ML20011B244
| ML20011B244 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 12/02/1981 |
| From: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Charemagne Grimes Office of Nuclear Reactor Regulation |
| References | |
| LD-81-088, LD-81-88, NUDOCS 8112080187 | |
| Download: ML20011B244 (100) | |
Text
C-E Power Systems Tel 203/688-1911 Cornbustion Engineenne, Inc. Telet 99297 1000 Prospe:t Hill Roed Wincsor. Connecticut 06095 POWER H SYSTEMS Docket No: STN-50-470F December 2, 1981 LD-81-088 31dQK Q f:~, *\ '
d, Mr. Christopher Grimes, Project Manager Standardization & Special Projects Branch y27'e[Q(/
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d Office of Nuclear Reactor Regulation -h, j,6 U. S. Nuclear Regulatory Commission x,, g4 Washington. O. C. 20555 1) g Subj ect: Appendix C to CESSAR-F D
Dear Mr. Grimes:
Enclosed please find a draf t input for CESSAR-F to describe reactor operation at a reduced power level with operation of two reactor coolant pumps diamet-rically opposed. The document will be submitted as Appendix C to CESSAR-F and is intended to provide all information necessary to justify operation with two pumps. The enclosed package includes, for your convenience, the ;
information previously transmitted to you in my letter of October 2,1981, which provided the initial submittal of information for two pump operation.
If you have any questions or require additional information, please contact me or Mr. G. A. Davis of my staff at (203)688-1911, Extension 2803.
Very truly yours, COMBUSTION ENGINEERING, INC.
& / ,A A. E. Scherer Director Nuclear Licensing AES:ctk o3
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APPENDIX C _
REACTOR OPERATION WITH TWO REACTOR COOLANT PUMPS .
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- Wifollowing W chapter-num6 s 6r,s'jif'CESSAP-FF7 Whtee('apfropriate, the sections refer to the ma p body of CESSAR y F as b'eing app 11gble'to.scart loop operation.
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TABLE OF CONTENTS APPYRDIX C Chapter Title Page Number C.1 TNTRUDUCTION AND GENERAL DESCRIPTION OF PLANT C-1 C.2 PLANT SITE CHARACTERISTICS C-3 C.3 DESIGN OFrSTRUCTURES, COMPONENTS AND C-3 SYSTEMS 3.1 GENERAL DESIGN CRITERIA C-3 3.2 CLASSIFICATION C-3 3.3 WIND AND TORNADO LOADINGS C-3 3.4 WATER LEVEL (FLOOD) DESIGN C-3 .
3.5 MISSILE PROTECTION C-3 3.6 PROTECT 13N AGAINST DYNAMIC EFFECTS C-3 ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING 3.7 SEISMIC DESIGN C-4 3.8 DESIGN OF CATEGORY ONE STRUCTURES C-4 3.9 MECHANICAL SYSTEMS AND COMPONENTS C-4 3.10 SEISMIC DESIGN OF CATEGORY ONE C-4 INSTRUMENTATION AND ELECTRICAL EQUIPMENT 3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND C-4 ELECTRICAL EQUIPMENT C.4 REACTOR C-4 4.1 REACTOR C-4 4.2 FUEL DESIGN C-4 4.3 NUCLEAR DESIGN C-5 4.4 THERMAL AND HYDRAULIC DESIGN C-7 4.5 REACTOR INTERNALS MATERIALS C - 11 4.6 FUNCTIONAL DESIGN OF REACTIVITY C - 11 CONTROL SYSTEM C.5 REACTOR COOLANT SYSTEM AND CONNECTED C .11 SYSTEMS C.6 ENGINEEREO SAFETY FEATURES C - 11 C.7 INSTRUMENTATION AND CONTROLS C - 18 C.8 ELECTRIC POWER C - 18 C.9 AUXILIARY SYSTEMS C - 18 C.10 STEAM POWER CONVERSION SYSTEMS C - 18 C.11 RADIOACTIVE WASTE MANAGEMENT C - 18 C.12 RADIATION PROTECTION ,
C - 18 C.13 CONDUCT OF OPERATIONS C - 18 C.14 INITIAL TEST PROGRAM C - 18 C.15 ACCIDENT ANALYSIS C - 19 15.0 ORGANIZATION OF METHODOLOGY C - 19 15.1 INCREASE IN HEAT REMOVAL BY THE C - 30 SECONDARY SYSTEM 15.2 DECREASED HEAT REMOVED BY THE C - 35 SECONDARY SYSTEM 15.3 DECREASED REACTOR COOLANT FLOW C al 15.4 REACTIVITY AND POWER DISTRIBUTION C - 48 ANOMALIES 15.5 INCREASE IN REACTOR C0OLANT SYSTEM C - 57 INVENT 0RY 15.6 DECREASE IN REACTOR C0OLANT INVENTORY C - 59 15.7 RADI0 ACTIVE MATERIAL RELEASE FROM A C-tS i SUBSYSTEM OR COMPONENT C.16 TECHNICAL SPECIFICATIONS C - 66 C.17 QUALITY ASSURANCE C - 66 i . . _ _ _ _ _ _ _
List of Tables C.4.4-1 " Thermal and Hydraulic Parameters" C.4.4-4 " Reactor Vessel Best Estimate Pressure Losses and Coolant Temperature Two Loop Opposed RCP Operation" C.4.4-7 "RCS Design Minimum Flow Two Loop Opposed RCP Operation" C . 6.3. 3. 2- 2 "C-E Systea 80 Standard Plant Design Two Loop General System Parameters and Initial Conditions Large Break ECCS Performance" C.6.3.3.2-5 "C-E System 80 Standard Plant Design Two Loop Operation Variables Plotted as a Function of Time for Large Break Analysis" C.15.0-2 " Chapter 15 Subsection Designation" C.15.0-4 " Reactor Protection System Trips Used in Safety Analysis" C.15.0-5 " Initial Conditions" C.15.0-6 " Single Failures" (two sheets)
C.15.3.1-1 " Sequence of Events for Total Loss of Reactor Coolant Flow" C.15.3.1-2 " Assumed Initial Conditions for Total Loss of Reactor Coolant Flow" C.15.3.3-1 " Sequence of Events for the Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting from Turbine Trip" (2sneets)
C.15.3.3-2 " Assumed Initial Conditions for the Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting from Turbine Tri p" C.15.3.3-3 "Parambters Used in Evaluating the Radiological Consequences of a Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting from Turbine Trip" (3 sheets) ,
C.15.3.3-4 " Secondary System Mass Release to the Atmosphere for the Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting from Turbine Trip Event" 11
r 1
List of Tables
- (continued)
C.15.4.1-1 " Sequence of Events for the Sequential CEA Withdrawal from Low Power" C.15.4.1-2 " Assumptions and Initial Conditions for the Low Power CEA Withdrawal Analysis" C .15. 4. 2- 1 " Sequence of Events for the Sequential CEA Withdrawal Event" C.15.4.2-2 " Assumptions and Initial Conditions for the Sequential CEA Withdrawal" J C.15.4.8-1 " Initial Reactor States Considered for the CEA Ejection Event"
.i C.15.4.8-2 " Assumptions Used for the CEA Ejection Analysis - Maximum Power Beginning of Cycle Initial Conditions" C.15.4.8-3 "Sunmary of Results for the CEA Ejection Event" C.15.6.2-1 " Assumed Input Parameters and Initial Conditions for the Double-i Ended Break of the Letdown Line Outside Containment Upstream of the Letdown Line Control Valve" C.15.6.3-1 " Assumptions and Initial Conditions for tne Steam Generator Tube Rupture" C.15.6.3-2 " Assumptions and Initial Conditions for the Steam Generator Tube i Rupture with a Loss of Offsite Power" l
l i
4 iii
List of Figures C.4.3-1 " Maximum Increase in Relative Assembly Power" C.4.4-1 " Core Wide Planar Power Distribution for Sample DNB Analysis" C.4.4-2 " Rod Radial Power Factors in Hot Quadrant for Sample DNB Analysis" C,4.4-3 " Average Void Fractt:ns and Qualities at the Exit of Different Core Regions" C.4.4-4 " Axial Distribution of Void Fraction and Quality in Subchannel Adjacent to the Rod with Minimum DNBR" C.4.4-5 " Sensitivity of Minimum DNBR to Small Changes in Reactor Coolant Conditions" ,
C.6.3.3.2-A "Two Loop Operation: Reactor Power" C . 6. 3. 3. 2-B "Two Loop Operation: Peak Clad Temperature" C.6.3.3.2-C "Two Loop Operation: Peak Local Clad Oxidation" C.6.3.3.2-D "Two Loop Operation: Hot Spot Gap Conductance" C.6.3.3.2-E "Two Loop Operation: Clad, Centerline, and Average Fuel Temperature for Hottest Node" C.6.5.3.2-F "Two Loop Operation: Hot Spot 1%et Transfer Coefficient" C.6.3.3.2-G "Two Loop Operation: Hot Rod Internal Gas Pressure" C.15.3.1-1 " Total Loss of Reactor Coolant Flow Core Power vs Time" C.15.3.1-2 " Total Loss of Reactor Coolant Flow Core Average Heat Flux vs
, Time" C.15.3.1-3 " Total Loss of Reactor Coolant Flow RCS Pressure vs Time" r
l C.15.3.1-4 " Total Loss of Reactor Coolant Flow Core Average Coolant Temperatures vs Time" l
! C.15.3.1-5 " Total Loss of Primary Coolant Flow Reactivity vs Time" C.15.3.1-6 " Total Loss of Reactor Coolant Flow Fraction vs Time" C.15.3.1-7 " Total Loss of Reactor Coolant Flow Right Hand and Left Hand Steam Cenerator Pressures vs Time" iv
List of Figures C.15.3.1-8 " Total Loss of Reactor Coolant Flow CE-1 Minimum D.NBR vs Time" C.15.3.3-1 " Single Reactor Coolant Pump Rotor Seizure with Loss af Offsite Power Resulting From Turbine Trip: Core Power Vs Time" C.15.3.3-2 " Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting From Turbine Trip: Core Average Heat Flux Vs Time" C.15.3.3-3 " Single Reactor Coolant Pump actor Seizure with Loss of Offsite Power Resulting From Turbine Trip: RCS Pressure Vs Time" C.15.3.3-4 " Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting From lurbine Trip: Core Average coolant Temperatures vs Time" C.15.3.3-5 " Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting From Turbine Trip: Reactivity vs Time" C.15.3.3-6 " Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting From Turbine Trip: Core Flow Fraction vs Time" C.15.3.3-7 " Single Reactor Coolant Pump Rotor Seizure witn Loss of Offsite Power Resulting From Turbine Trip: Steam Generator Pressure vs Time" C.15.3.3-8 " Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting From Turbine Trip: CE-1 Minimum DNBR vs Time" C.15.3.3-9 " Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting From Turbine Trip: Steam Generators Water Mass vs Time" C.15.4.1-1 " Sequential CEA Witndrawal at Low Power Core Power vs Time" C.15.4.1-2 " Sequential CEA Withdrawal at low Power Core Average Heat Flux vs Time" .
l C.15.4.1-3 " Sequential CEA Withdrawal at low Power Reactor Coolant System l Pressure vs Time" C.15.4.1-4 " Sequential CEA Withdrawal at low Power Minimum DNBR vs Time" C.15.4.1-5 " Sequential CEA Hitharawal at Low Power Core Average Coolant
! Te.nperatures vs Time" l
l C .15. 4.1- 6 " Sequential CEA Witndrawal at Low Power Steam Generator Pressure l vs Time" l
C.15.4.1-7 " Sequential CEA Witndrawal at Low Power Linear Heat Generation Rate vs Time" l v i
l
List of Figures C.15.4.2-1 " Sequential CEA Withdrawal at Power-Core Power vs Time" C.15.4.2-2 " Sequential CEA Withdrawal at Power-Core Average Heat Flux vs Time" C.15.4.2-3 " Sequential CEA Withdrawal at Power-Reactor Coolant System Pressure vs Time" C.15.4.2-4 " Sequential CEA Withdrawal at Power-Minimum DNBR vs Time" C.15.4.2-5 '" Sequential CEA Withdrawal at Power-Core Average Coolant Temperatures vs Time" C .15.4. 2- 6 " Sequential CEA Witndrawal at- Power-Steam Generator Pressure vs Time" C.15.4.2-7 " Sequential CEA Withdrawal at Power-Peak Linear Heat Generation Rate" C.15.4.8-1 "CEA Ejection: Core Power vs Time" C.15.4.8-2 "CEA Ejection: Peak Core Power Density vs Time" C.15.4.8-3 "CEA Ejection: Core Average-Heat Flux vs Time" C.15.4.8-4 "CEA Ejection: Peak Hot Channel Heat Flux vs Time" j C.15.4.8-5 "CEA Ejection: Hot and Average Channel Fuel and Clad Temperature vs Time" C.15.4.8-6 "CEA Ejection: Reactivity vs Time" I
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C.1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT C.
1.1 INTRODUCTION
This appendix addresses those aspects of safety importance for limited power operation of the C-E System 80 NSSS in a mode for which only two of the four reactor coolant pumps are operational in diametrically opposed cold legs, i.e., part loop operation.
The condition under.which part loop operation is shown to be acceptable is limited to the specific case defined herein where
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part loop operation may be elected by the applicant as a means .to minimize extended interruptions in power production when one reactor coolant pump is inoperable. The specific application would permit operation for a short period of time while a coolant pump is being readied for repair (i.e., less tnan three weeks per cycle).
Except where noted, all sections of CESSAR-F remain applicable for part loop operation.
C.1.2 GENERAL PLANT DESCRIPTION The principal site characteristics, System 80 scope and NSSS component size and arrangements are not different for operation in a part loop mode. The dasign criteria which have been established particularly for the design of the reactor core, internals, primary coolant system, and protection and monitoring systems, remain applicable to part loop operation (Sections 1.2.3-1.2.11 of CESSAR-F.
Although most of the design # ta and operating conditions corresponding to four-pump, 'ull power operation are unchanged in the part loop mode, there ar e some basic dif ferences in the procedures and parameters upon which the safety analyses for part loop operation are based. Thesa are as follows:
- a. Procedure The general procedure envisioned is the following: The plant is presumed to have te:n operating at or near full power. It shuts down to act standby as specific preparations are made for part loop operation. A set of two diametrically opposed pumps are chosen to be turned off and adjustments are made in setpoints. The plant returns to criticality and remains at less iban twenty percent power until xenon is stabilized. Pcwer is then gradually raised to the maximum allowable. Calibration of the ex-core nuclear instrumentation occurs at twenty percent power and at specified intervals thereafter as the power level is raised to a maximum of 48%.
C-1
- b. Operating Conditions The operating parameters and safety features associated with part loop operation are presented below. Additional assumptions and parameters will be noted in the appropriate sections.
Parumeter Value Power Level 48% of Thermal Core Flow Rate 50%
Axial Shape Index 10.2 Reactor Power Cutback Feature Disabled Primary Coolant Temperature-T(inlet) 564 1 5'F Primary Coolant Temperature-T(outlet) 624 1 10*F Operating Pressure 2250 1 50 psi Maximum Peak Linear Heat Rate 7 kw/ft
- c. Safety Systems Features and Setpoints The basic safety features and setpoints which apply to part loop operation are as follows:
CFA Position: Part Length CEAs will be fully withdrawn and the Power Dependent Insertion Limits (PDILs) will allow fifteen percent insert; ion of the lead regulating bank at full part loop power with more insertion permitted at lower power levels.
CPC Low Flow Trip: A large CPC DNBR and linear heat rate penalty factor will be applied when either operating pump reaches 90% speed. This ensures a trip at 90% pump speed during part loop operation. Adjustments will be made for non-operating pumps and the sheared snaft trip.
