Text

LOCA Analysis at 1500 MW Using Exxon WREM-IIA PWR ECCS Evaluation Model
	ML19254E829
Person / Time
Site:	Fort Calhoun
Issue date:	09/30/1979
From:	Krajicek J; SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:	;
Shared Package
ML19254E787	List: ML19254E790; ML19254E806; ML19254E813; ML19254E822; ML19254E829;
References
	XN-NF-79-089, XN-NF-79-89, NUDOCS 7911020368
	Download: ML19254E829 (66)
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FORT CALHOUN LOCA ANALYSES AT 1500 MWT USING ENC WREM-IIA PWR ECCS EVALUA"10N MODEL l

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XN-NF-79-89 ISSUE DATE: 10/17/79 FORT CALHOUN LOCA ANALYSES AT 1500 MWT USING ENC WREM-IIA PWR ECCS EVALUATION MODEL Prepared by J. E. KrajicelV Concur:

vi 44m I 7//r//f J. N. Morgan, Maflager

/

Nucidar Safet/ Engineering Concur:M7/'/

7dM79 G. J. BusselmM Kanager Neutronics and Fuel Management Concu

WV G. A. So @ls Engin/

anage Nuclear Fue eering Concur uil

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a L. J. Federico, Manager Nuclear Fuels Projects Approved:

W. S. Nechodom, Manager Licensing and Compliance ERON NUCLEAR COMPANY,Inc.

1263 291

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I NUCLEAR REGULATORY COMMISSION DISCL/ AER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical report was derived thrnngh researtfi and devel<>pment programs sponsored by Exxon Nuclear Company, Inc.

It is being sub mitted by Exxon Nuclear to the USNRC as part of a t chnical contri bution to facilitate safety analyses by hcensees of the USNRC which utilize Exxon Nuclear fabricatrtf relost fuel or other technical services providal by Exxon Nuclear for hoht water power reactors and it is true and correct to the best of Exxon Nuclear's k nowledge, mformation, aruf belief. The information cc.ita#ned herein may be used by the USNRC in its review of this report, and by hcensees or appbcants before the USNRC which are customers of Exxon Nuclear in their demonstration of compliance with the USNRC's regulatioos.

Wit hout derogatmg from the f uegoing, neit her Evuon Nui le.ir nor any person acting on its bel.alf.

A Makes any warranty, ex6. ass or implietf, with respect to the accuracy, completeness, usef ulness of the infor e

mation contained in this document, or that the use of any information, apparatus, method, or process disclosal in this document will not inf ringe privately owned rights, or B.

Assumes any habilities with respect to the use of, or for darrages resulting from the use of, any informaton, an-p ar a tus. met hod, or process disclosed m this document.

1263 292 g

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i XN-NF-79-89 The author wishes to express his appreciation to the individuals listed below for their efforts in performing various phar,es of the Fort Lolhoun LOCA analyses as well as their conments and suggestions.

I D. J. Braun G. C. Cooke I-R. D. Hyman S. E. Jensen W. V. Kayser D. R. Swope I

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Xh-NF-79-89 FORT CALHOUN L1CA ANALYSES AT 1500 MWT USING ENC WREM-IIA PWR ECCS EVALUATION MODEL TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

AND

SUMMARY

1 2.0 BREAK SPECTRUM ANALYSIS 4

2.1 IDENTIFICATI' vAUSES AND ACCIDENT DESCRIPTION..

4 2.2 THERMAL ANALYSIS 5

2.2.1 Method of Analysis..'............

5 2.2.2 Large Break LOCA Analysis Modeling......

6 2.3 BREAK SPECTRUM RESULTS 7

3.0 LOCA ECCS AXIAL POWER PROFILE ANALYSIS..........

39 4.0 LIMITING BREAK HEATUP ANALYSIS RESULTS INCLUDING R0D INTERNAL PRESSURE UNCERTAINTIES AND EXPOSURE SENSITIVITY.

46 4.1 ECCS CALCULATIONS FOR CE FUEL............

46 4.2 HEATUP ANALYSIS MODELS AND ASSUMPTIONS 46 4.3 FUEL ROD INTERNAL PRESSURE UNCERTAI 'TY ANALYSIS AND HEATUP RESULTS FOR BEGINNING-0F-LIFi 47 4.4 EXPOSURE SENSITIVITY...

47 5.0 SENSITIVITY ANALYSIS 54 5.1 RESULTS 54

6.0 CONCLUSION

S 56

7.0 REFERENCES

57 1263 294

XN-NF-79-89 iii LIST OF TABLES P

Table Page 2.1 Fort Calhoun Large Break Events.............

10 2.2 Fort Calhoun large Break Reratts 11 2.3 Fort Cal houn FlR Data..................

