ML20079E029

Omaha Public Power District Fort Calhoun Station Unit 1 Loss of Coolant Accident Analysis

ML20079E029
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Site:	Fort Calhoun
Issue date:	09/13/1991
From:	OMAHA PUBLIC POWER DISTRICT
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Oh1AllA PUBLIC POWER DISTRICT '

FORT CALilOUN STATION UNIT NO.1 LOSS OF COOLANT ACCIDENT ANALYSIS SEPTEhiBliR 13, 1991 Page 1 of 101

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REACTOR COOLANT SYSTFM PIPE RUPTURE (L()SS OF COOL ANT ACCIDENT)

A safety analysis of postulated pipe ruptures within the Reactor Coolant System (RCS)boundaryhasbeenperformed. This analysis has included cases of the Loss-of-Coolant Accident (LOCA) resulting from a spectra of small end large pipe ruptures including the Maximum flypothetical Accident (MilA) case of the double-ended break of the largest cold leg pipe.

The objective of the analysis was to determine the conditions of the RCS, core, and containment in the ever.t of a postulated LOCA, and to verify that

.the Emergency Core Cooling System (ECCS) has the capability to mitigate each LOCA, including the MilA.

1. GENERAL 1.1 Accident Description ALOCAresultsfromaruptureoftheReactorCoolantSystem(RCS)orof any line connected to that system up to the first isolation valve. The charging pumps have the capability to make up for leakage resulting from ruptures of a small cross section, thus permitting an orderly shutdown.

The coolant released would remain in the containment.

For.a postulated break, a reactor trip is initiated when the pressurizer pressure-low setpoint is reached. A S6fety injection Actuation Signal (SIAS) is actuated by either a containment pressure hijk or a pressurizer  !

pressure low-low signal. The consequences of the accident are limited in i two wayst:

i

1. Reactor t'ip and safety injection (of borated water) supplement void formation in causing a rapid reduction of the nuclear power to a j residual level enrresponding to the delayed fission product decay. I For-a postulated Large Break LOCA, credit for Cf.A insertion to keep the reactor subtritical post 4LOCA is not taken since large break l n forces may degrade the trip function of the CEAs.-

The requirement for post-LOCA subcriticality is met by maintaining a sufficiently high boron concentration in the Safety Injection and .

RefuelingWaterTank(SIRWT). <

Page 2 of 101 u.. - _ . - . - - . - - _ - -. - . - ..-. a_. - . .. - ,:

2. Injection of borated wa"- nsures sufficient flooding of the core to prevent excessive fuel . eratures.

Before the reactor ; rip occurs, the reactor is in an equilibrium condition, i.e., the heat generated ir the core is being removed via the secondary system. After reactor trip and turbine trip, core heat and heat from the vessel and internals is transferred to the RCS fluid, and then to the containment and the secondary system dependent upon the break size.

The reactor coolant pumps are tripped at the initiation of the accident due to an assumed loss of offsite power or remain running until operator actions trip the pumps. The effects of the RCP coastdown are included in W the blowdown analyses. Without a loss of offsite power, a Steam Generator Isolation _ Signal (SGIS), which occurs as a result of containment pressure-high, terminates normal feedwater flow by closing the main feedwater isolation valves. With the assumed loss of offsite power, main feedwater is lost with the coastdown of the electric motor driven main feedwater pumps. If offsite power is available, the steam is dumped to the condenser, although not credited in the analysis; if offsite power is not available, the steam is assumed to be dumped to the atmosphere via the main steam safety valves. Makeup to the steam generators is initiated by the auxiliary feedwater pumps if steam

+ generator level falls below the auxiliary feedwater system actuation a setpoint. If not terminated previously by High Pressure Safety injection (HPSI) pump flow, the core uncovery transient is turned around when the O RCS pressure falls below approximately 755 psia and the Safety injection

- Tanks inject borated water.

1.2 Performance Criteria for Emeroency Core Coolingjiystem A

The reactor is designed to withstand the thermal effects caused by a loss-of-ccolant accident including the double-ended severance of the largest Reactor Ccoiant System pipe. The reactor core and internals together with the Emergency Core Cooiing System are designed so that the

' reacter can be "afely shutdown and that the essential heat transfer r

geometry of the core is preserved following the accident. The Emergency Page 3 of 101

Core Cooling System, even when operating during the injection mode with the most severe single active failure, is designed to meet-the requirements of 10CFR50.462 , " Acceptance Criteria for Emergency Core CoolinD Systems for Light Water Power Reactors." The requirements are:

1. Peak clad temperatures (PCT) do not exceed 2200*F.

2. The amount of cladding that chemically reacts with the coolant does not exceed 1% of the zircaloy cladding sur ounding the fuel, excluding the cladding surrounding the plenum volume.

3. Oxidation of the cladding does not exceed 17% of the orig nal d cladding thickness, which precludes embrittlement problems.

4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5. After initial operation of the ECCS, core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period required by the long-lived radioactivity remaining in the core.