Pump Configuration: Measures will be taken to assure that non-operating pumps cannot be restarted.
Feed Pump: Feedwater flow will be restricted to less than 65%.
Steam Flow: Turbine and bypass valves restricted to prevent steam flow of greater than 65%.
Specific cnanges to setpoints, technical specifications, and procedures will be provided to each applicant.
C-2
i C.1.3 COMPARIS0N TABLES Except for the above noted modifications in operating conditions (Section C.1.2) and characteristics identified in the appropriate sections, the specific comparisons of System 80 features to other recer.t designs reflected in CESSAR-F remain applicable for part loop operation.
C.1.4 - C.1.9 Sections 1.4 through 1.9 of CESSAR-F directly apply to part loop operation. -
C.2 SITE CHARACTERISTICS The information presented in Chapter 2 is applicable to part loop operation.
C.3 DESIGN OF STRUCTURES COMPONENTS AND SYSTEMS C.3.1 GENERAL DESIGN CRITERIA The positions on the GDCs presented in Section 3.1 of CESSAR-F are applicable to part loop operation.
C.3.2 CLASSIFICATION The classification of components presented in Section 3.2 of CESSAR-F is application to part loop operation.
C.3.3 WIND AND TORNADO LOADINGS The information presented in Section 3.3 of CESSAR-F is applicable to part loop operation.
C.3.4 WATER LEVEL (FLOOD) DESIGN The information presented in Section 3.4 of CESSAR-F is applicable to part loop operation.
C.3.5 MISSILE PROTECTION The information presented in Section 3.5 of CES5AR-F is applicable to part loop operation.
C.3.6 PROTECTION AGAINST DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PTPING The information presented in Section 3.6 of CESSAR-F is applicable to part loop operation.
C-3 i
i C.3.7 SEISMIC DESIGN Part loop operation does not af fect the criteria and analytical methods used in these calculations. The information presented in Section 3.7 of CESSAR-F is applicable to part loop operation.
. C.3.8 DESIGN OF CATEGORY ONE STRUCTURES The information presented in Section 3.8 of CESSAR-F is applicable to part loop operation.
C.3.9 MECHANICAL SYSTEMS AND COMPONENTS The values of the key physical parameters (e.g., temperature and pressure) associated with part loop operation are not substantially different from those present during four loop operation. The physical parameters were surveyed with the
^
result that they remain applicable. The dynamic loads under two pump operation were reviewed and found only slightly worse than those for full power operation.
C.3.10 SEISMIC DESIGN OF CATEGORY ONE INSTRUMENTATION AND ELECTRICAL' EQUIPMENT The information presented in Section 3.10 of CESSAR-F is applicable to part loop operation.
C.3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL EQUIPMENT No normal, transient or accident condition for part loop operation produces environments more severe than those in Section 3.11 of CESSAR-F. Therefore, the information presented in 3.11 is applicable to part loop operation.
C.4 REACTOR C.4.1 REACTOR Part loop operation will be compatible with the fuel, reactor components and arrangements as already" described for System 80.
The information presented in Section 4.1 of CESSAR-F is applicable to part loop operation.
C.4.2 FUEL DESIGN Part loop operation will be compatible witn the fuel, reactor -
components and arrangements as already described for System 80.
The information presented in Section 4.2 of CESSAR-F is applicable to part loop cperation.
C-4
C.4.3 NUCLEAR DESIGN The specifically designated sections describe the effects of part loop operation on the core neutronics parameters (i.e., absolute values and changes in these parameters from the fcur pump configuration). All other provisions of Section 4.3 of CESSAR-F, including figures and tables, remain applicable to part loop operation, provided part loop operation is for a short period of time, i.e., cumulative time less than three weeks per cycle.
C.4.3.1 Design Bases The bases for the nuclear design of the fuel and reactivity control systems presented in Section 4.3.1 of CESSAR-F apply to part loop operation.
C.4.3.2 Description C.4.3.E.1 Nuclear Design Description The nuclear design is fixed at beginning of cycle and does not charge as a result of entering tne part loop mode. The inputs to tN nuclear analyses do change from four pump to part loop operation. In carticular, the power level, flow rate and flow distribution change from a core physics perspective. (See Section C.1 for a description of the operating conditions during part loop operation.)
C.4.3.2.2 Power Distribution The relative radial power distribution will change when the reactor is switched from four pump to part loop (two opposite pumps). There are two small but distinct effects. The first is an inward shift of power due to the differential change in Doppler reactivity upon entering the part loop mode at about half of the four pump rated power. The second is a shift of power due to the asymmetric flow distribution. The combination of these two effects produces the percent increases in assembly power shown on Figure C.4.3-1. These are maximum increases corresponding to the entire cycle. This figure shows the full core power distribution folded into an octant.
There is essentially no change in axial shape upon entering the part loop mode because the axial distribution of moderator density is preserved by maintaining a nearly constant ratio of core power to core flow. In addition, the two pump opposed loop axial shapes are within the allowable range of operation.
The reactivity insertions and cnanges in power distribution calculated for use in the CEA ejection and CEA drop analyses are bounded by those for four pump operation.
C-5
i C.4.3.2.3 Reactivity Coefficients The coefficients of core reactivity (i.e., fuel temperature and power moderator pressure) are virtually unchanged when in part loop operation. The moderator temperature coefficient (MTC),
which is notably sensitive to vari characteristics, becomes 0.1 x 10'ggions F lessinnegative core operating due to the slightly higher level of soluble boron (about 80 PPM). This variation is insignificant compared to tne bounding values of MTC used in the safety analyses (see Figures 4.3-46 and 4.3-47 of CESSAR).
C.4.3.2.4.1 Reactivity Control The soluble boron level for criticality at 48% power in the part loop mode is 80 PPM higher than in four pump mode for 100% power.
C.4.3.2.4.2 Power Level and Power Distribution Control The information in Section 4.3.2.4.2 of CESSAR-F applies to part loop operation.
C.4.3.2.4.3 Shutdown Reactivity Control The reactivity allowances for part loop operation are no more adverse than presented in Section 4.3.2.4.3 of CESSAR-F.
C.4.3.2.5 Control Element Assembly Pattern and Reactivity Worth The data in Section 4.3.2.5 of CESSAR-F applies to part loop operation with two exceptions: (1) only the lead bank is allowed in the active core during part loop operation; (2) the maximum rod radial peaking factors listed in Table 4.3-8 of CESSAR-F snould be raised by 3% for part loop operation.
C.4.3.3 Analytical Methods The analytical methods and procedures described in Section 4.3.3 of CESSAR-F are applicable to part loop operation.
C.4.3.3.4 Local Axial Power Peaking Augmentation Section 4.3.3.4 of CESSAR-F applies to part loop operation.
l C- 6
}
X.X MAXIMUM INCREASE IN 2.4 1.7 1.5 0.8 ASSEMBLY POWER
) ..
C-00 C-LO C-LO B-HI B-LO B-HI 2.1 2.4 2.6 2.3 1.8 0.9 C-00 C-LO A B LO A B LO B LO 1.6 1.3 2.1 2.6 2.6 1.8 1.2 B-LO B-LO A B HI A B-Hi 2.3 2.0 1.9 2.1 1.4 1.4 A B-HI A B-HI A 0.7 0.9 1.1 1.2 1.0 A B-HI A B HI 0.2 0.8 0.9 0.7 B LO B-HI A 0.4 0.3 0.2 A B HI 0.0 0.0 A
0.0 C-E MAXIMUM INCREASE IN RELATIVE ASSEMBLY POWER Figure
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FROM FULL POWER (FOUR PUMPS) TO FULL PART LOOP S b hd / POWER (TWO OPPOSING PUMPS) C 4.3-1
C.4.4 THERMAL AND HYDRAULIC DESIGN The specifically designated sections below explain the major changes to the steady state thermal and hydraulic analysis of the reactor core, the analytical methods and the experimental bases which justify operation of the plant at reduced power on two reactor coolant pumps in diametrically opposed cold legs. All other portions of Section 4.4 of CESSAR-F, including figures and tables, remain applicable for part loop operation.
C.4,4.2.4 Void Fraction ' Distribution The exit void fractions and qualities in different regions of the core are shown in Figure C.4.4-3. Tne axial distribution of void fraction and quality in the subchannel adjacent to the rod with the minimum DNBR is shown in Figure C.4.4-4. The information presented in these figures is based on the reactor operating conditions and engineering factors given in Table C.4.4-1, for radial power distributions in Figures C.4.4-1 and C.4.4-2, and for the 1.26 peaked axial power distribution in Figure 4.4-
- 3. For these conditions, only subcooled boiling occurs in the core and in the limiting subchannel (channel with tne minimum DNBR).
C.4.4.2.6.1 Reactor Vessel Flow Distribution The reactor coolant bypass flow distribution percentages are unchanged from those given in Section 4.4.2.6.1 and Table 4.4-3.
C.4.4.2.6.2 Reactor Vessel and Core Pressure Drops The irrecoverable pressure losses from the inlet to the outlet nozzles are calculated using standard loss coefficient methods and information from System 80 reactor flow model tests.
Reverse flow pressure losses at the two inactive inlet nozzles
(
are obtained by standard loss coefficient methods.
Pressure losses at 48% power, 50% of design minimum grimary coolant flow and an operating pressure of 2250 lb/in , are listed in Table C.4.4-4 together with the coolant temperature -
used to calculate each pressure loss. The calculated pressure losses include both geometric and Reynolds number dependent effects.
l l C.4.4.2.6.3 Hydraulic Loads on Internal Components
! The significant hydraulic loads which act on the reactor I internals during steady state operation are given in Table 4.4-
- 5. These loads already reflect the limiting case for any mode of part loop or full flow operation at 500 F. Therefore, operation in the two loop RCP opposed mode at 500'F will not
, exceed the values of Table 4.4-5.
l C-7 i
C.4.4.2.9.1 Pressure Drop Uncertainties The reactor vessel pressure losses listed in Table C.4.4-4 are based on best estimate values of the pressure loss coefficients for the various flow path segments within the vessel. The uncertainty on the vessel overall loss coefficient at the 2 cr level is + 10%.
C.4.4.2.9.2 I&draulic Loads Uncertainty The statements made in Section 4.4.2.9.2 are applicable to the two loop RCP opposed data of Table C.4.4-5.
C.4.4.3.1.1 Configuration of the RCS Table C.4.4-7 lists the design minimum flow through each flow path in the RCS for the two loop RCP opposed operating mode.
C.4.4.4.2.1-A Inlet Flow and Core Exit Pressure Distribution The core inlet flow distribution is required as input to the TORC thermal margin code (refer to Section C.4.4.4.5.2). The inlet e flow distribution for two loop operation with two opposed pumps operating was determined by adjusting the four loop flow distribution data from the System 80 reactor flow model test. A core map of incremental differences between two loop and four loop relative flow rates was constructed and used to generate a two loop opposed inlet flow distribution for System 80. Minimum inlet flow rates are found to occur in the two symmetric quadrants containing reverse flow inlet nozzles.
The core exist pressure distribution for two loop opposed RCP operation is equal to ths four loop case.
C 4.4.4.4 Core Thennal Responses l
Steady state core parameters are 1summarized in Table C.4.4-1 for ncimal two pump operation. Figure C.4.4-5 shows the sensitivity of the minimum DNBR to small changes in pressure, inlet temperature and flow from the conditions specified in Table C.4.4-
- 1. The 1.26 top peaked axial power distribution (Figure 4.4-3) is used to generate the sensitivity data.
The response of the core to anticipated operational occurrences is discussed in Chapter C.15. The response of the core at the j design overpower is not applicable for the System 80 cores since
- the reactor protective system prevents the design basis limits l from being exceeded.
l C-8
The supervisory and the protective systems will ensure that the design bases in Section 4.4.1 are not violated during two pump operation for any steady state operating condition of inlet temperature, pressure, flow, power and core power distribution, and for the anticipated operational occurrences discussed in Chapter C.15.
~
C.4.4.4.5.2 Thermal Margin Analysis Thermal margin analysis of System 80 reactor core for part loop operation have been performed using TORC (Refernce 3) and CETOP-D (Reference 4) computer codes with CE-1 critical heat flux correlation (References 1 and 2). Appropriate adjustments were made to the input of the codes to reflect part loop power, flow and core inlet flow distribution. Table C.4.4-1 contains a list of pertinent thermal hydraulic design parameters used for both the safety analysis and the generation of reactor protection system setpoint information.
Investigations nave been made to ensure that various correlations used in thermal hydraulic analyses for fcur loop operation are valid during two loop operation. These correlations have been discussed in Section 4.4.
The detailed core thermal margin calculations are performed using the TORC computer code are used to develop and to support tne simplified design core thermal margin calcuational scheme.
Simplified calculations are performed using the CETOP-D computer code. The method used for the simplified design calculations is discussed in detail in Reference 4. In summary, the method is to use one limiting not assembly radial power distribution for all analyses, to raise or lower the hot assembly power to provide the proper maximum rod radial power factor, and to use the core average mass velocity in .all fuel assemblies except the hot assembly. The appropriate reduction for the hot assembly mass velocity is determined based upon detailed TORC calculations and the core inlet flow distribution. ,
C.4.4.5.3 liydraulic Instability Analysis Section 4.4.5.3 is also applicable to the description of flow
! stability in the System 80 part loop operation. A Study of flow
! stability in the C-E PWRs using the CE-HYDNA code (Reference 5) was conducted to cover a wide range of operating conditions which encompass System 80 part loop operation. The results from that study indicate that flow is stable throughout the range of opersdng conditions examined.
i C-9
l' References for Section C.4.4
- 1. " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Grids, Part 1, Uniform Axial Power Distributions",
September 1976, CENPD-162-P-A (proprietary), CENPD-162-A (non-proprieta ry)
- 2. " Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Grids, Part 2, Non-Uniform Axial Power Distributions",, June .1976, CENPD-207-P (proprietary), .CENPD-207 (non-proprietary)
- 3. " TORC Code: A Computer Code for Determining the Thermal Margin of a Reactor Core", July 1975, CENPD-161-P (proprietary), CENPD-161 (non-proprietary)
- 4. "CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3", Docket No. 50-361, 50-362, CEN-160(S)-P, Rev.1-P, September 1981
Power & Light Co. , January 16, 1978, and enclosure,
" Assessment of Core Flow Stability for C-E PWRs", .
CEN-64( A)-P (proprietary) and CEN-64( A), (non-proprietary),
July 1977 9
C - 10
[
i TABLE C.4.4-1 THERMAL AND HYDRAULIC PARAMETERS (Sheet 1 of 2)
Reactor Parameters 4-loop 2-loop Core Average Characteristics at Full Power: (100% Power) (48% Power)
Total core heat output, Mit 3,800 1,824 Total core heat output, million Btu /h 12,970' 6,225 Average fuel rod energy deposition fraction 0.975 0.975 Hot fuel rod energy deposition fraction 0.975 0.975 1
l Primary system pressure, psia 2,250 2,250 Reactor inlet coolant temperature, 'F 565 564 Reactor outlet coolant temperature, *F 621 618 Core exit average coolant temperature, F 623 620 i Average core enthalpy rise, Btu /lbm 82 78 Design minimum primary coolant flow rate, 445,600 222,800 gal / min Design maximum core bypass flow, % of primary 3.0 3.0
~
, Design minimum core flow rate, gal / min 432,000 216,000 l
Hydraulic diameter of nominal subchannel, in. 0.471 0.471 l
' 2 Core flow area, ft 60.8 60.8 Core avg mass velocity, million Ibm /h-ft 2 2.61 1.30 Core avg coolant velocity, ft/s 16.6 8.3 Core avg fuel rod heat flux, Btu /h-ft 2 184,400 88,500 Total heat transfer area, ft 68,600 68,600-Average fuel rod linear heat rate kw/ft 5.40 2.59 Power density, kw/ liter 95.6 45.9
, N9. of active fuel rods 54,956 54,956 L.