12 2.4 Dry Containment Data 14 3.1 Relative Total Peaking Versus Axial Peak Location with a Constant PCT.......................

41 4.1 Fort Calhoun Exposure Heatup Analyses Results for ENC and CE Fuel.........................

49 1263 295 m

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I iv XN-NF-79-89 LIST OF FIGURES I

Figure Page I

1.1 Limits on Total Peaking versus Axial Power Peak Location.

3 2.1 RELAP4-EM Blowdown System Nodalization for Fort Calhoun PWR 17 8

2.2 Axial Peaking Factor Versus Fuel Rod Length for Fort Calhoun ECCS Analysis 18 2.3 Blowdown System Pressure, 1.0 DECLG Break 19 2.4 Blowdown Total Break ' iction Flow Rate, 1.0 DECLG Break...

20 2.5 Blowdown Pressurizer Surge Line Flow Rate, 1.0 DECLG Break..

21 2.6 Single Intact Loop Accumulator Flow Rate, 1.0 DECLG Break 22 2.7 Double Intact Loop Accumulator Flow Rate, 1.0 DECLG Break 23 2.8 Average Channel Inlet Flow Rate, 1.0 DECLG Break.......

24 2.9 Average Channel Outlet Flow Rate,1.0 DECLG Break 25 2.10 Hot Channel Inlet Flow Rate, 1.0 DECLG Break 26 2.11 Hot Channel Outlet Flow Rate, 1.0 DECLG Bren 27 2.12 Blowdown Hot Rod Cladding Surface Temperature, Node 18, 1.0 DECLG Break 28 2.13 Blowdown Hot Rod Volumetric Average Fuel Temperature, Node 18, 1.0 DECLG Break 29 2.14 Hot Rod Blowdown Heat Transfer Coefficient, Node 20,1.0 DECLG Break 30 1

2.15 Hot Rod Blowdown Depth of Zirconium - Water Reaction, Node 18, 1.0 DECLG Break 31 2.16 Containment Backpressure Versus Time,1.0 Dt CLG Break 32 1263 296 I

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XN-NF-79-89 LIST OF FIGURES (Contd)

I Figure Page 2.17 Normalized Power, 1.0 DECLG Break............... 33 2.18 Reflood Core Flooding Rate,1.0 DECLG Break.......... 34 2.19 Reflood System Pressure, 1.0 DECLG Break 35 2.20 Reflood Downcomer Mixture Level, 1.0 DECLG Break 36 2.21 Reflood Core Mixare Level,1.0 DECLG Break.......... 37 2.22 T00DEE2 Calculated Cladding Surface Temperature,1.0 DECLG Break............................. 38 3.1 Axial Power Profile with Peak at X/L = 0.7 for Fort l

Calhoun ECCS Analysis..................... 42 5

3.2 Axial Power Profile with Peak at X/L = 0.8 for Fort Calhoun ECCS Analysis..................... 43 3.3 Axial Power Profile with Peak at X/L = 0.9 for Fort Calhoun ECCS Analysis..................... 44 3.4 Limits on Total Peaking Versus Axial Power Peak Location 45 4.1 Cladding Surface Temperature During Heatup for ENC Fuel at BOL 50 4.2 Cladding Surface Temperature During Heatup for CE Fuel at BOL 51 4.3 Cladding Surface Temperature During Heatup for ENC Fuel at E0L...

52 4.4 Cladding Surface Temperature During Heatup for CE Fuel at E0L...

53 8

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peaking factor, F, of 2.53 throughout life for CE fuel and for ENC f' el.

u The exposures used bound the maximum anticipated burnups for the respective fuel types.

The corresponding linear heat generation rate limit is 15.22 kw/ft at 1500 MWt.

The axial power profile analysis determined the ECCS limits for reactor operation by performing ECCS calculations with the peak power loca-tions at 70, 80 and 90 percent of the active core height for the limiting break geometry (1.0 DECLG case). The axial power profile analysis results are presented in Section 3.0 and surmiarized in Figure 1.1.

Figure 1.i shows that at or above an axial peak location of 70 percent of the co' e r

I height, the total peaking factor must be reduced as shown to assure that the PCT above the 70 percent location does not exceed that calculated at the 70 percent elevation.

The 1.0 DECLG case was calculated to be the most limiting break.

With a total peaking, F, of 2.53 and an axial power peak at 70% core height, the corresponding maximum peak clad temperatures (PCT's) were calculated to be 2179 F for ENC fuel and 2192 F fer CE fuel.

These PCT values were calculated for end-of-life (E0L) conditions. At BOL the corresponding PCT's for the limiting 1.0 DECLG break were calculated to be 2015 F for ENC fuel and 2029 F for CE fuel.

The effects of input revisions, incorporated after the break spectrum calculations were completed, are presented in Section 5.0.

