Text

Fuel Upgrade Program LOCA Sar
	ML20058C358
Person / Time
Site:	Seabrook
Issue date:	08/31/1993
From:	; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20058C283	List: ML20058C279; ML20058C299; ML20058C316; ML20058C338; ML20058C358;
References
	NUDOCS 9312020431
	Download: ML20058C358 (100)
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SEABROOK STATION LOCA SAFETY ANALYSIS REPORT August 1993 TABLE OF CONTENTS: LIST OF TABLES iii LIST OF FIGURES iv 1.O INTRODUCTION 1 1.1 Classification of Faults 1 1.1.1 Condition IV Limiting Faults 1 1.1.2 Condition III Infrequent Faults 1 1.2 Loss-of-Coolant Accident Acceptance Criteria 1 2.0 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES. (LARGE BREAK LOCA) 2 2.1 Identification of Causes and Frequency Classification 2 2.2 Sequence af Events and Systems Operations 2 2.3 Description of Large Break LOCA Transient 4 2.4 Core and System Performance 5 2.4.1 Mathematica1.Model 5 2.4.2 Large Break LOCA Evaluation Model 5 2.4.3 Input Parameters and Initial 7 Conditions 2.5 Kesults 8 Tables, Section 2 10 Figures, Section 2 18 i

SEABRCOK STATION LOCA SAFETY ANALYSIS REPORT 3.0 LOSS OF REACTOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES WHICH i ACTUATES THE ECCS (SMALL BREAK LOCA) 79 3.1 Identification of Causes and Frequency Classification 79 3.2 Sequence of Events and Systens Operations 79 3.3 Description of Small Break LOCA Transient 80 3.4 Core and System Performance 81 3.4.1 Mathematical Model 81 3.4.2 Small Break LOCA Evaluation Model 81 3.4.3 Input Parameters and Initial Conditions 82 3.5 Results 82 Tables, Section 3 84 Figures, Section 3 88

4.0 CONCLUSION

115

5.0 REFERENCES

115 i ii 1

SEABROOK STATION LCCA SAFETY ANALYSIS REPORT LIST OF TABLES: TABLE 2-1 INPUT PARAMETERS USED IN THE LARGE BREAK LOCA ECCS ANALYSIS ]0 TABLE 2-2 LARGE BREAK LOCA CONTAINMENT DATA 11 TABLE 2-3 LARGE BREAK LOCA - CASES ANALYZED 13 TABLE 2-4 LARGE BREAK LOCA RESULTS - TIME SEQUENCE OF EVENTS 14 TABLE 2-5 LARGE BREAK LOCA RESULTS - FUEL CLADDING DATA 15 TABLE 2-6 LARGE BREAK LOCA C =0.6 MINIMUM 3 SAFEGUARDS LOCA REFLOOD MASS AND ENERGY RELEASE RATES 16 TABLE 2-7 LARGE BREAK LOCA C =0.6 mil 4IMUM 3 SAFEGUARDS LOCA BLOWDOWN MASS AND ENERGY RELEASE RATES 17 TABLE 3-1 INPUT PARAMETERS USED IN THE SMALL BREAK LOCA ECCS ANALYSIS 83 TABLE 3-2 SMALL BREAK LOCA - CASES ANALYZED 84 TABLE 3-3 SMALL BREAK LOCA RESULTS - TIME SEQUENCE OF EVENTS 85 TABLE 3-4 SMALL BREAK LOCA RESULTS - FUEL CLADDING DATA 86 iii

SEABROOK STATION l LOCA SAFETY ANALYSIS REPORT l l LIST OF FIGURES: FIGURE 2-1A LARGE BREAK LOCA SEQUENCE OF EVENTS 18 FIGURE 2-1B. LARGE BREAK LOCA CODE INTERFACE 19 FIGURE 2-2A PUMPED SAFETY INJECTION FLOW VS. RCS PRESSURE (MINIMUM SAFEGUARDS) 20 FIGURE 2-2B PUMPED SAFETY INJECTION FLOW VS. RCS PRESSURE (MAXIMUM SAFEGUARDS) 21 FIGURE 2-3 (A-N) LARGE BREAK LOCA RESULTS (Co = 0.4) 22 -A RCS PRESSURE -B COLD LEG BREAK MASS FLOW RATE l -C CORE POWER (FRACTION OF NOMINAL) -D CORE MAS: FLOW RATE (TOP AND BOTTOM) -E ACCUMULAT:? MASS FLOW RATE -F REFLOOD CORE AND DOWNCOMER WATER LEVELS -G BREAK ENERGY RELEASED TO CONTAINMENT l -H FLUID VELOCITY PAST CLAD HOT SPOT l -I FLUID QUALITY AT HOT SPOT -J HOT ROD HEAT TRANSFER COEFFICIENT -K CLAD HOT SPOT FLUID TEMPERATURE -L HOT ROD PEAK CLAD TEMPERATURE -M CONTAINMENT PRESSURE -N CONTAINMENT CONDENSING WALL HEAT TRANSFER COEFFICIENT l FIGURE 2-4 (A-N) LARGE BREAK LOCA RESULTS (Co = 0.6) 36 l 0.8) 50 FIGURE 2-5 (A-N) LARGE BREAK LOCA RESULTS (C = 3 FIGURE 2-6 (A-N) LARGE BREAK LOCA RESULTS (Co = 0.6, MAXIMUM SAFEGUARDS) 64 l FIGURE 2-7 K(Z) CURVE 78 l l 1 l 1 iv i

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT LIST OF FIGURES (Continued): FIGURE 3-1 SMALL BREAK LOCA CODE INTERFACE 87 FIGURE 3-2 SMALL BREAK LOCA HOT ROD POWER SHAPE 88 FIGURE 3-3 PUMPED SAFETY INJECTION FLOW VS. RCS PRESSURE 89 FIGURE 3-4 (A-H) SMALL BREAK LOCA RESULTS (3 INCH BREAK) 90 -A RCS PRESSURE -B CORE MIXTURE LEVEL -C PEAK CLAD TEMPERATURE -D CORE EXIT STEAM MASS FLOW RATE -E HOT ?.DD HEAT TRANSFER COEFFICIENT -F CLAD HOT SPOT FLUID TEMPERATURE -G BREAK MASS FLOW RATE -H PUMPED SAFETY INJECTION MASS FLOW RATE FIGURE 3-5 (A-H) SMALL BREAK LOCA RESULTS (4 INCH BREAK) 98 FIGURE 3-6 (A-H) SMALL BREAK LOCA RESULTS (6 INCH BREAK) 106 v

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT l

1.0 INTRODUCTION

I 1.1 Classification of Faults 1.1.1 Condition IV Limiting Faults 1 American Nuclear Society (ANS) Condition IV occurrences are faults which are not expected to occur during the lifetime of Seabrook

Station, but are postulated because their consequences would 1

include the potential for the release of significant amounts of ) radioactive material. They are the most drastic occurrences which must be designed against and thus represent limiting design cases. Condition IV faults are not to cause a fission product release to I the environment resulting in an undue risk to public health and safety in excess of guideline values of 10 CFR 100. A single Condition IV f ault is not to cause a consequential loss of required functions of systems needed to cope with the f ault, including those of the emergency core cooling system (ECCS) and the containment. For the purposes of this report, only the following Condition IV l fault will be addressed: Major rupture of pipes containing reactor coolant up to and including double-ended rupture of the largest pipe in the reactor coolant system (large break loss-of-coolant accident). q 1.1.2 Condition III Infrequent Faults ANS Condition III occurrences are faults which may occur very infrequently during the lifetime of Seabrook Station. They will be I accommodated with the f ailure of only a small fraction of the fuel rods although sufficient fuel damage might occur to preclude l resumption of operation for a considerable outage time. The i release of radioactivity will not be sufficient to interrupt or i restrict public use of areas beyond the exclusion radius. A Condition III fault will not, by itself, generate a Condition IV fault or result in a consequential loss of function of the RCS or containment barriers. For the purposes of this report, only the following Condition III fault will be addressed: Loss of reactor coolant f rom small ruptured pipes or f rom cracks in large pipes which actuate the emergency core cooling system (snell break loss-of-coolant accident). 1.2 Loss-of-Coolant Accident Acceptance Criteria The Acceptance Criteria for a loss-of-coolant accid 9nt (LOCA) are described in 10 CFR 50.46 (Reference 1) as follows: A. The calculated peak fuel element clad temperature does not exceed the requirement of 22000F. 1

SEABROOK STATION l LOCA SAFETY ANALYSIS REPORT 1 B. The amount of fuel element cladding that reacts chemically with water or steam to generate hydrogen, does r not exceed one percent of the total amount of Zircaloy (or ZIRLO) in the fuel rod cladding. 5 C. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The. localized cladding oxidation limit of 17 percent is not exceeded during or after quenching. D. The core remains amenable to cooling during and af ter the break. E. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining'in the core. These criteria were established to provide a significant margin in emergency core cooling system (ECCS) performance following 6.LOCA. WASH-1400 (USNRC 1975) (Reference 2) presents a study in regards to the probability of occurrence of RCS pipe ruptures. 2.0 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LARGE BREAK LOCA) This section presents a description and results of the large break loss-of-coolant accident (LOCA) in conformance with 10 CFR 50.46 and Appendix K of 10 CFR 50 (Reference 1). 2.1 Identificatien of causes and Frequency Classification A LOCA is the result of a pipe rupture of the reactor coolant system (RCS) pressure boundary. For the analyses reported here, a major pipe break (large break) is defined as a rupture with a total 2 cross-sectional area equal to or greater than 1.0 ft. This event is considered an American Nuclear Society (ANS) Condition IV event, J a limiting fault, in that it is not expected to occur during the lifetime of Seabrook Station, but is postulated as a conservative design basis. 1 2.2 Sequence of Events and Systems Operations The time sequence of events following a large break LOCA is presented in Figure 2-1A. Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. Loss-Of-Offsite Power j (LOOP) is assumed coincident with the occurrence of the break. The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached. A safety injection signal is-generated when the appropriate setpoint (high containment pressure or low pressurizer pressure) is reached. These countermeasures 2 +

SEABROOK STATION LCCA SAFETY ANALYSIS REPORT will limit the consequences of the accident in two ways: A. Reactor trip and borated water injection supplement void formation in causing rapid reduction of power to the residual level corresponding to fission product decay heat. 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 core remains suberitical.

In addition, the insertion of control rods to shut down the reactor is neglected in the large break analysis. B. Injection of borated water provides for heat transfer from the core and prevents excessive clad temperatures. For large break LOCAs, the most limiting single failure with respect to peak clad temperature (PCT) has been shown by experience to be that which reduces safety injection while producing the lowest containment pressure. The lowest containment pressure would be obtained only if all containment spray pumps cperated subsequent to the postulated LOCA. Therefore, for the purposes of large break LOCA analyses, the most limiting single failure would be the loss of one residual heat removal (RHR) pump with full operation of the spray pumps. Credit could be taken for two saf.ety injection pumps (SIPS), two centrifugal charging pumps (CCPs) and one RHR (low head) pump for a large break. However, the Seabrook large break LOCA analysis conservatively assumes both maximum containment safeguards (lowest containment pressure) and minimum emergency core cooling system (ECCS) safeguards (the loss of one complete train of ECCS components which includes one RHR pump, one SIP and one CCP), which results in the minimum delivered ECCS flow available to the RCS. Both Emergency Diesel Generators (EDGs) are assumed to start in the modeling of the containment spray pumps. Modeling full containment heat removal systems operation is required by Branch Technical Position CSB 6-1 and is conservative for the large break LOCA. This assumption is consistent with the current methodology j for large break analyses. j In the large break analysis, one ECCS train starts and delivers flow through the injection lines (one for each loop) with one branch injection line spilling to containment pressure. To minimize delivery to the reactor, the branch line chosen to spill is selected as the one with the minimum resistance (based on a 10 gpm flow imbalance of the SIP and CCP branch lines). In addition, the SIP and CCP performance curves were degraded by 5% and the RHR pump performance curve was degraded 8.75%. Therefore, in the large break ECCS analysis presented here, single failure is conservatively accounted for via the loss of an ECCS train, and the spilling of the minimum resistance injection line while assuming full active containment heat removal system operation. 3

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT 2.3 Description of Large Break LOCA Transient Before the break occurs, the unit is in an equilibrium condition: i.e., the heat generated in the core is being removed via the secondary system. During blowdown, heat from fission product decay, hot internals and the vessel, continues to be transferred to the reactor coolant. At the beginning of the blowdown phase, the entire RCS contains subcooled liquid which removes heat from the core by forced convection with some fully developed nucleate boiling. Shortly after break initiation, departure from nucleate boiling is calculated to occur using a critical heat flux correlation consistent with Appendix K of 10 CFR 50. Thereafter, the heat removal from the clad surf ace is calculated based upon the heat transfer coefficient appropriate to the regime, which is a function of the local fluid properties and rod heat flux. Radiation heat transfer from the clad to both steam and droplets, as well as other fuel rods, is at,o considered. The heat transfer between the RCS and the secondary system may be in either direction, depending on their relative temperatures. In the case of continued-heat addition to the secondary system, the secondary system pressure increases e.nd the main steam safety valves may actuate to limit the pressure. Makeup water to the secondary side is automatically provided by the auxiliary feedwater system. The safety injection signal actuates a feedwater isolation signal which isolates main feedwater flow by closing the main feedwater control valves and also initiates auxiliary feedwater flow by starting the auxiliary f eedwater pumps. The secondary flow aids in the reduction of RCS pressure. When the RCS depressurizes to 600 psia, the accumulators begin to inject borated water into the reactor coolant loops. The conservative assumption is made that all of the accumulator water injected during the bypass period is subtracted from the RCS after the bypass period terminates (called end-of-bypass). End-of-bypass occurs when the expulsion or entrainment mechanisms responsible for the bypassing are calculated not to be effective. The rejection of the accumulator water delivered prior to end-e>f-bypass is again consistent with Appendix K of 10 CFR 50. Since LOOP is assumed, the reactor coolant pumps trip at the inception of the accident. The effects of pump coastdown are included in the blowdown analysis. The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2300 psia) falls to a value approaching that of the containment atmosphere. Prior to or at the end of the blowdown, termination of bypass occurs and refill of the reactor vessel lower plenum begins. Refill is completed when _ emergency core cooling water has filled the lower plenum of the reactor vessel, which is bounded by the bottom of the fuel rods (called bottom of core (BOC) recovery time). 4

