ML17353A276

Proposed Tech Specs,Adding to Approved COLR Analysis Methodology Used for SBLOCA Analysis in Anticipation of Thermal Uprate to 2,300 Mwt for Both Units & Increasing Current Margin to Calculated PCT

ML17353A276
Person / Time
Site:	Turkey Point
Issue date:	07/26/1995
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17353A275	List: ML17353A274 ML17353A276
References
NUDOCS 9507310226
Download: ML17353A276 (55)

Text

G C (Continued)

Factor Limit Report, the Peaking Factor Limit Report shall be provided to the NRC Document Control desk with copies to the Regional Administrator and the Resident Inspector within 30 days of their implementation, unless otherwise approved by the Commission.

The analytical methods used to generate the Peaking Factor limits shall be those previously reviewed and approved by the NRC. If changes to these methods are deemed necessary they will be evaluated in accordance with 10 CFR 50.59 and submitted to the NRC for review and approval prior to their use is determined to involve an unreviewed safety question or if the change if a change would such require amendment of previously submitted documentation.

CORE OPERA G S 0 6.9.1.7 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

1. Axial Flux Difference for Specifications 3.2.1.

2. Control Rod Insertion Limits for Specification 3.1.3.6.

3. Heat Flux Hot Channel Factor - Fo(Z) for Specification 3/4.2.2.

The analytical methods used to determine the AFD limits shall be those previously reviewed and approved by the NRC in:

1. MCAP-10216-P-A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL F~

SURVEILLANCE TECHNICAL SPECIFICATION," June 1983.

2. MCAP-8385, "POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES

- TOPICAL REPORT,'eptember 1974.

The analytical methods used to determine the K(Z) curve shall be those previously reviewed and approved by the NRC in:

l. MCAP-9220-P-A, Rev. 1, 'Westinghouse ECCS Evaluation Model - 1981 Version,'ebruary 1982.

2. WCAP-9561-P-A, ADD. 3, Rev. 1, 'BART A-1: A Computer Code for the Best Estimate Analysis of Reflood Transients - Special Report:

T b e Modeling M ECCS Evaluation Model."

9. +~ HSf R~

The analytica me o s used to determine the Rod Bank Insertion Limits shall be those previously reviewed and approved by the NRC in:

1. WCAP-9272-P-A, "Mestinghouse Reload Safety Evaluation Methodology,"

July 1985.

The ability to calculate the COLR nuclear design parameters are demonstrated in:

1. Florida Power 5 Light Company Topical Report NF-TR-95-01, "Nuclear Physics Methodology for Reload Design of Turkey Point 8 St. Lucie Nuclear Plants'.

TURKEY POINT - NITS 3 L 4 6-20 AMENDMENT NOS. ~AND ]g8 9507310226 950726 PDR ADQCK 05000250 P . PDR

L-95-193 INSERT A

3. "WCAP-10054-P-A, (proprietary) and WCAP-10081-NP-A, (non-proprietary), "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", October, 1985.".

4. WCAP-10054-P-A Addendum 2, (proprietary), "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model", August, 1994."

ATTACHMENT 4 LICENSING REPORT SMALL BREAK LOSS-OF~OOLANT ACCIDENT ANALYSIS FOR THE TURKEY POINT UNITS 3 AND 4 UPRATING PROGRAM

TABLE OF CONHPPIS SMALLBREAK LOCA ANALYSIS 1 ~0 IntpvductIon ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 2 2.0 Input P'diameters and Assumptions ................... 2 3.0 Description on Analyses / Bvaluations Performed ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 4.0 Acceptance Criteria for Analyses / Hvaluations............ 6 5 0 Results

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 7 6.:0 Conclusions ................................. 9 7 .0 References t ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ '@ 10 Tables ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ....14 p lgOICS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 17 2

~J LIST OF TABLES Table 1: Small Break LOCA Accident Analysis Input Parameters for Turkey Point Table 2: Small Break LOCA Puel Cladding Results Table 3: Small Break LOCA Analysis Time Sequence of Events