The effect of these revisions were found to be small and in the direction of greate'r conservatism.

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4 XN-NF-79-89 2.0 BREAK SPECTRUM ANALYSIS 2.1 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION The analyses for large breaks specified by 10 CFR 50.46 (16),

" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Power Reactors", is presented in this section.

The results of the loss of coolant accident analysis are shown in Tables 2.1 and 2.2, which indicate compliance with the Acceptance Criteria.

The analytical techniques used are in compliance with Appendix K of 10 CFR 50, and are as described in XN-75-41, Volumes I and II, and supplements II); ENC-WREM-IIA model is described in XN-76-44(17), XN-76-36 (18), XN-NF-78-30 (3), XN-NF-78-25 (I9),

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and XN-76-27 plus supplements The detailed system models are as given in the Fort Calhoun example problem report, XN-NF-79-45 (15)

For the purpose of loss-of-coolant accident (LOCT) analyses, a LOCA is defined as a hypothetical rupture of the Reactor Primary 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.

Should a major break occur, depressurization of the Reactor Coolant System results in a pressure decrease in the pressurizer.

A reactor trip signal occurs when the pressurizer lower pressure trip setpoint is reached.

Reactor trip and scram were conservatively neglected for the large breck analyses. A Safety Injection System signal is actuated when a

i263 301

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XN-NF-79-89 the appropriate setpoint (high containment pressure) is reached.

These countermeasures will limit the consequences of the accident in two ways:

1. Reactor trip and borated water injection complements void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

2. Injection of borated water provides heat transfer from the reactor core and prevents excessive clad temperatures.

At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liquid which transfers heat from the core by forced convection coolir.g. After the break develops, the time to departure from nucleate boiling (DNB) is calculated consistent with Appendix K of 10 CFR 50(12)

Post DNB core heat transNr (both transition an j film boiling occurring) is also calculated in accordance with Appendix K of 10 CFR 50.

As the core becomes uncovered, both turbulent and laminar forced convection to steam are considered as core heat transfer mechanisms.

When the Reactor Coolant System pressure falls below 255 psia, the accumulators begin to inject borated water.

The conservative assumption is made that accumulator ECC water bypasses the core and goes out through the break until the termination of bypass.

This conservatism is consistent with Appendix K of 10 CFR 50.

2.2 THERMAL ANALYSIS 2.2.1 Method of Analysis For breaks greater than 0.5 ft, the RELAP4-EM Code (1,2,17) 2 is used to calculate the transient depressurization of the Reactor Coolant System as well as to describe the mass and enthalpy of flow out of the break.

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XN-NF-79-89 6

A specialized calculation (RELAp4-EM/ HOT CHANNEL) is used to calculate cladding temperatures using time dependent boundary conditions in the upper and lower plenum volumes from the basic blowdown analysis.

Beyond the point of refill to the bottom of the core, a specialized calculation (REFLEX) is applied to determine the reflooding rate and system conditions.

After end-of-bypass (E0BY), the program T00DEE2 is used to calculate peak clad temperatures.

2.2.2 Large Break LOCA Analysis Modeling The Fort Calhoun nuclear power plant is a 2X4 Combustion Engineering pressurized water reactor with a dry containment.

The reactor coolant system is nodalized into control volumes representing reasonably

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homogeneous regions, interconnected by flow-paths or " junctions" as described in XN-NF-77-45 (15)

The nodalized system blowdown model schematic is given in Figure 2.1.

One percent of the steam generator tubes were assumed to be uniformly plugged.

The unbroken loop was assumed symmetrical and modeled the same as the broken loop except for the break nodalization and the pressurizer.

Pump performance curves characteristic or the Fort Calhoun pumps as supplied by the Omaha Public Power District were used in the analysis.

System input parameters are gisen in Table 2.3.

The reactor core is modeled with heat generation rates determined fro'1 reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The axial power 1263 303

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XN-NF-79-89 profile used for the break spectrum analysis is a top skewed curve with an axial peakinc Tactor of 1.52 as given in Figure 2.2.

The values for the primary coolant system core inlet temperatures and the steam generator secondary side pressure were set for the Fort Calhoun plant based upon information provided by the utility.

The values of the core inlet temperature and the steam generator secondary side pressure are 545 F and 838 psig, respectively.

The containment backpressure for the analysis of the postulated LOCA was evaluated in accordance with the discussion presented in XN-75-41, Supplement 5, Section 4.6.

A containment analysis was per-formed using the computer code CONTEMPT-LT, Version 22, modified as des-O) cribed in Supplement 5, Revision 1, of XN-75-41 The condensing heat transfer coefficient is modeled in accordance with Branch Technical Position CSB 6-1, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation" 02)

The contaiment parameters used in the con-tainment analysis to determine the ECCS backpressure are presented in Table 2.4.