't SEABROOK STATION LOCA SAFETY ANALYSIS REPORT l The reflood phase of the transient is defined as the time period lasting from BOC recovery until the reactor vessel has been filled with water to the extent that the core temperature rise has been terminated. From the latter stage of blowdown to the beginning of reflood, the safety injection accumulator tanks rapidly discharge borated cooling water into the RCS thus contributing to the filling of the reactor vessel downcomer. The downcomer water elevation head provides the driving force required for the reflooding of the reactor core. The RHR pumps, CCPs and SIPS aid in the filling of the downcomer and subsequently supply water to maintain a full downcomer and complete the :eflooding process. Continued operation of .he ECCS pumps supplies water during long-term cooling. Core temperatures have been reduced to long-term steady state levels associated with dissipation of residual heat generation. After the water level of the refueling water storage tank (RWST) reaches a minimum allcwable value, coolant for long-term cooling of the core is obtained by switching l to the cold leg recirculation phase of operation. Spilled borated l water is drawn from the engineered safety features (ESF) containment sumps by the RHR pumps and returned to the RCS cold legs. The containment spray pumps are manually aligned to the containment emergency sumps and continue to operate to further reduce containment pressure and temperature. 2.4 Core and System Performance 2.4.1 Mathematical Model The requirements of an acceptable ECCS evaluation model are resented in Appendix K of 10 CFR 50. J.4.2 Large Break LOCA Evaluation Model The analysis of a large break LOCA transient is divided into three phases: (1) blowdown, (2) refill, and (3) reflood. There are three distinct transients analyzed in each phase, including the l thermal-hydraulic transient in the

RCS, the pressure and l

temperature transient within the containment, and the fuel and clad temperature transient of the hottest fuel rod in the core. Based on these considerations, a system of interrelated computer codes was developed for the analysis of the LOCA. A description of the various aspects of the LOCA analysis methodology is given by Bordelon,

Massie, and Zordan (1974)

(Reference 3). This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with the Acceptance Criteria. The SATAN-VI, WREFLOOD, BASH and LOCBART codes, which are used in the LOCA analysis, are described in detail by Bordelon et al. (1974) (Reference 4); Kelly et al. (1974) (Reference 5); Young et al. (1987) (Reference 6); and Bordelon et al. (1974) (Reference 3). 5 l I i

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT Code modifications are specified in References 7, 8, 9, and 10. These codes assess the core heat transfer geometry and determine if the core remains amenable to cooling through and subsequent to the blowdown, refill, and reflood phases of the LOCA. SATAN-VI calculates the thermal-hydraulic transient, including the RCS pressure, enthalpy, density, and the mass and energy flow rates in the RCS, as well as steam generator energy transfer between the primary and secondary systems as a function of time during the blowdown phase of the LOCA. SATAN-VI also calculates the accumulator water mass and internal pressure and the break mass and energy flow rates that are assumed to be vented to the containment during blowdown. At the end of the blowdown phase, the mass and energy release rates during blowdown are transferred to the COCO code, detailed in Reference 11, for use in determination of the containment pressure response during the first phase of the LOCA. Additional SATAN-VI output data f rom the end-of-blowdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCBART code. At the end of the blowdown, information from SATAN-VI on the state of the system is transf erred to the WREFLOOD code which calculates the time to BOC recovery, RCS conditions at BOC and mass and energy release from the break during the reflood phase of the LOCA. 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

pressure, the WREFLOOD and COCO codes are interactively linked.

The BOC conditions calculated by WREFLOOD and the containment pressure transient calculated by COCO are used as input to the BASH code. Data from both SATAN-VI code and the WREFLOOD code out to BOC are input to the LOCBART code which calculates core average conditions at BOC for use by the BASH code. 4 The BART code (Reference 12) has been coupled with a loop model to form the BASH code, in which BART provides the entrainment rate based on the core flooding rate. The BASH code provides a realistic thermal-hydraulic simulation of the reactor core and RCS during the reflood phase of a large break LOCA. Instantaneous values of the accumulator conditions and safety injection flow at the time of completion of lower plenum refill are provided to BASH by WREFLOOD. Figure 2-1B illustrates how BASH has been substituted for WREFLOOD in calculating transient values of core inlet flow, enthalpy, and pressure for the detailed fuel rod model, LOCBART. A detailed description of the BASH code is available in Reference 4. The BASH code provides a sophisticated treatment of steam / water f.ow phenomena in the reactor coolant system during core reflood. A dynamic interaction between core thermal-hydraulics and system behavior is expected, and experiments have shown this behavior. The loop model determines the loop flows and pressure drops in response to the calculated core exit flow determined by BART. The updated inlet flow calculated by the loop model is used by BART to calculate a new entrainment rate to be fed 6

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT i y back to the loop code. This process of transferring data between j

BART, the loop code and back to BART forms the calculational process for analyzing the reflood transient.

This coupling of the j BART code with a loop code produces a dynamic flooding transient, { which reflects the close coupling between core thermal-hydraulics and loop behavior. The cladding heat-up transient is calculated by LOCBART which is a combination of the LOCTA code with BART. A more detailed description of the LOCBART code can be found in References 3 and 8. During reflood, the LOCBART code provides a significant improvement i in the prediction of fuel rod behavior. In LOCBART the empirical FLECHT correlation has been replaced by the BART code. BART l employs rigorous mechanistic models to generate heat transfer i coefficients appropriate to the actual flow and heat transfer ] regimes experienced by the fuel rods. i The fuel rod model is initialized at steady state in the LOCA calculation to be consistent with fuel temperature and rod internal i pressure data provided by the more sophisticated Westinghouse fuel l temperature analysis model. The stored energy and rod internal pressure are key parameters and agreement with the fuel analysis data is obtained by matching pellet average temperatures and rod l internal pressures. Modeling features necesrary to account for the reactor barrel-baffle region and the reactor fuel assembly thimbles were included in this analysis as presented in Reference 2. 2.4.3 Input Parameters and Initial Conditions Important input parameters and initial conditions used in the analysis are listed in Tables 2-1 and 2-2. The safety injection { performance, as modeled for the various large break LOCA cases, is presented in Figures 2-2A and 2-2B. Cases analyzed are given in Table 2-3. The bases used to select the numerical values that are input parameters to the analysis have been conservatively determined from extensive sensitivity studies (Westinghouse 1974 (Reference 13); Salvatori 1974 (Reference 14); Johnson, Massie, and Thompson 1975 (Reference 15)). In addition, the requirements of Appendix K to 10 CFR 50 regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis. The assumptions which were made pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs, and include such items as the core peaking factors, the containment pressure, and the performance of the ECCS. Decay heat generated throughout the transient is also conservatively calculated as per the requirements of Appendix K to 10 CFR 50. The large break evaluation model assumes a chopped cosine power shape and a core design methodology will be applied 7

l SEABROOK STATION LOCA SAFETY ANALYSIS REPORT which assures that this shape remains bounding. Figure 2-7 contains the K(:) curve assumed in the analysis. Westinghouse ECCS analyses currently assume minimum safeguards for the safety injection flow, which minimizes the amount of flow to the RCS by a'ssuming that the spilling line is the branch line with the least resistance. This is the limiting single failure assumption when LOOP is assumed for most Westinghouse plants. l However, for some Westinghouse plants, including Seabrook Station, the current nature of the Appendix K ECCS evaluation model is such that it may be more limiting to assume the maximum possible ECCS flow delivery. The maximum safeguards case assumes operation of l both trains of ECCS pumps, minimum injection line resistances and enhanced ECCS pump performance. The worst break for Seabrook Station was reanalyzed assuming maximum safeguards. Examination of l the LOCA analysis results in Table 2-5 demonstrates that minimum safeguards assumptions result in the highest peak clad temperature for Seabrook Station. 2.5 Results Based on the results of the LOCA sensitivity studies (Westinghouse 1974 (Reference 13); Salvatori 1974 (Reference 14);

Johnson, Massie, and Thompson 1975 (Ref erence 15) ), the limiting large break was found to be the double-ended cold leg guillotine (DECLG).

Therefore, only the DECLG break is considered in the large break ECCS performance analysis. Calculations were performed for a range of Moody break discharge coefficients. The results of these calculations are summarized in Tables 2-4 and 2-5. The mass and energy release rates during reflood are shown in Table 2-6. The blowdown mass and energy release data are provided in Table 2-7. i l Figures 2-3 through 2-6 present the results of the cases analyzed for the large break LOCA. [ Figures A Reactor Coolant System Pressure (Calculated Core Pressure) Figures B Cold Leg Break Mass Flow Rate (Sum of Both Ends of the Guillotine Break) Figures C Core Power Transient (Fraction of the Nominal Power) Figures D Core Mass Flow Rate During Blowdown i Figures E Accumulator Mass Flow Rate During Blowdown (Sum of Injection into the Intact Loops) 8 . _ _ _ _ _. _ _. _.,. _ _ _. ~ _ _ _ _ _

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9 SEABROOK STATION LOCA SAFETY ANALYSIS REPORT Figures F Reflood Core and Downcomer Collapsed Liquid Water Levels Figures G Break Energy F21 eased to Containment as Calculated by SATAN Figures H Fluid Velocity Past Clad Hot Spot During Reflood Figures I Fluid Quality a; Calculated at the Hot Spot Figures J Hot Rod Heat Transfer Coefficient at the Hot Spot Figures K Fluid Temperature at the Clad Hot Spot Figures L Clad Temperature at the Hot Spot on the Hot Rod Figures M Containment Pressure Transient Figures N Containment Condensing Wall Heat Transfer Coefficient The limiting large break peak clad temperature PCT calculated is 18890F, which is less than the acceptance criteria limit of 22000F. This PCT, from the C = 0.6 minimum safeguards case, bounds all 3 fuel types and fAatures analyzed. Addition of the SoF penalty for the increase in Tm associated with RTD bypass elimination yields a total PCT of 1894 F. This licensing basis PCT remains below the 22000F acceptance criterium of 10 CFR 50.46. The maximum local metal-water reaction is well below the embrittlement limit of 17 percent as required by 10 CFR 50.46. The total core metal-water reaction is less than 1.0 percent for all breaks analyzed, corresponding to less than 1.0 percent hydrogen generation as required by 10 CFR 50.46. The clad temperature transient is terminated at a time when the core gecmetry is still amenable to cooling. As a result, the core temperature will continue to drop and the ability ec remove decay heat generated in the fuel for an extended period of time will be provided. 9 +

u SEABROOK STATION LOCA SAFETY ANALYSIS REPORT TABLE 2-1 INPUT PARAMETERS USED IN THE LARGE BREAK LOCA ECCS ANALYSIS License Core Power * (MWt) 3411 Peak Linear Power for Fuel Rods" (kW/ft) 13.6 Total Peaking Factor, F 2.5 o Axial Peaking Factor, Fz 1.515 Hot Channel Enthalpy Rise Factor, 7AH 1.65 Hot Assembly Average Power, Pu 1.469 Power Shape Chopped Cosine Fuel Assembly Array 17x17 ZIRLO or Zircaloy-4, 0.374

Fuel Rods Accumulator Water Volume (f t / accumulator) 850 (Nominal) 3 Accumulator Gas Pressure, Minimum (psia) 600 Safety Injection Pumped Flow (SIPS and CCPs degraded 5%, RHR 'kgraded 8.75%,

See Figures 2-2A CCP flow imbalance = 10 gpm) and 2-2B Containment Parameters See Table 2-3 Initial Loop Flow (gpm/ loop) 93800 Vessel Inlet Temperature (oF) 557.66 Vessel Outlet Temperature (oF) 619.34 Reactor Coolant Pressure (psia) 2300.0 Steam -Pressure (psia) 955.7 Steam Generator 2be Plugging Level (%) 13** Refueling Water Storage Tank Temperature for Containment Spray (oF) 50.0 Refueling Water Storage Tank Temperature for Safety Injection (oF) 100.0 Fuel Backfill Pressure (psig) 275 (100 for IFBA) Low Pressurizer Pressure Setpoints (psia): Reactor Trip 1860 Safety Inject.'on Signal 1665 Safety Injection Delay Time (sec) 30 S&fety Injection Spilling Containment Pressure (psig) 0.0 Feedwater Iso 19 tion Delay after Reactor Trip (sec, 0.0 Steamline Isolation Delay after Reactor Trip (sec) 0.0 Blowdown Containment Pressure (psia) 36.5 Two percent is added to this power to account for calorimetric error. Analysis is performed at 13% SOTP in order to support operation with B% SGTP. 5% SGTP is allocated for steam generator tube crush during a combined LOCD seismic event. 10 _-__-__m_

9 SEABROCK STATION LOCA SAFETY ANALYSIS REPORT TABLE 2-2 LARGE BREAK LOCA CONTAINMENT DATA Net Free Volume

2,974,000 ft' Initial Conditions Pressure 14.6 psia Temperature 90.O'F RWST Temperature 50.00F Temperature Outside Containment 50.00F Spray Temperature 50.00F Soray Syste-Spray ystem Flow Rate 6000 gpm Starting Time for Spray 47 see Temperature of Spray Water 50eF The contai=ent net free volu:r.e modeled in the analyses was adjusted from this value to account for contaiment purge effects.