TABLE C.4.4-1 (Cont'd)
(Sheet 2 of 2)
Power Distribution Factors:
Rod radial power factor 1.55 1.55 Nuclear power factor 2.28 2.28 Total heat flux factor 2.35 2.35 Maximum augmentation factor 1.059 1.059 Maxir:2.em pp lengtn, in. 0.761 0.761 Fngineering Factors:
Engineering heat flux factor 1.03 1.03 Engineering enthalpy rise factor 1.03 1.03 Pitch, Bowing and Clad Diameter Enthalpy 1.05 1.05 Rise Engineering factor on linear heat rate 1.03 1.03 Characteristics of Rod and Channel with .
Minimum DNBR Maximum fuel rod heat, flux, Btu /h-ft 2 433,000 208,000 Maximum fuel rod linear heat rate, kw/ft 12.7 6.1 U0 maximum steady state temperature, F 3,200 1875 2
Outlet temperature, F 652 644 1
Outlet enthalpy, Btu /lbm 699 684 Minimum DNBR at nominal conditions 1.79 3.08 (CE-l Correlation) l I
i l
l
_: . - _~ c - - - - --
/
1 e
TABLE C.4si:4 Reactor Vessel Best Estimate i Pressure Losses and Coolant Temperature Two Loop Opposed RCP Operation i
Temperature Component Pressurg)
(1b/in Loss (OF)
Inlet Nozzle and 900 Turn 14.1 564 Downcomer, Lower Plenum, 4.0 564 and: Support Structure i
i Fuel Assembly 4.3 592 Fuel Assembly Outlet to 4.2 620
. Outlet Nozzle .
TOTAL PRESSURE LOSS 26.6 l
l t
l l -
i l
l l
i l
i
s TABLE C .4.4-7 RCS Design flinimum Flows Two Loop Opposed RCP Operation LlowPath F1ou (1bm/hr)
Total flinimum RCS Flow 82.1 x 106 Core Bypass Flow (Design flaximum) 2. 5 x 106 6
Core Flow 79.6 x 10 0
Hot Leg Flow 41.0 x 10 Cold Leg Flows Active RCP (forward) 54.1 x 10 6 Locked Rotor RCP (reverse) -13.1 x 10 6 h
6
/
0.621 0.799 0.858 0.876 1.009 1.130 1.194 1.202 ASSY. AVG.R0D RADIAL POWER FACTOR----- 0.616 0.887 1.103 1.092 1.188 1.219 gkMApO R U
Qg- - 1.002 1.213 1.314 1.252 1.367 1.388 0.704 1.093 0.948 1.131 0.867 1.199 1.257 1.101 1.333 1.063 1.256 0.989 1.372 1.409 0.620 1.266 1.176 1.155 0.972 1.115 0.938 1.060 1.003 1.550 1.309 1.290 1.057 1.236 1.057 1.252 0.883 0.952 1.161 0.864 1.127 0.963 1.017 0.624 1.212 1.064 1.296 0.987 1.253 1.061 1.176 0.790 0.626 1.099 1.133 0.970 1.133 0.998 1.191 0.951 1.029 1.012 1.315 1.258 1.058 1.260 1.109 1.335 1.088 1.194 0.796 1.100 0.869 1.112 0.967 1.185 1.247 1.204 0.991 1.131 1.258 0.993 1.232 1.063 1.329 1.378 1.349 1.087 0.855 1.196 1.207 0.937 1.019 0.948 1.198 1.017 1.165 l
1.195 1.374 1.377 1.060 1.178 1.087 1.342 1.124 1.294 0.872- 1.215 1.255 1.068 0.623 1.035 0.995 1.159 0.874 1.202 1.3 85 1.407 1.258 0.792 1.197 1.087 1.287 0.969 C-E CORE WIDE PLANAR PDWER DISTRIBUTION Figure l
g FOR SAMPLE DNB ANALYSIS C.4.4-1
POISON ROD e?e"#2OR Q0000000.
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l l C-E i R0D RADIAL POWER FACTORS IN HOT QUADRANT Figure S3l@f8 / FOR-SAMPLE DNB ANALYSIS C.4.4-2
0.0 0.0 0.0 0.0 EXIT VOID FRACTION, % -21 -17 -16 -16
\
s 0.0 0.0 0. 0 0.0 0. 0 0.0 EXIT QUALITY, %- =-21 -16 -8 -12 -10 -10 0.0 0.0 0.0 0.0 0.0 0.0 0.0
-19 -12 -13 -11 -15 -11 -10 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
-21 -9 -10 -11 -13 -12 -14 -13 0.0 0.0 0.0 0.0 0. 0 0. 0 0.0 0. 0
-16 -15 -11 -15 -12 -14 -14 -20 I
0.0 0.0 0.0 0.0 0.0 0. 0 I I
-21 -12 -12 -13 -12 -13 i i
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-13 !
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_ _ _ _ _:_ _ _ _ q _ _ _ _ _ _ _ _ 4 _ _ _ _ 4. _ _ _ _ __ _ _ _ s _ _ _ ,;_ __ _ _ _
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i I
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i i I I ' I i
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i C-E i AVERAGE VOID FRACTIONS AND QUALITIES Figure l
gggg f/ AT THE EXIT OF DIFFERENT CORE REGIONS I
C 4.4-3 l
1
1 l
O i ,
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!iE QUALIW e -20 E'
W d -30 -
5 cr
-40 i , , ,
-50 l
0 0.2 0.4 0.6 0. 8 1'0 l
FRACTION OF ACTIVE CORE HEIGHT l
NOTE: VOID FRACTION ALONG THE ENTIRE SUBCHANNELIS ZERO t
C-E AXIAL DISTRIBUTION OF VOID FRACTION AND Figure t
Sf3 / QUALITY IN THE SUBCHANNEL ADJACENT TO THE ROD WITH MINIMUM DNBR C.4.4-4
i i 1 50% OF DESIGN MINIMUM FLOW g 2.9 E -
PRESSURE _
g 2.8 (PSIA) h200
, 2200 _
2.6 -
f I I 560 565 570 REACTOR INLET COOLANT TEMPERATURE, UF i 1 55% OF DESIGN MINIMUM FLOW g 3.1 -
z PRESSURE (PSIA) s 3.0
$ 2300 pg 2.9
'N 220
~
2.8 -
l t 1 560 565 570 REACTOR INLET COOLANT TEMPERATURE, UF i I i
605 0F DESIGN MINIMUM FLOW g 3.4 -
z PRESSURE _
@ 3*3 (PSIA)
E -
y 3.2 -
2300 2 ~ 2250 -
3.1 -
,2200 560 565 570 REACTOR INLET COOLANT TEMPERATURE, UF c-t SENSITIVITY OF MINIMUM DNBR Figure TO SMALL CHANGES IN S REACTOR COOLANT CONDITIONS C. 4.4-5
U a .
,m C.4.5 REACTOR INTERNALS MATERIALS -
- -J 1
N s
Part loop operation is compatible with the fuel, reactor ,
components and arrangements a3 already described forMystem 30. S The information presented in Section 4.5 of CESSAR Fils s
~- e applicable for part loop operation. 3, ;L'
~ "
y' ' ,4, *
~ -
C.4.6 '
FUNCTIONAL DESIGN OF' REACTIVITY' CONTROL SYSTEM -
, .'( 9 Part loop operation is compatible with the fuel, reactor h, f components and arrangements as already described for System 80g i .
The information presented in Section 4.6 is applicable for pari I
loop operation. - Y C.5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 7
The conditions for part loop operation are not significantly ,; e different from operating conditions specified for four loop .. .
. , ~ ,
operation. Therefore, the information of Chapter 5 of CESSAR-F i describing the structural integrity of the reactor vessel,'
primary coolant system and associated subsystems is applicable for part loop operation. -
s
-- #g
^
C.6 ENGINEERED SAFETY FEATURES J :<
.g ,
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C.6.1 MATERIALS .
Y The information presented in_. Sectio r,.i of CESSAR-F is
, %. Q' - -
r applicable to part loop operation. y %. .
C+ Q- N w C.6.2 CONTAINMENT SYSTEMS .
s s The conditions for part loop opbration were' reviewed aqd'it was concluded that the subcompartment shcrt tenn energy released. data in CESSAR-F is applicable. However, for peak containment pressure sizing calculations, energy released from the primary side is affected by part loop operation. Time dependent LOCA and MSLB data for part loop conditions will be generated. This data will be consistent with the four-loop data currently in CESSAR-F. .
C.6.3 EMERGENCY CORE COOLING SYSTEM C.6.3.1 Design Bases The design bases presented in Section 6.3.1 of CESSAR-F remain applicable to part loop operation C.6.3.2 System Design The design of the safety injection system in CESSAR-F remains unchanged for part loop operation.
C - 11
/
I ,. .
s <
r L
hN '-
i T. . x TWO' LOOP, [CCS P(RFORMANCE RESUL TS
'iC.6.3i3,i f '. ~ . ,
- C.).3.%1 ,-introdu'ction and-Sbnmary '
\ e . s TheECC5'oerfoba'nceevaluationfdemonstratingconformancewith i k- [' qg'\
m s
y .
~,,, . , *the10CFR5qQS,iAcceptance Crite-ia (Reference 1) is presented in
- V- ,CESSAR-F foli Jull power, four loap operation. The purpose of
' this r.upplementary evaluation is'to demonstrate that two loop op.eration at 48% power is also in conformance with 10CFR50.46 at a ' peak linear heat generation rate of 7.0 kw/ft. Conformance is summarized as follows r-Criterion (1) Peak ' Clad ' Temperature "The calculated maximum fuel element cladding temperature shall not exceed
' 2200*F".
s The evaluation demonstrates that the peak clad t'empe?a+ure is less than 2065*F.
- , w ,- ,
p b Criterion (2) Maximuk Cladding 0xidation "The calculated
'. total oxidation of tne cladding shall nowhere exceed 17% of the total cladding thickness before
^
,s oxidation". ,
Theevaluationdekonstratesthattnepeaklocal i\
clad oxidation percentage is less than 13.2%.
_ Criterion (3) Maximum Hydrogen Generation "The calculated
- total amount of nydrogen generated from the
. chemical reaction of the cladding with water or-steam shall not exceed 1% of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to reict".
The ev.aluation demonstrates that the peak core-wide ' oxidation is less than 0.799%.
Criterion (4) Coolable Geometry " Calculated changes in core geometry snall be such that the core remains amenable to cooling."
The clad swelling and rupture model whicn is part of the Evaluation Model (References 3 and 4) accounts for the effects of changes in core geometry if such changes are predicted to occur. With these core changes, core cooling was enough to lower temperatures. No further clad swelling and rupture can occur since the calculations were carried to the point at which the clad temperatures were decreasing and the system has been completely depressurized. Thus, a coolable geometry has been demonstrated.
C - 12
Criterion (5) Long Term Cooling "After any. calculated successful initial operation 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".
Analysis shows-that the rapid insertion of borated water from the ECCS will suitably limit the peak clad temperature and cool the core within a short _
period of time. Subsequently, the safety injection pumps will supply cooling water. from the refueling water tank or the containment sump to remove decay heat resulting from the long-lived radioactivity remaining in the core.
The method of analysis and results pertaining to two loop operation at 48% power are discussed in the following sections.
C.6.3.3.2 Large Break Analysis C.6.3.3.2.1 Mathenatical Model In the C-E ECCS large break evaluation model (Reference 4), the CEFLASH-4A (Reference 5) computer program is used to determine the primary system thermal-hydraulic behavior during the blowdown period, and the COMPERC-II (Reference 6) computer program is used to describe the system thermal-hydraulic behavior during the refill and reflood periods. The resulting transient parameters from these programs are input to the STRIKIN-II (Reference 7) program which is used to calculate the hot rod peak clad temperature and the peak local clad oxidation percentage.
In performing this analysis, the CEFLASH-4A and COMPERC-II programs were not utilized. The STRIKIN-II program was used to determine the hot rod clad temperature transient based on conservative assumptions for the blowoown and refill /reflood hydrauli cs. The details of these assumptions and the analytical tools used in this analysis are discussed in the subsections bel ow.
C.6.3.3.2.3 Core and System Parameters The significant core and system parameters considered in the analysis are presented in Table C.6.3.3.2-2.
The fission product decay heat generation curve used in this analysis considers previous operation at 102% with a subsequent shutdown and ascension to 50% power prior to the postulated LOCA. It was assumed that subsequent to shutdown from 102%
power, the plant achieves 50% power operation within one hour.
Thus, the decay heat generation includes botn decay heat contributions from 102% power operation after one hour plus the decay neat generation from 50% power operation. The decay heat C - 13 I
.- ---,.,, - ~. . - , < - , - - . ~ - .e . , -- . - - -
operation from 102% power operation is based on the standard ANS decay heat curve for infinite operation. The decay heat fraction from operation at 50% power was calculated from the buildup of the fission products for a finite operating period. - The
- assumption of infinite operation at 102% power plus the finite operation at 50% power is used to maximize the amount of heat generated for use in this evaluation. The composite decay heat curve was also increased by 20% as required by Appendix K to 10CFR50. The normalized power transient is shown in Figure C . 6. 3. 3 . 2- A .
During the blowdown period, the hot rod heat transfer was completely neglected. That is the blowdown was modelled as an adiabatic heatup period. The duration was conservatively chosen to be two seconds longer than the longest blowdown period predicted for any of the large breaks in the range 1.0 to 0.6 times the double-ended pipe arEd. Breaks smaller than 0.6 double--
ended area have been shown to be not limiting. Since hot rod heat transfer was totally neglected during blowdown, detailed CEFLASH-4A blowdown hydraulic calculations were not required.
Since blowdown heat transfer was ignored, the results of the analysis are a conservative representation of all large break sizes and locations.
During the refill period hot rod heat transfer included only rod-to-rod thermal radiation. During the reflood period heat transfer modes associated with less tnan one inch-per-second
, reflood rates were assumed. That is heat transfer consisted of rod-to-rod radiation and steam cooling. However, steam cooling was modeled assuming a constant, minimum heat transfer coefficient representing radiation to steam wnich is discussed in Reference 4, Section S III.D.6.b. This approach obviates the' need to perform detailed calculations of the reflood hydraulics.
FLECHT heat transfer coefficients from the full power analysis were applied directly to the average rod and the hot rod below the rupture location. At 48% initial power, the reflood rates I and heat transfer coefficients would be higher than the values
( employed from the full power analysis.
In modeling thermal rod-to-rod radiation, an enclosure was used l in which all of the fuel rods surrounding the hot rod were I considered to be at the same power as the hot rod. This radiation enclosure is much more limiting than can actually occur during operation thereby minimizing the benefit from thermal radiation. In this configuration, thermal radiation from the hot rod to immediate neighboring fuel rods is prevented. Thermal l radiation is only accounted for from the hot rod to the guide l
tube and more remote rods in the hot assembly. Since the power j distribution surrounding the hot rod is assumed to be uniform or
" flat", the results of this analysis are valid for any time in-life.
l l
l C - 14 I
L
l The initial operating thermal parameters assumed are those of the time-in-life in which the fuel stored energy is at a maximum.
In addition, the fuel rod was forced to rupture during the blowdown period to mbimize radiation cooling during the refill and reflood periods; thus, maximizing the peak clad temperature and local clad oxidation. This was done by adjusting the pressure differential across the clad to _ the value it would have if the system had fully depressurized within one second following the initiation of the transient.
C.6.3.3.2.4 Containment Parametidrs The containment parametors discussed in Section 6.2.1.5 of the CESSAR-F are not affected by two loop operation and remain applicable. Likewise the containment pressure determined for full power, four loop operation is also applicable for two loop operation.