2.3 BREAK SPECTRUM RESULTS Using the ENC WREM-IIA codes, transient system behavior is deter-mined by solving the governing conservation equations for mass, energy, and momentum.

Energy transport, flow rates, and heat transfer are determined from appropriate correlations.

Table 2.1 presents the timing and sequence of events as determined for the large break guillotine configuration with discharge coefficients of 1.0, 0.6 and 0.4 and the split break configuration with break areas of 6.28, 3.77, 2.51 and 0.5 square feet.

i 1263 304 g

8 XN-NF-79-89 In general, the transient events occur slower for smaller dis-charge coefficients or break sizes.

Table 2.2 presents the peak clad temperatures and maximum metal-water reaction results for the above spectrum of break cases.

This range of break sizes was determined to include the limiting case for peak clad temperature The analysis of the loss-of-coolant accident is performed at 102%

of 1500 MWt (1530 MWt).

The core power and other parameters used in the analyses are given in Table 2.3.

Since there is usually margin between the value of the peak linear power density used in this analysis and the value expected in operation, a lower peak clad temperature would be obtained by using the peak linear power density expected during operation.

For the result discussed below, the hot spot is defined to be the location of maximum peak clad temperature'.

This location is given in Table 2.2 for each break size analyzed.

Figure 2.3 through 2.22 present the results of the analysis for the limiting break (1.0 DECLG).

Unless otherwise noted, zero time corresponds to the time of break initiation.

The maximum peak cladding temperature of 2092 F was calculated for the double-ended cold-leg guillotine break configuration (CD = 1.0) with a total linear heat generation rate of 15.53 kw/ft (F = 2.53) for ENC fuel (102% of 15.22 kw/ft).

The maximum local metal-water reaction is less than 9% all well below the limits sets by the criteria of 10 CFR 50.46.

ENC has performed numerous analyses and sensitivity studies on PWR systems using the ENC ECCS evaluation model.

These studies have 1263 305

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9 XN-NF-79-89 1

demonstrated the adequacy of the system nodalization used.

In addition, these studies have shown that for transient conditions similar to those calculated for the Fort Calhoul reactor during the LOCA, the reactor coolant inlet pipe or cold leg is the worst break location.

NSSS vendor analyses have shown large breaks to be limiting for Fort Calhoun ECCS Analyses.

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TABLE 2.1 FORT CALHOUN LARGE BREAK E'/'M Event Time (Seconds)

DECLG DECLG DECLG 1.0 DECLS 0.6 DECLS 0.4 DECLS 0.08 DECLS 2

2 2

(C =1.0)

(C =0.6)

(C =0.4)

(6.28 ft )

(3.77 ft )

(2.51 ft )

(0.5 ft2)

D D

D Start 0.0 0.0 j

0.0 0.0 0.0 0.0 0.0 l

0.05 0.05 0.05 0.05 0.05 Initiate Break 0.05 0.05 Safety Injection Signal 0.55 0.66 0.31 0.56 0.61 0.72 2.77 Accumulator Injection, Intact Loop (s) 15.50 18.20 22.40 16.50 17.20 20.20 89.36 l

8.50 8.40 8.40 8.45 9.505 Pressurized Empties 8.40 8.45 l

25.89 19.53 20.39 23.38 93.57 End-of-Bypass 19.63 21.54 Safety Injection Flow, SIS 20.45 20.56 20.71 20.48 20.51 20.61 22.67 Start of Reflood 32.76 34.91 39.36 32.64 33.64 36.70 107.40 Peak Clad Temperature Reached (sec) 159.5 165.5 181.1 171.0 161.6 164.1 337.6 x

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TABLE 2.2 FORT CALHOUN LARGE BREAK RESULTS Event DECLG DECLG DECLG 1.0 DECLS 0.6 DECLS 0.4 DECLS 0.08 DECLS 2

2 2

2 (C =1.0)

(C =0.6)

(C =0.4)

(6.28 ft )

(3.77 ft )

(2.51 ft )

(0.5 ft )

D D

D Peak Cladding Temperature, F 2092 2041 1995 2051 2045 1972 1756 Peak Temperature Location, ft 7.97 7.97 7.97 7.97 7.97 7.97 7.47 Local Zr/H O Reaction (Max.), %

8.0%

6.5%

5.1%

6.5%

6.8%

5.3%

3.3%

2 Local Zr/H O Location, ft 7.47 7.47 7.47 7.47 8.22 8.22 7.47 2

Total H2 Generation, % of +otal d

Zr Reacted

<1 %

<1%

<1 %

<l%

<1 %

Hot Rod Burst Time, sec 33.03 35.84 41.49 32.83 33.89 51.3 228.6 Hot Rod Burst Location, ft 7.47 7.47 6.99 7.47 7.47 7.47 7.47 Linear Heat Generatien Rate, kw/ft at B0CREC 0.8065 0.7947 0.7708 0.8071 0.8009 0.7847 0.6242 X

