood Relative Core Midplane Entrainment, 0.4 DEG/RD Break.............................. 38 4.23 Blowdown HOT CHANNEL Heat Transfer Coefficient, 0.4 DEG/RD Break........................................... 39 4.24 Blowdown HOT CHANNEL Center Volume Quality,

0. 4 DEG/RD Bre ak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 4.25 Blowdown HOT CHANNEL Center Volume Coolant Temperature, 0.4 DEG/RD Break.............................. 41 4.26 Typical Hot Assembly Heatup Results, BOL, MAPLHGR = 13.0........................................ 42 l

l XN-NF-84-119

1.0 INTRODUCTION

l l

The results of a LOCA-ECCS analysis for the ENC fuel in the Susquehanna l Unit I reactor are summarized in this document. The results of this analysis are presented in terms of the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit as a function of fuel exposure. These calculations were performed with the generically approved Exxon Nuclear Company EXEM/BWR Evaluation Model(1,2) according to Appendix K of 10 CFR 50(3), and the results comply with the U.S. NRC 10 CFR 50.46 criteria.

A generic break spectrum analysis applicable to BWR 3 and BWR 4 reactors with modified low pressure coolant injection (LPCI) loop selection logic (LSL) has been previously reported (4). This generic analysis showed that the limiting break for BWR 3 and BWR 4 reactors with modified LPCI loop selection logic is a double-ended guillotine (DEG) break in the recirculation discharge (RD) piping with a discharge coefficient of 0.4. This spectrum analysis is applicable to the Susquehanna Unit 1 BWR 4 reactor which has a modified LPCI loop selection logic system.

i 2 XN-NF-84-119 2.0

SUMMARY

A limiting break calculation was performed for Susquehanna plant specific geometry ard conditions to establish boundary conditions for the heatup analyses. The exposure dependent MAPLHGR limit was determined for ENC fuel from beginning-of-life (80L) to an assembly exposure of 35 GWD/MTM using the limiting break boundary conditions. The resulting MAPLHGR limit for ENC fuel is presented in Figure 2.1. The MAPLHGR limit of Figure 2.1 is constant at 13.0 kw/ft for exposures from 0 to 19 GWD/MTM, the limit decreases linearly from 13.0 to 10.4 kw/ft for exposures from 19 to 30 GWD/MTM, and then it remains constant at a value of 10.4 kw/ft for exposures from 30 to 35 GWD/MTM.

The limit applies to the initial and subsequent reloads incorporating ENC 8x8 fuel of the design analyzed in Susquehanna Unit 1 at a reactor power of 3293 MWt.

All.of these calculations were performed according to Appendix K of 10 CFR 50 and the MAPLHGR of Figure 2.1 satisfies the requirements speciffed b'y 10 CFR 50.46 of the U.S. Code of Federal Regulations. Operation of the Susquehanna Unit I reactor with ENC fuel within the limit of Figure 2.1 assures that the Emergency Core Cooling System for Susquehanna Unit 1 will ~

meet the U.S. NRC acceptance criteria for breaks up to and including the double-ended severance of a reactor coolant pipe. 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.

u

1 3 XN-NF-84-119

3. The cladding temperature transient is terminated at a time the core geometry is still. amenable to cooling. The hot fuel rod cladding

. oxidation limit of 17% is not exceeded during or after quenching.

4. The system long term cooling capabilities provided for the previous core remains applicable to ENC fuel.

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r 5 XN-NF-84-119 3.0 JET-PUMP BWR ECCS EVALUATION MODEL 3.1 LOCA DESCRIPTION ,

A loss-of-coolant accident (LOCA) is defined as a hypothetical ,

rupture of the reactor coolant system piping, up to and including the double-ended rupture of the largest pipe in the reactor coolant system or of any line connected to that system up to the first closed valve. In the unlikely event a LOCA occurs in the Susquehanna plant, the reactor system coolant inventory loss would result in a high containment drywell pressure and reduced reactor vessel pressure. The concurrent high drywell pressure and low reactor vessel pressure provide a safety injection signal which brings coolant injection systems into operation to limit the accident consequences.

During the early phase of the LOCA depressurization transient, core cooling is provided by the exiting coolant inventory. In the litter stage of system depressurization and after depressurization has been achieved, the core spray provides core cooling and supplies liquid to refill the lower portion of the reactor vessel and reflood the core. The reflood process provides sufficient heat removal to terminate the core temperature tran-sient.