C.6.3.3.2.5 Break Spectrum As discussed in Section C.6.3.3.2.3, due to the conservative assumptions of this analysis the results are applicable for all large break sizes and locations.
C.6.3.3.2.6 Pesults and Conclusions The results of the analysis demonstrate- that a PLHGR of 7.0 kw/ft is acceptable for two loop operation at'48% power. A summary of the performance parameters is as follows: -
The peak clad temperature was calculated to be less than 2065 F, with a peak local clad oxidation less than 13.2%. The total core-wide clad oxidation percentage would be less than.799%, the value obtained for full power, four loop operation. The core-wide clad oxidation will be much less at 48% power operation, being a strong function of power level.
The transient behavior of the important NSSS parameters is shown in the figures listed in Table C.6.3.3.2-5.
Figure C.6.3.3.2- A shows the normalized reactor power level .
Despite the many conservative assumptions, the results comply with the 10CFR50.46 Acceptance Criteria (Reference 1). The peak clad temperature was calculated to occur during the " late j reflood" period and was due to the very conservative assumptions i in regard to the limited heat transfer imposed during this ~
period. It is estimated that the resulting peak clad temperature would have been significantly lower than those reported nerein had a more representative calculation been performed.
l l
C - 15 l
/
C.6.3.3.3 Snall Break Analysis Section ' 6.3.3.3 demonstrates that small breaks at 102% power, _
four loop operation, and with a PLHGR of_15.0 kw/ft are not as limiting as the large breaks described in Section 6.3.3.2. Two loop operation at 48% power and a PLHGR of 7.0 kw/ft will yield improved results since small break ECCS performance is affected primarily by the inital power with other initial system hydraulic parameters being secondary. As such the small breaks analysts of Section 6.3.3.3 applies conservatively to two loop operation.
~
C.6.3.3.4 Post-LOCA Long Term ' Cooling Operation at 48% power will increase the post-LOCA performance margins for the boron precipitation consideration of large breaks and the neat removal consideration of small breaks. This occurs
, because the lower core power will result in slower baron concentrating rates for large breaks and less decreased.RCS neat removal for small breaks. The conclusions of the long term cooling performance analysis reported in Section 6.3.3.4 therefore remain applicable and are a conservative estimate of performance for two loop operation at 48% power. 4 C.6.3.4 Tests and Inspectior:s The information in CESSAR-F remains applicable' to part loop operation.
C.6.3.5 Instrumentation The information presented in Section 6.4 of CESSAR-F remains applicable to part loop operation.
C.6.4 HABITABILITY The information presented in Secton 6.4 of CESSAR-F is applicable to part loop operation.
C.6.5 FISSION PRODUCTION REMOVAL AND CONTROL SYSTEMS The information presented in Section 6.5 of CESSAR-F is applicable to part loop operation.
C.6.6 IN-SERVICE INSPECTION ,
! The information presented in Section 6.6 of CESSAR-F is applicable to part locp operation.
l l
1 C - 16
REFERENCES FOR SECTION C.6.3
- 1. Acceptance Criteria for Emergency Core Cooling Systems for light Water Cooled Nuclear Power Reactors, Federal Register, Vol . 39, No. 3 - Friday, January 4, 1974.
- 2. (deleted)
- 3. " Calculative Methods for the C-E Small Break LOCA Evaluation Model", CENPD-137, August, 1974.
" Calculative Methods for the C-E Small Break LOCA Evaluation Model", CENPD-137, Supplement 1, January,1977 (Proprietary)
- 4. " Calculative Methods for the C-E Large Break LOCA Evaluation Model", CENPD-132, August 1974 (Proprietary)
" Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model", Supplement 1, CENPD-132, December 1974, (Proprietary)
" Calculational Methods for tne C-E Large Break LOCA Evaluation Model",
Supplement 2, CENPD-132, July 1975 (Proprietary)
- 5. "CEFLASH-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis", CENPD-133, April 19'!1 (Proprieta ry)
"CEFLASH-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis (Modification)", Supplement 2, CENPD-133, December 1974 (Proprietary)
- 6. "COMPERC-II, A Program for Emergency Refill-Reflood of the Core", CENPD-134, August 1974, (Proprietary)
"COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modification)", Supplement 1, CENPD-134, February 1975 (Proprietary)
- 7. "STRIKIN II, A Cylindrical Geometry Fuel Rod Heat Transfer Program", CENPD-135, April 1974 (Proprietary)
"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modification)", Supplement 2, CENPD-135, February 1975 (Proprietary).
"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program",
Supplement 4, CENPD-135, August 1976, (Proprietary).
"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program",
Supplement 5, CENPD-135, April 1977 (Proprietary).
l C - 17
TABLE C.6.3.3.2-2 C-E SYSTEM 80 STANDARD PLANT GENERAL SYSTEM PARAMETERS AND INITIAL CONDITIONS F0R TWO LOOP OPERATION LARGE BREAK ECCS PERFORMANCE Quantity Val ue Reactor Power Level 1938 MWT Average Linear Heat Rate 2.8 kw/ft Peak Linear Heat Rate 7.0 kw/ft 2
Gap Conductance at Peak Linear Heat Rate
- 808.1 BTV/hr-ft 0F Fuel Centerline Temperature at Peak Linear 2003 F Heat Rate
- Fuel Average Temperature at Peak Linear 1497 F Heat Rate
- Hot Rod Gas Pressure 1129.0 psia 6
System Flow Rate (Total) 82.0x10 lbs/hr 6
Core Flow Rate 79.55x10 lbs/hr Initial System Pressure 2250 psia Core Inlet Temperature 569 F Core Outlet Temperature 625.6 F
- These quantities correspond to the burnup (774 MWD /MTU, hot rod average) yielding the highest peak clad temperature.
i TABLE C.6.3.3.2-5 C-E SYSTEM 80 STANDARD PLANT DESIGN TWO LOOP OPERATION VARIABLES PLOTTED AS A FUNCTION OF '
TIME FOR LARGE BREAK ANALYSIS Variabl e Figure No.
' Normalized Reactor Power C .6.3.3. 2- A Peak Clad Temperature C.6.3.3.2-B Peak Local Clad 0xidation C.6.3.3.2-C Hot Spot Gap Conductance C.6.3.3.2-D Clad Temperature, Centerline Fuel Temperature and C . 6 . 3 . 3 . 2- E Average Fuel Temperature for Hottest Node Hot Spot Heat Transfer Coefficient C .6.3.3. 2-F Hot Rod Internal Gas Pressure C.6.3.3.2-G D
- - . _ - . _ - ---__m______ _ _ _ _ _ _ _ _ _ _ _ _ _ . , _ _ _ _ _ _ _ . , _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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0 200 400 600 800 1000 TIME, SECONDS C-E / / TWO-LOOP O PER ATION CN.'3 2 f ff f f ly ,, / / REACTOR POWER -A
2200 , , , , , ,
2000 - x -
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g 1200 - --
d 1000 - PEAK CLAD TEMPERATURE NODE -
--- PEAK OXIDATION NODE 800 - -
600 - -
400 O 100 200 300 400 500 600 700 TIME, SECONDS C-E ' TWO-LOOP OPER ATION Sf25278t // PEAK CLAD TEMPERATURE C352
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18 i i , , , i 16 PEAK OXIDATION NODE -
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0 0 100 200 300 400 500 600 700 TIME, SECONDS C-E / F re MO-LOOP OPER ATION C. 3.2 Eff5F81 // PEAK LOCAL CLAD OXIDATION _
I 1800 , i i i i i -
1@0 - -
O o;[ 1400 -
t E
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E co 1200 -
N E 1000 - -
b a
z 8 800 - -
o 600 400-- -
200 - -
0 0 100 200 300 400 500 600 700 TIME, SECONDS C-E I F re TWO-LOOP OPER ATION 3.2 SMP81 /! HOT SPOT GAP CONDUCTANCE C.
, I 2400 i i i , i i FUEL CENTERLINE 2100 - -
/
CLAD u_ 1800 g AVERAGE FUEL E
1500 k
f5 S
g 1200 - -
900 - -
600 - -
300 0
0 100 200 300 400 500 600 700 TIME, SECONDS C-E TWO-LOOP OPER ATION Figure g CLAD TEMPER ATURE CENTERLINE FUEL TEMPERATURE C.63.3.2
& AVERAGE FUEL TEMPERATURE FOR HOTTEST NODE -E
l 180 i i i i i i 160 _ _
oI i--
LL.
& 140 c
25 120 s
5 G
T - -
b 100 8
5 -
th 80 5
E Q 60 E
40 20 - -
0 0 100 200 300 400 500 600 700 TIME, SECONDS C-E / Fi ure TWO-LOOP OPER ATION C. 3.3.2 S$l@P8. / f HOT SPOT HEAT TR ANSFER COEFFICIENT _
/
1200 i i i ,
yPINITIAL = 1129 PSIA 1000 - -
RUPTURE = 16.8 SEC
$ 800 - -
E sf 8
O 600 EE 400 - -.
200 -
0 0 10 20 30 40 50 TIME, SECONDS l
C-E I TWO-LOO P OPER ATION """
ffiff8. / [ HOT ROD INFERNAL GAS PRESSURE C.6.3 3.2
C.7 INSTRUMENTATION AND CONTROLS For part loop operation, analyses will be performed to determine the appropriate modification in the PPS trip settings and COLSS/CPC variables which ensure that the PPS, CPC and COLSS systems will continue to provide their normal safety functions.
Before commencing part loop operation, the required PPS trip setpoints, CPC and COLSS variables will be adjusted. As the reactor power is increased, PPC, CPC and COLSS verification tests will be performed to ensure the adequacy of the setpoint and variable changes as well as their proper implementation prior to proceeding to high power levels. Procedure guidelines are provided to the applicant to be followed in the control room to adjust the setpoints, initiate part loop operation and recover for resumption to full power.
C.8 ELECTRIC POWER The information presented in Chapter 8 of CESSAR-F is applicable to part loop operation.
C.9 AUXILIARY SYSTEMS No additional analyses are required for fuel storage and handling, water systems and HVAC. The effects of part loop operation upon such process auxiliaries as the CVCS are considered within the present four loop capability and additional analyses are not needed. The information presented in Chapter 9 .
of CESSAR-F is applicable to part loop operation.
C.10 STEAM POWER CONVERSION SYSTEMS No additional analyses are required for the turbine generator and main steam system. The turbine bypass system will be disabled
- during part loop operation. Maximum steam flow will be restricted to less than 65% of normal flow at 100% power.
C.11 RADI0 ACTIVE WASTE MANAGEMENT The information presented in Chapter 11'of CESSAR-F is applicable to part loop operation.
C.12 RADIATION PROTECTION The information presented in Chapter 12 of CESSAR-F is applicable to part loop operation.
C.13 CONDUCT OF OPERATIONS This chapter is not in the scope of CESSAR.
- C.14 INITIAL TEST PROGRAM The applicants test program will include additional startup test
- requirements for part loop operation (2 pumps operating in oppo-siteloops).
C - 18 i
/
C.15. ACCIDENT ANALYSES C.15.0 ORGANIZATION AND METHODOLOGY This chapter presents analytical evaluations of the Nuclear Steam Supply System (NSSS) response to postulated disturbances in process variables and to postulated malfunctions or failures of equipment for part-loop operation. Such incidents (or events) are postulated and their consequences analyzed despite the many precautions which are taken in the design, construction, quality assurance, and plant operation to prevent their occurrence. The effects of these incidents are examined to determine their consequences and to evaluate the capability built into the plant to control or accommodate such failures and situations.
It is anticipated that only a short portion of operating time will be spent on part loop operation, therefore, the probability of the events presented here will be less than for four pump, full power operation.
C.15.0.1 CLASSIFICATION 0F TRANSIENTS AND ACCIDENTS C.15.0.1.1 Format and Content This chapter is structured according to the format and content suggested by Reference 1 and required by Reference 16.
C.15.0.1.2 Ever,t Categories Each postulated initiating event has been assigned to one of the following categories;
- a. Increased Heat Removal by Secondary System,
- b. Decreased Heat Removal by Secondary System,
- c. Decreased Reactor Coolant flow,
, d. Reactivity and Power Distribution Anomalies,
- e. Increase in RCS Inventory,
- f. Decrease in RCS Inventory,
- g. Radioactive Release from a Subsystem or Component,
- h. Anticipated Transients Without Scram (ATWS).
Definition of an appropriate evaluation basis and the acceptance criteria does not presently exist for ATWS, therefore, these events are not addressed in this chapter. The assignment of an initiating event to one of these eight categories is made according to Reference 16.
C - 19 i
l C.15.0.1.3 Event Frequencies Reference 16 subjectively classifies initiating events in the following qualitative frequency groups:
A. Moderate Frequency Events B. Infrequent Events C. Accidents C.15.0.1.4 Events'and Event Combinations The events and event combinations in this chapter are tnose identified by Reference 16, and are presented with respect to the event specific acceptance criteria specified therein. For eacn applicable acceptable criterion in an event category, only the limiting event or event combination is presented in analytical detail. Qualitative discussions are provided for all other events or event combinations explaining why they are not limiting.
For event combinations which require consideration of a single failure, the limiting failure is selected from those listed in Table C.15.0-6. Only low probability dependent failures (e.g.,
loss of offsite power following turbine trip) and independent pre-existing failures are considered credible and included in the Tabl e. Pre-existing failures are equipment failures existing prior to the event initiation and are not revealed until called upon during the event (e.g., a failure of an emergency feedwater pump). High probability dependent occurrences are always included in the event analysis, if they have an adverse impact (e.g., loss of main feedwater pumps following a loss of electric power).
C.15.0.1.5 Section Numbering The incidents analyzed in this chapter are presented in sections in accordance with Reference 16 and are numbered as described in Table C.15.0-2. .
l C.15.0.1.6 Sequence of Events Analysis l ,
The purpose of the Sequence of Events and Systems Operation Section is to provide the step-by-step sequence of events from event initiation to the final stabilized condition. The progression of events and system operation is essentially the same for four pump operation as for two pump operation (part-i loop) although the timing of various system actuations differs due to the restrictive operating space and plant protection system setpoints allowed during part loop operation. Therefore the sequence of events di sgrams for four pump operation apply for part-loop operation.
C - 20
i C.15.0.2 SYSTEMS OPERATION During the course of any event various systems may be called upon to function. Some of these systems are described in Chapter .7 and include those electrical, instrumentation and control systems designed to perform a safety function -(i.e., those systems which must operate during an event to mitigate the consequences) and those systems not required to perform a safety function.
The reactor Protection System (RPS) is described in Section 7.2 of CESSAR-F. Table C.15.0-4 lists the RPS trips for which credit-is taken in the analyses discussed in this section, including the setpoints and the trip delay times-associated with each trip.
The analyses take into consideration the response times of actuated devices after the trip setting is reached.
The reactor trip delay times shown in Table C.15.0-4 are defined as the elapsed time from the time the sensor output reaches the trip setpoint to the. time the trip breakers open. The sensor response is modeled by using the transfer function for.the particular sensor used.
The interval between trip breaker opening and the time at which the magnetic flux of the Control Element Assembly (CEA) holding coils has decayed enough to allow CEA motion is conservatively assumed to be 0.34 seconds. Finally, a conservative value of 3.66 seconds is assumed for CEA insertion, defined as the elapsed time from the beginning of CEA motion to the time of 907.
insertion of the CEAs in the reactor core.
The Engineered Safety Feature Actuation Systems (ESFAS) and electrical, instrumentation and control systems required for. safe shutdown are described in Sections 7.3 and 7.4, respectively, of CESSAR-F. The manner in which these -systems function during events is discussed in each event description. The instrumentation which is required to be available to the operator in order to assist him in evaluating the nature of the event and determining required action is described in Section 7.5 of CESSAR-F. The use of this instrumentation by the operator is discussed
, in each event description. '
Other systems called upon to function are described in Chapters 6, 9, and in the Applicant's SAR. The utilization of these systems is described in the Sequence of Events section-of each presentation.