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12 XN-NF-79-89 TABLE 2.3 Fort Calhoun PWR Data Primary Heat Output, MWt 1500*

7 Primary Coolart Flow, lbm/hr 7.233 x 10 3

Primary Coolant Volume, ft 12,031**

Operating Pressure, psia 2100 Inlet Coolant Temperature, F 545 3

Reactor Vessel Volume, ft 2986 3

Pressurizer Volume, Total, ft 900 3

Pressurizer Volume, L; quid, ft 500 Accumulator Volume, Total, ft3 (one of four) 1300 3

Accumulator Volume, Liquid, ft 825 Accumulator Pressure, psia 255 Steam Generator Heat Transfer Area, ft2 (onc of two) 47.184 0

Steam Generator Secondary Flow, lbm/hr 3.38 x 10 Steam Generator Secondary Pressure, p'la 853 Reactor Coolant Pump Head, ft 2 01 Reactor Coolant Pump Speed, rpm 1192 2

Moment of Inertia, Ibm-ft / rad 71,000 Cold Leg Pipe, I.D., in 24.0 Hot Leg Pipe, I.

D., in 32.0 Pump Suction Pipe, I. D., in 24.0

Primary Heat Output used in RELAP4-EM Model - 1.02 x 1500 = 1530 MWt.

- Includes total accumulator and pressurizer volumes.

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I 13 XN-NF-79-89 TABLE 2.3 (Continued)

Fuel Assembly Rod Diameter, in*

.442 Fuel Assembly Rod Pitch, in*

.580 Fuel Assembly Pitch, in*

8.180 Fueled (Core) Height, in*

128.0 2

Fuel Heat Transfer Area, ft 28,892 2

Fuel Total Flow Area, ft 32.579 Steam Generator Tube Plugging (Assumed uniform) 1%

ENC fuel parameters 1263 310 5

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14 XN-NF-79-89 TABLC 2.4 DRY CONTAINMENT DATA Containment Physical and Thermal Parameters 6 3 Net Free Volume 1.05 x 10 ft Outside Air Temperature

-17 F Initiation Time for.

Spray Flow 55.0 sec Fan Coolers 25.0 sec Containment Initial Conditions:

Temperature 85 F Pressure 14.7 psia Relative Humidity 80%

Containment Spray Water:

Temperature 40 F Flow Rate (Total, 3 pumps) 5100 gpm Fan Air Cooler Capacity (total 4 coolers)

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Vapor Temperature ( F)

Capacity (Btu /hr) 8 150

.50 x 10 8

185 1.38 x 10 8

244 3.00 x 10 8

288 4.37 x 10 8

320 5.40 x 10 Thermal Conductivity and Volumetric Heat Capacity Thermal Volumetric Conductivity Heat Capacity MateriJ s (Btu /hr-ft-F)

(Btu /ft3-F)

Steel 26.0 59.0 Structural Concrete 0.85 32.0 Paint for Steel Surfaces 1.5 57.6 Paint for Cor. crete Surfaces 0.3 43.2

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I 15 XN-NF-79-89 TABLE 2.4 (Continued)

I DRY CONTAINMENT DATA Containment Passive Heat Sinks SURFACE AREA DESCRIPTION MATERIAL THICKNESS FT2 I

1.

Containment Cylindrical Wall paint 3 mil 44,090 steel

.25 in concrete 3.875 in 2.

Containment Dome paint 3 mil 6,850 steel

.25 in 3

concrete 3 ft 4,

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Foundation Slab paint 3 mil 8,650 g

steel

.25 in g

concrete 15 ft 4.

Refueling Cavity Walls stainless 0.06 in 14,160*

steel concrete 4 ft 5.

Refueling Cavity Floor stainless 0.06 ft 400 steel concrete 16.5 ft I

6.

Misc. Concrete paint 6 mil 8,000*

concr'te

.75 ft paint 6 mil 7.

Misc. Concrete paint 6 mil ll,150*

concrete lft paint 6 mil 8.

Misc. Concrete paint 6 mil 53,600*

concrete 2.0 ft paint 6 mil 9.

Misc. Concrete paint 6 mil 9,260*

l concrete 5 ft a

paint 6 mil 10.

Misc. Steel paint 3 mil 10,960*

steel 1.0 in paint 3 mil I

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16 TABLE 2.4 (Continued)

DRY CONTAINMENT DATA Containment Passive Heat Sinks SURFACE AREA DESCRIPTION MATERIAL THICKNESS FT2

11. Misc. Steel paint 3 mil 5,700*

steel 0.25 in paint 3 mil

12. Ventilation Ducts galvanized 0.125 in 72,000*

steel

Tabulated surface area includes areas of both sides of the slab.