3.2 EXEM/BWR APPLICATION TO SUSQUEHANNA UNIT 1

. The ENC EXEM/BWR codes sere used for the LOCA ECCS analysis for Susquehanna Unit-1. The versions of the EXEM codes used in this MAPLHGR analysis are the same as used in the Generic Break Spectrum Analyses (4) except for a minor increase in the convergence tolerance for the thermal radiation solution in the'HUXY/BULGEX code. EXEM/BWR is comprised of the R00EX2, RELAX, FLEX, and HUXY/BULGEX computer codes.

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l 6 XN-NF-84-119 j

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The initial stored energies for the RELAX blowdown RELAX / HOT CHANNEL and HUXY/BULGEX calculations are determined with the RODEX2 code.

The RELAX code is used to calculate the reactor system behavior during the initial portion of the reactor system depressurization transient.

RELAX predicts' mass distribution, core and system thermal hydraulics, and break flow rates. The blowdown calculation provides core boundary conditions for the heatup calculation and reactor coolant system conditions for the initialization of the refill /reflood transient calculation using the FLEX code. Because the FLEX code does not model closure of the valves in the ;

'recirculdtion piping, the system blowdown code, RELAX, is run until both the low pressure core spray reaches rated flow and recirculation line valves are completely closed. This approach is necessary to properly model event times; the RELAX hot channel heat transfer is used in the heatup model only until the time that rated core spray flow is calculated. The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and decay heating required by Appendix K of part 10 CFR 50 to the U.S. Code of Federal Regulations. For the blowdown calculation, the reactor coolant system is nodalized into control volumes representing reasonably homogenous regions interconnected by junctions as shown in Figure 3.1. Reactor system data for the Susquehanna Unit 1 LOCA system analysis are summarized in Table 3.1. Pump performance curves characteristic of the Susquehanna recirculation pumps are used in the analysis.

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7 XN-NF-84-119 For the maximum power fuel assembly, a separate RELAX / HOT CHANNEL calculation is used to calculate the cladding-to-coolant heat transfer coefficients and the coolant thermodynamic properties. The HOT CHANNEL analysis uses time dependent plenum boundary conditions from the RELAX blowdown calculation. The calculated results from the H0T CHANNEL calcu-lation are used as input data to the subsequent hot assembly heatup calculation until time of core spray flow is calculated. Conservative haat transfer coefficients and fluid thermodynamic properties for the heatup calculation are assured by using the maximum stored energy in the H0T CHANNEL calculation for the generation of this information. The HUXY/BULGEX hot assembly heatup calculation computes the fuel temperature transient from its initiation through peak clad temperature (PCT). The RELAX / HOT CHANNEL nodalization is shown in Figure 3.2.

The FLEX system refill /reflood analysis predicts the latter segment of the reactor coolant system depressurization, the refilling of the lower plenum and the reflooding of._ the core. The time of hot-node-reflood is-determined by FLEX and is an input quantity to the hot assembly HUXY/BULGEX heatup calculation. The FLEX system refill / reflood nodalization is shown in Figure 3.3.

The HUXY/BULGEX heatup calculation computes the entire LOCA tem-perature' transient and uses: fuel stored energy, thermal gap conductivity and dimensions from RODEX2 as a function of power and exposure; time of rated spray, decay power, heat transfer coefficients and coolant thermodynamic

8 XN-NF-84-119 properties from RELAX; and time of hot-node-reflood from FLEX. These data are input to HUXY/BULGEX to determine the peak clad t'emperature (PCT) and the

- cladding oxidation percentage.

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9 XN-NF-84-119 Table 3.1 Susquehanna Unit 1 Reactor System Data Primary Heat Output, MWt (102% of rated) 3358.9

' Total Reactor Flow Rate, lb/hr 100 x 106 -

. Active Core Flow Rate, Ib/hr 90.26 x 106 Nominal Reactor System Pressure, psia 1031 Reactor Inlet Enthalpy, Btu /lb 522.1

R'ecirculation Loop Flow Rate, lb/hr 15.68 x 106 Steam Flow' Rate, lb/sec 13.78 x 106 Feedwater Flow Rate,'lb/sec '13.75 x 106 Rated Recirculation Pump' Head, ftt 710 Rated Recirculation Pump Speed,. rpm 1670 Moment-of Inertia, Ibm-ft / 2 rad 15710 Recirculation Suction Pipe I.D., in 25.34 Recirculation Discharge Pipe I.D., in 25.34 h

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13 XN-NF-84-119 4.0 ANALYSIS RESULTS 4.1 LOCA ANALYSIS-An ECCS analysis calculation of the limiting break has been performed for Susquehanna Unit 1, and the calculated results are summarized herein. The break spectrum calculations for BWR 3 and BWR 4 plants with LPCI

. loop. selection logic have been previously performed and reported (4) on a generic bases. In the generic analysis, the limiting break location was found to be in the recirculation discharge piping, and the configuration and size were found to be a double-ended guillotine (DEG) with a discharge coefficient of 0.4.