11

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT TABLE 2-2 (Continued) LARGE BREAK LOCA CONTAINMENT DATA Structural Heat Sinks T na Thickness Wall Tu, Area i [0F) [ft] [ F] [ inches) 3 1 50 70151 90 0.009 Paint / 0.375 Carbon Steel / 54.0 Concrete 2 50 33856 90 0.009 Paint / 0.50 Carbon Steel / 42.0 Concrete 3 90 106165 90 0.017 Paint / 53.5 Concrete 4 90 7873 90 0.199 Stainless Steel / 33.1 Concrete 5 90 74419 90 0.174 Carbon Steel 6 90 71157 90 0.018 Paint / 0.56 Carbon Steel 7 90 21912 90 0.018 Paint / 1.43 Carbon Steel 8 90 2580 90 0.009 Paint / 1.55 Carbon Steel 9 90 12804 90 0.009 Paint / 0.250 Carbon Steel / 167 Concrete 10 50 3575 90 0.009 Paint / 0.844 Carbon Steel 11 50 396 90 0.009 Paint / 0.816 Carbon Steel

1. - Containment Cylinder

2. - Containment Dome

3. - Miscellaneous Concrete 4.

- Refueling Canal

5. - Ducts and Trays

6. - Structural Steel 7.

- Polar Crane B. - Equipment Steel

9. - Containment Floor and Sump

10. - Equipment Hatch

11. - Personnel Hatch 12

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT TABLE 2-3 LARGE BREAK LOCA - CASES ANALYZED CASE I - Co= 0. 4, 3411 Edt Core Power, Fe=2. 5, FM=1. 65, P-BAR-HA=1.469, Minimum Safeguards. CASE II - C= = 0. 6, 3411 Edt Core Power, F =2.5, FM=1.65, o P-BAR-HA=1.469, Minimum Safeguards. This case was found to be limiting and bounds both ZIRLO and Zircaloy-4 cladding. CASE III - Ce= 0. 8, 3411 Edt Core Power, F =2. 5, FM=1. 65, e P-BAR-HA=1.469, Minimum Safeguards. CASE IV - C = 0. 6, 3411 Kdt Core Power, F =2. 5, FM=1. 65, e P-BAR-HA=1.469, Maximum Safeguards. All cases mode: 13% steam generator tute plugging (in order to bound eperation with it SGTP) and 2% reduction in thermal design flow (93800 gp:/ loop). 13

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT TABLE 2-4 LARGE BREAK LOCA RESULTS - TIME SEQUENCE OF EVENTS Case I Case II Case III Case IV C =0. 4 C =0. 6 Cc= 0. 8 Ce= 0. 6 3 3 Min SI Min SI Min SI Max SI Start of LOCA with LOOP (sec) 0.00 0.00 0.00 0.00 Reactor Trip Setpoint Exceeded (sec) 0.757 0.735 0.720 0.735 Safety Injection Setpoint Exceeded (sec) 2.09 1.67 1.45 1.67 Accumulator Injection Begins (sec) 20.0 15.0 12.5 15.0 End-of-Bypass (sec) 39.2 33.0 28.9 33.0 End-of-Blowdown (sec) 39.2 33.3 28.9 33.3 Pump Injection Begins (sec)" 32.1 31.7 31.5 31.7 Bottom of Core Recovery (sec) 54.5 47.8 43.8 47.4 Accumulators Empty (sec) 61.1 53.3 50.1 53.8 14

't SEABROOK STATION LOCA SAFETY ANALYSIS REPORT TABLE 2-5 LARGE BREAK LOCA RESULTS - FUEL CLADDING DATA Case I Case II Case III Case IV Co= 0. 4 Co= 0. 6 Co= 0. 8 Co= 0. 6 Min SI Min SI Min SI Max SI Peak Clad Temperature (oF) 1682 1894* 1823 1762 Peak Clad Temperature 6.25 Location (ft) 8.75 7.00 7.00 Peak Clad Temperature Time (sec) 281 128 92.6 59.8 Maximum Local Zr/H O 2 Reaction (%) 1.95 3.41 2.69 1.87 Maximum Zr/H O Reaction 2 Location (ft) 6.75 7.00 7.00 7.00 Total Zr/H O Reaction 2 (%) <1.0 <1.0 <1.0 <1.0 Hot Rod Burst Time (sec) 88.7 55.1 47.8 46.9 Hot Rod Burst Location (ft) 6.75 6.00 6.00 5.50 Includes 5.DCF evaluation penalty for Increased Temperature Uncertainty (25'F) associated with RTD Bypass Elimination. 15 e

SEABROOK STATION LOCA SAFETY. ANALYSIS REPORT TABLE 2-6 LARGE BREAK LOCA Cc=0.6 MINIMUM SAFEGUARDS LOCA REFLOOD MASS AND ENERGY RELEASE RATES Time Mass Flow Rate Energy Flow RateL' (sec) (1bm/sec) '(BTU / sec) - 47.80 0.0-

0. 0.

50.00 9.14 11859 56.78 47.08 59471 60.03 52.13 65527 70.08 66.95. 83306 72.88 71.12 88310 80.08 81.51 100781 90.08 94.39 116239' 90.88 95.35 117391 100.08 105.91 130062 108.88 115.17 141170 110.08 116.37 142610 120.08 126.90 154213 126.58 214.10 .182311 130.08 262.26 196636 140.08 323.03 212249 143.88 330.59 213119 150.08 337.12 212641. 160.08_ 342.59 210280 161.88 343.32 209762 170.03 389.81' 220922-180.03 421.50 227259 190.03 433.60 228914 200.03 429.70 226158 203.73 430.20 225608 210.03 429.79 224385 220.03 431.49 223140 230.03 433.81 222068 240.03 437.88 221466 250.03 440.41 220476 260.03 443.17 219641 270.03 445.21 218627 274.73 445.83 218063-280.03 446.64 217485 290.03 449.22 216715 300.03 450.67 215667 16

SEABROOK STATION' LOCA SAFETY. ANALYSIS REPORT TABLE 2-7 LARGE BREAK LOCA' C:=0.6 MINIMUM SAFEGUARDS LOCA BLOWDOWN MASS AND ENERGY RELEASE RATES. Time Mass Flow Rate Energy Flow Rate (sec) (1bm/sec) (BTU /sec) 0.0 2954-184950' 1.0 -2760 172772 2.0 2602 162860-3.0 2469 154548 4.0 2355 147421 5.0 2256 141197 6.0 2167 135669 7.0 2008 130709 8.0 2016 126227 9.0 1951 122150 10.0 -1892 118420 11.0 1837 '114993 12.0 1786 111835 13.0 1740 108908 14.0 1696 106182 15.0 1656 103638 16.0 '1618 101265 17.0 1582 99047 18.0 1549 96975 19.0 1518 95032 20.0 1489 93195-21.0 1461 -91463-22.0 1435 89831-23.0 1410 88288-24.0 1387 86827' 25.0 '1365 85447-26.0 1344 84146 27.0 1324 82910. 28.0 1306 81727 29.0 1287 80595 29.51 0.0 0.0 17

[L taEAR OCCuas atactoa f air COMPEm&Af t0 PRESSumt2em patssynes puMeEO SAptTV sNJECTION slGNAL iMe.t CONT patts on to petssunt2ER PRESE i punaPED SAFETY I4 JECT 40eu 3EGi45 iASSUMING Opps TE pouvEm Ava:LAstEs e L 0 ACCUMUL Atom sNJECTION = 0 CONTasseMENT ME AT REMCv&L SYSTEM sWITIATeON (AssumiWG oppstTE PowEm ava LASLEt Ow N u ENo os sveAss y a END OF BLOWOOWN

== neu g a PUMPED SA8ETY 48WECT30N SEGlast IASSUMtWG LOS$ OF OFFSITE powtas i L l L t soTto" 0F Coat mEcovEnv - O CONTAleeMENT ME AT AEMOV AL SYSTEM teelT8 ATION (A33yMineG LOSS OF OFFSITE POWEma R i 8 ACCUMuLATOREEarTV L 0 0 0 CCSE OyENCMED T L o = G swtTCM TO COLD LEG RECIRCULATIQes ON AwST L0ur LEVEL ALAmu tsEMI.Aut0MAftCI T E a M C o o SuriTCM TO LOesc.TEau RECtRCULAfl0N IMA8ev&L ACTIOss N G 1F LARGE BREAK LOCA SEQUENCE OF G WESTINGHOUSE ELECTRIC CORPORATION EVENTS Nucieer and Advanced Technology Divisions FIGURE 21 A 18

I l E0etNu W ILL l StFLOOD l l fog 80C#tt ' 8t' ' " I l CALuAft3 et a00. AwACm n00. Am CALnAfts et me. matm n00. AaB et A51[MT 000 ftwtuTual. GLOCuGE. h.t.c. 47 A55CET B00 TDFERAtual. GLOCIACE. AL50 CALCE ATES Coat Ttw tRATWat (LOCTA QuLT) j AaB n.s.t. e 7 115t % f. Cett wil vtLOCITT. twat!Tv. "II#E CORI FLOOC1hG Raft. IRLif Cgs [ (NTMALPT CONCITIONS Satan l AT 90CREC SASM l CALCULAfts CDet FLOOCl e 30Catt. Raft. aC5 CONCIT10ml Dualms CAL M Afts it$. CDAE. RCS CONCITIOR$ ALFLOOD ET A11t R T AT 90CAtt FLul0 CONCIT10R$ o RCS M IT10n3 m13. twtetY atttAst ACCU R ATOs. $1 Flow.

I I" ImTO CONTA! M ht CONTAI M NT Pl[55WRC CALMAft1 etFILL. FLODBlut MTt 45 255. ENERCY atL! alt Raft FBom BCS OWBlut strL000.

(WatrLD001. CAL M Afts CONTAI M NT PRC15utt (C0CO) dtFLOOC/C0C0l CAL M Aft 1 CORTA! M NT Patssung (CDC0 thiLY) LARGE BREAK LOCA CODE O WESTINGHOUSE ELECTRIC CORPORATION INTERFACE Nuclear and Advanced Technology Drvisions FIGURE 2-1B 19

600 500 4 400 N E$ oj300 5E en j200 100 0 0 100 200 300 400 500 Pressure (psio) PUMPED SAFETY INJECTION FLOW O WESTINGHOUSE ELECTRIC CORPORATION vs. RCS PRESSURE Nuclear and Advanced Technology Divisions (MINIMUM SAFEGUARDS) FIGURE 2-2A 20

s { 700 600 500 E< E $ 400 s E $300 c E 200 N \\ 100 0 0 100 200 300 400 500 Pressure (psic) PUMPED SAFETY INJECTION FLOW O WESTINGHOUSE ELECTRIC CORPORATION vs. RCS PRESSURE Nuclear and Advanced Technology Divsions (MAXIMUM SAFEGUARDS) FIGURE 2-2B 21

r l i l I ] i 2500. l l 2250. 2000. l l 1750. 1500. -5 1 <n 0. 1 i W 1250 \\ ,000. ct l a. \\ 750. \\ 500. \\ 250. w 0. O. 5. 10. 15. 2 0.. 25. 30. 35. TIWE (SEC) l i l l RCS PRESSURE l WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Devaions (Co = 0.6) FIGURE 2-3A 22

.7E+05 .6E+05 .5E+05 6 =d 4E+05 N< C 3l: 9 .3E+05 u. M $2 .2E+05 1E+05 O. O. 5. 10. 15. 20. 25. 30. 35. TIME (SEC) COLD LEG BREAK MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvaions (C = 0.6) o FIGURE 2-3B 23

l l i I l 1. 1 I 9 I j .8 1 3 i Z 2 .7 i O l i z u.O '5 i z i 9w i o i .5 CE Cw 4 3O i .3 6 j 1 m I l l l 0. 0 5 10 15 20 25 30 35 TIME (SEC) l CORE POWER 0 WESTINGHOUSE ELECTRIC CORPORATION (FRACTION OF NOMINAL) Nuclear and Advanced Technology Drvisions (C = 0.6) o l FIGURE 2-3C 24 l 1

I 9 1

8000, i,

l\\ lit fi 6000. I i 1 4000. i 6 I w l m ig i 2 2000. .i a \\ (j N \\ s I' i 0. m - I if w ~ M '[I Q -2000-. .g 2 \\ l -4000. -6000. -8000. 1 0. 5. 10. 15. 20. 25, 30. 35. TlWE (SEC) CORE MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION (TOP AND BOTTOM) Nuclear and Advanced Technology Dtvsions (Co = 0.S) FIGURE 2-3D 25 9

l 4000. 3500. l 3000. G w 2500. m I T 3 G I i H 2000. i I l Z 3oif 1500. m m 2 1000. 500. l 1 1 0. O. 5. 10. 15. 20. 25. 30. 35. TIME (SEC) 1 O WESTINGHOUSE ELECTRIC CORPORATION ACCUMULATOR MASS FLOW RATE Nuclear and Advanced Technology Divsions (Co = 0.6) FIGURE 2-3E 26 l