1.3 , Thermal Analysis Description The analysis specified by 100FR50.46, is presented in Sections 2 and 3 for large and'small breaks, respectively. The results of the large and small break loss-of-coolant ucident analyses which are summarized in Tables 2-3 and 3-3 show compliance with the above Acceptance Criteria.

The highest PCT calculated was for a double-ended cold leg guillotine (DECLG)' break with a Moody discharge coefficient (C o ) of 0.4.

The large break analysis is based on the initial reactor conditions shown L in Table 2-1. The analytical techniques ur u b.e in compliance with Appendix K of 10CFR50, and are described in the topical report, l " Westinghouse ECCS Evaluation Model - Summary."2.3 The description of the various' aspects of the LOCA analysis methodology is provided in u

References 4, 5, 6, and 26.

These documents describe the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the-Acceptance Criteria.

Page 4 of 101

= . = . . .

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The method of analysis to determine peak clad temperature is divided into two types of analysis: 1) large break LOCA, and 2) small break LOCA.

The methods of analysis for the large and small break LOCAs are described below and results are given.

2. LARGE BREAK LOCA ANALYSIS 2.1 Description of Analysis and Assumn11ons Should a major break occur, depressurization of the Rea tor Coolant System results in a decrease in pressurizer pressure. A reactor trip signal eccurs when the pressurizcr pressure-low trip setpoint is reached.

lhe Safety injection System is actuated when either the pressurizer pressure low-low setpoint or the containment pressure-high setpoint is reached. At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling.

Af ter the break develops, the time to departure from nucleate boiling is calculated, consistent with the requirements of Appendix K of 10CfR50.'

,qereafter, the core heat transfer is unstable, with both nucleate boiling and film boiling occurring. As the core becomes uncovered, both turbulent and laminar forced convection and radiation are considered as core heat transfer mechanisms.

When the Reactor Coolant System pressure falls Lelow approximately 255 psia, the Safety injection Tanks begin to inject borated water. A conservative assumption is made that injected water bypasses the core and goes out through the break until the termination of bypass. This conservatism is consistent with the requirements of Appendix K of 10CFR50. The termination of bypass is defined as the commencement of a continuous flow of liquid down the downcomer into the lower plenum.

Sensitivity studies,' for large break LOCA were performed which covered a break spectrum utilizing a double-ended cold leg guillotine (DECLG) break with various values of discharge coefficient (Co), a double-ended hot leg guillotine (DEHLG) break, a double-ended pump suction guillotine (DEPSG) break, and a range of split-type break sizes ranging from a 1.0 f t' area to a full double-ended area of the cold leg. This study determined that, Page 5 of 101

for Westinghouse plant designs, the DECLG type break was both the most limiting type and location. Furthermore, Combustion Engineering's Analysis8of the Fort Calhoun plant has shown that the DECLG type break was both the most limiting type _and location for that unit. Therefore, the spectrum of breaks was limited to'only DECLG types for fort Calhoun Station. The initial large break spectra was analyzed with a core which contained 424 burnable shims, it was performed at 102 percent of 1500 MW , (the core licensed power), with a vessel flow rate of 196,000 gpm, a t

6%' steam generator plugging level in each generator, and a 30.9 second delay in delivery of pumped Si ficw with loss of.offsite power.

A range of-axial core power distributions was studied, as required for LOCA analysis (Appendix K of 10CFR50 ), and the distribution resulting in 1

the most severe calculated consequences selected for analysis. The power distributions considered hot spot elevations from mid-core to 8.75 ft, andan'AxialShapeIndex(ASI)rangeof0.0to-0.16asiu. The axial

-core power distribution in Figure 2-2 was found to be limiting. This power distribution, determined by Westinghouse methods, was found to be similar to that used in the most recent LOCA analysis performed by Combustion Engineering.

To determine the limiting plant conditions, Loss of Offsite Power, and No

-Loss of Offsite Power for both the minimum and maximum SI flow conditions with modeling_of pumped SI/SI Tank flow interactions were analyzed. No Loss of Offsite Power with Minimum Safeguards was found to produce the most severe consequences.

= The large break LOCA analysis presented here was performed with the 1981 BART for CE NSSS Evaluation Model (EM).26 .important features of this

~

[ + a L -model are described in other Westinghouse Evaluation Models.12.0,u.15 j The 1981 + BART for CE NSSS Evaluation Model incorporates the core heat transfer portions of the BART computer code to obtain a mechanistic core heat transfer model. The 1981 + BART EM was approved by the NRC for use 2

in PWRs with Westinghouse fuel 12 The CE NSSS Evaluation Model s has been submitted to the NRC for review as a topical report under separate transmittal.

Page 6 of 101

2.2 Method of Larae Break Analysis The large break LOCA transient can be conveniently divided into three periods: blowdown, refill, and reflood. Also, three physical parts of the transient are analyzed for each period: the thermal-hydraulic transient in +he reactor coolant system, the containment pressure and temperature, and the fuel and cladding temperatures of the hottest rod.