The LOCA system behavior is determined by the system geometry and break size. Core parameters have only a secondary effect on system event

-times. For these reasons, the system analysis described in this report is

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applicable to future cycles of Susquehanna Unit I unless system modifications or revised operating conditions negate the plant' conditions used herein.

~ Major event times and results for. the limiting break for Susque-hanna Unit 1 are shown in Tables 4.1 and 4.2; Figures 4.1 through 4.19 are

,. system blowdown. parameters; Figures 4.20 through 4.22 are system ' refill parameters.

4.2 MAPLHGR RESULTS The MAPLHGR results for the Susquehanna Unit I reactor are based on the results of the limiting break calculation described in Section 4.1 which is a double-ended guillotine (DEG) with a discharge coefficient of 0.4 in the

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I 14 XN-NF-84-119 recirculation discharge piping. This system analysis used plant specific l data applicable to Susquehanna Unit 1. MAPLHGR results are obtained using LOCA system analysis boundary conditions but require an additional RELAX /H0T CHANNEL calculation and a series of HUXY/BULGEX calculations at various fuel exposures.

i A bounding HOT CHANNE'. calculation has been performed for this

- MAPLHGR analysis in which the fuel stored energy was the maximum for the exposure range of interest for ENC 8x8 fuel. This bounding HOT CHANNEL calculation provides heat transfer coefficients, fluid temperature and fluid quality at the plane of interest for the HUXY/BDLGEX calculations. These HOT CHANNEL calculated parameters are shown in Figures 4.23 through 4.25. Figure 4.26 shows a typical clad temperature trace as calculated by the HUXY/BULGEX code.

The HUXY/BULGEX calculated results and the corresponding MAPLHGR

! limits for ENC 8x8 reload fuel are shown in Table 4.2 and Figure 1.1. These results conform to the U.S. NRC requirements specified by 10 CFR 50.46. Table 4.2 shows .the average burnup (not planar burnup) of the hot assembly, MAPLHGR, peak local metal-water reaction and peak cladding temperature.

It should be noted that about 500F of margin to the 22000F PCT limit is available to account. for possible future plant operational changes or other items that might impose a small change in the analysis results.

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15 XN-NF-84-119 Table 4.1 Susquehanna Unit 1 Limiting Break Event Times i

Event Time (sec)

Start 0.00 i'

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-Steam Flow Stops 5.05

- Low Mixture Level (LI) 12.8 Jet-Pumps Uncover. 15.8

- Recirculation Suction Uncovers 25.2 Lower Plenum Flashes (Quality _> 0.) 21.9 HPCI' Flow Starts 33.6 LPCS Flow Injection Starts 80.2 Rated Spray Calculated 121.9 Depressurization Ends

'(Vessel pressure reaches 1 atmosphere) 263.7 Start of'Reflood-(high density fluid '

enters core) 287.5 -

i Peak Clad Temperature Reached 291.0 1

9 1

3 16 XN-NF-84-119 .

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Table 4.2 Susquehanna MAPLHGR Results for ENC 8x8 Reload Fuel Assembly Average Local Peak Clad Burnup MAPLHGR MWR* Temperature (GWD/MTM) (kw/ft) (%) (OF)

O. 13.0 1.9 2074

5. 13.0 2.0 2093

10. 13.0 2.1 2116

15. 13.0 2.2 2136

19. 13.0 2.3 2147

25. 11.5 1.6 1977

30. 10.4 1.0 1846

35. 10.4 1.2 1852

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[

5.0 REFERENCES

1. " Generic Jet Pump BWR 3 LOCA Anlaysis Using the ENC EXEM Evaluation Model," XN-NF-81-71(A), Supplement 1, Exxon Nuclear Company, September 1982.

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

3. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 4, 1974.

4. "Genaric LOCA Break Spectrum Analysis for BWR 3 and 4 with Modified Low Pressure Coolant Injection Logic," XN-NF-84-ll7(P), Exxon Nuclear Com-pany, December 1981 w

k E

'e 1

b

- e . - - -

XN-NF-84-119 Issue Date: 12/14/84 SUSQUEHANNA UNIT 1 LOCA-ECCS ANALYSIS MAPLHGR RESULTS Distribution D.J. Braun J.C. Chandler R.E. Collingham S.F. Gaines S.E. Jensen J.E. Krajicek J.N. Morgan L.A. Nielsen H.G. Shaw G.A. Sofer R.B. Stout D.R. Swope PP&L/H.G. Shaw (40)

Document Control (5)

ML20112H515

Contents

Text

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

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ML20112H515

Text

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

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