I LIST OF FIGURIN Figure 1: Small Break Hot Rod Power Shape Figure 2: Small Break Pumped Safety Injection Flow Rate - 1 HHSI Pump Figure 3: Code Interface Description for the Small Break LOCA Model Figure 4: RCS Depressurization Transient, Limiting 3-Inch Break, High Typ Figure 5: Core Mixture Level, 3-Inch Break, High Typ Figure 6: Peak Cladding Temperature - Hot Rod, 3-Inch Break, High Typ Figure 7: Top Core Node Vapor Temperature, 3-Inch Break, High gyp Figure 8: BCCS Pumped Safety Injection - Intact Loop, 3-Inch Break, High Ty Figure 9: ECCS Pumped Safety Injection - Broken trop, 3-Inch Break, High T~yp Figure 10: Cold Leg Break Mass Flow, 3-Inch Break, High Typ Figure 11: Hot Rod Surface Heat Transfer CoeKcient - Hot Spot, 3-Inch Break, High T~yp Figure 12: Fluid Temperature - Hot Spot, 3-Inch Break, High Thyp Figure 13: RCS Depressurization Transient, 2-Inch Break, High Typ Figure 14: Core Mixture Level, 2-Inch Break, High gyp Figure 15: Peak Cladding Temperature - Hot Rod, 2-Inch Break, High Typ Figure 16: RCS Depressurization Transient, 4-Inch Break, High T~yp Figure 17: Core Mixture Level, 4-Inch Break, High Typ Figure 18: Peak Cladding Temperature - Hot Rod, 4-Inch Break, High gyp

LIST OF FIGUR1<Q (cont.)

Figure 19: RCS Depressurization Transient, 3-Inch Break, Low TyQ Figure 20: Core Mixture Level, 3-Inch Break, Low TAyo Figure 21: Peak Cladding Temperature - Hot Rod, 3-Inch Break, Low Ta

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SMALLBMDKLOCA ACCIDENT ANALYSIS

1.0 INTRODUCTION

This report contains information regarding the small break Loss-of-Coolant Accident (LOCA) analysis aad evaluations performed in support of the uprating program for Turkey Point Units 3 and 4. The purpose of analyzing the small break LOCA is to demonstrate conformance with the 10 CPR 50A6 (Rt:ference 1) requirements for the conditions associated with the upratiag. Important input assumptions, as well as analytical models and analysis methodology for the small break LOCA, are contained in subsequent sections. Analysis results are provided in the form of Tables and Pigures, as well as a more detailed description of the limiting transient. It was determined that no design or regulatory limit related to the small break LOCA would be exceed due to the upzated power and assumed plant parameters.

2.0 INPUT PARAMETERS AND ASSVMPHONS The important plant conditions and features are listed in Table 1. Several additioaal considerations that are not identified in Table 1 are discussed below:

Pigure 1 depicts the hot zod axial power shape modeled in the small break LOCA analysis. This shape was chosen because it represents a distribution with power concentrated in the upper regions of the core (the axial offset is + 20%). Such a distribution is limiting for small break LOCA since it minimizes coolant swell while maxiznizing vapor superheating and fuel rod heat generation at the uncovered elevations. The chosen power shape has been conservatively scaled to a flat K(Z) envelope based oa the peaking factors given in table 1.

Piguze 2 provides the degraded High Head Safety Injection (HHSI) flow versus pressure curve modeled in the small break LOCA analysis. The flow from oae HHSI pump only is assumed in this analysis.

3.0 DESCRIPTION

OP ANALYSES / EVALUATIONSPHRPORMHD Por small breaks, the NOTRUMP computer code (References 2 and 3) is employed to calculate the tzansieat depzessurization of the reactor coolant system (RCS), as well as to describe the mass and energy of the fluid flow through the break. The NOTRUMP computer code is a one-dimeasioaal general network code incorporating a number of advanced features. Among these advanced features are: calculation of thermal aon-equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, micture level tracking logic in multiple-stacked fiuid nodes, regime-dependent drift fiux calculations in multiple-stacked fluid nodes and regime-dependent heat trmsfer correlations. The NOTRUMP small break LOCA Emergency Core Cooliag System (ECCS) Evaluation Model was developed to determine the RCS response to design basis small break LOCAs, and to address NRC concerns expressed in NUTMEG-0611 (Reference 4).