These considerations lead to the use of a system of computer codes designed to model the LOCA transient. A detailed description of the various aspects of the LOCA analysis is given in WCAP-8339.2 The SATAN-VI' code evaluates the thermal-hydraulic transient during blowdown. The EREFLOOD w6 code, using output from the SATAN-VI code, computes the time to bottom of core recovery (BCR), RCS conditions at BCR and mass and energy release from the break during the reflood ptcse of the LOCA. The C0C0" code is used to model the containment pres .'e transient. Since the mass flow rate to the containment depends upon the core flooding rate and the local core pressure, which is a function of the containment backpressure, the EREFLOOD and C0C0 codes are interactively linked. The BCR conditions calculated by WREFLOOD and the containment pressure transient calculated by C0C0 are used as input to >

the LOCBART code.

in the 1981 + BART for CE NSSS model, the cladding heat-up transient is

~

calculated by LOCBART which is a combination of the LOCTA6.16 code with 12 the BART code In LOCBART, the empirical FLECHT correlation has been replaced by the BART methodology. BART methodology employs mechanistic models to generate heat transfer coefficients appropriate to the actual flow and heat transfer regimes experienced by the fuel rods. During reflood, the LOCBART code provides a significant improvement in the prediction of fuel rod behavior. A more detailed description of the LOCBART code can be found in References 12 and 17.

The calculation of peak cladding temperature is performed by modeling the hottest fuel assembly (from the reactor core) in LOCBART. The Improved Page 7 of 101

Fuel Performance Modela l is used to generate the initial input parameters for fuel rod conditions for LOCBART. The hottest fuel assembly is subdivided into three regions:

1. The hottest rod

2. Adjacent rod to the hottest rod

3. The average fuel channel in the hot assembly.

While the peak cladding temperature occurs on the hottest rod, the adjacent rod and average rod temperatures provide the required conditions for heat transfer within the hot assembly.

Figure 2-1 shows the interaction of the 1981 + BART for CE NSSS model and the relationship of the computer codes to the LOCA sequence of events.

In this analysis an additional node was included in the RATAN lower plenum model to accurately describe the Fort Calhoun configuration and WREFLOOD was expanded to accurately model both cold legs in the broken loop.

2.3 Results of Larae Break Analysis A break spectrum sensitivity analysis with a range of Moody discharge coefficients (Co = 0.4, 0.6 and 0.8) was run assuming a loss of off site power, minimum SI flow, F g i = 1.75 with non-Integral Fuel Burnable Absorber (IFBA)fuelrods. The time sequence of events for this break spectrum is given in Table 2-2 and the associated peak cladding temperatures (PCT) and hot spot metal reactions are su e rbed in Table 2-3. Based upon the results of Table 2-3, an additional spectra of analyses in which FnI = 1.80 (in order to bound Cycle 14 and future cycle operation) and a matrix of minimum and maximum SI flows, with and without the loss of offsite power were analyzed. The limiting (Co = 0.4) break size remains valid for the increased F p I value since the increase in PCT would be consistew. for all the break sizes previously analyzed.

The core axial power shape sensitivity was also examined, in the break spectrum sensitivity analysis, with a set of shapes covering a range of ASIS from 0.0 to -0.16 asiu. The hot spot elevations correspondingly ranged from 5.33 ft to 8.75 ft. The limiting shape was iound to occur at the ASI limit of .16 asiu, as shown in Figure 2-2.

Page 8 of 101

] .

The second set of analyses also included the case of no loss of_offsite power to establish the limiting plant conditions. Both Minimum and Maximum SI conditions were evaluated with the reactor coolant pumps running and as noted above, the Minimum Si case found to be limiting with a PCT.of 2066*F, Table 2-4. The limiting discharge coefficient Co = 0.4

-and the limiting power shape were included in both cases. For the

-Maximum Safeguards (Maximum SI) calculations, no loss of Safety injection was assumed and nominal Si pump flows were used. Maximum SI calculations were then performed for the case of loss of offsite power. The maximum PCT for the cases of Maximum Si analyzed was 2032 F, as indicated in Table 2-4. Thus, the plant is Minimum SI, no loss of offsite power limited for the large break LOCA.

'A study was performed on the_IFBA fuel rods to establish the limiting fuel type.- As typically observed with )! fuel designs, the non-IFBA fuel

-is limiting, with the IFBA fuel providing a PCT benefit of not less than

-15'F'over non-IFBA fuel for the corresponding conditions. An evaluation was also performed by OPPD comparing Westinghouse and CE fuel rod temperatures over the appropriate range of power and burnup, it was determined that Westinghouse fuel was' limiting due to higher fuel rod

. temperatures', i.e. higher' stored energy.

figures _2-3 through 2-18 present the limiting (Co = 0.4)- LOCA case (PCT =

20660 ) of Minimum SI flow, no loss of offsite power and Fa' = 1.80.