Systems which may function but are not required to perform safety functions are described in Section 7.7 of CESSAR-F. These include various control systems and the Core Operating Limit Supervisory System (COLSS). In general, normal automatic operation of these control systems is assumed unless lack of operation would make the consequences of the event more adverse.
In such cases, the particular control system is assumed to be inoperative, in the manual mode, until the time of operator i action.
C - 21
)
C.15.0.3 CORE AND SYSTEM PERFORMANCE C.15.0.3.1 Mathematical Model The Nuclear Steam Supply System (NSSS) response to various events was simulated using digital computer programs and analytical methods most of which are documented in Reference 2 and have been approved for use by the NRC by Reference 3.
C.15.0.3.1.1 Loss of Flow Analysis Method The method used to analyze incidents which are initiated by a decrease in reactor coolant flowrate is the static method documented in topical report CENPD-183 (Reference 4). The only deviation from that method was the use of the CETOP-D computer code (Reference 14) with the CE-1 CHF correlation to calculate both the time and value of the minimum DNBR during the transient.
C.15.0.3.1.2 CEA Ejection Analysis Method The method used for analysis of the reactivity and power distribution anomalies initiated by a CEA ejection (Section C.15.4.8) is documented in Reference 13, Topical Report CENPD-190-A, which was approved by the NRC for reference in license applications on June 10, 1976.
C.15.0.3.1.3 CESEC Computer Program The CESEC digital computer program (References 5 through 11,17 through 19) provides for the simulation of the Nuclear Steam Supply Systen (NSSS). The program calculates tne plant response for non-LOCA (loss of coolant accident) initiating events for a wide range of operating conditions.
The CESEC program, which numerically integrates the one-dimensional conservation equations, assumes a node flow-path network to model the NSSS. The primary system components considered in the code include the reactor vessel, the reactor core, the primary coolant loops, the pressurizer, the steam generators, and the reactor coolant pumps. The secondary system components, include the secondary side,of the steam generators, the main steam system, the feedwater system, and the various steam control valves. In addition, the program models some of the control and plant protection systems.
The code self-initializes for any given, but consistent, set of reactor power level, reactor coolant flow rate, and steam generator power sharing. During tne transient calculation, the time rate of cnange in system pressure and enthalpy are obtained from the solution of the conservation equations. Tnese derivatives are tnen numerically integrated in time, under the assumption of thermal equilibrium, to give the system pressure and nodal enthalpies. The fluid states recognized by the code are subcooled and saturated; superheating is allowed in the pressurizer. The fluid in the reactor coolant system is assumed to be homogeneous.
C - 22
i C.15.0.3.1.4 C0AST Computer Program The C0AST computer program is used to calculate the reactor coolant flow coastdown transient for any combination of active and inactive pumps and forward or reverse flow .in hot or cold legs. The program is described in Reference 11 and was referenced in Reference 2.
The equations of conservation of momentum are written for each of the flow paths of the C0AST model assuming unsteady one-dimensional flow of an incompressible fluid. The equation of conservation of mass is written for the appropriate nodal points. Pressure losses due to friction, bends, and shock losses are assumed proportional to the flow velocity squared. Pump dynamics are modeled using a head-flow curve for a pump at full speed and using four-quadrant curves, which are parametric diagrams of pump head and torque on coordinates of speed.versus flow, for a pump at other than full speed.
C.15.0.3.1.5 STRIKIN-II Computer Program The STRIKIN-II computer program is used to simulate the heat conduction within reactor fuel rods and its associated surface heat transfer. The STRIKIN-II program is described in Reference 12.
The STRIKIN-II computer program provices a single, or dual,'
closed channel model of a core flow channel to calculate the clad and fuel temperatures for an average or hot fuel rod, and the extent of the zirconium water reaction, for a cylindrical geometry fuel rod. STRIKIN-II includes:
A. Incorporation of all major reactivity feedback mechanisms B. A maximum of six delayed neutron groups C. Both axial (maximum of 20) and radial ~(maximum of 20) segmentation of the fuel element D. Control rod scram initiation on hi,gh neutron power.
C.15.0.3.1.6 CETOP-D Computer Program The CETOP-D computer program is used to simulate the fluid conditions within tne reactor core and to predict the existence of DNB on the fuel rods. The CETOP-D program is described in Reference 14 and applies the modeling techniques described in Reference 15. CETOP-D is also discussed in Section C.4.4.4.5.2.
C.15.0.3.1.7 Reactor Physics Computer Programs Numerous computer programs are used to produce the input reactor physics parameters required by the NSSS simulation and reactor core programs previously described. These reactor physics computer programs are as described in Chapter 4 of CESSAR-F.
C - 23
C.15.0.3.2 Initial Conditions The events discussed in this chapter have been analyzed over a range of initial values for the principal process variables. The ranges were chosen to encompass steady state operational configurations for part-loop operation.
Analysis over a range of initial conditions is compatible with the monitoring function performed by the COLSS (which is described in Chapter 7 of CESSAR-F) and the flexibility of plant operation which the COLSS allows. This flexibility is produced by allowing parameter trade-offs by monitoring the principal process variables, synthesizing the margin to fuel thermal design limits, and displaying to the reactor operator the core power operating limit. The required margin to DNB incorporated in COLSS is currently estaclished by the total loss of forced reactor coolant flow as described in Appendix 15A of CESSAR. The required margin to DNB is based on the total loss of forced reactor coolant flow since this initiating event produces the most rapid loss of margin to DNB before reactor trip and the maximum loss of margin to DNB after reactor trip. The peak linear heat generation rate incorporated in COLSS is established by the loss of coolant accident (LOCA). The range of values of each of tha principal process variables that was considered in analyses of events discussed in this chapter is listed in Table C.15.0-5.
C.15.0.3.3 Input Parameters The parameters used in the analyses are consistent with those listed in the preceding section and are primarily based on first-core values.
C.15.0.3.3.1 Doppler Coefficient The effective fuel temperature coefficient of reactivity (Doppler Coefficient) as discussed in Section C.4.3 is mul,tiplied by a weighting factor to conservatively account for higher feedback effects in the higher power density portions of the core and to account for Uncertainties in determining the actual fuel -
temperature reactivity effects. The Doppler weighting factor, which is specified for each analysis, is 0.85 for cases where a less negative Doppler feedback produces more adverse results and 1.15 for cases where a more negative Doppler feedback produces more adverse results.
The effective fuel temperature correlation is discussed in Section C.4.3. This correlation relates the effective fuel temperature, which is used to correlate Doppler reactivity, to the core power.
C - 24
C.15.0.3.3.2 Moderator Temperature Coefficient The range of moderator' temperature coefficients of reacgivity at Beginnigg of0 Life (B0L) operating conditions is 0.0 ao/ F to--
2.1x 10- Ao/ F and tne corre (E0C) conditions is -1.5x10 Ap /gpondjng rangegp/
F to -3.5x10- at F, eng of gycle all evaluated at a core average temperature of 594 F.
In addition, the moderator coefficient varies with changes in coolant temperature and tne inserted Control Element Assembly (CEA) worth. The most unfavorable value of the moderator coefficient is assumed for a particular analysis.
C.15.0.3.3.3 Shutdown CEA Reactivity The shutdown reactivity is dependent on the CEA worth available upon trip, the axial power distribution, the position of the regulating CEAs, and the time in cycle life. The minimum total negative reactivity worth of the CEAs available for a reactor trip at part-loop power operation) and zero power is taken as 10.0% Ao and 8.9% Ao, respectively, except where noted in individual discussions of events. These values include the most reactive CEA stuck in the fully withdrawn position and the effects of cooldown from hot full power (part loop operation) to hot Zero power temperature conditions.
The shutdown reactivity worth versus position curve which is employed in tne Chapter C.15 analyses is shown in Figure 15.0-2 of the CESSAR. This shutdown worth versus position curve was calculated assuming a more conservative rate of negative reactivity insertion than is expected to occur during the majority of operations, including power maneuvering.
Accordingly, it is a conservative representation of shutdown reactivity insertion rates for reactor trips which occur as a result of the events analyzed.
C.15.0.3.3.4 Effective Delayed Neutron fraction The effective neutron lifetime and delayed neutron fraction are functions of fuel burnup. For each analysis, the values of the neutron lifetime and the delayed neutron fraction are selected consistent with the time in life analy Nd.
C.15.0.3.3.5 Decay Heat Generation rate Analyses based upon full (part-loop) power initial conditions conservatively assume a decay heat generation rate based upon an infinite reactor operating period at full (part-loop) power.
C - 25
C.15.0.4 RADIOLOGICAL CONSEQUENCES Several of the events discussed are accompanied by the release of steam or liquid from the reactor coolant system or main steam system. The methodology and important input parameters used to assess the radiological consequences of these releases, are-discussed below.
The CESEC computer code (described in section C.15.0.3.1.3), in combination with hand calculations, was used to determine the mass and energy releases as a function of time. These data are then used as input to the calculation of radiological release to the atmosphere for determining thyroid and whold body doses at the exclusion area boundary.
The assumptions used for calculating radiological releases to the atmosphere follow.
- 1. The initial primary system activity level is based on the maximum activities in tne reactor coolant due to continuous full power operation with 1% failed fuel. This activity level corresponds to a cancentration of 2.09 x 10-3 is Curies /lbm, dose equivalent I-131.
- 2. The i x 10 gitial secondary Ci/lbm, systed activity dose equivalent I-131. level is equal to 4.54
- 3. Primary-to-secondary steam generator tube leakage 'is included in the calculation of activity releases to atmosphere from the steam generators. The " technical specification leakage" discussed in the analyses of Chapter 15 is a 1 gpm primary-to-secondary tube leak.
, 4. Events for which Reference 26 requires consideration of
" iodine spiking" the following are used:
A. For iodine spiking generated by the event, the iodine appearance rate is increased by a factor of 500.
B. For an abnormally high iodine concentration due to a previous iodine spike, a reactor coolant activity of 2.72 X 10-j Ci/lbm dose equivalent I-l l 131 is assumed.
The dose at the site exclusion area boundary (EAB) is calculated as follows:
- 1. Multiply the total primary system mass release by the primary system activity level and divide by the appropriate Decontamination Factor (DF). This gives the total number of dose equivalent 1-131 curies released from the primary system.
C - 26
l
- 2. For the applicable secondary system releases, multiply the total secondary system mass release by the secondary system activity level and divide by tne appropriate DF to obtain the equivalent I-131 curies released to the environment.
- 3. The curies of dose equivalent I-131 released to the environment can be converted to a thyroid _ dose by multiplying by the following factors:
- a. Breathing rate = 0.347 x 10-3 3 ~
- ec 3
- b. Atmospheric dispersion factor (X/Q) = 2.00 x 10 6 rem
- c. 1-131 dose conversion factor = 1.48 x 10
- Combinino these parameters gives an effective dose conversion factor equal to 1.027 rem /Ci. Thus, the total thyroid dcse is calculated by multiplying the total activity release (dose equivalent I-131 curies) by the eftcetive dose conversion factor (1.027 rem /Ci).
- 4. Mditional assumptions used in the determination of radiological releases to the atmosphere forr certain events are:
a.. For pipe breaks outside containment in piping connected-to the reactor coolant system, the release to atmosphere accounts for the _ formation of steam
- resulting from depressurization of the reactor i coolant. ;
- b. For pipe breaks or valve malfunctions outside containment in the main steam system which result in eventual dry-out of a steam generator, radioactive nuclides within tne steam generator are assumed to be released to atmosphere with a decontamination factor .
(DF) equal to 1. ,
[
+
9 C - 27
/
REFERENCES FOR SECTION C.15.0
- 1. NRC Regulatory Guide 1.70, Revision 2, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," September 1975.
- 2. " Combustion Engineering Standard Safety Analysis Report," CESSAR Docket t No. STN-50-470, December 1975.
- 3. " Combustion Engineering Standard safety Analysis Report (CESSAR System 80 Nuclear Steam Supply System Standard Nuclear Design Preliminary Design Approval," PDA-2, Docket No. STN 50-470, NRC, December 31, 1975.
- 4. "C-E Methods for loss of Flow Analysis," C-E, CENPD-183, July .1975.
- 5. "CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," CENPD-107, April 1974, Proprietary Information.
- 6. "ATWS Model Modi fications to CESEC," CENPD-107, Supplement 1, September 1974, Proprietary Information.
- 7. "ATWS Mad?ls Modi fication to CESEC" CENPD-107, Supplement 1, Amendment 1-P, November 1975, Proprietary Information.
- 8. "ATWS Model for Reactivity Feedback and Effect of Pressure on Fuel,"
CENPD-107, Supplement 2, September 1974, Proprietary Information.
- 9. "ATWS Model Modifications to CESEC," CEhPD-107, Supplement 3, August 1975.
- 10. "ATWS Model Modifications to CESEC," CENPD-107, Supplement 4-P, December 1975, Proprietary Information.
- 11. "C0AST Code Description," CENPD-98, April 1973, Proprietary Information
- 12. "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program,"
CENPD-135, April 1974 (Proprietary).
"STRIKIN-II, A cylindrical Geometry Fuel Rod Heat Transfer Program (Modification)," CENPD-135, Supplement 2, December D74 (Proprietary).
"STRIKIN-II, A Cylindrical Geometry Fuel Rod He'at Transfer Program,"
CENPD-135, Supplement 4, August 1976 (Proprietary).
- 13. "C-E Method for Control Element Assembly Ejection Analysis," C-E, CENPD-190-A, January 1976.
- 14. "CETOP-D, Code Structure and Modeling Methods For San Onofre Nuclear Generating Station (Proprietary) Units 2 and 3," Docket Nos. 50-361, 50-362, CEN-160(s)-P, Rev.1-P, Septea6er 1981.
- 15. " TORC Code--Verification and Simplified Modeling Methods," CENPD-206-P, January 1977, Proprietary Information.
C - 28
- 16. NUREG-75/087, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," as revised through December 31, 1978.
- 17. "ATWS Model Modifications to CESEC," CENPD-107, Supplement 5, June 1976, Proprietary Information.
- 18. "CESEC-Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," CENPD-107, Supplement 6, August 1978.
- 19. (Later - CESEC III Documentation)
C - 29
..}
TABLE C.15.0-2 4
i CHAPTER 15 SUBSECTION DESIGNATION 1 ..
. Each subsection is identified as C.15.W.X.Y. With trailing zeros omitted where: ,.
l W = 1. Increase in heat removal by the secondary system 2 Decrease in heat removal by the secondary system -
3 Decrease in reactor coolant system flowrate 4 Reactivity and pcwer distribution anomalies -
5 Increase in reactor coolant inventory 4
6 Decrease in reactor coolant inventory.
7 Radioactive release-from a subsystem or component 8 Primary System. Pressure deviation
~ X = 1,2, etc. Event Title from Ref. 26 Y=1 Identification of causes and frequency classifications
, 2 Sequence of events and systems operation 3 Analysis of effects and consequences 4 Conclusions i
l i-i
- -. .- . , . . . . . . ., - , . - . - . . . - -. ..-- .- - u. ,- - - - . , . - - . - - . -
M<>++S - , - ,, .
$<#4 's TEST TARGET (MT-3) 1.0 SuaEM y
- EE u L" Ea I.8 1.25 1.4 1.6
= s~
- 4 +4%
%, 7/#-
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<. >4 , ,
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TABLE C.15.0-4 REACTOR PROTECTION SYSTEM TRIPS USED IN THE SAFETY ANALYSIS Analysis Trip Delay Event RPS Setpoint Time (c)
High Logarithmic Power Level 2% of 3800 Mwt 550 ms Variable Overpower 17% or 58% (a)
High Pressurizer Pressure 2450 psia 550 ms All Chapter 15 Low Pressurizer Pressure 1870 psia 550 ms Events Except as Low steam Generator Pressure P -260 psia f) 550 ms sg Mentioned Below Low Steam Generator Water Level 55% wide range (b) 550 ms High Steam Generator Water Level 90% narrow 550 ms range (e)
Low DNBR 1.19 550 ms High Lccal Power Density 21 kw/ft(d) 750 ms feedwater and High Pressurizer Pressure 2475 psia 550 ns '
Steam Line Low Steam Generator Water Level 50% wide range (b) 550 mc Breaks Inside Containment.