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The RCS model is nodalized into volumes iatezconnected by flow paths. The broken loop is modeled explicitly, while the intact loops are lumped together into a second loop. Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum. The multi-node capability of the program enables explicit, detailed spatial repzesentation of various system components which, among other capabilities, enables a proper calculation of the behavior of the loop seal during a small break LOCA. The reactor coze is represented as heated contzol volumes with associated phase separation models to permit transient mixture height calculations.

Fuel cladding thermal analyses are performed with a version of the LOCTA-IV code (Refezeace 5) using the NOTRUMP calculated core pressure, fuel zod power history, uncovezed coze steam flow and mixture heights as boundary conditions (see Figure 3).

gmhsh A spectrum of 2-inch, 3-inch, and 4-inch equivalent diameter cold leg bzeaks, was performed usiag the analytical model described above. A sensitivity of the limiting transient to the RCS vessel avezage temperature was also performed.

The most limiting single active failure assumed for a smaH break LOCA is that of an emergency power train failure which results ia the loss of one complete tzain of ECCS components. In addition, a Loss-of-Offsite Power (LOOP) is assumed to occur coincident with reactor trip. This means that credit may be taken'or at most two high head safety injection (Sl) pumps aad one Iow head, or zesidual heat removal (RHR), pump. However, in the analysis of the smaH break LOCA presented heze, only the miaiznum delivered ECCS flow fimm a single high head SI pump with degraded flow was assumed.

The small break LOCA analysis performed for the Turkey Point Units 3 and 4 uprating pzogzam utilizes the NRC-appzoyed NOTRUMP Hvaluation Model (Refezences 2 and 3), with appropriate modifications to model pumped SI and accumulator injection in the broken loop as weH as an impzoved condensation model (COSI) for the pumped SI into the broken and intact loops (Reference 6 and 7).

The small break LOCA analysis performed for the Turkey Point upzating program assumes SI is delivered to both the in~ and broken loops at the RCS backpzessuze.

Prior to break initiatioa, the plant is assumed to be ia a fuH upzated power (102%) equilibzium condition, i.e., the heat generated in the core is being removed via the secondary system. Other initial plant conditions assumed in the analysis aze given in Section 2.0 and Table 1. Subsequent to the break opening, a period of zeactor coolant system blowdown ensues in which the heat from fission product decay, the hot zeactor internals, and the reactor vessel continues to be transfezzed to the RCS Quid. The heat transfer between the RCS and the secondary system may be in either diction and is a function of the relative temperatures of the primary and secondary. In the case of continuous heat addition to the secondary during a period of quasi-equilibrium, an increase in the secondazy system pressure zesults in steam relief via the steam generator safety valves, which were modeled with 3 pezcent accumulation and 3 percent tolerance.

Should a smaH break LOCA occur, depressuzization of the RCS causes fIuid to flow into the loops from the pressurizer resulting in a pressure aad level deczease in the pressurizer. The reactor trip signal

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subsequently occurs when the pressurizer low-pzessure zeactor trip setpoint, conservatively modeled as 1805 psia (including uncertainties), is reached. LOOP is assumed to occur coincident with reactor trip.