Figures 2-19 through 2-60 provide the principal parameters for the

! initial break _ spectra-(Co = -.04, 0.6 and 0.8) usinal Maximur si loss of

~o ffsite power, and Fgi = 1.75 as input assumptions. The following parameters-are provided for each break case:

1. Core Power. Transient

2. Core Pressure Transient

3. Break Flow Rate

4. Containment Pressure-

5. Safety Injection Tank' Flow (Blowdown)

L' 6. Pumped ECCS Flow (Reflood)

7. Reflood Core and Downcomer Levels L 8. Core Reflood Rate l 9. Core Mass Velocity Page 9 of 101 l

l

- .. . . . . - . . - ~ --

10. Core Flow (Top and Bottom)

11. Core Fluid Quality

12. Core Heat Transfer

.13. Core f_luid Temperature

14. Peak Cladding Temperature

'In addition, Break Energy Release (Blowdown), Figure 2-6 and Containment Wall Heat Transfer Coefficient, Figure 2-7, are provided for the limiting case

3. SMALL BREAK LOCA ANALYSIS 3.1_ Dngr_Jotion of Analysis and Assumptions The Fort Calhoun srall break LOCA analysis was performed using the Westinghouse ECCS Small Break Evaluation model" which utilizes the NOTRUMP'*2 *** and LOCTA-IV" computer codes. The small break was analyzed with a core which contained 424 burnable shims. Figure 3-1 shows the interaction of the computer codes used to evaluate the small break cases.

The core power level utilized in the small break LOCA analysis was 102%

of 1500 MWt,- the licensed' power. The peak linear heat rate and peaking factor used in the analyses are also given in Table 3-1. Additional s assumptions for. the analysis of the Fort Calhoun Station were; operation at a total primary system flow rate of 196,000 gpm, a 6% steam generator tube plugging in each of the two steam generators, and a 30.9 second delay in delivery of pumped ECCS flow assuming Loss of Offsite Power coincident with Reactor Trip. Eight of the ten Main Steam Safety Valves were assumed to be operable. They were a,sumed to be set at 3% above the Technical Specification setpoint value and require an addit'ional 3%

accumulation before being assumed fully opea.

Figure 3-2 presents the_ axial power shape utilized to perform the small break analysis. This axial power shape was chosen because-it provides distribution of power versus core height which will maximize PCT using

.16 asiu for the Technical Specification Axial Shape Index to provide the most top peaked power distribution.

Page 10 of 101

A spectrum of cold leg break sizes, .022, .049 and .087 f t ,2 was analyzed in order to determine the most limiting break size. These breaks were analyzed following the method presented in Section 3.2. Additionally, the no loss of offsite power with the trip 2/ leave 2 RCP strategy case 4 was studied and found to be non-limiting.

3.2 Method of Small Break Analysis n ,23,ts computer coaes are used to The NOTRUMP20,21.22.23,26 and LOCTA-IV perform the analysis of Loss-Of-Coolant Accidents for small breaks _in the 2 ReactorCoolantSystem(RCS). The NOTRUMP computer code, approved for use by the Nuclear Regulatory Commission (NRC),-is used to calculate the

. transient depressurization of the RCS as well as to describe the mass and enthalpy of.the flow through the reactor core and through the break.

This code is a state-of-the-art one-dimensional general network code incorporating a number of advanced features. Among these features are the' utilization of nonequilibrium thermal calculation in all fluid volumes, flow regime-dependent drift flux calculations with counter-current l flooding limitations, mixture level tracking logic in multiple-stack fluid nodes and regime-dependent heat _ transfer

-correlations. The NOTRUMP small break LOCA emergency core cooling-system evaluation model was developed to determine the RCS response to design basis small break LOCAs and to address the NRC concerns expressed in NUREG-0611, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse-Designed Operating Plants."

In-NOTRUMP, the RCS is subdivided into fluid filled control volumes (fluid nodes)-and metal nodes interconnected by flowpaths and heat L.- -transfer links. -The transient behavior of the system is determined from l

L the governing conservation equations of mass, energy, and momentum applied to these nodes. Both the broken loop and the intact loop.are e modeled explicitly with two cold legs and two reactor coolant pumps in L each providing the option for analysis of the " trip 2 leave 2" RCP strategy. A detailed description of the NOTRUMP code is provided'in

References 21,22,23, ana 26.

Page 11 of 101

1 In the NOTRUMP model", the reactor core is represented as a vertical stack of heated control volumes with an associated bubble rise model to i

i permit a transient mixture height' calculation. The multi-node capability.

I of the program enables the explicit and detailed spatial representation of.various system components. In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant accident.

Clad thermal analysis is performed with the LOCTA-IV" computer code which uses as input-the RCS pressure, fuel rod power history, steam flow i past the uncovered part of the core, and_ mixture height history from the NOTRUMP hydraulic calculations as input. For all computations, the NOTRUMP and LOCTA-IV calculations were terminated when the core mixture level returned to the top of the core.

A schematic representation of the computer code interfai:es is given in Figure 3-1.--

3.3' Results of Small Break Analysis The small break LOCA analyses were performed with the assumptions contained in Table 3-1. The time sequence of events and results of key 2

parameters of for the .022, .049'and .087 ft breaks analyzed are shown.