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39 XN-NF-79-89 3.0 LOCA ECCS AXIAL POWER PROFILE ANALYSIS The break spectrum LOCA ECCS analysis presented in Section 2.0 identified the limiting break for the Fort Calhoun Reactor to be a guillotine break in the pump discharge line with a discharge coefficient of 1.0 (1.0 DECLG break).

The break spectrum analysis was perforoed with a power profile having the axial peak located at 0.7 of core height and a total peaking factor of 2.53.

This section expands that analysis to define the ECCS limits for axial power distribution and to show the sensitivity of peak cladding temperature (PCT) to axial power peaking.

Three axial power shapes were investigated in this study using the ENC WREM-IIA ECCS evaluation model.

The profiles used are shown in

~

Figures 3.

to 3.3.

The three profiles have power peaks located at 0.7, 0.8, and 0.9 of core height.

The results of the analyses are summarized in Table 3.1 and it..tigure 3.4.

Figure 3.4 shows the limits on tetti peaking as a function of core height.

As the location of the axial peak rises in the core, it is necessary to reduce the peak linear heat generation rate (LHGR), since it takes longer for the ECCS fluid to cool the higher core elevations in the reflood portion of the transient.

The total peaking obtained at the 0.7 of core height elevation is also applied to lower core elevaticas as shown in Figure 3.4.

This is conservative as the dominant influence on PCT in the axial studies is the time to terminate the temperature rise for the high-powered regions of the core.

With power peaking lower in the core, reflood 1263 ;36

I 40 XN-NF-79-89 temperaturt transients will be terminated earlier with a consequent reduction in PCT at the assumed peak power limit.

Reactor operation with total peaking less than or equal to the value defined by the F versus core height curve in Figure 3.4 assures that q

10 CFR 50.46 criteria are not exceeded.

I I

E 1263 337 I

I I

~

I

41 XN-NF-79-89 TABLE 3.1 RELATIVE TOTAL PEAKING VERSUS AXIAL PEAK LOCATION WITH A CONSTAffT PCT Peak Location Relative (inches) (x/L)

Peaking 79.08 0.7 1.0 92.26 0.8 0.945 105.44 0.9 0.874 9

1263 338 f

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Figure 3.1 Axial Power Profile With Peak at X/L = 0.7 for Fort w

Calhoun ECCS Analysis M

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46 XN-NF-79-89 4.0 LIMITING BREAK HEATUP ANALYSIS RESULTS INCLUDING R0D INTERNAL PRESSURE UNCERTAINTIES AND EXPOSURE SENSITIVITY This section presents the results of the fuel rod internal pressure uncertainty and exposure sensitivity analyses. The results of these analyses support an ECCS allowable total peaking, F, of 2.53 for both q

ENC and CE fuel types.

4.1 ECCS CALCULATIONS FOR CE FUEL A complete LOCA ECCS calculation for the limiting 1.0 DECLG break was performed for CE fuel. The calculational models, and assump-tions were identical to those for the ENC fuel calculation (Section 2.0).

The only difference in the calculations was that the core was modeled with CE fuel rather than ENC fuel.

Since there are only relatively small differences between these fuel types, the blowdown and heatup tran-sients for both ENC and CE fuel types are very similar.

Beginning-of-li fe peak cladding temperatures differed by only 14 F for the two fuel designs.

4.2 HEATUP ANALYSIS MODELS AND ASSUMPTIONS ECCS heatup calculations require consideration of rod internal pressure uncertainties in order to conservatively maximize the calculated flow blockage in the T00DEE2 hot rod heatup calculations.

The heatup analysis models are identical to those detailed and referenced for Prairie Island (9).

The computer codes used in performing the calculations are the RELAP4-EM/

HOT CHANNEL (version ENC 28F) and the T00DEE2/MAY79 codes.

1263 343

I 47 XN-NF-79-89 Table 2.3 sumarizes the key system parameters for the exposure heatup analysis.

These parameters apply to the analyses for both ENC and CE fuel. The core upper and lower plenum boundary conditions used for the BOL and exposed fuel H0T C!IANNEL calculations are from ~he respective limiting break blowdown calculations, and the reflood rate versus time for the T00DEE2 calculati6ns are from the limiting break reflood calculations.

4.3 FUEL R0D INTERNAL PRESSURE UNCERTAINTY ANALYSIS AND HEATUP RESULTS FOR BEGINNING-0F-LIFE Table 4.1 provides the heatup analysis results for ENC and CE fuel types at beginning-of-life.