1 l + i i j i i 1 1 1 1 22 l l l .I DOWNCOMER 1 20 [ [ [ TOP OF CORE 16 w 1a 7) j' l QUENCH FRONT !'10-

~-

I i a i 1 CORE ....m.... .1... g. l 0 l a j 0 50 100 150 200 250 300 ,i TtWE (SEC) i i i i a 4 l i i i i i l 4 REFLOOD CORE AND DOWNCOMER l WESTINGHOUSE ELECTRIC CORPORATION WATER LEVELS j Nuclear and Advanced Technology Divsions (Co = 0.6) i j FIGURE 2 3F i e e j 27 i

l i l i \\ l i 4E-08 .35E+08 i i l .3E+08 \\ 2 } $.25E-08 l 3 i s i cn 1 j t t i e .2E-08 ( c l Wz w 15E-08 l \\ 1E-08 .5E-07 g w l 0. D. 5. 10. 15. 20. 25. 30. 35. TlWE (SEC) i l l BREAK ENERGY RELEASED [ l WESTINGHOUSE ELECTRIC CORPORATION TO CONTAINMENT Nuclear and Advanced Technology Drvisions (Co = 0.6) FIGURE 2-3G i 1 28 i

l l. l l .9 .8 1 f .7 l .6 5 ~ jf P 3 C g h4 l' r A, u d -p .3 l ~ .2 .1 0. 0 50 100 150 200 250 300 TIME (S) l FLUID QUALITY 6 WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvisions (Co = 0.6) FIGURE 2-3H 29

i l I l i 4 l ) 80 i 1 e0 y 40 ihr G 20 I x \\ u-J O h -20 9 o t o 3 -40 LJ t -60 -80 I -100 I O 50 100 150 200 250 300 l TIME (S) l FLUID VELOCITY PAST CLAD 6 WESTINGHOUSE ELECTRIC CORPORATION HOT SPOT Nuclear and Advanced Technology Drvisions (Co = 0.6) FIGURE 2-31 l 30

e 103 u 1 ca L .x= sa* 102 4 cn l ,i 1 L I Lw h / ,H a U 1p W 1 I \\ b l 10 l u Ch l Z l Cm l W W Cu 100 0 50 100 150 200 250 300 TIME (S) i HOT ROD HEAT TRANSFER O WESTINGHOUSE ELECTRIC CORPORATION COEFFICIENT l Nuclear and Advanced Technology Drvsions (Co = 0.6) FIGURE 2 3J 31

i - =_. l l l 1800. I l 1600. 1400. 'g y u. i 1200. w ma 1000. s C" J E 800. s 7 ru 000. 400. 200.O. 50. 100. 150. 200. 250. 300. TIME (S) CLAD HOT SPOT FLUID O WESTINGHOUSE ELECTRIC CORPORATION TEMPERATURE Nuclear and Advanced Technology Drvisions (Co = 0.6) FIGURE 2 3K 32

4 i 1900. \\^ 1800. N/\\ k 1700. {1600. g Y u 1500. m ) i o 1 +c 1400. m w i CL I 1300. w 1200. l I \\ 1100. J 'O. 50. 100. 150. 200. 250. 300. TIME (S) HOT ROD PEAK CLAD TEMPERATURE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divaions (C = 0.6) o FIGURE 2-3L 33

I i l l 45 40 f a x 35 - 7 \\ b N n. l 1

25 E

.E2 i C 2 l 0 20 15 l i 10 l 0 100 200 300 400 500 l Time (sec) i l i CONTAINMENT PRESSURE e WESTINGHOUSE ELECTRIC CORPORATION Nucisar and Advanced Technology Drvisions (Co = 0.6) l FIGURE 2 3M l 34 l

I i l l l 800 f e if r l l 600 E I N ( r l t l N 3 l s E 40c - o t

G E

$o L2 200 8z 0 0 100 200 300 400 500 Time (sec) CONTAINMENT CONDENSING WALL WESTINGHOUSE ELECTRIC CORPORATION HEAT TRANSFER COEFFICIENT Nu:Isar and Advanced Technology Dmsens (Co = 0.6) FIGURE 2-3N 35 J

l l i i l i I 2500. 2250. l 2000. J 1750. r 5 1500. m e. w l 1250. \\< mwc \\ n. 1000. 750. 500. \\ 250. N 0. '1 O. 5. 10. 15. 20. 25, 30. TIWE (SEC) l l l l l RCS PRESSURE WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Desions (Co = 0.8) FIGURE 2 4A 36

9 .BE+05 7E.6E+05 ^ .5E-05 3s .4E-05 = N c O .3E-05 d 8 j .2E+05 g 1E+05 h n-0. - 1E+05 O. 5. 10. 15. 20. 25. 30. TIME (SEC) COLD LEG BREAK MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divisions (Co = 0.8) FIGURE 2-4B i 37

1. .9 .8 3<z5 7 i O z LL.O .6 z9r .5 C's Cu) 4 3:O o. I .3 .2 i l 1 i T l i l 0. 0 5 10 15 20 25 30 TlWE (SEC) CORE POWER WESTINGHOUSE ELECTRIC CORPORATION (FRACTION OF NOMINAL) Nuclear and Advanced Technology Drvsions (Cp = 0.8) FIGURE 2-4C 38

i 1 l 1 4500. 4000 -^ l / 3500. 6 l 3000. T l 3 2500. ^ y i l gr 9 2000. l u. m m<2 1500. 1000. g 500. O. O. 5. 10. 15. 20. 25. 30. TIWE (SEC) l ACCUMULATOR MASS FLOW RATE 9 WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvsions (Co = 0.8) FIGURE 2-4D l f 39 l

i I i i S000. i i, I II Ill 111 1 6000. ilt lll 1 11 4000. i; 8 1 6 w I 2000. 3 \\ s s \\ 0-1 f 5 ,/ 3 / g t I m -2000. ij m \\ If e n l{. \\ l -4000.

\\

i l f W -6000. i -8000. O. 5. 10. 15. 20. 25 30. TtuE (SEC) 1 l CORE MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION (TOP AND BOTTOM) Nuclear and Advanced Technology Drvsions (Co = 0.8) FIGURE 2-4E 40 l I

1 i

l a

i 4 i I l i 1 i 22 DOWNCOMER 18 l l l ,h TOP OF CORE 5 [l"F I I b l Tf l OUENCH FRONT CORE B-l l 6 l I 4 0 50 100 150 200 250 300 TiuE (SEC) REFLOOD COF.E AND DOWNCOMER O WESTINGHOUSE ELECTRIC CORPORATION WATER LEVELS Nuclear and Advanced Technology Drvisions (C = 0.8) o FIGURE 2 4F 41

1 45E-08 4E-08 l .35E-08 ) .3E+08 2.25E-08 a3 \\ .2E-08 >-0 E l z 15E-08 f w 1E-08 w .5E-07 0. ^ 0. 5. 10. 15. 20. 25. 30. TIME (SEC) BREAK ENERGY RELEASED O WESTINGHOUSE ELECTRIC CORPORATION TO CONTAINMENT Nuclear and Advanced Technology Dusions (Co = 0.8) FIGURE 2-4G 42

1. .9 .8 .7 ] .6 C / I ao .5 hi l% l d ^ ^ 4 }\\ gM p# 'V' 3 .3 l .2 0 50 100 150 200 250 300 TIME (S) s FLUID QUALITY 9 WESTINGHOUSE ELECTRIC CORPORATION y (Co = 0.8) l Nuclear and Advanced Technology Dumions FIGURE 2-4H f$1,o 43 .'

Tp, 1

60 a - n-w 40 t i r i s. h 20 en \\ N ( Y ~ b o i C -20 o I \\ 3w> -40 -60 L -80 0 50 100 150 200 250 300 TIME (S) FLUID VELOCITY PAST CLAD O WESTINGHOUSE ELECTRIC CORPORATION HOT SPOT Nucle'ar avi Actvanced Technology Drvisions (Co = 0.8) f [ FIGURE 2-41 44

103 L m W L l[/ Co ~ 102 co l u i u S -W / ) 10 1 u en Z c i oc + i CW 100 ~ 0 50 100 150 200 250 300 TIME (S) l l l HOT ROD HEAT TRANSFER l WESTINGHOUSE ELECTRIC CORPORATION COEFFICIENT l Nuclear and Advanced Technology Drvsions (Co = 0.8) FIGURE 2-4J l 45 l

I i t i 1800. 1600. / ) I 1400. l 5rt i 1200. ig u ma 1000. r Cm B y 800. p uu \\ 600. 400. s ? ~OO 'O. 50. 100. 150. 200. 250. 300. TIME (S) CLAD HOT SPOT FLUID O WESTINGHOUSE ELECTRIC CORPORATION TEMPERATURE Nuclear and Advanced Technology Divisions (Co = 0.8) FIGURE 2-4K 46

4 1 1900. ^ ^ 1800. S \\ l 1700. / \\ \\ l 1600. c / \\ 1500.,; j ;i W 1400- \\ xo l 1300. A / m w / ct 1200. r i \\ [ } w* 1100. s 1000. 900. 800. O. 50. 100. 150. 200. 250. 300. TIME (S) HOT ROD PEAK CLAD TEMPERATURE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divisions (Co = 0.8) FIGURE 2-4L 47

t l l 45 l 40 N \\ o l O i e \\ 5 30 b$ c" 25 E .E E I c ) O l U 20 l i l l l l 15 I 10 O 100 200 300 400 500 Time (sec) l l CONTAINMENT PRESSURE 6 WESTINGHOUSE ELECTRIC CORPORATION NucIsar and Acfvanced Technology Divisions (C = 0.8) l o FIGURE 2-4M l 48 l

s l l l l 1000 n u_ g 800 m l l E i N <t) 600 I 5 v %.c 8 400 O f u 2 m E* 5 200 o I V O O 100 200 300 400 500 Time (sec) l CONTAINMENT CONDENSING WALL O WESTINGHOU.SE ELECTRIC CORPORATION HEAT TRANSFER COEFFICIENT Nuclear and Advanced Technology Duisions (Co = 0.8) FIGURE 2-4N 49

l l i 2500. I 2250. l 2000. 1750. Eg 1500. k w C o 1250. l m mw C \\ 1 1000. 750. x 500. 1 i 250. s\\ l w O. l 0. 5. 10. 15. 20. 25. 30. 35. 40. l TIWE (SEC) RCS PRESSURE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvisions (C = 0.4) o i FIGURE 2-5A l [ 50 l

.6E+05 .5E+05 @.4E+05 m 2d N $.3E+05 3: 9 u. Om $.2E+05 1E+05 7 0. O. 5. 10. 15. 20. 25. 30. 35. 40. IlWE (SEC) COLD LEG BREAK MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divisions (Co = 0.4) FIGURE 2-5B 51

i 2 1. .9 i l i .8 l 3 .7 2E2Oz .6 u. O zg .5 ro<C i g .4 4 Cw i 3l:o .3 c. .2 i 1 ~N 1 0' O 5 10 15 20 25 30 35 40 1 TIME (SEC) i CORE POWER O WESTINGHOUSE ELECTRIC CORPORATION (FRACTION OF NOMINAL) Nuclear and Advanced Technology Drvisions (Co = 0.4)- FIGURE 2-5C 52

8000. i i I f I 6000. i i I i II l L 4000. g \\ \\ \\ 1 n0 \\ 2000. l m t s w N r ~u ~Y f ~, L. hP f 9 $ -2000. I ip l' 1 2 f -4000. -6000. 1 ~ 'O. 5. 10. 15. 20. 25. 30. 35. 40. ItuE (SEC) CORE MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION (TOP AND BOTTOM) Nuclear and Advanced Technology Divsions (C = 0.4) o i FIGURE 2-50 53

1 l 4000. 3500. l 3000. G 2500. T 35 Q 2000. 1 3 9 I 1500. $<2 1000. 500. O. O. 5. 10. 15. 20. 25. 30. 35. 40. TlWE (SEC) ACCUMULATOR MASS FLOW RATE 6 WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divisions (C = 0.4) o FIGURE 2-5E 54

i t i i i I 4 27.5 l

25. \\

I I i 22.5

20. \\

^ -A / i 17.5 DOWNCOMER TOP OF CORE f j

15. ' ~ ~

2 ~' '~~- -

72.5

-[ QUENCH FRONT 10. 7.5 -- -. r j r i CORE pf t I 2.5 0 50 200 ggn TluE (SEC) i ) i 1 i i i REFLOOD CORE AND DOWNCOMER WESTINGHOUSE ELECTRIC CORPORATION WATER LEVELS Nuclear and Advanced Technology Dumions (Co = 0.4) FIGURE 2-5F 55 4 -a p-t p + *

l l

.35E-08 .3E+C8 I .25E+08 2 O

d 2E-08 o

~ >-0 15E+08 zw \\ 1E+08 \\ A .5E+07 \\ x N N 0.0. 5. 10. 15. 20. 25. 30. 35. 40. TIWE (SEC) BREAK ENERGY RELEASED O WESTINGHOUSE ELECTRIC CORPORATION TO CONTAINMENT Nuclear and Advanced Technology Dmssons (Co = 0.4) FIGURE 2-5G 56

i 1. l l k .9 3 t .8 l i i .7 W ] .6 c I i (N oo j ( (. .5 \\ \\ .4 \\Ni .3 l ? ~~ 0 50 100 150 200 250 300 TIME (S) i FLUID QUALITY O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Meanced Technology Drvsions (C = 0.4) o FIGURE 2-5H I 57