2

'.in Tables 3-2 and 3-3, respectively. The_.049 ft break is shown to be the limiting break. This is the largest break terminated by the HPSI pumps; with each Safety Injection-Tank injecting less than 2 ft3 of SI flow. The peak clad temperature for the .049 ft 2break was 1444'F, occurring for both the IFBA and non-IFBA fuel rods .at the beginning of

- cycle, 0 MWD /MT'J. The maximum local zirconium oxidation was 0.40% which

'isLless than the 1% core average criteria as required by 10CFR50.46.

These results' indicate that a coolable geometry would be maintained for small break LOCAs. Figures 3-3 througi. 3-20 show RCS pressure, core mixture-height, peak clad temperature, steam _ flow rate, rod film coefficient, and hot spot fluid temperatures versus time for each break size.

Page 12 of 101

e REFERENCES

1. " Acceptance Criteria for Emer9ency Core Cooling Systems for Light Water >

Cooled Nuclear Power Reactors," 10CFR50.46 and Appendix K of 10CFR50.

Federal Register Vol. 39,-Number 3, January 4, 1974. [

2. .Bordelon, F. M., Massie, H. W., and Zordan, T. A., " Westinghouse ECCS Evaluation Model - Summary," WCAP-8 39, July 1974.

3. 'Bordelon F. M..;et al., "The Westinghouse ECCS Evaluation Mod 71 and Supplementary Information," WCAP-8471-P-A (Proprietary), WCAP-8472 '

(Non-Proprietary). January 1975.

4. "Westinghcuse ECCS Evaluation Model February 1978 Version,"

WCAP-9220-P-A (Proprietary) and WCAP-9221-A (Non-Proprietary), February 1978.

5. " Westinghouse ECCS Evaluation Model 1981 Version," WCAP-9220-P-A, Rev. I (Proprietary) and WCAP-9221-A, Rev. 1 (Non-Proprietary), February 1982.

6. Ferguson, K. L., and R.-M. Kemper, "ECCS Evaluation Model for

-Westinghouse Fuel Reloads of Combustion Engineering NSSS," WCAP-9528 (Proprietary), June 1979.

7. Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340 (Proprietary)andWCAP-8356(Non-Proprietary), July 1974.

8. " Fort Calhoun Station Unit No. 1 Updated Safety Analysis Report", Omaha Public Power District, July 1989.

9. Bordelon, F. M.,-et al._, " SATAN-VI Program: Comprehensive Space-Time Dependent;AnalysisofLoss-of-Coolant",WCAP-8302(Proprietary)and WCAP-8306(No'n-Proprietary). June 1974.

10. Kelly,'R D., et. al., " Calculation Model for Core Reflooding After A Loss-Of-CoolantAccident(WREFLOODCode),"WCAP-8170(Proprietary)and WCAP-8171(Non-Proprietary), June 1974.

11. Bordelon,-F. M. and E. T. Murphy, " Containment Pressure Analysis Code (C0CO),"WCAP-8327(Proprietary)andWCAP-8326(Non-Proprietary), June 1974.

12. Young, M. Y., et al., "BART-A1: A Computer Code for the Best Estimate Analysis'of Reflood Transier.ts," WCAP-9561-P-A (Proprietary) and WCAP-9565-A (Non-Proprietary), March 1984.

Z

13. Chiou, J. S. et. al., " Spacer Grid Heat Transfer Effects During Reflood,"

WCAP-10484 (Proprietary) and WCAP-10485 (Non-Proprietary), March 1986, i

h Page 13 of 101

- . - - . . ~ . - .- - - . - - . - - -

14. LetterfromCecil0. Thomas _(NRC) toe.P. Rahe (Westinghouse)," Acceptance forReferencingofLicensing_TopicalReportWCAP-10484(P)/10485(NP),

" Spacer Grid _ Heat Transfer Effects During Reflood." i

15. Young,:M. Y., " Addendum To: BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients (Special Report: Thimble Modeling la Westinghouse ECCS Evaluution Model," WCAP-9561-P-A, Rev. 1,. Addendum 3 (Proprietary) 'and WCAP-9561-NP-A, Rev.1, Addendum 3 (Non-Proprietary),

July 1986.

16. Bordelon, F..M., et. al., "LOCTA-1V Program: Loss of Coolant Transient Analysis,"WCAP-8301(Proprietary)andWCAP-8305'(Non-Proprietary), June 1974.

17. Kabadi, J.N., et al., "The 1981 Version of the Westirighouse ECCS Evaluation Model Using the BASH Code Addendum 2; Bash Methodology improvements and Reliability Enhancements," WCAP-10266-P-A, Rev. 2, Addendum 2, May 1988.

18. Weiner,:R. A.,_et al., " Improved fuel Performance Models for-Westinghouse fuel Rod Design-and Safety Evaluations," WCAP-10851-P-A, August'1988.

19. Lee, N., et al, " Westinghouse Small Break LOCA ECCS Evaluation Model using the NOTRUMP code," WCAP-10054-P-A (Proprietary) and WCAP-10081-A (Non-Proprietary), August 1985.-

20. Meyer, P. E.,'et. al., "NOTRUMP: A Nodal Transient General Network Code,"

WCAP-10076(Proprietary), March 1982.