A safety injection signal is generated when the pressurizer low-pmsuze safety injection setpoint, conservatively modeled as 1615 psia (including uncertainties), is reached. Safety injection is delayed 35 seconds after the occuzzence of the low pressure condition. 'Hds delay accounts for signal initiation, diesel generator start up and emergency power bus loading'consistent with the assumed loss of offsite power coincident with reactor trip, as well as the time involved in aligning the valves and bringing the HHSI pump up to full speed. These countermeasures limit the consequences of the accident in two ways:

I Reactor trip and bozated water injection supplement void formation in causing a zapid axiuction of nuclear power to a zesidual level cozzesponding to the delayed fission and fission product decay. No credit is taken in the LOCA analysis for the boron content of the injection water. (However, an average RCS/sump mixed boron concentration is calculated to ensure that the post-LOCA coze remains subcritical). In addition, czedit is taken in the small break LOCA analysis for the insertion of Rod Cluster Control Assemblies (RCCAs) subsequent to the zeactor trip signal, while assuming the most reactive RCCA is stuck in the full out position. A zod drop time of 3 seconds was assumed while also considering an additional 2 seconds for the signal processing delay time. Therefoze, a total delay time of 5 seconds fzom the time of reactor trip signal to full rod insertion was used in the small break LOCA analysis.

2). Injection of borated water ensures sufficient flooding of the core to prevent excessive cladding temperatures.

During the earlier part of the small break transient prior to the assumed loss-of-offsite power coincident with reactor trip, the loss of flow through the break is not suKcient enough to overcome the positive core flow maintained by the zeactor coolant pumps. During this period, upward flow thzough the coze

. is maintained. However, foHowing the reactor coolant pump tap (due to a LOOP) and subsequent pump coastdown, a partial period of core uncovezy occurs. Ultimately, the small break transient analysis js terminated when the ECCS flow provided to the RCS exceeds the break flow rate.

The core heat removal mechanisms associated with the small break transient include not only the break itself and the injected ECCS water, but also that heat transferzed from the RCS to the steam generator secondazy side. Main Peedwater (hGzW) is assumed to be isolated coincident with the safety injection signal, and the MFW pumps coast down to 0% flow in 10 seconds. A continuous supply of makeup water is also provided to the secondary using the amCiliazy feedwater (APW) system. An APW actuation signal occurs coincident with the safety injection signal, resulting in the assumed delivery of full AFW system flow 120 seconds following the signal. The heat transferred to the secondary side of the steam generator aids in the reduction of the RCS pzessuze.

Should the RCS depressurize to approximately 600 psig, as in the case of the limiting 3-inch break and the 4-inch break, the cold leg accumulators begin to inject bozated water into the reactor coolant loops.

In the case of the 2-inch break however, the vessel mixture level is recovered without the aid of accumulator injection.

Upon completion of the small break LOCA analysis, an evaluation was performed for automatic containment spray actuation during small break LOCA. This evaluation accounts for the fact that Turkey Point Units 3 and 4 may be subject to SI interruption for up to 2 minutes while switching over to cold leg recirculation. The results of this evaluation are discussed in Section 5.0.

4.0 ACCEPTANCE CRIH<2HA FOR ANALYSES / EVALUATIONS The Acceptance Criteria for the LOCA are described in 10 CFR 50.46 (Reference 1) as follows:

The calculated maximum fuel element cladding temperature shall not exceed 2200'F, 2). The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation, 3). The calculated total amount of hydrogen generated from the chemical zeaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated ifall of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volumewere to react, 4). Calculated changes in core geometry shall be such that the coze remains amenable to cooling, 5). After any calculated successful initialoperation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

Criteria 1 through 3 are explicitly covered by the small break LOCA analysis at upzated conditions.

For criterion 4), the appropriate core geometzy was modeled in the analysis. The zesults based on this geometry satisfy the PCT criterion of 10 CFR 50.46 and consequently, demonstrate the core remains amenable to cooling.

For criterion 5), Long-Term Coze Cooling (LTCC) considerations aa: not dizectly applicable to the small break LOCA analysis.

The criteria were established to provide a significant margin in emergency. core cooling system (BCCS) performance following a LOCA.

5.0 RESULTS In order to determine the conditions that produced the most limiting small break LOCA case (as determined by the highest calculated peak cladding temperature), a total of four cases were examined.

These cases included the investigation of variables including break size and RCS temperature to ensure that the most severe postuhted small break LOCA event was analyzed. The following discussions

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provide insight into the analyzed conditions.