1 i l l 200 ? 150 G 100 N L l" "~ 8 ' 50 -] [T I y y 1 f 1 u O J 0 ) ( w> ] -50 -100 0 50 100 150 200 250 300 TIME (S) FLUID VELOCITY PAST CLAD O WESTINGHOUSE ELECTRIC CORPORATION HOT SPOT Nuclear and Advanced Technology Dmssons (Co = 0.4) FIGURE 2 51 58

103 u m g L I i m i E J 1 a 102 cc \\e i I \\' I N t h h I v w u g a J j o by VM, 1 10

A n' z

c p \\ 2 .[ h + 1 -cw 100 0 50 100 150 200 250 300 TIME (S) HOT ROD HEAT TRANSFER O WESTINGHOUSE ELECTRIC CORPORATION COEFFICIENT Nuclear and Advanced Technoiogy Div sions (Co = 0.4) FIGURE 2-5J 59

l 1600. 1400. f i A 1200-7 mv s W n i L 1000. 7 u ce D 1 I E 800. u Q. r. ) u '\\ 600. t i k 400. 3i 200. O. 50. 100. 150. 200. 250. 300. TIME (S) l CLAD HOT SPOT FLUID 0 WESTINGHOUSE ELECTRIC CORPORATION TEMPERATURE Nuclear and Advanced Technology Drvisions (C = 0.4) o FIGURE 2-5K 60

1 i 1800. l l 1600-l M/ [ 1400. i w m l o 1200. s C l m 1 w 1 i 1 j r

1000, w

f l 800. / \\ 600. O. 50. 100. 150. 200. 250. 300. l TIME (S) l t HOT ROD PEAK CLAD TEMPERATURE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divisions (Cp = 0.4) FIGURE 2-5L 61 l

45 l 40 N g 35 N mO o $ 30 \\ a h \\ c $ 25 .E3 EoU 20 15 10 0 100 200 300 400 500 Time (see) CONTAINMENT PRESSURE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Orvsions (Co = 0.4) FIGURE 2-5M 62

800 C ? I600-N I reb -6 j400 c '5 8 O bE C $200 E k 0 0 100 200 300 400 500 Time (sec) CONTAINMENT CONDENSING WALL O WESTINGHOUSE ELECTRIC CORPORATION HEAT TRANSFER COEFFICIENT Nuclear and Advanced Technology Drvisens (Co = 0.4) FIGURE 2-5N 63 i

2500. 2250. 2000. 1750. l 1 55 1500. L w I 3 1250' \\ w T I 1000. N 750. x 500. i \\ 250. w 0. 1 0. 5. 10. 15. 20. 25. 30. 35. TIWE (SEC) d RCS PRESSURE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divisions (C = 0.6, idAX SI) o FIGURE 2-6A 64

I .7E+05 .6E+05 .5E+05 G wwTg.4E+05 N<C 3 9 3E+05 u. m i w<2 .2E+05 1E+05 O. O. 5. 10. 15. 20. 25. 30. 35. TlWE (SEC) i 1 COLD LEG BREAK MASS FLOW RATE i 0 WESTINGHOUSE ELECTRIC CORPORATION j i Nucisar and Advanced Technology Drvaions (Co = 0.6, MAX SI) l FIGURE 2-6B 65 j

I l 1. l i I .9 .8 3 I .e i5 l 2 Oz .5 u.O z9 .5 wo<C l b .4 1w 5 2 .3 l .2 I i i I l l 1 l l 0. 0 5 10 15 20 25 30 35 TlWE (SEC) CORE POWER O WESTINGHOUSE ELECTRIC CORPORATION (FRACTION OF NOMINAL) Nuclear and Actvanced Technology Dmssons (Co = 0.6, MAX SI) FIGURE 2-6C 66

8000. ll lin II 6000. I I I 4000. I G I Id i T 2000. ir i md i L \\ b 0. .A a' f NF v; -2000. 7 2 g \\ -4000. -6000. -8000.' O. 5. 10. 15, 20. 25. 30. 35. TIWE (SEC) I 1 l CORE MASS FLOW RATE 6 WESTINGHOUSE ELECTRIC CORPORATION (TOP AND BOTTOM) Nuclear and Advanced Technology Dwisions (C = 0.6, MAX SI) o FIGURE 2-6D l 67 1 i

4000. 3500. 3000. G $T 2500 7 F s 3: 2000. 9 u. j 1500. ~ 1000. 500. O. O. 5. 10. 15. 20. 25. 30. 35. TlWE (SEC) ACCUMULATOR MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Duisions (Co = 0.6, MAX SI) FIGURE 2-6E 68

1 I i i s I } d 1 22 j _-f-DOWNCOMEP 20 t l 18 j TOP OF CORE l 16 _ -- d i i ) l I ~ i 6 14 l l QUENCH FRONT j h 12-a 1 ( l 1 d 10 - l CORE l 6 'I l 1 4 0 50 100 150 200 250 300 l TIWE (SEC) REFLOOD CORE AND DOWNCOMER 9 WESTINGHOUSE ELECTRIC CORPORATION WATER LEVELS Nuclear and Advanced Technology Drvaions (Co = 0.6, MAX SI) FIGURE 2-6F 69

4E+0B .35E+0B .3E+08 I 3.25E+0B ed I ? m [ .2E+08 3 e N E 5 15E+0B 1E+08 \\ .5E+07 NI I l 0. O. 5. 10. 15. -20. 25. 30. 35. T1WE (SEC) l l BREAK ENERGY RELEASED 9 WESTINGHOUSE ELECTRIC CORPORATION TO CONTAINMENT Nuclear and Advanced Technology Drvsons (Co = 0.6, MAX SI) { FIGURE 2-6G i 70 r

l \\ 1 = \\ ~ %Cp+W 0. .5 t- _1-a Cao -1.5 -2. -2.5 -3. 0 50 100 150 200 250 300 TIME (S) 1 I FLUID QUALITY WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvaions (Co = 0.6, MAX SI) FIGURE 2-6H i = l 71 l l l

i i 80 60 40 i i ~ WM w 20 t NW u 0 i -20 ~o 1 o J -40 u { -60 t i -80 -100 0 50 100 150 200 250 300 TIME (S) i FLUID VELOCITY PAST CLAD O WESTINGHOUSE ELECTRIC CORPORATION HOT SPOT 1 Nuclear and Advanced Technology Divisions (Co = 0.6, MAX SI) FIGURE 2-6l 72

103 k a. m l w k mzxo* 102 ji! ? 1 n Lw \\ sr h 0 _V o r 1 k 2u 10 L O Z l c 2w wc u 100 0 50 100 150 200 250 300 TIME (S) HOT ROD HEAT TRANSFER O WESTINGHOUSi ELECTRIC CORPORATION COEFFICIENT Nuclear and /.dvanced Technolo;y Dimens (C = 0.6, MAX SI) o FIGURE 2-6J 73

l i i l l l 1800. 1600. ( l 1400. t. 1200. u 1000. t^TN i \\ E 800. ru 600. g. q 400. \\ 200. O. 50. 100. 150. 200. 250. 300. TIME (S) ~ CLAD HOT SPOT FLUID O WESTINGHOUSE ELECTRIC CORPORATION TEMPERATURE Nuclear and Advanced Technology Divaions (Co = 0.6, MAX SI) FIGURE 2-6K i 74

1800. 1600. 3N A s 1400. g4 L 1200. ) LJ Ma 1000. \\ I-C E 800. E LJ 600. 400. 200. O. 50. 100. 150. 200. 250. 300. TIME (S) I HOT ROD PEAK CLAD TEMPERATURE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Dusions (Co = 0.6, MAX SI) FIGURE 2-6L 75

1 i l l 45 l l l ^ 40 l-l ) 35 9sO \\ E3 30 E N E CL c \\

25

\\ C O U l 20 15 10 0 100 200 300 400 500 Time (sec) l l CONTAINMENT PRESSURE 0 WESTINGHOUSE ELECTRIC CORPORATION i Nuclear and Advanced Technology Duisions (Co = 0.6, MAX SI) FIGURE 2-6M 76 j i ~

J 800 i c EP i600 A 1 $eb -8 E 400 .e o a Yo C J200 E E k 0 0 100 200 300 400 500 Time (sec) CONTAINMENT CONDENSING WALL O WESTINGHOUSE ELECTRIC CORPORATION HEAT TRANSFER COEFFICIENT Nuclear and Advanced Technology Divaions (Co = 0.6, MAX SI) FIGURE 2-6N 77

i I l l I 1.2 i l 1 i ( $0.5 "o i u E 1o 20.6 i E O E Elevation (ft) K(r) z.4 0.0 1.0 o 6.0 1.0 12.0 0.925 0.2 F = 2.5 q 0 0 2 4 6 8 10 -12 Elevation (ft) 1 l l K(z) CURVE e WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Dumion FIGURE 2 7 l 78 1

SEABROOK STATION l LOCA SAFETY ANALYSIS REPORT I l 3.0 LOSS OF REACTOR COOLANT FROM SMALL RUPTURED PIPES OR FROM CRACKS IN LARGE PIPES WHICH ACTUATES THE ECCS (SMALL BREAK LOCA) This section presents a description and results of the small break l loss-of-coolant accident (LOCA) in conformance with 10 CFR 50.46 and Appendix K of 10 CFR 50 (Reference 1). 3.1 Identification of Causes and Frequency Classification A LOCA is the result of a pipe rupture of the reactor coolant system (RCS) pressure boundary. Ruptures of small cross-section will cause expulsion of the coolant at a rate which can be accommodated by the high head safety injection pumps and which would maintain an operational water-level in the pressurizer permitting the operator to execute an orderly shutdown. The coolant which would be released to containment contains the fission products present in it. ? A small break, as considered in this section, is defined as a rupture of the RCS piping with a cross-sectional area less than 1.0 ft, in which the normally operating charging system flow is 2 not sufficient to sustain pressurizer level and pressure. This event is considered an American Nuclear Society (ANS) Condition III event, which is a fault which may occur very infrequently during the life of Seabrook Station. 3.2 Sequence of Events and Systems Operations Should a small break LOCA occur, depressurization of the RCS causes fluid to flow into the loops from the pressurizer resulting in a pressure and level decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low-pressure trip setpoint is reached. Loss-of-Offsite Power (LOOP) is assumed to occur coincident with reactor trip. A safety injection signal is t l generated when the pressurizer low-low-pressure setpc.at. is I reached. These counter measures limit the consequences of the accident in two ways: A. Reactor trip and borated water injection supplement void formation in causing rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay. No credit is taken in the LOCA analysis for the boron content of the injection water. j However, an average RCS/ sump mixed boron concentration is calculated to ensure that the post-LOCA core remains suberitical. In addition, in the small break LOCA l

analysis, credit is taken for the insertion of Rod Cluster Control Assemblies (RCCAs) subsequent to the l

reactor trip signal, while assuming the most reactive l RCCA is stuck in the full out position. i 79

i SEABROOK STATION LOCA SAFETY ANALYSIS REPORT B. Injection of borated water ensures suf ficient flooding of the core to prevent excessive clad temperatures. For small break LOCAs, the most limiting single active failure is j the one that results in the minimum emergency core cooling system (ECCS) flow delivered to the RCS. This has been determined to be the loss of an emergency power train which results in the loss of one complete train of ECCS components. This means that credit can be taken for only one centrifugal charging pump (CC P), one safety l injection pump (SIP), and one residual heat removal (RHR) (or low head) pump. During the small break transient, one ECCS train is assumed to start and deliver flow through the injection lines (one for each loop) with one branch injection line (RHR and SIP) spilling tc the RCS backpressure. Because the line diameter may be less than the break size for small breaks, one charging pump injection line spills to containment backpressure. To minimize delivery to the reactor, the branch line chosen to spill is selected as the one with the minimum recistance. In addition, the l SIP and CCP performance curves were degraded by 5%, the RHR pump l performance curve was degraded 8.75%, and a 10 gpm flow imbalance was assumed for the high head safety injection pumps. l 3.3 Description of Small Break LOCA Transient Before the break occurs the plant is in an equilibrium condition, l i.e., the heat generated in the core is being removed via the secondary system. Af ter the small break LOCA is initiated, reactor trip occurs due to a low pressurizer signal. During the earlier part of the small* creak transient, the effect of the break flow is not strong enough to overcome the flow maintained bv the reactor coolant pumps through the core as the pumps coast oown following LOOP. Upward flow through the core is maintained. However, the core flow is not sufficient to prevent a partial core uncovery. Subsequently, the ECCS provides sufficient core flow to cover the core. During blowdown, heat from fission product decay, hot internals, and the vessel continues to be transferred to the RCS. The heat j transfer between the RCS and the secondary system may be in either direction depending on the relative temperatures. Continued heat j addition to the secondary system results in increased secondary system pressure which leads to steam relief via the main steam i safety valves. The safety injection signal isolates normal feedwater flow by closing the main feedwater control and bypass valves. Makeup to the secondary is automatically provided by the l auxiliary feedwater pumps. Loss-of-Offsite

Power, assumed i

concurrent with reactor trip, initiates auxiliary feedwater flow by starting the auxiliary feedwater pumps. The secondary flow aids in the reduction of RCS pressure. 80