21. Meyer, P. E., "NOTRUMP, A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Non-Proprietary), August 1985.

22. Bajorek, S. M.,'et al, " Addendum to the. Westinghouse Small Break ECCS Evaluation Model Using the N9 TRUMP Code for the Combustion Engineering NSSS", WCAP-10054-P-A, Addendum 1, March 1987.

23. -Rupprecht, S. D., et al, " Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," WCAP-11145-P-A (Proprietary),

andWCAP-11372-A(Non-Proprietary), October 1986.

24. Report on Small Break Accidents for Westinghouse NSSS, WCAP-9600, June

-1979.

25. LetterfromW.J. Johnson (Westinghouse)toThomasE.Murley(NRC),

"10CFR50.46 Annual Notification for 1989 of Modifications in the Westinghouse ECCS Evaluation Models", NS-NRC-89-3463, October 5, 1989.

E6. -Akers, J., et al, " Westinghouse ECCS Evaluation Model for Analysis of CE-NSSS",WCAP-13027-P(Proprietary)andWCAP-13028(Non-Proprietary), July 1991.

Page 14 of 101 '

TABLE 2-1 FORT CALHOUN LARGE BREAK LOCA ANALYSIS i

INPUT PARAMETERS AND ASSUMPTIONS NSSS Power - 102% of 1500 Mwt 1530 Mut Peak Linear Heat Rate - at-102% of 1500 Mwt 15.5 Kw/ft.

Radial Peaking Factor (F ') = 1.80 Maximum Allowable Peaking Factor (F,) = 2.545 Axial Power Distribution See Figure 3-2 Reactor Coolant System Pressure = 2100 psia Reactor Coolant System' Flow Rate = 196,000 gpm

[ Reactor Inlet-Temperature = 545 *F 3

Reactor Trip Signal (Including uncertainties) = 1728 psia, Pressurizer Pressure LOW SI Signal (Including uncertainties) = 1578 psia, Pressurizer Pressure LOW-LOW Safety Injection Tank Water Volume = 825 f t*/ Tank Safety Injection Tank Minimum Pressure = 255 psia Steam Generator Tube Plugging Level = 6% (Uniform)

TABLE ~2-2 FORT CALHOUN LARGE BREAK LOCA' ANALYSIS LARGE BREAK'SE00ENCE OF EVENTS MIN. SI FLOW MIN. SI FLOW MIN. SI FLOW DECLG ' Cm=0.4 DECLG Cm=0.6 DECLG Cm=0.8 Start 0.0 0.0 0.0 Rx Trip Signal 0.60 0.59 0.59 S.I. Actuation Signal 0.97 0.77 0.67

? 5.I. Tank Injection 22.80 16.80 14.00 5

g Pump Injection Begins 31.87 31.67 31.57 E, End of Bypass 28.92 20.59 17.48

!! End of Blowdown' 28.92 20.59 17.48 Bottom of Core Recovery 39.34 31.73 28.52 S.I. Tanks Empty 94.92 90.14 88.01 Note: All times are in seconds.

' TABLE 2-3 FORT CAltl00N LARGE BREAK LOCA ANALYSIS BREAK SPECTRUM SENSITIVITY ANALYSIS RESULTS MIN. SI FLOW MIN.SI FLOW MIN. Si. FLOW F',, = 1. 7 5 F' = 1.75 F'o = 1.75 RESULTS DECLG Ca0.4 DECLG Cm=0.6 DECLG Cm=0.8 Peak Clad Temperature (*F) 1981. 1869. 1815. '

Peak Clad Temp. Elevation (Ft.) 9.25 9.25 9.25 Peak Clad Temperature time (Sec) 113.9 98.3 86.8 Max local Zr/H,0 Reaction (%) 2.98 2.88 2.38 Total Zr/H 2O Reaction (%) <1.0 <1.0 <1.0 E,

- Hot Assy. Burst Time (sec.) 47.4 69.5 61.1 S

Hot Assy. Burst Elevation (Ft.) 8.75 9.00 8.75 Blockage on Hot Rod (%) 41.0 35.2 38.8

, TABLE 2-4 FORT CALHOUN LARGE BREAK LOCA ANALYSIS LARGE BREAK-LOCA.RESULTS MIN.'SI FLOW, MIN. SI FLOW, . MAX. SI FLOW, F/ = 1.80 = F/ = 1.80 F/ = 1.80 LOSS-0F 0FFSITE N0 LOSS OF OFFSITE LOSS OF OFFSITE.-

POWER, POWER, POWER, RESULTS DECLG C,=0.4 DECLG C,=0.4 ' ..DECLG C,=0.4 Peak Clad Temperature' (*F) . 2006. 2066. 2032.

Peak Clad Temp. Elevation (Ft.) 9.50 9.25 9.50 E Peak Clad Temperature Time (Sec) 118.5 94.4 117.1

[ Max Local Zr/H,0 Reaction (%) 3.38 5.77 3.66 g Total Zr/H O 2 Reaction (%) <1.0 <1.0 ' <1.0 Hot Assy.. Burst Time (sec.). 51.2. 51.1 51.2 .