Pirst, a break spectrum based on higll RCS Tgyg was performed as this was expected to yield more limiting PCT results than low RCS Tpyg Tile limiting break for the Turkey Point Units was found to be a 3-inch diameter cold leg bzeak. The results of Reference 8 demonstzate that the cold leg break location is limiting with respect to postulated cold Ieg, hot leg and pump suction leg break locations.

The PCT attained during the transient was 1688'F (refer to Table 2). While Table 3 provides the key transient event times.

A summary of the transient response for the limiting high T~ 3-inch break case is shown in Figures 4 through 12 These figures present the response of the following parameters:

RCS Pressure Transient, (Figure 4) 2). Core Mixture Level, (Piguze 5) 3). Peak Cladding Temperature, (Figure 6) 4). Top Coze Node Vapor Temperature, (Pigure 7) 5). Safety Injection Mass Flow Rate for the Intact and Broken Loops, (Figures 8 and 9) 6). Cold Leg Break Mass Plow Rate, (Figure 10)

Hot Rod Surface Heat Transfer Coefficient at the Hot Spot, (Figure 11) 8). Fluid Temperature at the Hot Spot, (Figure 12)

Upon initiation of the limiting 3-inch break, there is a slow depzessurization of the RCS (see Piguze 4).

During the initial period of the small break transient, the effect of the bzeak flow rate is not sufficient to overcome the flow rate maintained by the zeactor coolant pumps as they coast down. As such, normal upward flow is maintained through the coze and coze heat is adequately removed. Pollowing reactor trip, the removal of the heat generated as a zesult of fission products decay is accomplished via a two-phase mixture level covering the core. Prom the core mixture level and cladding temperature transient plots for the 3-inch break calculations given in Pigures 5 and 6, respectively, it is seen that the peak cladding temperature occurs near the time when the core is most deeply uncovered and the top of the core is being cooled by steam. This time is characterized by the highest vapor superheating above the mixture level (refer to Piguze 7).

A comparison of the flow provided by the safety injection system to the intact and broken loops to the total break mass flow rate at the end of the transient (as given in Pigures 8, 9 and 10, respectively),

shows that at the time the transient was terminated, the total safety injection flow rate that was delivered to the intact and broken loops exceeds the mass flow rate out the break (70.1 ibm/sec versus 61.2 ibm/sec). In addition, the inner vessel core mixture level has recovered the top of the core (Figure 5).

Figures 11 and 12 provide additional information on the hot rod surface heat transfer coefficient at the hot spot and fluid temperature at the hot spot, respectively.

After this point in the transient, there is no longer a concern of exceeding the 10 CFR 50.46 criteria as described in Section 4.0 since:

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1). The RCS pressure is gradually decaying, and 2). The net mass'inventory is increasing.

As the RCS inventory continues to gradually increase, the core mixture level will continue to increase and the fuel cladding temperatures will continue to decline. The 3-inch high Tp small break LOCA transient is terminated.

Studies documented in Reference 8 have determined that the limiting small break LOCA transient occurs for breaks of less than 10 inches in diameter. To ensure that the 3-inch diameter break was the most limiting, calculations were also performed with break equivalent diameters of 2 inches and 4 inches.

The results of each of these cases are given in Tables 2 and 3. Plots of the following parameters for each case are also given in Figures 13 through 15 for the 2-inch break case and Figures 16 thmugh 18 for the 4-inch break.

1). RCS Pressure Transient, 2). Core Mixtuze Level, and 3). Peak Cladding Temperature.

The PCI.'s for the 2-inch and 4-inch breaks were 1656'F and 1583'F, respectively (see Table 2). The PCTs for each of these cases was calculated to be less than that for the 3-inch break case based on high Thyp conditions.

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Reduced operating temperature typically zesults in a PCT benefit for the small bzeak LOCA. However, due to competing effects and the complex nature of small break LOCA transients, there have been some instances where more limiting results have been observed for the reduced opezating temperature case.