The respective heat,, transients for ENC and CE fuel are given in Figure 4.1 and 4.2.

These heatup results include accounting for the upper bound uncertainties on rod internal pressure.

The calculated PCT's are 2015 F for ENC fuel and 2029 F for CE fuel. Maximum local metal water reaction is less than 9.0% in both cases. These results are within the Appendix K allowable limits of 2200 F and 17% and support an ECCS allowable total peaking of 2.53 for BOL.

4.4 EXPOSURE SENSITIVITY The most limiting fuel exposure case is EOL.

This case is most limiting as a consequence of burnup dependent fission gas release which has the two-fold effect of increasing stored energy due to reduced pellet-to-clad conductance and causing high flow blockage due to high rod internal pressure.

The heatup analysis results for ENC and CE fuel types at EOL are given in Table 4.1 along with those for BOL.

The E0L us3 344 I

I

48 XN-NF-79-89 PCT's are 2179"F for ENC fuel and 2192 F for CE fuel.

The respective heatup transients are given in Figurts 4.3 and 4.4.

flaximum local metal water reaction is less than 9.0% in both cases. Again, these results are within the 10 CFR 50.46 limits so that an ECCS allowable total peaking of 2.53 (15.22 kw/ft) is supported throughout life for the respective fuel types.

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TABLE 4.1 Fort Calhoun Exposure Heatup Analyses Results for ENC and CE Fuel ENC FUEL CE FUEL BOL E0L BOL E0L Total Peaking, F 2.53 2.53 2.53 2.53 q

Peak Clad Temperature (PCT),

F 2015 2179 2029 2192.0 Max. Local Zr/H 0 - Reaction, percent 6.4 8.7 7.6 11.5 2

Hot Rod Burst Time, sec 35.4 61.4 32.4 29.6 8

Hot Rod Burst Location, ft 7.47 7.72 7.47 7.47 Time of PCT, sec 197 257 196 250 PCT Location, ft 8.22 8.47 8.22 8.22

?

N Max. Zr/H O Reaction Locat'on, ft 8.22 8.47 8.22 3.22

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TIME - SECONOS Figure 4.1 Cladding Surface Temperature During Heatup for ENC Fuel at BOL

FORT CALHOUN 1.0 DECLG CE FUEL B0C HOT R0D HEATUP

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oo Figure 4. 2 Cladding Surface Temperature During Heatup for CE Fuel at BOL M

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TIME - SECONDS Figure 4.3 Cladding Surface Temperature During Heatup for ENC Fuel at E0L i

--- ----i- ---i-

1 Fort Calhoun 1.0 DECLG E0C Exp CE Fuel HR Heatup I

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W

XN-NF-79-89 54 5.0 SENSITIVITY ANALYSES _

Large break ECCS sensitivity calculations were performed to evaluate the effects of two minor input system parameter changes identified after the break spectrum analysis was performed: (1) the ENC fuel cladding outside diameter (0.C) was increased from 0.440 in. to 0.442 in.

i and (2) the two hot leg volumes in the reactor primary coolant system were reduced by approximately 15 cubic feet each.

The sensitivity cal-culations were performed for the previously defined limiting break (1.0 DECLG) using the ENC WREM-IIA ECCS evaluation models (9,15),

The ENC fuel cladding 0.D. change occurred after the break spectrum calculations discussed in Sectirn 2 were completed, but was incorporated in the axial power profile and exposure analyses.

After all of the

CS calculations were completed, an incorrect length dimension for the hot leg straight pipe was identified. The hot leg length was accordingly corrected resulting a reduction in the total primary coolant system volume of less than 0.3 percent. A negligible impact was expected to result from this small system volume change, however, the limiting break was recalculated to verify this effect.

5.1 RESULTS The results for the sensitivity studies were compared with 1.0 DECLG base case from the break spectrum analysis.

The increase in cladding diameter was calculated to decrease PCT by 43 F.

The reduction in hot leg volume decreased PCT by an additional 5 F.

These sensitivity studies confirm that the effects were small and in a conservative direction for both revisions.

Thus, the ECCS analysis calculations remain valid and i263 351

I XN-NF-79-89 55 in conformance with 10 CFR 50.46 and Appendix K when these minor revisions are implemented.

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1263 352 I

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XN-NF-79-89 56

6.0 CONCLUSION

S For breaks up to and including the double-ended severance of a reactor coolant pipe, the Fort Calhoun Emergency Core Coolint System will meet the Acceptance Criteria as presented in 10 CFR 50.46 with the cycle 6 ccre aad future anticipated ENC reload fuel of similar design.

That is:

1.

The calculated peak fuel element clad temperature does not exceed the 2200 F 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.

The cladding temperature transient is terminated at a time when the core geometry is still ?menable to cooling.