Hot Assy. Burst Elevation (Ft.) 8.75 8.75 8.75 Blockage on Hot Assembly (%) 39.0 37.6 38.8 i'

e.- __ _ _

. TABLE 3-1 FORT.CALHOUN SMALL BREAK LOCA ANALYSIS INPUT PARAMETERS AND ASSUMPTIONS NSSS ver - 102%'of 1500'Mwt = 2530 Mwt Peak Linear Heat Rate - at.102% of 1500 Mut = 15.5 Kw/ft Radial Feaking Factor (F/) = 1.80 Maximuin Allowable Peaking Factor (F,) = 2.545 Axial Power Distribution See Figure 3-2

- Reactor Coolant System Pressure = 2100 psia G Reactor Coolan: System Flow Rate - 196,000 gpm E

Reactor Inlet Temperature = 545 "F C

~

Reactor Trip Signal (Including uncertainties) = 1728 psia, Pressurizer Pressure LOW SI Signal (Including uncertainties) = 1578 psia, Pressurizer Pressure LOW-LOW

!afety Injection Tank Watec Volume = 825 f t*/ Tank Safety Injection Tenk Minimum Pressure = 255 psia Steam Generator Tube Plugging Level = 6% (Uniform)

MSSV Setpoint Uncertainties = 134 Nominal setpoint pressure 13% Valse accumulation pressure 1

TABLE 3 FORT'CALHOUN SMALL BREAK LOCA ANALYSIS SMALL BREAK SE0VENCE OF EVENTS COLD LEG BREAK COLD LEG BREAK COLD LEG BREAK- 4

.022 50 FT .049 50 FT .087 S0 FT Start 0.0 Sec 0.0 Sec 0.0 Sec Reactor trip Signal 23.0 Sec 10.6 Sec 7.2 Sec SI Actuation Signal 36.6 Sec 17.2 Sec 10.5 Sec y Pumped SI Begins 67.5 Sec 48.1 Sec' 41.4 Sec

=

[ Top of Core Uncovered 2178.5 Sec 1995.1 Sec 7.10.7 Sec S.I. iank Injection Begins NA NA 932.9 Sec 5 PCT Occurs 3075.8 Sec 1898.0 Sec 1022.1 Sec Top of Ocre Recovered 4713.8 Sec 3100.3 Sec 1368.9 Sec

I-TABLE 3-3 FORT CALHOUN SMALL BREAK LOCA ANALYSIS SMALL BREAK RESULTS BOC IFBA BOC IFBA BOC IFBA

COLD LEG BREAK' COLD LEG BREAK COLD LEG BREAK RESULTS .022 50 FT .049 50 FT .087 50 FT }

Peak Clad Temperature (*F) 1076. 1444. 1166.

Peak Clad Temp. Elevation (Ft.) 10.25 10.25 10.00 Peak Clad Temperature time (Sec) 3075.8 1893.0 1022.1 7

f Max Local Zr/H 2O Reaction (%) 0.05 0.40 0.03 Z Max Local Zr/H,0 Rxn Elev (Ft.) 10.25 10.25 10.00 Total Zr/H O 2 Reaction (%) <1.00 <1.00 <1.00 i E

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-36 Containment Pressure OECLG (CD 0.6) l' age Si of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-37 Sah y Injection Tank Flow (1410wdown)

DECLG (C0 0.6) e Page 58 of 101

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)

FORT CALHOUN LOSS OF COOLANT' ACCIDENT.-REPORT Figure 2-38'-

Pumped ECCS Flow-(Reflood)-

DECLG (CD-0,6)

Page 59 of 101

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FORT CALHOUN-LOSS OF-COOLANT ACCIDENT REPORT Figure 2 Reflood Transient Core and Downcomer Levels DECLG (CD-0.6)

Page 60 of 10L

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Page 61 of 101

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l Page 62 of 101-g g .y - , , e-,., - +~ .-e,-,

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FORT-CALHOUN-LOSS OF COOLANT ACCIDENT REPORT-Ffgure 2-42 Core Flow (Top.and Bottom)

DECLG (C0 0.6)-

Page 63 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-43 t Core Fluid Quality l DECLG (CD=0.6) l Page 64 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-44 Core Heat Transfer Coefficient DECLG (CD-0.6)

Page 65 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-45 Core Fluid Temperature

. DECLG (CD-0.6)

Jage 66 of 101

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m --: 3

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-46 Peak Cladding Temperature DECI.G (CD-0.6)

Page 67 of 101

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Page 68 of 10t L

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FORT CALHOUN LOSS.-0F COOLANT' ACCIDENT REPORT j.