For this reason, a small break LOCA transient based on a lower bound RCS vessel average tempezatuze was pezfoIIIled.

The temperature window analyzed was based on a nominal vessel average temperature of 574.2'F, with

+ 3'F for an operating window and k 8.5'F to bound uncertainties. The break spectrum was performed at the high vessel average temperature (585.7'F), as this case was expected to yield limiting results. Then, a sensitivity analysis for the low vessel average temperature (562.7'F) was performed based on the limiting 3-inch break case from the break spectrum analyses previously described.

Plots of the fonowing parameters are given in Figures 19 through 21 for the 3-inch break case at low Tp conditions:

RCS Pressure Transient, 2). Core Mixture Level, and Peak Cladding Temperature.

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The PCT for the 3-inch break case based on low vessel average temperature was 1619'F (see Table 2).

Therefore, the PCT for this case was calculated to be less than that for the 3-inch break case with high vessel average temperature conditions.

The evaluation for containment spray actuation in small break LOCA resulted ia no change to the predicted small bzeak LOCA PCT for the various cases analyzed. The DRFA fuel stack height above the lower coze plate was explicitly modeled for the various cases analyzed.

6.0 CONCLUSION

S A full break spectrum Small Break LOCA analysis supporting the uprated Turkey Point core with the high nominal vessel average temperature,.Tokyo = 585.7'P, was performed. Peak claddiag temperatures of 1656'F, 1688'P, and 1583'F were calculated for the 2-inch, 3-inch, and-4-inch cold leg breaks, respectively, thus identifying the 3-inch equivalent diameter break as limiting. A sensitivity to low nominal vessel average temperature, Thyg 562.7'P, was performed. The calculated peak cladding temperature was 1619 P, identifying the 3-inch equivalent diameter cold leg break, high nominal vessel average temperature, as the limiting case.

The analyses presented in this section show that the high head safety injection subsystems of the Emergency Core Cooliag System, together with the heat removal capability of the steam generator, provide sufficient core heat removal capability to maintain the calculated peak cladding temperatures below the required limit of 10 CFR 50.46 which is defined in Section 4.0.

7.0 RFMHQKCES "Acceptance Criteria for Emergency Core Cooliag Systems for Light Water Cooled Nuclear Power Reactors," 10 CPR 50.46 and Appendix K of 10 CFR 50, Pederal Register, Volume 39, Number 3, January 1974, as amended in Pederal Register, Volume 53, September 1988.

2. Meyer, P.E., "NOTRUMP - A Nodal Transient Small Break and- General Network Code,"WCAP-10079-P-A, (proprietary) and WCAP-10080-NP-A (non-pzoprietary), August 1985.

3. Lee, N. et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A (proprietary) and WCAP-10081-NP-A (non-proprietary), August 1985.

4. "Generic Evaluation of Peedwater Transients and Small Brcak Loss-of-Coolant Accidents in Westinghouse - Designed Operating Plant," NU1 EG-0611, January 1980.

5. Bordelon, P. M. et al., "LOCTA-IVProgram: Loss-of-Coolant Transient Analysis", WCAP-8301 (proprietary) and WCAP-8305 (non-proprietary), June 1974.

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6. Thompson, C. M, et al., "Addendum to the Westinghouse Small Break LOCA Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and the COSI Condensation Model", WCAP-10054-P, Addendum 2 (proprietary) and WCAP-10081-NP (non-proprietary),

August 1994.

7. Shimeck, D. J., "COSI SVSteam Condensation Experiment Analysis", WCAP-11767-P-A (proprietary), March 1988.

8.. Rupprecht, S. D. et al, "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code", WCAP-11145-P-A (proprietary) and WCAP-11372-NP-A (non-proprietary), October 1986.