The hot fuel rod cladding oxidation limits of 17% are not exceeded during or after quenching.

4.

The system long term cooling capabilities provided for previous cores remains applicable for ENC fuel.

These Acceptance Criteria are satisfied if the Fort Calhoun reactor is operated at 1500 MW(t) within the maximum LHGR of 15.22 kw/ft and the exposure limits and axial profile limits given in Figures 1.1 and 1.2.

It is anticipated that operation within the allowed maximum LHGR limit and allowed axial offset limits will neutronically preclude total peaking above the 70% of core height level from reaching the limits shown by Figure 1.1 l263 353

~~

XN-NF-79-89 57

7.0 REFERENCES

1.

Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model, XN-75-41:

a.

Volume I, July 1975 b.

Volume II, August 1975 c.

Volume III, Revision 2, August 1975 d.

Supplement 1, August 1975 e.

Supplement 2, August 1975 f.

Supplement 3, Augus,1975 g.

Supplement 4, August 1975 h.

Supplement 5, Revision 5, October 1975 i.

Supplement 6, October 1975 j.

Supplement 7, November 1975 2.

Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-II, XN-76-27:

3.

Exxon Nuclear Company, Exxon Nuclear, Mmpany WREM-Based Generic PWR ECCS Evaluation Model Update ENC.! REM-IIA, XN-NF-78-30.

4.

U. S. Nuclear Regulatory Commission, WREM, Water Reactor Evaluation Model_, NUREG-75/056, Revision 1, May 1975.

5.

Exxon Nuclear Company, Exxon Nuclear Company ECCS Evaluation of a 2-Loop Westinghcuse PWR With Dri Containment Using the ENC WREM-II ECCS Model - Large Break Example Problem, XN-NF-77-25 (A)

September 1978.

6.

Exxon Nuclear Company, Big Rock Point Example LOCA Analysis Using The Exxon Nuclear Company Non-Jet Pump BWR Evaluation Model -

Large Break Example Problem, XN-NF-78-25, Revision 1, September 1978.

7.

U. S. Nuclear Regulatory Commission, letter, T. A. Ippolito (NRC) to W. S. Nechodom (ENC), SER for ENC RELAP4-EM update, March 1979.

8.

U. S. Nuclear Regulatory Commission, letter, T. A. Ippolito (NRC) to W. S. Nechodom (ENC), SER for ENC WREM-IIA Evaluation Model, March 30, 1979.

9.

Exxon Nuclear Company, Exposure Sensitivity Study for ENC XN-1 Reload Fuel at Prairie Island Unit 1 Using the ENC-WREM-IIA PWR Evaluation Model, XN-NF-79-18, March 1979.

[263 354

XN-NF-79-89 58

10. Block, J. A., and Wallis, G. B., Effect of Hot Walls on Flow in a Simulated PWR Downcomer During a LOCA, CREARE-TN-188, May 1974.

11. Block, J. A., and Crowley, C. J., Hot Wall Experiments in Simu-lated Multiloop PWR Geometry, CREARE-TN-202, February 1975.

12. U.S. Nuclear Regulatory Commission, Minimum Containment P essure Model for PWR ECCS Performance Evaluation, Branch Technical Posi-tion CSB 601.

13. Exxon Nuclear Company, ECCS Large Break Spectrum Analysis for Prairie Island Unit 1 Using ENC WREM-IIA PWR Evaluation Model, XN-NF-78-46, November 1978.

14. NRC memo, N. Lauben to Z. Rosztocey, dated December 12, 1978.

15. {xyca Nuclear Company, Fort Calhoun LOCA Analyses at 1500 MWT Using CNC WREM-IIA PWR ECCS Evaluation Model - Large Break Example Problem, XN-NF-79-45, May 1979.

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

17. Exxon Nuclear Company, Revised Nucleate Boiling Lockout for ENC WREM-Based ECCS Evaluation Models, XN-76-44, September 1976.

18. Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generi.-

PWR ECCS Evaluation Model ENC-WREM-II) Loop PWR With Ice Condenser, Large Break Example Problem, XN-76-36, August 1976.

19. Exxon Nuclear Company, Big Rock Point Example LOCA Analysis Using The Exxon Nuclear Company Non-Jet Pump BWR Evaluation Model -

Large Break Example Problem, XN-NF-78-25, Revision 1, September 1978.

I 1263 355 g

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ML19254E829

Contents

Text

1.0 INTRODUCTION

SUMMARY

6.0 CONCLUSION

7.0 REFERENCES

6.0 CONCLUSION

7.0 REFERENCES

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ML19254E829

Text

1.0 INTRODUCTION

SUMMARY

6.0 CONCLUSION

7.0 REFERENCES

6.0 CONCLUSION

7.0 REFERENCES

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