- a. 17% of full power above the initial power or 58% of full power, whichever i s l ower. (See discussion in Section 7.2 of CESSAR-F. )
- b. Percent of distance between the wide range instrument taps above the lower tap. (See Chapter 5 for details.)
- c. The trip delay times are in milliseconds.(ms). (See Secten 7.2 for details.)
- d. Setpoint value is set below the value at which fuel centerline melting would occur. (See Section 4.4.)
- e. Percent of distance between the narrow range instrument taps above the l ower ta p. (See Chapter 5 for details.)
- f. P = steam generator pressure at event initiation.
sg
TABLE C.15.0-5 INITIAL CONDITIONS Parameter Units Range Core Power % of 3800 Mwt 0 - 50 Radial 1-pin peaking to 1.60 (all reds out) factor (with uncertainty) -
to 1.77 (lead bank inserted)
Axial Shape Index -0.2 < ASI < + 0.2 Reactor Vessel Inlet Ccalant 6 Flowrate % of 157.4 x 10 lbm/hr 50-60 Pressurizer Water Level % distance between 45 to 55 upper tap and lower tap above lower tap Core Inlet Coolant Tenperature F 564 + 5 F Reactor Coolant System Pressure psia 2200-2300 Steam Generator Water Level % distance between 77-87 upper tap and lower tap above lower tap (wide range) area under axial shape in lower half of core ASI = - area under axial shape in upper half of core (1) total area under axial shape
I 1
TABLE C.15.0-6 (Sheet 1 of 2)
SINGLE FAILURES STEAM BYPASS CONTROL SYSTEM l
- 1. Failure to Modulate Open l
- 2. Failure to Quick Open i
- 3. Cae Bypass Valve Fails to Quick Close
- 4. Excessive Steam Bypass Flow
- 5. Failure to Generate Automatic Withdrawal prohibit Signal During Steam Bypass $peration
- 6. Failure to Generate the reactor Power Cutback Signal REACTIVITY CONTROL SYSTEMS
- 7. Regulating Group (s) Fail (s) to Insert or Withdraw
- 8. A single CEA Stuck *
- 9. A CEA Subgroup Stuck *
- 10. Failure to Initiate or Execute the Reactor Power Cutback (not applicable - disabled)
- 11. CEA's Witndraw upon Automatic Withdrawal Prohibit and/or CEA Withdrawal Prohibit FEEDWATER CONTROL SYSTEM
- 12. Failure of Reactor Trip Override
- 13. Failure of High Level Override TURBINE-GENERATOR CONTROL SYSTEM
- 14. Failure to Modulate the Turbine Control Valves
- 15. Failure to Trip the Turbine i
~
PRESSURIZER PRESSURE CONTROL SYSTEM (PPCS)
- 16. Failure of Spray Control Valves to Open
- 17. Failure of Spray Control valves to Close
- 18. Failure of Backup Heaters to Turn On
- 19. Failure of Backup Heaters to Turn Off Control Element Drive Mechanism does not respond to control signal.
Release of CEA(s) on trip is not inhibited.
O l
l l
TABLE C.15.0-6 (Cont 'd) i PRESSURIZER LEVEL. CONTROL SYSTEM
- 20. Backup Charging Pump Fails to Turn On
. 21. Backup Charging Pump Fails to Turn Off
- 22. Letdown Flow Control Valve Fails to Close
- 23. Letdown Flow Control valve Fails to Open MAIN FEEDWATER SYSTEM ,
- 24. One MFIV Fails to Close
- 25. One Back flow Check Valve Fails to Close ,
MAIN STEAM SYSTEM
- 26. One MSIV Fails to Close
- 27. One Atmospheric. Dump Valve Fails to Open
- 29. Failure of Any One Emergency Feed Pump to Start EMERGENCY CORE COOLING SYSTEM
- 30. Failure of One HPSI or LPSI Pump 4 ELECTRICAL POWER SOURCES 3 2. . Loss of Offsite power after turbine trip
- 32. Failure of one emergency generator to start, run, or load
- 33. Failure of one breaker to achieve fast transfer to backup power supply l
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u
.i
C.15.1 INCREASED HEAT REMOVAL BY SECONDARY SYSTEM C.15.1.1 DECREASE IN FEEDWATER TEMPERATURE C.15.1.1.1 Identification of Event and Causes The identification of event and causes of a decrease in feedwater temperature are the same as described in Section 15.1.1.1 of the CESSAR.
C.15.1.1.2 Sequence of Events and Systems Operation The sequence of events and system operation are_ the same as described in Section 15.1.1.2 of tne CESSAR.
C.15.1.1.3 Analysis of Effects and Consequences The effects and consequences of a decrease in feedwater temperature event and a decrease in feedwater temperature event plus a single failure will be no more adverse than that given in Section 15.1.1.3 of the CESSAR. Tne results are not more adverse because the core protection calculators ensure that specified acceptable fuel design limits will not be violated for the increased main steam flow event and ensure that only a limited number of fuel pins will experience DNB for the event in combination with loss of offsite power.
C.15.1.1.4 Conclusions Tne decreased feedwater temperature event results in a DNBR greater that 1.19 throughout the transient. The event in combination with a loss of offsite power results in only a limited number of fuel pins in DNB. For both cases, the RCS pressure remains well below 2750 psia.
C.15.1.2 INCREASE IN FEEDWATER FLOW C.15.1.2.1 Identification of Event and Causes The identification of event and causes of an increase in feedwater flow are the same as described in Section 15.1.2.1 of the CESSAR.
C.15.1.2.2 Sequence of Events and Systems Operation The sequence of events and system operations are the same as described in Section 15.1.2.2 of the CESSAR.
C.15.1.2.3 Analysis of Effects and Consequences The effects and consequences of an increased feedwater flow event and an increased feedwater flow event plus a single failure will be no more adverse than that given in Section 15.1.2.3 of the CESSAR. The results are not more adverse because the core protection calculators ensure that specified acceptable fuel C - 30
/
design limits will not be violated for the increased main steam flow event and ensure that only a limited number of fuel pins will experience DNB for the event in combination witn loss of offsite power.
C.15.1.2.4 Conclusions The increased feedwater flow event results in a DNBR greater than 1.19 throughout the transient. The event in combination with a loss of offsite power results in only a limited number of fuel pins in DNB. For both cases, the RCS pressure remains below 2750 .
psia.
C.15.1.3 INCREASED MAIN STEAM FLOW C.15.1.3.1 Identification of Event and Causes The identification of event and causes of an increased main steam flow are simmilar to those described in Section 15.1.3.1 of the CESSAR. Steam flow will, at a maximum, increase to 65". of nominal full power steam flow rate, based on the operating restrictions presented in Section C.1.2.
C.15.1.3.2 Sequence of Events and Systems Operation An increase in main steam flow causes a decrease in the temperature of tne reactor coolant, an increase in core power and heat flux, and a decrease in reactor coolant system and steam generator pressures. Detection of these conditions is accomplished by the low pressurizer pressure, the low steam generator pressure and the low steam generator water level and the high reactor power alarms. If th2 transient were to result in an approach to specified acceptable fuel design limits, trip signals generated by the core protcction calculators would assure that low departure from nucleate V, ling ratio (DNBR) or high local power density limits are not exceeded.
, C.15.1.3.3 Analysis of Effects and Consequences The effects and consequences of an increased main steam flow event and an increased main steam flow event plus a single failure will be no more adverse than that given in Section
- 15.1.4.3 of the CESSAR. The results are not more adverse because
! the core protection calculators ensure that specified acceptable fuel design limits will not be violated for the increased main steam flow event and ensure that only a limited number of fuel pins will experience DNB for the event in combination with loss of offsite power.
C - 31
1 C.15.1'3.4
. Conclusions The increased main steam flow event results in a DNBR greater than 1.19 throughout the transient.' The event in combination with a loss of offsite power results in only a limited number of fuel pins in DNB. For both cases, the RCS pressure remains well below 2750 psia.
C.15.1.4 INADVERTENT OPENING 0F A STEAM GENERATOR RELIEF OR SAFETY VALVE i
C.15.1.4.1 Identification of Event and Causes The identification of event and causes of an inadvertent opening of a steam generator relief or safety valve are the same as described in Section 15.1.4.1 of the CESSAR.
-C.15.1.4.2 Sequence of Events and Systems of Operation
! The sequence of events and-systems operation are similar to those
- described in Section 15.1.4.2 of the CESSAR. The timing of-
, events for part loop will differ from CESSAR (15.1.4.2) due to the initial conditions defined for part loop operation (Table
, C.15.0-5).
C.15.1.4.3 Analysis of Effects and Consequences The effects and consequences will be no more adverse than that given in Section '15.1.4.3 of the CESSAR. The results are not more adverse because the core protection calculators ensure that 4
specified acceptable fuel design limits will not be violated for-the increased main steam flow event and ensure that only-a limited number of fuel pins will" experience DNB for the event in
- combination with loss of offsite power.
C.15.1.4.4 Concl usions The 10SGADV event results in a DNBR greater than 1.19 througnout the transient. The event in combination with a loss of offsite power results in only a limited number of fuel pins in DNB. For both cases, the RCS pressure remains well below 2750 psia, ensuring that tne integrity of tne RCS is maintained.
C - 32
-C.15.1.5 STEAM SYSTEM PIPING FAILURES INSIDE AND OUTSIDE CONTAINMENT' C.15.1.5.1 Identification of Event and Causes The identification and causes of steam system piping failures inside and outside containment are the same as that described in-Section 15.1.5.1 of CESSAR.
C.15.1.5.2 Sequence of Events and Systems Operation The sequence of events and system operation are the same as that described in Section 15.1.5.2 of CESSAR. However, the times for the sequence of events will differ due to the lower reactor coolant flow and due to the restricted operating space defined by' the . initial conditions in Table C.15.0-5.
C.15.1.5.3 Analysis of Effects and Consequences ,
The effects and consequences for steam system' piping failures inside and outside containment will be no more adverse than those presented in Section 15.1.5.3 of CESSAR. Due to restrictions on operating conditions for part loop ' operation (Table C.15.0-5) and due to an increo;0 in shutdown rod worth at.each-power level, the effects for part loop operation will be no more adverse'than for those presented for four pump operation.
C.15.1.5.4 Conclusion For the large steam line break in combination witn a single failure and stuck CEA, with or without a loss of offsite power, fission power remains suf ficiently low following reactor trip to preclude fuel damage as a result of post-trip return to power.
For a large steam line. break during zero power operation in combination with a loss-of offsite power and the maximum technical specification tube leakage, the two-hour inhalation -
thyroid dose at the exclusion area boundary (EAB) is well within 10 CFR 100 guidelines.
The maximum potential for radiological , releases due to fuel failure occurs for small steam line breaks outside containment in combination with a stuck CEA. For these cases, the maximum potential for degradation in fuel performance occurs prior to and during reactor trip. Witn the assumption of one gallon per minute steam generator tube leakage, the resultin'g two-hour inhalation thyroid dose at the EAB is within the 10 CFR 100 guidelines.
Potential fuel failure is sufficiently limited to ensure that the core will remain in place and intact with no loss of core cooling capabilities.
C - 33
C.15.2 DECREASE 0' HEAT REMOVAL BY THE SECONDARY SYSTEM C.15.2.1 LOSS OF EXTERNAL LOAD-
.-C.15.2.1.1 Identification of Event and Causes The identification of event and causes for the loss of external-i- load event during part loop operation is the same as that described in Section 15.2.1.1 of tne CESSAR.
~
. C.15.2.1.2 Sequence of Events and Systems Operation 1
The sequence of events and systems operation forht e loss of load
event during part loop operation is similar to tqet described in Section 15.2.1.2 of tne CESSAR. The only difference being that ir. part loop operation only two reactor coolant oumps will be operating.
C.15.2.1.3 Analysis of Effects and Consequences 4
The results of the loss of load event and the loss of load with a
! single failure event during part loop operatica are no more
- limiting than the results presented in Section 15.2.1.3 of the CESSA0.. Due to the lower core power allowed during part-loop l operation, a less severe reactor coolant heatup, and pressuri-zation can occur compared to the results of Section 15.2.1 of the CESSAR. Similar to the Loss of Condenser Vacuum (LOCV) event, the loss of load event results in an -increase in DNBR above the initial value prior to reactor trip due to the RCS Thus, this event does not result in fuel i pressurization.
cladding degradation. The limiting single failure with respect
! to fuel performance is the loss of offsite power following turbine trip. This event with a loss of offsite power results I in an event identical to the Loss of Flow (LOF) event discussed in Section C.15.3.1. Results of tne LOF event are directly appli-cable to the loss of load witn loss of offsite power following turbine trip event.
C.15.2.1.4 Concl usions ,
for the loss of load event and the loss of load with a concurrent '
i single failure event, the peak RCS pressure remains below 110%
- design pressure, thus ensuring primary system integrity. The DNBR remains above 1.19, thus ensuring fuel cladding integrity, i
I i
C - 34
4 C.15.2.2 TURBINE TRIP C.15.2.2.1 Identification of Event and Causes The identification of event and causes for the turbine trip event during part loop operation is the same as that described in Section 15.2.2.1 of tne CESSAR.
C.15.2.2.2 Sequence of Events and Systems Operation The sequence of events and systems operation for the turbine trip event during part loop operation is similar to that described in Section 15.2.2.2 of the CESSAR. The only difference being tnat in part loop operation only two reactor coolant pumps will be operating.
C.15.2.2.3 Analysis of Effects and Conseouences The results of the turbine trip event and the turbine trip with a single failure event during part loop operation are no more limiting than the results presented in Section 15.2.2.3 of the CESSAR. Due to the lower core power allowed during part loop operation, a less severe reactor coolant heatup and pressurization can occur compared to the results of Section
. 15.2.2.3 of the CESSAR. Similar to the loss of Condenser Vacuum i (LOCV) event, the turbine trip event results in an increase in DNBR above the initial value prior to reactor trip due to the RCS pressurization. Thus, this event does not result in fuel cladding degradation. The limiting single failure with respect to fuel performance is the loss of- offsite power following turbine trip. This event with a loss of offsite power results in an event identical to the Lo:s of Flow (LOF) event discussed in Section C.15.3.1. Resalts of the LOF event are directly applicable to the turtiine trip following loss of offsite power event.
C.15.2.2.4 Conclusions For the turbine trip event and the turbine trip with a concurrent single failure event, the peak RCS pressure remains below 110%
design pressure, thus ensuring primary system integrity. The DNBR remains above 1.19, thus ensuring fuel cladding integrity.
e C - 35
C.15.2.3 LOSS OF CONDENSER VACUUM C.15.2.3.1 Identification of Event and Causes The identification of event and causes for the Loss of Condenser Vacuum (LOCV) event during part loop operation is the same as that described in Section 15.2.3.1 of the CESSAR.
C.15.2.3.2 Sequence of Events and Systens Operation The sequence of events and systems operation for the LOCV event during part loop operation is similar to that in Section 15.2.3.2 of the CESSAR. The only difference being that in part loop operation only two reactor coolant pumps will be operating.