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Table 1 SMALLBREAK LOCA ACCIDENT ANALYSIS INPUT P FOR TURKEY POINT Parameter High Tavo (Low T~vo)

Reactor coze rated thermal power', (MWt) 2300 Peak linear power", (kw/ft) 14.9 Total peaking factor (F< ) at peak 2.50 Power See Figure 1 shape'm 1.70 1.515 15x15 DRFA Fuel'ccumulator water volume, nominal (6'/acc.) 892 Accumulator tank volume, nominal (ft'/acc.) 1200 Accumulator gas pmsuze, miniznum (psig) 600 Pumped safety injection flow See Figure 2 Steam generator tube plugging level (%)', 20 Thermal Design Flow/loop, (gpm) 85,000 Vessel average temperature w/ uncertainties, ('F) 585.7 (562.7)

Reactor coolant pressure w/ uncertainties, (psia) 2320 Min. aux. feedwater flowzate/loop, (lb/sec)'.26 Two percent is added to this power to account for calorimetric error.

This represents a power shape corresponding to a one-line segment peaking factor envelope, K(z),

based on Fz~ = 2.50.

DRFA fuel type modeled in the small break LOCA analysis.

Maximum plugging level in any one or all steam generators.

Flowrates per steam generator.

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Table 2 SMALLBRE~ LOCA ANALYSIS EWHL CLADDING RESULTS Break Spectrum, (High TQ Rioah Hach Peak Cladding Temperature ('F) 1656 1688 1583 Peak Cladding Temperature Location (ft) 11.75 11.75 11.50 Peak Cladding Temperature Time (sec) 2627 1188 Local Zr/H,O Reaction, Max (%) 2.0188 1.5535 0.6679 L'ocal Zr/H,O Reaction Location (ft) 11.75 11.50 11.25 Total Zr/H,O Reaction (%) < 1.0 < 1.0 Hot Rod Burst Time (sec) No Burst No Burst No Burst Hot Rod Burst Location (ft) N/A N/A N/A Results for the limiting 3-inch break size HC4Xhvo ~wKhvo Peak Cladding Temperature ('F) 1688 1619 Peak Cladding Temperature Location (ft) 11.75 11.50 Peak Cladding Temperature Time (sec) 1188 1229 Local Zr/H~O Reaction, Max (%) 1.5535 1.1034 Local Zr/H,O Reaction Location (ft) 11.50 11.50 Total Zr~O Reaction (%) < 1.0 Hot Rod Burst Time (sec) No Burst No Burst Hot Rod Burst Location (ft) ,

N/A N/A 15

Table 3 SMALLBREAK LOCA ANALYSIS 'IIME SEQUENCE OP EVE%IS Break Spectrum, (High Tg Lizh Elk@i Break Occurs (sec) 0.0 0.0 0.0 Reactor Trip Signal (sec) 40.6 17.0 10.4 Safety Injection Signal (sec) 58.9 30.4 21.4 Top Of Core Uncovered (sec) 1402 482 278'25 Accumulator Injection Begins (sec) N/A 1040 Peak Cladding Temperature Occurs (sec) 2627 1188 668 Top Of Core Covered (sec) 4554 2363 965 Results for the limiting 3-inch break size Hc1LKJLYG ~EXhvo Break Occurs (sec) 0.0 0.0 Reactor Trip Signal (sec) 17.0 14.4 Safety Injection Signal (sec) 30.4 21.8 Top Of Core Uncovered (sec) 482 526 Accumulator Injection Begins (sec) 1040 1086 Peak Cladding Temperature Occurs (sec) 1188 1229 Top Of Core Covered (sec) 2363 2343 Momentary core uncovery occurred at 213 seconds during prelude to loop seal clearing. The beginning of the subsequent extended core uncovery at 278 seconds is the time listed.