SEABROOK STATION LOCA SAFETY ANALYSIS REPORT i When the RCS depressurizes to approximately 600 psia, the cold leg accumulators begin to inject borated water into the reactor coolant loops. However, for most small breaks the vessel mixture level starts to increase, covering the fuel with ECCS pumped injection before accumulator injection begins. 3.4 Core and System Performance e 3.4.1 Mathematical Model The requirements of an acceptable ECCS evaluation model are presented in Appendix K of 10 CFR 50. 3.4.2 small Break LOCA Evaluation Model 2 For small breaks (less than 1.0 f t ) the NOTRUMP digital computer code (References 16 and 17) is employed to calculate the transient depressurization of the RCS as well as to describe the mass and energy of the fluid flow through the break. The NOTRUMP computer code is a state-of-the-art one-dimensional general network code incorporating a number of advanced features. Among these are t calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes and regime-dependent drift flux calculations l with multiple-stacked fluid nodes and regime-dependent heat transfer correlations. The NOTRUMP small break LOCA ECCS evaluation model was developed to determine the RCS response to design basis small break LOCAs, and to address NRC concerns expressed in NUREG-0611 (Reference 18). i The RCS model is nodalized into volumes interconnected by flowpaths. The broken loop is modeled explicitly, while the intact loops are lumped into a single second loop. Transient behavior of I the system is determined from the governing conservation equations i of mass, energy, and momentum. The multinode capability of the program enables

explicit, detailed spatial representation of various system components which, among other capabilities, enables a proper calculation of the behavior of the loop seal during a t

l LOCA. The reactor core is represented as heated control volumes with associated phase separation models to permit transient mixture height calculations. Clad thermal analyses are performed with the LOCTA-IV code (Reference 19) using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture heights as boundary conditions (Figure 3-1). Figure 3-2 depicts the hot rod axial power shape used to perform the small break LOCA analysis. The shape was chosen because it represents a distribution with power concentrated in the upper regions of the core. Such a distribution is limiting for small break LOCA because it minimizes l l i l 81 i wr -ewmi-- .ye,y g ,,--,,..,-w-+, - - m m,.-9 y -e e,-- -gs-yg+ -yy, <y., g

SEABROOK STATION LOCA SAFETif ANALYSIS REPORT t coolant level swell, while maximizing vapor superheating and fuel rod heat generation at the uncovered elevations. The small break analysis assumes that the core continues to operate at full power until the control rods are completely inserted. For conservatism, it is assumed that the most reactive RCCA does not insert. i 3.4.3 Input Parameters and Initial Conditions Important input parameters and initial conditions used in the analysis are listed in Table 3-1. The safety injection' .i performance, as modeled in the small break analysis, is presented in Figure 3-3. Cases analyzed are given in Table 3-2. 3.5 Results NUREG-0737 Section II.K.3.31 (Reference 20) requires a plant-specific small break LOCA analysis using an Evaluation - Model l revised per Section II.K.3.30. In accordance with NRC generic l letter 83-35 (Reference 21), generic analyses using NOTRUMP (References 16 and 17) were performed and are presented in References 22 and 23. Those results demonstrate that in a comparison of cold leg, hot leg and pump suction leg break locations, the cold leg break of less than 10 inches in diameter is l limiting. Calculations were made for the 3, 4 inch and 6 inch break sizes. The results of these calculations are summarized in Tables 3-3 and 3-4. It was determined that, because of the low calculated PCT, rod l burst and blockage effects would not have a significant effect on the small break results for Seabrook Station. Therefore a fuel' assembly burnup sensitivity study was not required. Figures 3-4 through 3-6 present the results of'the cases analyzed for the small break LOCA. Figures A Reactor Coolant System Pressure (Calculated Core Pressure) l Figures B Core Mixture Level i i l Figures C Hot Spot Clad Temperature Figures D Core Exit Steam Mass Flow Rate ] l Figures E Hot Rod Heat Transfer Coefficient j Figures F Fluid Temperature at the Clad Hot Spot Figures G Break Mass Flow Rate Figures H Pumped Safety Injection Mass Flow Rate 82 r

I SEABROOK STATION LOCA SAFETY ANALYSIS REPORT l I l 1 i The limiting small break PCT is 10820F, which is less than the 'I acceptance criteria limit of 22000F. This PCT, from the 4-inch case, bounds all fuel types and features analyzed. Addition of the i BoF penalty for the increase in Tm associated.with RTD bypass elimination yields a total PCT of 10900F. This licensing basis PCT l remains below the 2200 F acceptance criterium of 10 CFR 50.46. The maximum local metal-water reaction is well below the embrittlement limit of 17 percent as required by.10 CFR-50.46. j i The total core metal-water reaction is less than 1.0 percent for ~ I all breaks analyzed, corresponding to less than 1.0 percent l hydrogen generation as required by 10 CFR 50.46. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. As a result, the core temperature will continue to drop and the ability to remove decay i heat generated in the fuel for an extended period of time will be j provided. I l I l l l I i l l l l 83 L _ ~..

SEABROOK STATION-. LOCA SAFETY ANALYSIS REPORT TABLE 3-1 INPUT PARAMETERS USED IN THE SMALL BREAK LOCA ECCS ANALYSIS l License Core Power (MWt) 3411* l Total Peaking Factor, Fo 2.5 i l Hot Channel Enthalpy Rise Factor, FAH 1.65 i i Power Shape See Figure 3 ' (+20% A.O.) Fuel Assembly Array 17x17'ZIRLO or Zircaloy-4, 0.374

Fuel Rods Accumulator Water Volume l

I ( f t'/ accumulator) 850 (Nominal) i Accumulator Gas Pressure, Minimum (psia) 600 Safety Injection Pumped Flow (SIPS and CCPs degraded 5%, RHR degraded 8.75% i CCP flow imbalance = 10 gpm) See Figure 3-3 Total System Flow Rate-(lbm/sec) 38788 l Vessel Inlet Temperature (CF) 557.64 Vessel Outlet Temperature (OF) 619.34 Reactor Coolant Pressure (psia) '2300.0 i i Steam Pressure (psia) 955.7 [ Steam Generator Tube Plugging. Level'(%)' 13** l Minimum Refueling Water Storage Tank-l Temperature ( F) 50.0 l Fuel Backfill Pressure (psig) 275 ~! (100 for IFBA) Low Pressurizer Pressure Setpoints (psia): Reactor Trip 1860 l Safety Injection Signal 1665 Rod Drop Time (sec) 4.4 Safety Injection Delay Time (sec) 30 Feedwater. Isolation Delay after Reactor. l Trip (sec) 2.0 Feedwater Isolation Valve Closure Time (sec) 0.0 'No percent is added to this power to account for calorimetric error. Analysis is performed at 13% SGTP, however, the large break analysis limits operation to 8% SGTP. 5% SGTP is allocated for steam generator tube-crush during a combined LOCA/ seismic event. I i l 84 l I-

1 SEABROOK STATION LOCA SAFETir ANALYSIS REPORT' 1 TABLE 3-2 SMALL BREAK LOCA - CASES ANALYZED i I CASE I - 3-Inch Break, 3411 MWt-Core Power,. F =2. 5, e FAH=1.65, P-BAR-HA=1.469. 3 I l CASE II - 4-Inch Break, 3411 MWt Core Power, Fo=2. 5, FAH=1.65, P-BAR-HA=1.469. This case was.found to be. limiting and bounds both ZIRLO ' and j Zircaloy-4 cladding. CASE III - 6-Inch Break, - 3411 MWt Core ^ Power, ' F =2.5, i a FAH=1.65, P-BAR-HA=1.469. l i ) All cases model 13% steam generator tube plugging (but'are limited by the l large break a-='cesis to operation with 8% SOTP) and 2% reduction in thermal design flow 4Ht,' gpm/ loop). j I I a r l l 1 1 I 85 1 2 i , ~,, ~, - y -n,- e e, .a r

b SEABROOK' STATION LCCA SAFETY ANALYSIS REPORT -] 1 1 TABLE 3-3 SMALL BREAK LOCA RESULTS - TIME SEQUENCE OF EVENTS i i ) Case I Case II Case III Seconds 3-inch 4-inch 6-inch i Start of LOCA 0.00 0.00 0.00 Reactor Trip Setpoint Exceeded 20.0 11.8 7.4 Safety Injection Setpoint Exceeded 28.9 20.2 13.5 i Pump Injection Begins 58.9 50.2 43.5 Start of Auxiliary Feedwater Delivery 95.0 86.8 82.4. Initial Loop Seal j Venting 420 283 .344 Loop Seal I..

overy N/A 321 166' l

Loop Seal Recovery N/A -345 172 j l Boil-off Core Uncovery 1210 687 220 I Accumulator Injection i Begins (1) 925 363 Peak Clad Temperature 1710 942 421 j Top of Core Recovered (2) 1459 436 l l Si Flow Rate Exceeds l Break Flow Rate 1775 1335 (3) i i l l (1) System pressure never drops below the accumulator cut-in pressure (600 psia). I (2) Although the core is not yet covered in this case, SI flow exceeds break flow - and the clad temperature transient is over. (3) Although SI flow has not yet matched break flow, the core is covered, the clad temperature transient has ended, and the total RCS mass is increasing. i i 86 l I ~.

I SEABROOK STATION LOCA SAFETY ANALYSIS REPORT 1 TABLE 3-4 SMALL BREAK LOCA RESULTS - FUEL CLADDING DATA l Case I Case II Case III 3-inch 4-inch 6-inch Peak Clad Temperature (oF) 1052 1090* 962 Peak Clad Temperature Location (ft) 11.25 11.00 11.00 l l Peak Clad Temperature Time (sec) 1710 942 421 Maximum Local Zr/H O 2 Reaction (%) 0.0492 0.0396 0.0335 j l Maximum Zr/H;O Reaction Location (ft) 11.25 11.25 11.00 Total Zr/H;O Reaction (%) <1.0 <1.0 <1.0 Rod Burst None None None l l Includes B.00F evaluation penalty for Increased Temperature Uncertainty (25 F) associated with RTD Bypass Elimination. ) l t l l l 87 i S v

N L l 0 0 T C R T U CORE PRESSURE. CORE A FLOW. MIXTURE LEVEL. AND FUEL ROD 1 P POWER HISTORY 3 l 0 < TIME < CORE COVERED l l 1 l ) i i ) ) 1 SMALL BREAK LOCA CODE O WESTINGHOUSE ELECTRIC CORPORATION INTERFACE Nuclear and Advanced Technology Drvsions FIGURE 31 88 l

l l l 14 i i 12 / i i ^ 10 t / \\ m6 i i b l E* / n / l e j O l x a 6 ( l 4 i I i i 2 0 2 4 6 8 10 12 Elevation (ft) i l SMALL BREAK LOCA HOT ROD WESTINGHOUSE ELECTRIC CORPORATION POWER SHAPE Nucisar and Advanced Technology Drvisions FIGURE 3-2 l I 89 s

100 \\ 80 'O 8 N E 60 2 5 5 E 40 E O 2 20 0 0 500 1000 1500 2000 2500 3000 Pressure (psio) PUMPED SAFETY INJECTION FLOW O WESTINGHOUSE ELECTRIC CORPORATION vs. RCS PRESSURE Nuclear and Advanced Technology Dusions FIGURE 3-3 90

2500 l l i 2000-l l l l l 2 1500-D CL v W l 5 D Du g 1000-l 500-N O O 500 1000 1500 2000 TIME (SEC) RCS PRESSURE - O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divisions (4 INCH BREAK) FIGURE 3-4A 91 i

l l l l f 35 1 i 30-I I j u ~ l l u>w 25-ue 2 -x TOP OF CORE y.__ 1 I 20 4 ( i 15 O 500 1000 1500 2000 TIME (SEC) 1 CORE MIXTURE LEVEL l I WEST 8NGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divtsions (4 INCH BREAK) FIGURE 3-48 92 l

l l l 1100. - ~' 1000. 900. L ~ 800. y E E 700. EW 00. f N I N f m 500. 400200. 400. 600. 800. 1000. 1200. 1400. 1600. I TlWE (S) i i PEAK CLAD TEMPERATURE 0 WESTINGHOUSE ELECTRIC CORPORATION l Nuclear and Advanced Technology Divsions (4 INCH BREAK) 1 FIGURE 3-4C 1 l 93 ? I

i 1250 1 1000-3 1 5 750-ta==<c: m o a' 500-tn 2 250-l W I O 0 500 1000 1500 2000 TIME (SEC) CORE EXIT STEAM MASS O WESTINGHOUSE ELECTRIC CORPORATION FLOW RATE Nuclear and Advanced Technology Divsions (4 INCH BREAK) FIGURE 3-4D 94 i

l l 105 l l l u h 10 4 ,x Y W i i t o l 103 Q I U i a u b k i E . k l 1, f' s 102 C' I 1 I 5 nm l 1 10200. 400. 600. 800. 1000. 1200. 1400. 1600. TIME (S) HOT ROD HEAT TRANSFER O WESTINGHOUSE ELECTRIC CORPORATION COEFFICIENT Nuclear and Advanced Technology Desions (4 INCH BREAK) FIGURE 3-4E 95

l t l I l l 850. ? h 800. l l 750. f i l - 700. l l L wa 1 i .o 650. i C l w l a i r 1 l w 600. i J -l 550. 3 N i l 500. l ? j b m 450200. 400. 600. 800. 1000. 1200. 1400. 1600. j TlWE (S) 1 l l 1 l CLAD HOT SPOT FLUID WESTINGHOUSE ELECTRIC CORPORATION TEMPERATURE Nuclear and Advanced Technology Omsions (4 INCH BREAK) FIGURE 3-4F l 96 ---w, m

l l 3000 i l l I - 2000-. tn N 1 O a V u -<c: wo a t_ sn I en< I 1 I ~ { 0 0 500 1000 1500 2000 TIME (SEC) BREAK MASS FLOW RATE WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Otvsions (4 INCH BREAK) FIGURE 3-4G I 97

oc. i 50-i ^ m i N 2 0 60-- C a c:: O ' 40-m m 2 20-i i 0 O 500 1000 1500 2000 j TIME (SEC) PUMPED SAFETY INJECTION MASS O WESTINGHOUSE ELECTRIC CORPORATION FLOW RATE Nuclear and Advanced Technology Drvisions (4 INCH BREAK) FIGURE 3-4H j l 98 1

2500 J i 3 2000-A 5 mE 1500-y trom m u' 7 1000-500 O 500 1000 1500 2000 2500 TIME (SEC) RCS PRESSURE i 0 WESTINGHOUSE ELECTRIC CORPORATION I Nuclear and Advanced Technology Drvsions (3 INCH BREAK) FIGURE 3-5A 99

35 30 ) u. v 1 -u i >u.25-TOP OF CORE c:= X 2 i 20-i l 15 O 500 1000 1500 2000 2500 TIME (SEC) l CORE MIXTURE LEVEL WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvisions (3 INCH BREAK) FIGURE 3-5B 100

i*: i 1100. j / 1000. N s 900. r C \\ l w 800. Ex dr i W { r 700. i 600. f i ~ 500.1000. 1200. 1400. 1600. 1800. 2000. 2200. 2400. 2600. I TlWE (S) r PEAK CLAD TEMPERATURE l WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Divisions (3 INCH BREAK) FIGURE 3-SC 4 4 101 I

i I i j 250 J I 200-2 l ~ 1 150-3 C i 4 f l 100-I ^ 50-i 0 0 500 1000 1500 2000 2500 TIME (SEC) CORE EXIT STEAM MASS O WESTINGHOUSE ELECTRIC CORPORATION FLOW RATE Nuclear and Advanced Technologv Drvisions (3 INCH BREAK) FIGURE 3-5D 102

a W 10' 1 1 i i i t i i l I 1 l Y S t i E 103 so to i I I I ) l l 4I L u u au Ui \\ E 102 i a ~ w c i 1 bd i l I ^ 3 = :_: 4.- ...i..i 1000. 1200. 1400. 1600. 1800.