C.15.2.3.3 Analysis of Effgcts and Consequences The results of the LOCV event during part loop operation are no more limiting than the results presented in Section 15.2.3.3 of the CESSAR. Due to the lower core power allowed during part loop operation, a less severe reactor coolant heatup and pressurization can occur compared to the results presented in Section 15.2.3 of the CESSAR. For the same reasons identified in the LOCV event, described in Section 15.2.3 of the CESSAR, no single failure will result in a larger peak RCS pressure. The LOCV event results in an increase in DNBR above the initial value prior to reactor trip due to the RCS pressuri zation. Thus, this event does not result in fuel cladding degradation. Furthermore, since LOCV produces an increase in DNBR, a greater thermal margin exists than is required to preclude a DNBR below 1.19 when the most limiting single failure, as identified in Section 15.2.3.3 of the CESSAR, is considered. Consequently, neither the event or the event plus a single failure will result in fuel cladding degradation.
C.15.2.3.4 Concl usions For the LOCV event and the LOCV with a concurrent single failure event, the peak RCS pressure remains below 110% design pressure, thus ensuring primary system integrity.. The DNBR remains above 1.19, thus ensuring fuel cladding integrity.
I l
C - 36
[
.I C.15.2.4 MAIN STEAM ISOLATION VALVE' CLOSURE C.15.2.4.1 Identification of Event and Causes The identification of event and causes for the Main Steam Isolation Valve (M5IV) Closure event during 'part loop operation are the same as that described in Section 15.2.4.1 of the CESSAR.
C.15.2.4.2 Sequence of Events and Systems Operation The sequence of events and systems operation for the MSIV closure event during part loop operation is similar to that described in Section 15.2.4.2 of the CESSAR. Tne only difference being that in part loop operation only two reactor coolant pumps will be operating.
C.15.2.4.3 Analysis of Effects and Consequences The results of the MSIV closure event and the MSIV closure with a single failure event during part loop operation are no more limiting than the results presented in Section 15.2.4.3 of the CESSAR. Due to the lower core power allowed during part loop operation, a less severe reactor coolant heat up and pressuri-zation can occur compared to the results in Section 15.2.4 of the CESSAR. Similar to the Loss of Condenser Vacuum (LOCV) event, tne MSIV closure event results in an increase in DNBR above the initial value prior to reactor trip due to the RCS pressur-ization. Thus, this event does not result in fucl cladding degradation. The limiting single failure with respect to fuel performance is the loss of offsite power following turbine trip. This event with a loss offsite power results in an event nearly identical to the loss of Flow (LOF) event dis-cussed in Section C.15.3.1. Results of the LOF event are directly applicable to the MSIV closure with loss of of fsite power following turbine trip event.
C.15.2.4.4 Concl usions For tne MSIV closure event and the MSIV closure with a concurrent single failure event, the peak RCS pressure remains below 110%
design pressure, thus ensuring primary system integrity. The DNBR remains above ;.19, thus ensuring fuel cladding integrity.
i i
C - 37
i '
C.15.2.5 STEAM PRESSURE REGULATOR FAILURE This event does not apply to the CESSAR System 80 design and therefore is not presented.
C.15.2.6 LOSS OF NON-EMERGENCY A-C POWER TO THE STATION AUXILIARIES' C.15.2.6.1 Identificatien of Event and -Causes i
i The identification of event and causes for the loss of non-1 emergency A-C power to tne station auxiliaries (LOAC) during part loop operation is the same as that described in Section 15.1.6.1 of the CESSAR.
C.15.2.6.2 Sequence of Events and Systems Operation The sequence of. events and systems operation for the LOAC event during part loop operation is the same as that described in Section 15.2.1.2 of the CESSAR.
i C.15.2.6.3 Analysis of Effects and Consequences The results of the LOAC event and the LOAC with a single failure event during part loop operation are no more limiting than the results presented in Section 15.2.6.3 of the CESSAR with respect to RCS pressurization. Due to the lower core power allowed during part loop operation, a less severe reactor coolut heatup, and pressurization can occur compared to the results presented in Section 15.2.6 of the CESSAR. The LOAC event results in an event t l identical to the loss of Flow (LOF) event discussed in Section i
C.15.3.1 witn respect to fuel performance. Results of the LOF event are directly applicable to the LOAC event.
C.15.2.6.4 Conclusions
! For the LOAC event and the LOAC with a concurrent single failure
, event, the peak RCS pressure remains below 1107. design pressure, ,
! thus ensuring primary system integrity. Tne DNBR remains above
! 1.19, thus ensuring fuel cladding integrity.
l C.15.2.7 LOSS OF NORMAL FEEDWATER 4
C.15.2.7.1 Identification of Event and Causes
! The identification of events asnd causes for the loss of normal
- feedwater flow (LFW) event during part loop operation is the same
! as that described in Section 15.2.7.1 of tne CESSAR.
i C.15.2.7.2 Sequence of Events and Systems Operation
, The sequence of events and systems operation section for the LFW j event during part loop operation is similar to tnat described in l Section 15.2.7.2 of the CESSAR. The only difference being that .
l in part loop operation only two reactor coolant pumps will be i operating.
i e
C - 38
C.15.2.7.3- Analysis of Effects and Consequences
-The results of the LFW event and tne LFW with a single failure event during part loop operation are no more limiting than the results presented in Section 15.2.7.3 of the CESSAR with respect to RCS pressurizttion. Due to the lower core power allowed during part loop operation, a less severe reactor coolant heatup and therefore pressurizatin can occur compared to the results of Section 15.2.7 of CESSAR. Similar to the loss of Condenser Vacuum (LOCV) event the LFW event results in an increase in'DNBR above the initial value prior to reactor trip due to the RCS pressurization. Thus, this event does not result in fuel cladding degradation. The limiting single failure with respect to fuel performance is the loss of of fsite power following turbine trip. This event with a loss of offsite power results in an event less severe than the loss of Flow (LOF) event discussed in Sectin C.15.3.1 due to the initial increase in DNBR.
C.15.2.7.4 Concl usions For the LFW event and the LFW with a concurrent single failure event, the peak RCS pressure remains below 110% design pressure, thus ensuring primary system integrity. The DNBR remains above 1.19 tnus ensuring fuel cladding integrity.
C.15.2.8 LOSS OF FEEDWATER INVENTORY C.15.2.8.1 Identification of Event and Causes The Loss off Feedwater Inventory (LFI) event is initiated by a break in the main feedwater system (MFS) pipirq as described in Appendix 15B.2 of the CESSAR.
C.15.2.8.2 Sequence of Events and Systems Operation The sequence of events and systems operation section for the LFI
- event during part loop operation is similar to that for the LFI
- event presented in Appendix 158.4 of the CESSAR. The event i progresses slower with the lower core power required during part I
loop operation as compared to four pump, full power operation shown in Appendix 15B of the CESSAR, tnereby delaying the timing of the system actuations and parameter trends.
C.15.2.8.3 Analytis of Effects and Consequences i
The results of thee LFI event are no more limiting than the l results presented in Appendix 15B of the CESSAR. Due to the i lower core power allowed during part loop operation, a less severe reactor coolant neatup and pressurization can occur compared to the results of Appendix 15B of the CESSAR. This i sensitivity is shown in Figure ISB-S of the CESSAR. Due to RCS pressu-ization, the LFI event results in an increase in DNBR I above the initial value prior to turbine trip on reactor trip.
i l
l
! C - 39 L
With a loss of of fsite power following turbine trip, the DNBR decreases due to the reactor coolant pump coastdown, however, the results are less severe than those of the Loss of Flow (LOF) event discussed in Section C.15.3.1 with respect to fuel performance due to the initial increase in DNBR.
C.15.2.8.4 Concl usions For the LFI event the peak RCS pressure remains sufficiently low to unsure primary system integrity. The DNBR remains above 1.19, thus ensuring fuel cladding integrity.
S C - 40
l C.15.3 DECREASED REACTOR C0OLANT FLOW C.15.3.1 TOTAL LOSS OF REACTOR COOLANT FLOW C.15.3.1.1- . Identification of Events and Causes The identification of event and causes for the total-loss of reactor coolant flow is tne same as that descriDed in Section 15.3.1.1 of the CESSAR.
C.15.3.1.2 Sequence of Events and Systems Operation The sequence of events and systems operation are the same as that presented in Section 15.3.1.2 of tne CESSAR. The timings of the sequence of events will be different due to the limiting conditions for part loop operation. The timings are shown in Table C.15.3.1-1. Also, the primary safety valves do not open for this event during part loop operation.
C.15.3.1.3 Analysis of Effects and Consequences A. Mathematical Model The NSSS response to a total loss of reactor coolant flow was simulated using the CESEC computer program described in Section C.15.0.3. The minimum DNBR was calculated using the CETOP-D computer code (see Section C.15.0.3) which uses the CE-1 CHF correlation described in Reference 1 of Section C.4.4.
B. Input Parameters and initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a total loss of flow are discussed in Section C.15.0. The parameters, wH ch are unique to the analysis, discussed below, are listed in Table C.15.3.1-2.
The initial conditions were chosen to minimize the initial thermal margin and thus the minimum transient DNBR. Tne conditions chosen were maximum inlet temperature, minimum RCS pressure, minimum core flow rate, maximum core power, maximum radial peaking factor and a top peaked axial power di stribution. (Tabl e C.15.3.1-2)
I j C. Results The dynamic behavior of important NSSS parameters following
- i. a total loss uf reactor coolant flow is presented in Figures C.15.3.1- 1 to C .15.3.1-8.
l.
- The loss of offsite power causes the plant to experience a i simultaneous turbine trip, loss of main feedwater, condenser
! inoperability and a two reactor coolant pump coastdown. The loss of steam flow due to closure of the turbine stop valves 1
- C - 41 L
results in a rapid increase in the steam generator pressure.- The main steam safety valves open at 6.7 seconds and continue to cycle during the first 30 minutes of tne transient. A sharp' reduction in primary to secondary heat transfer follows .which in conjunction with the loss of
- - forced reactor coolant flow, causes a rapid heat up of the primary coelant.
The RCS pressure reaches a maximum of 2297 psia at 6.8 seconds (Figure C.15.3.1-3). This is less-than 110% of design pressure. At 11.8 seconds the secondary
- pressure reaches its maximum value of 1317 psia (Figure 15.3.1-7). This pressure is c,o less than 110% of design .
- pressure.
Subsequently, the RCS pressure descreases rapidly as the combination of reactor trip and main : steam safety valves opening reduce tne reactor coolant system energy. - After 30 minutes, the operator commences'cooldown using the Auxiliary Feedwater System and the atmospheric dump valves.
The minimum CE-1 DNBR calculated to occur during the transient is 1.61 (Figure C.15.3.1-8); thus, no fuel pins ,
are assumed to experience DNB for this event.
C.15.3.1.4 Conclusions The maximum RCS and secondary system pressures remain within 110%
of their design values following the total loss aof forced reactor coolant flow. The minimum DNBR calculated to occur duriag the transient is above 1.19 which ensures that the specified' acceptable fuel design limit is not violated.
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C - 42 1
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TABLE C.15.3.1-1 SEQUENCE OF EVENTS FOR TOTAL LOSS OF REACTOR COOLANT FLOW i
Time -Setpoint Success j
(Sec.) . Event Or Value Patn ,
i
- 0.0 Loss of Offsite Power
! - Turbine Trip l - Diesel Generator Starting Signal
- - Reactor Coolant .
I Pumps Coast Down
- Main Feedwater is Lost 1.7 _ Low DNBR Trip Signal 1.19 Projected Reactivity-Generated Control 2.19 CEA's Begin to Drop Reactivity:
6.7 Steam Generator Safety 1282 Secondary 4 Valves Open, psia System
! Integrity 6.8 Maximum RCS Pres:ure, psia 2297 e
l 11.8 Maximum Steam Generator 1317 Pressure, psia 1800.0 Operator Igitiates Plant -100 Secondary System Cooldown, F/ hour Integrity i
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TABLE C.15.3.1-2 ASSUMED INITIAL CONDITIONS FOR TOTAL LOSS OF REACTOR COOLANT FLOW Parantter Val ue Core Power Level, MWt 1908.7 Core Inlet Coolant Temperature, OF 569 Reactor Coolant System Pressure, psia 2200 Steam Generator Pressure, psia 1182.7 Core Mass Flow,106 lbm/hr 78.7 Core Minimum DNBR 2.16 I4aximum Radial Power Peaking Factor 1.63
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C.15.3.2 FLOW CONTROLLER MALFUNCTION CAUSING FLOW C0ASTDOWN This event is categorized as a Boiling heter Reactor event in SRP 15.3.2 and, therefore will not be analyzed.
C.15.3.3 SINGLE REACT'JR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF 0FFSITE POWER C.15.3.3.1 Identification of Event and Causes The identification of event and causes for the single reactor coolant pump rotor seizure with loss of offsite power is the same as that described in Section 15.3.3.1 of the CESSAR.
C.15.3.3.2 Sequence of Events and Systems Operation The sequence of events and systems operation are the same as that described in Section 15.3.3.2 of the CESSAR. The timings of the sequence of events will be different and are presented in Table C.15.3.3-1.
C.15.3.3.3 Analysis of Effects and Conseque"ces C.15.3.3.3.1 Core and System Performance A. Matnematical Model The NSSS response to a single reactor coolant pump rotor seizure with loss of offsite power resulting from turbine trip was simulated using the CESEC computer program described in Section C.15.0.3. The DNBR was calculated using the CETOP-D computer code (see Section C.15.0.3) whicn uses the CE-1 CHF correlation described in Reference 1 of Section C.4.4.
B. Input Parameters and Initial Conditions The ranges of initial conditions considered are given in Section C.15.0. Table C.15.3.3-2 gives the initial conditions used in this analysis. The rationale for selecting the values of the initial conditions which have a first order effect on the analysis follows. The initial conditions were chosen to minimize the initial thermal margin and thus the minimum transient DNBR.
The conditions chosen were maximum inlet temperature, minimum RCS pressure, minimum core flow rate, maximum core power, maximum radial peaking factor and a top peaked axial power distribution.
The most positive moderator temperature coefficient and the minimum available scram CEA worth tend to maximize the heat flux after a reactor trip occurs, increasing the RCS heat up. The operator initiation of plant cooldown at 30 minutes maximizes the offsite doses.
C 43
i During this event two sources of radioactivity contribute to the offsite doses, the initial activity in the steam generators and the activity associated with the assumed one gallon pu minute steam generator tube leak. The initial secondary activity is assumed to be at 0.1x Ci/gm dose equivalent I-131. The initial activity assumed to be present in the reactor coolant leaking through the steam generator tubes is 4.6n Ci/gm (see Table C.15.3.3-3).
C. Results The dynamic behavior of important NSSS parameters following a single reactor coolant pump rotor seizure with a loss of offsite power is presented in Figures C.15.3.3-1 tnrougn C.15.3.3-9.
Table C.15.3.3-1 summarizes the significant results of the event. Refer to Table C.15.3.3-1 wnile reading this section.
The single reactor coolant pump rotor seizure event results in a flow coastdown in the affected loop, a consequent reduction in flow through the core, an increase in the average coolant temperature in the core, a corresponding reduction in the margin to DNB, and an increase in the primary system pressure. A low DNBR reactor trip is generated by the core protection-calculators. The reactor trip causes a turbine trip signal to occur. The CEAs begin to drop into the core at 1.18 seconds. At this time the generator trips and the loss of offsite power occurs. The flow in the unaffected cold leg increases until the loss of offsite power occurs. At this time the flow in the unaffected cold leg begins to decrease as a result of the reactor coolant pump coastdown. The loss of offsite power also causes a loss of main feedwater and condenser inoporability. Tne turbine trip with the SBCS and the condenser unavailable leads to a rapid buildup in secondary system pressure and temperature. This increase in pressure is shown in Figure C.15.3.3-7. The opening of the main steam safety valves (MSSVs) limits this pressure increase. The maximum secondary system pressure is 1296 psia whicn is less than 1107 of design pressure.