Figure 2-48' Core Pressure Transient DECLG (CD-0.8) l-Page 69 of 101

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T i r1E ISEC1 FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-49 Break Flow Rate DECLG (CD=0.8)

Page 70 of 101

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FORT CALHOUN LOSS OF-COOLANT ACCIDENT REPORT Figure 2-50 Containment Pressure DECLG-(CD-0.8) l Page 71 of 101

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TIME (SEcl f

p i

i FORT CALHOUN l LOSS OF COOLANT ACCIDENT REPORT l:

Figure 2-51 Safety Injection Tank Flow (Blowdown)

.DECLG (CD-0.8)

Page 72 of 101-

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-52 Pumped ECCS Flow (Reflood)

DECLG (CD=0.8)

Page 73 of 101

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FORT-CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-53 Reflood Transient -

Core and Downcomer Levels DECLG (CD=0.8)

Page 74 of 101

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TIMC t$CC) i FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-54 I--

Reflood Transient Core Flooding Rate

-DECLG (CD-0.8) .

Page 75 of 101

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FORT-CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-55 Core Mass Velocity DECLG (CD-0.8) d Page 76 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-56 Core Flow (Top and Bottom)

DECLG (CD-0.8)

Page 77 of 10 '.

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 2-57 Core Fluid Quality DECLG (CD-0.8) i Page 78 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure-2-58 Core Heat Transfer Coefficient DECLG (CD=0.8)

Page 79 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT-Figure 2-59 Core Fluid Temperature DECLG (CD-0.8)

Page 80 of 101

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L FORT CALHOUN LOSS-0F COOLANT ACCIDENT REPORT t:

1: l l- Figure 2-60 1.

Peak Cladding Temperature DECLG (CD 0.8)

Page al of 101

l N L 0 0 T _ C N T CORE PRESSURE CORE U #

FLOW.WIXTURE LEVEL

$ AND FUEL ROO POWER HISTCRY O< TIME < CORE COVERED I f' .

l l

l l FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-1 Code Interface Description for Small Break Model Page 82 of 101

- o - 4 FORT CALHOUN SMALL BREAK LOCA

-0.16 ASI POWER SHAPE 1 --

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/ ,

p 0.6 D

A X 0.4 p 0.2 W

l k 0 -

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i Page 83 of 101

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l FORT CALHOUN LOSS ~OF COOLANT ACCIDENT REPORT L Figure 3-3 l

Small Break LOCA (.022 ft2 )

RCS Depressurization 9

Page 84 of 101 a

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% FORT CALHOUN l> LOSS OF COOLANT ACCIDENT-REPORT 1

L_ Figure 3-4 l Small Break LOCA (.022 ft2 )

L- Core Mixture Height L

Page 85 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-5 2

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Peak Cladding Temperature Page 86 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-6 Small Break LOCA ( 022 ft2 )

Steam Flowrate Page 87 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDCNT REPORT Figure 3-7 Small Break LOCA (.022 ft2 )

j Hot Rod Film Coeffi-ient Page 88 of 101 l

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f FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-8 2

Small Break LOCA (.022 ft )

Hot Spot Fluid Temperature Page 89 of 101

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RCS Depressurization 4

Page 90 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-10 Small Break LOCA ( 049 ft2 )

Core Mixture Height Page 91 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-11 2

Small Break LOCA (.049 ft )

Peak Cladding Temperature Page 92 of 101

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT figure 3-12 Small Break LOCA (.049 ft2 )

Steam flowrate Page 93 of 101

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TIME is 1

FORT CALHOUN LOSS 0F. COOLANT ACCIDENT REPORT

. Figure 3-13 2

Small Break LOCA (.049 ft )

Hot Rod Film Coefficient Page 94 of 101-

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FORT CALHOUN LOSS OF- COOLANT ACCIDENT REPORT figuro 2-14 2

Small Break LOCA (.049 ft )

Hot Spot Fluid Temperature Page 95 of 101

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TIMC ISCCI FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-15 Small Break LOCA (.'087 ft2 )

RCS Depressurization Page 96 of 101

St.

se.

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TOP OF CORE (19.8 feet) k.

u 22 r

y

20. " >

18. Y 16.

y j t

II. 298 488. 688. 968. 1980. 1200. 1400.

TINC (SCC)

FORT CALHOUN LOSS-0F COOLANT ACCIDENT REPORT  :

Figure 3-16 2

Small Break LOCA ( 087 ft )

Core Mixture Height Page 97 of 101

1100.

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i 1(00.

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i /

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600. / i 500.

400.

700. 800. 900. 1000. I100. 1200.

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-17 Small Break LOCA (.087 ft2 )

Peak Cladding Temperature i

Page 98 of 101

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TINC iSCCI FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT figure 3 2 Small Break LOCA (.087 ft )

Steam Flowrate Page 99 of 101

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v S,

w 102 E i Y

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00. 750. 600. 850, 900. 950. 1000. 1050, 1100. 1150, 1200.

tint isi FORT CALHOUN LOSS OF COOLANT ACCIDENT REPORT Figure 3-19 small Break LOCA (.087 ft2 )

Hot Rod Film Coefficient Page 100 of 101

900, etc.

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h600, b

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FORT CALHOUN LOSS OF COOLANT ACCIDENT REP 0RT _

Figure 3-20 2

Small Break LOCA (.087 f t )

Hot Spot Fluid Temperature f

Page 101 of 101

. . . ....