2000, 2200.

2400. 2600. TIME (S) HOT ROD HEAT TRANSFER O WESTINGHOUSE ELECTRIC CORPORATION COEFFICIENT Nuclear and Advanced Technology Divisions (3 INCH BREAK) FIGURE 3-SE 103

i 800. 750. 700. 9 h 650. u l l l 5 h' w 600. r \\ 550. y 500. l l 450. t 1000. 1200. 1400. 1600. 1800. 2000. 2200. 2400. 2600. TIME (S) l CLAD HOT SPOT FLUID WESTINGHOUSE ELECTRIC CORPORATION TEMPERATURE Nuclear and Advanced Technology Duisions (3 INCH BREAK) FIGURE 3-5F l l l l 104

i 2000 i l ~ 1500-2 xso a l d< 1000 -- c: =o a L. W W<s 500-- I L 0 0 500 1000 1500 2000 2500 TIME (SEC) BREAK MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvsions (3 INCH BREAK) FIGURE 3-5G 105

i I I l j S0 l l 60-I f m m N 2c: t.u < 40-- cr E O u. ( ww 2 20-0 i 0 500 1000 1500 2000 2500 TIME (SEC) PUMPED SAFETY INJECTION MASS O WESTINGHOUSE ELECTRIC CORPORATION FLOW RATE Nuclear and Advanced Technology Drvisions (3 INCH BREAK) FIGURE 3-5H 106

i 25CO i i l l l 2000-i 2 15C0-m 1 i c. t v ueaw W i u Cc. 1000-i 500-1 i d i 0 O 200 400 600 800 TIME (SEC) RCS PRESSURE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvnions (6 IN0H BREAK) FIGURE 3-6A 107

l 35 l l 30-h f \\ f 5 25-- t h c i TOP OF CORE l l wxo E 20-2 1 15-- l i 10 0 200 400 600 800 TIME (SEC) CORE MIXTURE LEVEL O WESTNGHOUSE ELECTRIC CORPORATION Nuchiar and Advanced Technology Drvisions (6 INCH BREAK) j i FIGURE 34B l 1 108 i t

1000 900 800 C. W 3 700 E 600 N 500 i 400 100 200 300 400 -500 600 700 TlWE (S) l PEAK CLAD TEMPERATURE 6 WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Drvsions (6 INCH BREAK) l FIGURE 3-6C l 109 I

2000 1 1500-l m m N2 8 1000-v u no a' 500-m m< 2 l H}l 21-A r' F ' iI O-i -500 O 200 400 600 800 TIME (SEC) CORE EXIT STEAM MASS O WESTINGHOUSE ELECTRIC CORPORATION FLOW RATE l Nuclear and Advanced Technology Drvisions (6 INCH BREAK) FIGURE 3-6D 110 I

l l l l 10' m,_ i l Y Y $ 103 r'A (1 i \\ I a E' lin i s,.1 /w! m. U P1 M1 / a l I w t, l h m I U tn E 102 l 5 W s ~ H Al (r7 101100. 200. 300. 400. 500. 600. 700. TiWE (S) HOT ROD HEAT TRANSFER O WESTINGHOUSE ELECTRIC CORPORATION COEFFICIENT Nuclear and Advanced Technology Drvisions (6 INCH BREAK) FIGURE 3-6E 111

i i l 800 j i 750 700 650 b i 600 l u l E N oc 550 Eru \\ 500 l 450 400 350 100 200 300 400 500 600 700 l TlWE (S) i CLAD HOT SPOT FLUID WESTINGHOUSE ELECTRIC CORPORATION TEMPERA.TURE l Nuclear and Advanced Technology Duisions (6 INCH BREAK) l FIGURE 3-6F i 112

9l l 8000 l 1 6000-n W I 2 cm .a V i I y < 4000-l c: i l uo a t W W<s 2000-b iy&=,m 0 l 0 200 400 600 800 TIME (SEC) BREAK MASS FLOW RATE O WESTINGHOUSE ELECTRIC CORPORATION Nuclear and Advanced Technology Duisions (6 INCH BREAK) FIGURE 3-6H 113 l

200 ^ l 80 i i n m N 2 1 = 60-- u. 5 i =c 40__ m m< 2 4 20-- i i 0 1 0 200 400 600 800 i TIME (SEC) PUMPED SAFETY INJECTION MASS O WESTINGHOUSE ELECTRIC CORPORATION FLOW RATE Nuclear and Advanced Technology Divisions (6 INCH BREAK) FIGURE 3-6G 114

o I l SEABROOK STATION l LOCA SAFETY ANALYSIS REPORT j I l l l

4.0 CONCLUSION

The analyses presented in Sections 2 and 3 show that the Seabrook Station emergency core cooling system provides sufficient core flooding to meet the requirements of 10 CFR 50.46 in the event of a large break or small break loss-of-cooling accident.

5.0 REFERENCES

1. " Acceptance Criteria for Emergency Core Cooling System for Light Water Cooled Nuclear Power Reactors,

10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register 1974, Volume 39, Number 3.

2. U. S. Nuclear Regulatory Commission 1975, " Reactor Safety Study - An Assessment of Accident Risks in U. A. Commercial Nuclear Power Plants

WASH-1400, NUREG-75/014.

l l 3.

Bordelon, F.

M.;

Massie, H.

W.; and

Zordan, T.

A. l " Westinghouse ECCS Evaluation Model - Summary,* WCAP-8339, l July 1974. l 4.

Bordelon, F.

M. et al.,

SATAN-VI Program:

Comprehensive l

Space, Time Dependent Analysis of Loss-of-Coolant

WCAP-8302 (Proprietary) and WCAP-8306 (Non-Proprietary),

June 1974. 5.

Kelly, R. D. et al.,

" Calculation model for core Reflooding Af ter. Loss-of-Coolant Accident (WREFLOOD Code),

WCAP-8170 (Propr.etary) and WCAP-8171 (Non-Proprietary), June 1974.

i 6.

Young, M.

Y. et al, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-10266-P-A Rev. 2 (Proprietary) and WCAP-11524-A (Nonproprietary), March 1987. 7.

Rahe, E. P.

(Westinghouse), letter to J. R. Miller (USNRC), Letter No. NS-EPRS-2679, November 1982. 8.

Rahe, E.

P., " Westinghouse ECCS Evaluation Model, 1981 Version, " WCAP-9920-P-A (Proprietary Version), WCAP-9221-P-A (Non-Proprietary version), Revision 1, February 1982. 9. Bordelon, F. M., et al., " Westinghouse ECCS Evaluation Model - Supplementary Information, " WCAP-8471- (Proprietary) and WCAP-8472 (Non-proprietary), April 1975. 10. Special Report NS-NRC-85-3025(NP), "BART-WREFLOOD Input Revision." 115

u SEABROOK STATION LOCA SAFETY ANALYSIS REPORT 11.

Bordelon, F.

M., and Murphy, E. T., " Containment Pressure Analysis Code (COCO), " WCAP-8327 (Proprietary) and WCAP-8326 (Non-Proprietary), June 1974. 12.

Collier, G.,

et al, 'BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients" WCAP-9561, Janur y 1980. 13. " Westinghouse ECCS - Evaluation Model Sensitivity Studies," WCAP-8341 (Proprietary) and WCAP-8342 (Non-proprietary), July 1974. Plant Sensitivity 14. Salvatori, R., " Westinghouse ECCS Studies,* WCAP-8340 (Proprietary) and WCAP-8356 (Non-proprietary), July 1974. 15.

Johnson, W.

J.;

Massie, H.

W.; and Thompson, C. M. " Westinghouse ECCS Four Loop Plant (17x17) Sensitivity Studies," WCAP-8565-P-A (Proprietary) and WCAP-8566-A (Non-Proprietary), July 1975. 16.

Meyer, P.

E., "NOTRUMP - A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A (Proprietary) and WCAP-10080-A (Nonproprietary), August 1985. 17.

Lee, N.

et al., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, " WCAP-1005'. P-A (Proprietary) and 10081-A (Nonproprietary), August 1985. 18. " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants," NUREG-0611, January 1980. 19.

Bordelon, F. M. et al.,

'LOCTA-IV Program: Loss-of-Coolant Transient Analysis," WCAP-8305, June

1974, WCAP-8301 (Proprietary), June 1974.

20. " Clarification of TMI Action Plan Requirements", NUREG-0737, November 1980. 21. NRC Generic Letter 83-35 from D. G. Eisenhut, " Clarification of 'IMI Action Plan Item II.K.3.31", November 2, 1983. 22. 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-A (Non-Proprietary),

October 1986. 23. " Westinghouse ECCS Evaluation Model Sensitivity Studies," WCAP-8341 (Proprietary), and WCAP-8342 (Non-Proprietary), July 1974. 116

r l i i i l l l l I i 1 Proposed Core Operating Limits Report ( ) I

I l l l 1.0 CORE OPERATING LIMITS REPORT This Core Operating Limits Report for Seabrook Station Unit 1, Cycle 3 ] has been prepared in accordance with the requirements of Technical Specification 6.3.1.6. ) i ] The Technical Specifications affected by this report are: l I 1) 2.2.1 Limiting Safety System Settings I 2) 3.1.1.1 Shutdown Margin limit for MODES 1, 2, 3, 4 1 3) 3.1.1.2 Shutdown Margin limit for MODE 5 4) 3.1.1.3 Moderator Temperature Coefficient i 5) 3.1.3.5 Shutdown Rod Insertion Limit i 6) 3.1.3.6 Control Rod Insertion Limits i 7) 3.2.1 Axial Flux Difference l B) 3.2.2 Heat Flux Hot Channel Factor 9) 3.2.3 Nuclear Enthalpy Rise Hot Channel Factor 2.0 OPERATING LIMITS i 1 l The cycle-specific parameter limits for the specifications listed in j section 1.0 are presented in the following subsections. These limits l have been developed using the NRC-approved methodologies specified in Technical Specification 6.8.1.6. ] I ( 2.1 Limiting Safety System Settings: (Specification 2.2.1) i 2.1.1 Cycle Dependent Overtemperature AT Trip Setpoint Parameters and Function Modifier: i 1.145 l 2.1.1.1 K = 1 2.1.1.2 K = 0.020/*F 2 l l 0.001/psig 2.1.1.3 K = 3 l 2.1.1.4 Channel Total Allowance (TA) = N.A. 2.1.1.5 Channel Z = N.A. l

l I l l 2.1.1.6 Channel Sensor Error (S) = N.A. 2.1.1.7 Allowable Value - The channel's maximum Trip Setpoint shall-not exceed its computed Trip Setpoint by more than 2.2% of AT span. 2.1.1.8 F (AI) is a function of the indicated difference between top 3 l and bottom detectors of the power-range neutron ion chambers I l with gains to be selected based on measured instrument response during plant startup. test. F (AI) is specified in 3 Figure 1.1. l 2.1.2 Cycle Dependent Overpower AT Trip.Setpoint Parameters and l Function Modifier: 2.1.2.1 K. = 1. 0 8 0 (1. 085 with modified Fa ( AI) NCH PC card) i 2.1.2.2 K, = 0.020/*F for increasing average temperature and K = 0.0 5 for decreasing average temperature. i -0.00196/*F for T > T", and K. = 0. 0 for T s T", where: 2.1.2.3 K = 6 T = Average temperature (*F), and I Indicated T,, at RATED THERMAL POWER (Calibration T" = temperature for AT instrumentation, s 588.5"F). 2.1.2.4 Channel Total Allowance (TA) = N.A. 2.1.2.5 Channel Z = N.A. 2.1.2.6 Channel Sensor Error (S) = N.A. 2.1.2.7 Allowable Value - The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.1%* of AT span. (2.0% with modified F ( AI) NCH PC card) 2 2.1.2.8 F (AI) is a function of the indicated difference between top 2 and bottom detectors of the power-range neutron ion chambers with gains to be selected based on measured instrument

response during plant startup tests. F (AI) is specified in ] 2 Figure 1.2. 2.2 Shutdown Margin Limit For MODES 1,. 2, 3, AND 4: (Specification 3.1.1.1) l i l A) The Shutdown Margin shall be greater than or equal to 1.3% AK/K in MODES 1, 2, and 3. I B) The Shutdown Margin shall be greater than or equal to 1.8% AK/K in MODE 4. 1 2.3 Shutdown Margin Limit For MODE 5: (Specification 3.1.1.2) The Shutdown Margin shall be greater than or equal to 1.8% AK/K. 2.4 Moderator Temperature Coefficient: (Specification 3.1.1.3) j 2.4.1 The Moderator Temperature Coefficient (MTC) shall be less j positive than +0.5 x 10** AK/K/ F for all the rods withdrawn, Beginning of Cycle Life (BOL), for power levels up to 70% RATED THERMAL POWER with a linear ramp to O AK/K/*F at 100% RATED THERMAL POWER. l 2.4.2 MTC shall be less negative than -4.2 x 10~* AK/K/ F for End of Cycle Life (EOL), ARO, Rated Thermal Power conditions. 1 2.4.3 The 300 ppm ARO, Rated Thermal Power MTC shall be lesc negative than -3.3 x 10-* AK/K/ F (300 ppm Surveillance Limit). l 2.5 Shutdown Rod Insertion Limit: (Specification 3.1.3;5) 1 l 2.5.1 The shutdown rods shall be fully withdrawn. The fully l withdrawn position is defined as the interval within 225 steps withdrawn to the mechanical fully withdrawn position inclusive. l l l