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K(z) URVE 2 ~ 5 2

CD 1 ~ 5 hC 4J Q

~ 5 6 8 10 CORE ELEVATIOM [ ft]

Figure 1: Small Break Hot Rod Power Shape

FLOW TO INTACT LOOP FROM I HHRI PUMP FLOW TO BROKEN LOOP FROM 1 HHSI PUMP 400 300 I

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Figure 2: Small Break Pumped Safety Injection Flow Rate - 1 HHSI Pump

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C CORE PRESSUREI CORE R PMV, KXTOBE L8VE4, T U LND WBL BOD A P01ER KSRRY P

0 < TDE < CORE RE-COVERED Figure 3: Code Interface Description for the Small Break LOCA Model

2500 2000 n 1500

~ 1000 0

500 0

0 1000 2000 3000 4000 Time (s)

Figure 4: RCS Depressurization Transient, Limiting 3-Inch Break, High TG

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~30

~ 25 TOP'F COR

~ 20 t5 1000 2000 3000 4000 Time (s)

Figure 5: Core Mxture Level, 3-Inch Break, High Tokyo

1800 1600

~ 1400

$ 200 L

C7

~ 1000 800 600 400 1000 2000 3000 4000 Time (s)

Figure 6: Peak Cladding Temperature - Hot Rod, 3-Inch Break, High To

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1400 1200

$ 000 800 CL E

I 600 400 1000 2000 3000 4000 Time (sj Figure 7: Top Core Node Vapor Temperature, 3-Inch Break, High Tokyo

50 40 E

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~ 20 CO 10 0

1000 2000 3000 4000 Time (s)

Figure 8: ECCS Pumped Safety Injection - Intact Loop, 3-Inch Break, High Tokyo

25 20 E

15 O )0 5

1000 2000 3000 4000 Time (s)

Figure 9: ECCS Pumped Safety Inttection - Broken Loop, 3-Inch Break, High TG

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1600

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~E 1200 1000 o 800 600 cn 400 C) 200 0

1000 2000 3000 4000 Time (s)

Figure 10: Cold Leg Break Mass Flow, 3-Inch Break, High T~vc

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0 10 1000 2000 3000 4000 Time (8)

Figure 11: Hot Rod Surface Heat Transfer Coefficient - Hot Spot, 3-Inch Break, High TgyG

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1800 1600

~ 1400

~ 1200 C7

~ 1000 800 600 400 1000 2000 3000 4000 Time (s)

Figure 12: Fluid Temperature - Hot Spot, 3-Inch Break, Egh TG

2500 2000

$ 500 1000 500 0 1000 2000 3000 4000 5000 Time (s)

Figure 13: RCS Depressurization Transient, 2 Inch Break, High TG

~ 30 CP

~ 25 TOP OF C RE 2Q 15 1000 2000 3000 4000 5000 Time (s)

Figure 14: Core Mixture Level, 2-Inch Break, High TvG

4 Fa

1800 1600

~ 1400 1200 L

~ 1000 800 600 400 0 1000 2000 5000 4000 5000 Time (s)

Figure 15: Peak Cladding Temperature - Hot Rod, 2-Inch Break, High To

2500 2000

n. 1500 ca 1000 Q

500 1000 2000 3000 Time (s)

Figure 16: RCS Depressurization Transient, 4<<Inch Break, High To

35 30 25

)

CD CD TOP OF CORE 20 OK 15 10 0 1000 2000 3000 Time (s)

Figure 17: Core Mixture Level, 4-Inch Break, High TgyG

1600 1400

~ 1200

~ 1000 C7 800 600 400 200 1000 2000 5000 Time (s}

Figure 18: Peak Cladding Temperature - Hot Rod, 4-Inch Break, High TG

5/

,go

~ ~

2500 2000

n. 1500 L

cn 1000 CL 500 1000 2000 3000 4000 Time (s)

Figure 19: RCS Depressurization Transient, 3-Inch Break, Low To

~30

~ 25 TOP OF COR

~ 20 15 1000 2000 3000 4000 Time (s)

Figure 20: Core Mixture Level, 3-Inch Break, Low TgyG

K ~.

- 4~ .l v ~

'4 r

1800 1600

~ 1400

~ 1200 C7 F000 800 600 400 0 1000 2000 3000 4000 Time (s)

Figure 21: Peak Cladding Temperature - Hot Rod, 3-Inch Break, Low To

4

~ ~