5 2.6 Control Rod Insertion Limits: (Specification 3.1.3.6) t 2.6.1 The control rod banks shall be limited in physical insertion as specified in Figure 1. 2.7 Axial Flux Difference: (Specification 3.2.1) i 2.7.1 For operation with the Fixed Incore Detector Alarm OPERABLE, the indicated AFD must be within the Acceptable Operation Limits specified in Figure 2.1. i 2.7.2 For operation with,the Fixed Incore Detector Alarm inoperable, the indicated AFD must be.within the Acceptable Operation Limits specified in Figure 2.2. 2.8 Heat Flux Hot Channel Factor: (Specification 3.2.2) 2.8.1 F *7" 2.50 = o 2.8.2 For operation with the Fixed Incore Detector Alarm OPERABLE, K (~' is specified in Figure 3. p 2.8.3 For operation with the Fixed Incore Detector Alarm inoperable, K(Z) is specified in Figure 4. 2.9 Nuclear Enthalpy Rise Hot Channel Factor: (Specification 3.2.3 ) The limits on Ffu are specified in Figure 5. The limits apply to F", measured using either the fixed or movable incore detectors since a bounding measurement error has been allowed for in determination of the design DNBR limit value.

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EPTABLE

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r. - - - - 1, - 'c - 1. - 'c - i. - ' - 4. - ' - 4. - 'r - 4, - 'r - -

10 - r c e ...{..p..{.p..{..p_q..[.p-}..p.}_.p.{_.p.j..p_j__p J.._p_j.p.. O ^.................. -60 0 -40 -30 0 0 0 10 20 30 40. 50 60 i Axial Flux Difference ( %Al) J i 4 i i 1 - 1 Axial Flux Difference Limits as a Function of Rated Thermal Power: SEABROOK STATION CYCLE 4 for Operation With CORE OPERATING LIMITS REPORT Fixed incore Detector System Alarm inoperable FIGURE 2.2 1 ,p,--

j r I i ~i t 1.2 _ _4.. _ p _ _;_ _ ;...p t. _ _p ; _ _p _ ; - _l_ _ ; _ _p _ ; - _p _4. _ ;_ _4.._p _4...;. _4._.p.. l .r 1.0 _.i..s.._i__ __u. __u. __u_...u N... s. s__,__ _.r_,.__ .r .r .r - - i.- - F. - 4.- - F. - 4.- - t. - -l- - t. - -l.- - t. - '.- - l. - -l.- - l. - -l.- - i. - -!.-- i. --F. - 0.s - t n u,a --.--.-.-.---+-.--+.---+----+-.+.-.--+.-.4.-.--<.-.-.--*-- o o i $ o.s _. _4.. _ p _4... p _4._. p..;. ; _.;_ _ ; _ _;.. ; _ _;.. ; _.p _ ; _ _p. 4..p. 4.. p.4.. _ p.. o . r i .i .i.i z __t__.. ,.__.__t.__..t.____2.__. __u_ __u __i.. 6 n_a. 1

c 1

0.4 -- -,.. - - -,.--.-.--r-,.-- -,.-- -,. .- i.... ._.[ _p..{._p..p_p__;__[__l_.} 4__. _2_ _ _ _2_ _p _ j._p _ q _ p _.j._p _ _ r ...... i Basis: FOT - 2.50 t ...... i-......... 0.2 ---.-.*-*.--*-.--*.-.--+.-.--+.---+.--+.-.--+n-.--*.-.-*.----*.-.*-- i ...j._p_ 4._.p_.j..[__p -[__p _)__p _;__l__j _p _j _l__j._l__4.__p..j__p.. .i l i...................... i i 1 i i e' a i i i a' i 0.0 0 1 2 3 4 5 6 7 8 9 10 11 12 Core Height (Feet) i K(Z)- Normalized F,(Z) SEABROOK STATION CYCLE 4 As A Function of Core Height CORE OPERATING LIMITS REPORT l FIGURE 3 4 1 1 i a

k' l I 1 1.2 i ................ i....... _ _4,. ;. _ q_.;. _ ;- _ ;.. ;.. ; - _p _ ; _ _;_ _ ; _ p _ ; _ _p. 4.p _4...;...;.p.;--; _ _ i -J.-.'-.'-8.-8.-8.-.'-'.-.'-'.,-2-'.-'-2.-.'-3.-.'-d.-.5-d.-.'J.-.'-- 1.0 - i ., _,...r.,.. _. _,. _ _ -.r.,. I i --r__ .r. .r o - - i. -.F - 4,-- F -i.-- F -,l-- t - '- - t --l- - t. -,l-- f, - -l- - i. -.l-- i - -:- - i. -.F - i. --F - - 0.8 - ~, s - -., - -. -. - -.. - -. -. - - +. - -. - - + -. - - +. - - - -. - -. - - .-.--4.-.-- u. e i g E0.s--- '.. - i.-- l - -l- - l-- -l-- F. --l- - i, -:- - l. --l- - +. --l.- - 4. - -l - 4. - -h - 4. - -F. - 4. --F. - - i = o z . _ 2. _ _.i. s. _ _. a. _ i. a, _ _ i,. _i_ _ i. _ _ _ _ i. _ _i.,.__i__,__i._,..._c.....u_,. __u_. re x i. 0.4 - i i _ _ q. _ ;. _ q _ _ p. s,. p - _;.. t. _ _;i__t,. ..i__ 2_.._.2._.__3,..-.;__q._p_4.._p - j j Basis: FOT = 2.50 i .i .i i....... i 4 0.2 --i-.*-.--*-i.--*- --+..--+.-. +----+-.--+.----4.----*.-*-4.----- i i ...- i... i..... .i _. j. _ p _.j. _ j...;.. j...;. _ ). _l.. }. _p _ ;. __j_.;_.j__;..j._l..j__p..[._}... 0.0 s' . '......... i....... .. i s'. ' e' i i a' i i a' i O 1 2 3 4 5 6 7 8 9 '10 11 12 Core Hei:ht (Feet) t K(Z)- Normalized F,(Z) SEABROOK STATION CYCLE 4 As A Function of Core Height - CORE OPERATING LIMITS REPORT With Fixed incore Detector Alarrn inoperable FIGURE 4 I r I l

l l 1.55 1 t I L 8 e t I s a i e e s .. 3 _ .tp.-r-7-- i-- r-,a- - - - T - - -'i - - r - T -,t- - r - -r - r - - - O 1oo ,..j...(..L..J...(..[..)...( 1...)...(..j L.J_..L__L.._. f o so 1 5*A --r--r--,.---r-r- r-5- r-3 eo r--- o 70 1.53 - . _ t.. _ ;. _. t.... J. _.. t. 2...... t.

2... _L. _ t,. _.

2.. _.a.. e i a 7 60 s -*--+-"---'--*--'---i--d------+--- .i._'-_.e,.._,i...i_..i...,a...,i_..i,..,i..,_..'....

50 1.52 -

i ...._ ___,i... e.._ - - r - -! - _..- _ t - - ' _ _ '.... i - _ '....'. - _ i - - '.. - - i, - - - 1.51 --2 = so r--i- .a i i e i i e A 2o - - - r - -,i .. i a i --v-, --m,---r,-,---c---r. . to 1.5 - --r-- -r- ..._4_.. ..,....__r ........r... Y o e i e i a e -d,--, 1.49 - a-i ...u r--- 1.48 -_.._t... z.. 1.. e i 1 I t I 1 ,. 3 ~~- --~~- g-~. a 1.47 -'- ~ i wm 3 7.,. _ 7 3 r- .--*s**- C

4. - -

y 3-- T3 146 - --J,--',---', -. i =z ..v.. ~....... u e r--- -- 1.45 -- --c-r-,.-- c--r--,i .--. - --i r-2. .i_.. l 2.. _i.. _ t _. a...t.. t.. ;... i... a. .t. .. s...a..e 4... s 4 e 3 s a a 1,44 --....._-_.w__ .._n..s..-...w.. x. s_ ~- t a e e a -t 4 Q . 3...s.. r -,a- - c - r - -,a -- -- T- -r 1 i-r- -.-i

4. -

LL 1 43 ---4.--'-'i-J - - 'a - L - - i - - 's - -. -. ........._..i.. ......_._...e ...e...i.._w _.i i G 1 a 1.42 ----n- ,l---r-,$-- r-r-,4-- r--r-I I 4 4 I ) ---r- --r _ _ _ l...L.. L.. J...L _ L.. _ J. _L.

1.. J.. _ L. _.

a. J. L.. i i i i i i i e i ---*---"--'e~~~i ~ - - - ---*--~~--*---~~-',------"-'s 1.41 - ---*i..i...ir -,.- i- - r - - i - - r- - T - -,e i e i r-1--9--- -1 --r-L.. J, ..L. _1, _ _ J.. _ L. _1.. J. _. L. _ J,.. J. _. L.. 1.4 e s i s ._.2 __..._'s....a.. 4...s..4.._a_._4 .s... 4._.a...i..a. a ...,i.,i...,i. _,9.,8. _ _ 9.,1.. 1r..,4__,0_-_4,..,3..,i...B._,l...,. 1 e 1,39 r .f...f...I....I..4..._g..i.._I....t__.I...I... ...f. f._f.._f_... e e e e e a a a i e i i r i e a d------*- d-- '--'--'-d--'---'--- --',i_._'i,...'i..,e__',.__i,..,i...i,,_.,i..e,.._'..,i___i..i,....i,...,i...,i... 1.38 - i ie 3, 9 I D f I I f 1 f f f f f 1.37 -- - - i -.- - r - ,9--,-- r-f. f.- - 4r-t--.-- r-1--.-- r--- 1 .._4..sw .L..J._.u..L..s_..L._A...a...L .a..a___L_.a._.u..e_.. --,I-- ---r--*---,4 I l I I 4 8 8 8 8 i 3 6 8 8 ---r,--- h-1.36 - ---*-.,---.-.-----r-.---r--e--..,._..._r-,--.--r----.--T-,.--r-t-- r.- r--- i 1.35 4 i 4 a i e i 0 2 4 6 8 10 12 14 16 18 4 Cycle Average Burnup (GWD/MTU) e 1 i W Y r-T M. l-All-Rods-Out f. SEABROOK STATION CYCLE 4 Nuclear Enthalpy Rise Hot Channel Factor CORE OPERATING LIMITS REPORT Versus Power Level FIGURE 5 i I i i i

10CFR50.46 LOCA Model Assessments on the PCT Margin Utilization j l

1 l i i l i f 10CFR50.46 LOCA MODEL ASSESSMENTS ON Tile PCT MARGIN UTILIZATION i The new small break Loss of Coolant Accident (LOCA) analyses for Seabrook Station utilize the I NOTRUMP Emergency Core Cooling System (ECCS) analysis model ( see proposed Technical Specification 6.8.L6.b2). Westinghouse Electric Corporation (Westinghouse) has identified the following non-conservatisms/conservatisms associated with the NOTRUMP model. ) l 1. The Seabrook Station LOCA analysis assumes no Safety injection (SI) into the broken loop. Sensitivity calculations indicate that a Peak Clad Temperature (PCT) penalty of 150 F is typical to address this affect. To ofTset this affect, a more realistic condensation model based on COSI i test results has been evaluated by Westinghouse. The COSI test facility is a 1/100 scale representation of the cold leg SI injection pods in a Westinghouse designed PWR. This model demonstrates that analysis performed with spillage to the broken loop using the original condensation model is conservative compared to the improved condensation model with injection j into the broken loop. i 2. The Seabrook Station LOCA analysis was performed with an error in the NOTRUMP drift flux floe regime map. The net impact of this error is estimated to decrease the calculated PCT in the i range of 13 F to 55 F. Thus, the reference small break LOCA PCTs will be modified slightly to account for the above items. l The modified PCTs will be reported to the NRC pursuant to the requirements of 10CFR50.46. l t l i ,. _}}

ML20058C358

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ML20058C358

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