The increasing temperature of the secondary system leads to a reduction of tne primary to secondary heat transfer.
Concurrently, the failed reactor coolant pump and the reactor coolant pump coasting down (Figure C.15.3.3-6) result in a
! decreased RCS flow which further reduces the heat transfer capability of the RCS. This decrease in heat removal from the RCS leads to an increase in the core coolant temperatures as shown in Figure C.15.3.3-4. The core coolant temperatures peak shortly after the time of reactor trip.
C - 44
Tne increase in RCS temperature leads to an increase in RCS pressure, as shown in Figure C.15.3.3-3 caused by the thermal expansion of tne RCS fluid. Tne RCS pressure reacnes a maximum value of 2379 psia at 6.8 seconds whicn is less than 110% of design pressure. After this time, tne RCS pressure decreases rapidly due to the declining core heat flux (see Figure C.15.3.3-2), in combination with tne opening of tne MSSVs. Opening of tne MSSYs limits the peak temperature and pressure of the secondary system. The MSSVs cycle until the emergency feedwater begins entering the steam generator in the unaffected loop. Emergency feedwater begins entering tne steam generator in the unaffected loop at 994 seconds, thus, ennancing the RCS cooldown and the subsequent reduction in pressure.
During the first few seconds of the transient, tne combination of decreasing flow rate, and increasing RCS temperatures results in a decrease in :ne minimun DNER. The transient minimum DNBR of 0.66 occurs at 2.8 seconds as indicated in Table C.15.3.31.
Figure C.15.3.3-8 shows the variation of tne minimum DNBR with time. The negative CEA reactivity inserted after reactor trip causes a rapid power and neat flux decrease whicn causes the DN3R to increase again. For tnis event no more than 5.0 percent of tne fuel pins are calculated to experience DNB. All fuel pins which experience DNB are conservatively assumed to fail.
The offiste doses for this event result from steam released through the main steam safety valves (MSSVs) and atmospneric dump valves (ADVs).
At 30 minutes, tne operator is assumed tc use the ADVs to begin cooldown. Table C.15.3.3-1 snows the integrated steam release from the MSSYs and the ADVs. The radiological release produced by tne transient is less severe than that presented in Section 15.3.3 of the CESSAR. The two nour thyroid innalation dose at the exclusion area boundary is less tnan 29.5 rem.
C - 35
C.15. 3.3. 3.2 Radiological Consequences A. Physical Mcdel To evaluate the consequences of tne single reactor coolant pump !
i rotor seizure with a loss of offsite power event, it is assumed that the condenser is not available for the ' entirety of the
- t ransient. For the.first thirty minutes of the event the cooldown is performed via the main steam safety valves.
! Afterwards, the cooldown is performed manually by the operator via the atmospheric dump valves. ,
, B. Assumptions, Parameters, and Calculational Methods Tne major assumptions, parameters, and calculational methods .used to evaluate the radiological consequences of the single reactor coolant pump rotor seizure are presented in Tables C.15.3.3-3 and C.15.3.3-4. Additional clarification is provided in Section 15.3.3.3.2B of the CESSAR.
C. Identification of Uncertainties and Conservatisms in the Evaluation of the Results .
The uncertainties and conservatisms in the assumptions used to evaluate the radiological consequences of the single reactor coolant pump rotor seizure with a loss of offsite power are the same as that described in Section 15.3.3.3.2C of tne CESSAR.
4 C.15.3.3.4 Conclusions The maximum RCS and steam generator pressures due to a single reactor coolant pump rotor seizure in combination with loss of j offsite power following generator trip event remain less than l
110% of tneir design values. The radiological release is less severe than that presented in Section 15.3.3.4 of the CESSAR, a
. small fraction of 10 CFR 100 guidelines. ,
i .
t 4
i C - 46
TABLE C.15.3.3-1 (Sneet 1 of 2)
SEQUENCE OF EVENTS FOR THE SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF 0FFSITE POWER RESULTiNG FROM TURBINE TRIP Total Integrated Setpoint Steam Flow Time or To Atmosphere Success (Sec.) Event Value (lbm) Path 0.0 Seizure of a Single Reactor Coolant Pump 0 0.69 Low DNBR Trip Signal 1.19 0 Reactivity Generated, projected Control 1.18 CEAs Begin to Drop 0 Reactivity Into-the Core Control 1.18 Turbine Trip / Generator 0 Trip / Loss of Offsite Power Occurs 2.8 Mnimum Transient DNBR 0.66 0 6.8 Maximum RCS Pressure, 2379 0 psia 10.9 Main Steam Safety 1282 0 Seconda ry Valves Open, System Unaffected loop, psia Integrity 12.1 Main Steam Safety 1282 476 Secondary Valves Open, Affected System Loop, psia Integrity 16.4 Maximum Steam Generator 1296 3903 Pressure, unaffected Loop, psia 16.5 Maximum Steam Generator 1288 3983 Pressure, Affected Loop 948.6 Low Water Level EFAS 20 84300 Secondary Setpoint Reached in System the Steam Generator, Integrity Unaffected Loop, Percent of Wide Range L
1 TABLE C.15.3.3-1 (Cont'd)
(Sheet 2 of 2)
SEQUENCE OF EVENTS FOR THE SINGLE REACTOR C0OLANT PUMP ROTOR SEIZURE WITH LOSS OF 0FFSITE POWER RESULTING FROM TURBINE TRIP Total Integrated Setpoint Steam Flow Time or To Atmosphere Success Event Value (lbm) Path (Sec.)
949.0 Steam Generator 1218 85527 Secondary Safety Valves Close, System Affected and Unaf- '
Integrity fected Loop, psia 993.6 Emergency Feedwater 119 85527 Secondary Begins Entering Steam System Generator, Unaffected Integrity Loop, Ibm /sec 1800.0 Abnospheric Dump -100 85527 Secondary Valves Opened to System Initiate P Integri ty Cooldown, F/ fant hour 7200.0 Total Steam Release 588,965 to Atmosphere, lbm 10321.2 Shutdown Cooling Entry 400/350 Reactor ConditionsReached,gCS Heat Pressure, psia / Temp. F Removal
. l TABLE C.15.3.3-2 ASSUMED INITIAL CONDITIONS FOR THE SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP Parameter Value Core Power Level, MWt 1908.7 Core Inlet Coolant Temperature, F 569 Reactor Coolant System Pressure, psia 2200 Steam Generator Pressure, psia 1182.7 Core Mass Flow,106 lbm/hr 78.7 Maximum Radial Power Peaking Factor 1.63 Maximum Axial Power Peak 1.34 Minimum DNBR 2.16 Doppler Coef ficient Multiplier 0.85 CEA Worth on Trip, % Ao - 10.0 (Most Reactive CEA fully withdrawn)
Moderator Temperature Coefficient b.0 v
t i TABLE C.15.3.3-3 (Sheet 1 of 3)
< PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP Parameters Value ,,, ,, ,
A. Data and Assumptions Used to Evaluate the
- Radicactive Source Term
- a. Power Level, MWt 4200 I
- b. Burnup 2 year
- c. Percent of Fuel Calculated to 5.0 ,
Experience DNB, % ap
- d. Reactor Coolant Activity 4.6 u C1/gm i Before Event (based on 4100 MWt) Table 11.1.1-2 of CESSAR.
- c. Secondary System Activity Section C.15.0.4 Before Event
- f. Primary System Liquid 525,600 Inver. tory, Ibm 1 g. Steam Generator Inventory
- Liquid, lbm per stcam generator 176,749 >
j - Steam, lbm per steam generator 13,798 B. Data and Assumptions Used to Estimate
- Activity Released from the Secondary System
- a. Primary to Secondary Leak Rate, gpm 1.0 (total)
- b. Total Mass Release Through the Main 588,965 l'
Steam Safety Valves and Atmospheric Dump Valves (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
A 4
7 ---i-t-i-s=e-- -pF- , y me-v-y -- -- w P- -y ,
r -y --e-.- --w.s-em,c- %*- e e+----,,--, -
- -y--;-+,-- w-- - , - - - - + - - - - - - - -t& e--- e
i TABLE C.15.3.3-3 (Cont'd)
(Sheet 2 of 3)
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP Parameters Value
- c. Reactor Coolant System Activity After Event, Ci Isotope I-131 9.55 (+5)
I-132 1.40 ( + 6)
I-133 1.92 (+ 6)
I-134 2.08 ( + 6)
I-135 1.79 (+6)
Kr-85M 2.40 (+7)
Kr-85 7.63 (+5)
Kr-87 4.40 (+7)
Kr-88 6.29 (+7)
Xe-131M 6.72 (+5)
Xe-133 1.93 (+8)
Xe-135 3.46 (+7)
Xe-138 1.54 (+8)
- d. Percent of Core Fission Products Refer to Section Assumed Release to Reactor Coolant C.!5.3.3.3.2B
- e. Iodine Partition Coefficient in tae 100.0 Steam Generators (Primary Liquid)
- f. Iodine Partition Coefficient in the 1.0 Steam Generators (Primary Steam) 9 Credit for Radioactive Decay in No Transit to Dose Point
- n. Loss of Offsite Power ,
Yes
, i TABLE C.15.3.3-3 (Cont 'd)
(5neet 3 of 2)
PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SINGLE REACTOR C00LAN' PbMP ROTOR SEIZURE WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP Parameters Value ,_, ,
C. Dispersion Data
- 1. Distance to Exclusion Area Boundary, m 500
- 2. Lnsta9ce to Low Population Zone 3000.0 Outer Boundary, m 3
- 3. Atmospheric Dispersion Factor, sec/m 2.00 x 10-3 D. Dose Data
- 1. Method of Dose Calculation Section C.15.0.4
- 2. Dose Conversion Assumptions Section C.15.0.4
- 3. Control Room Design Parameters See Applicant's SAR i
4 4
e I
a
i TABLE C.15.3.3-4 i SECONDARY SYSTEM MASS RELEASE I-Tf .'ME ATriOSHPHER FOR THE SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH LOS5 OF OFF5ITE POWER RESULTING FROM TURBINE TRIP EVENT Time Integrated Safety Integrated Leakage to (Sec.' Valve F1ow (1bm) Seconda ry (gall ons ) , ,, ,, , ,
?
0.0 0.0 0.00 2.0 0.0 0.03 3.0 0.0. 0.05 i
5.0 0.0 0.08 10.0 0.0 0.17 ,
i 20.0 6775.1 0.33 40.0 21646.6 0.67 60.0 21665.3 1.00
.i 120.0 29720.0 2.00
{
180.0 31545.2 3.00
^
240.0 37448.1 4.00 ;
300.0 44907.4 5.00 480.0 52807.2 8.00 i
I 600.0 63478.7 10.00 949.0* 85527.0 15.82 1800.0** 85527.0 30.00
- Main Steam Safety Valves close
- **0perator takes control of plant and begins cooldown utilizing the
- Atmospheric Dump Valves i
F e aww ,vn -1,w-wr- ew+,w--,- - , - - - - - - - - - - , , - - + - , - - - - - - - . --~-w -- c,- - - , - ,- ,---w- ,a , - - r , - , - , -- -- n --- -----s---- --
/
50 i i i i i 5 40 - -
o_
d 2
$ 30 - -
E of y20 - -
2 u
8 10 -
0 0 20b 4b0 6b0 8b0 10$0 1200 TIME, SECONDS C-E SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE Figure g WITH LOSS OF OFFSITE POWER RESULTING FROM C.15.3.3
-1 TURBINE TRIP - CORE POWER vs TIME
50 , i i i '
5 o.
--; 40 - -
2 8
o
@ 30 M'
cd
!E E 20 --
8 5
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- 0- 400 600 800 1000 1200 200 TIME, SECONDS C-E / SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE Figure gggg f[ WITH LOSS OF OFFSITE POWER RESULTING FROM C.15.3.3 TURBINE TRIP - CORE AVER AGE HEAT FLUX vs TIME -2
_ x. . ,
l' 2400 i i i i ,
\
2300 5
f 2200 N -
51 m \
E 2100 -
i %
2WO -
1900 l
t 1800 0 200 4b0 600 800 1000 1200 TIME, SECONDS C-E Figure SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE 1 ggg WITH LOSS OF OFFSITE POWER RESULTING FROM C.15.3.3
-3 TURBINE TRIP - RCS PRESSURE vs TIME
3 i i l
660 , , , , ,
O of 9s 640 - -
tie
- E W
620 -- -
8 o
U T
E5 600 - H0T -
E h' x
~
8 T AVG 580 -
T COLD 560 i i i i i 0 200 400 600 800 1000 1200 TIME, SECONDS C-E SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE Figure WITH LOSS OF OFFSITE POWER RESULTING FROM C.15.3.3 E TURBINE TRIP - CORE AVG. COOLANT TEMP's vs TIME -4
5.0 i i ,
i i i DOPPLER REACTIVITY 0 MODERATOR REACTIVITY
<l e
{'-5.0 b
o 05 m
TOTAL REACTIVITY
-10.0 -
CEA REACTIVITY 1 I i i I
-15.0 0 200 400 600 800 1000 1200 TIME, SECONDS C-E SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE Figure WITH LOSS OF OFFSITE POWER RESULTING FROM C.15.3.3 E TUR BINE TRIP - REACTIVITY vs TIME -5
/
O.50 i i i i i 0.40 - -
5 C
g 0.30 -
E Bi b
g 0.20 - -
o 0.10 -
0 0 2b0 400 600 8b0 1000 1200 TIME, SECONDS C-E SINGLE REACTOR COOLANT UPMP ROTOR SEIZURE Figure g WITH LOSS OF OFFSITE POWER RESULTING FROM C.15.3.3
-6 TURBINE TRIP - CORE FLOW FRACTION vs TIME
1300 , , , , ,
4 1280
. /
1260 f -
i E a ; j y 1240 - f f o_
5 Y I
I 3 1220 - l .
w 5
o k
m 1200 -
0; 1180 -
1 1160 - -
1 1140 i 0 2b0 4b0 6b0 8b0 1000 1200 TIME, SECONDS I
C-E SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE Figure g WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP - S.G. PRESSURE vs TIME C.15.3.3
-7
o i 2.4 i i 1 i i
- 2. 2 - -
- 2. 0
- 1. 8 - -
5 5 1. 6 5
1 '
5 1.4 - -
1 1
a
- 1. 2
- 1. 0 0.8 - -
- 0. 6 O 1 2 3 4 5 TIME, SECONDS C-E I SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE WITH C kI3.3 EM@P8 / LOSSg;0FgI"El1ggJUbfdk"hsli!E TUEINE 28
,l '
250,000 i i i i i i
- E 3 200,000 -
vi
\ AFFECTED LOOP Q 1 3 L ' -
150,000
. 5 5
~
o
$ UNAFFECTED LOOP 100,000 - -
50,000 0 200 4b0 6b0 800 1000 1200 TIME, SECONDS C-E SINGLE REACTOR COOLANT PUMP ROTOR SEIZURE Figure g WITH LOSS OF OFFSITE POWER RESULTING FROM TURBINE TRIP - S.G. WATER MASS vs TIME C.15.3.3
-9
C.15.3.4 REACTOR COOLANT PUMP S'iAFT BREAK WITH LOSS OF OFFSITE PO'ER C.15.3.4.1 Identification of Event and Causes The identification of event and causes for a single reactor coolant pumn shaft break with loss of offsite power is the same as that described in Section 15.3.4.1 of the CESSAR.
C.15.3.4.2 Sequence of Events and Systems Operations The sequence of events and systems operations is similar to that for the reactor coolant pump rotor seizure event, Section C.15.3.3. The difference is that for the shaft break event, the reactor is tripped on differential pressure across either steam generator, whereas for the pump rotor seizure event the reactor is tripped by the CPC on a projected low DNBR condition.
The flow coastdown for a rotor seizure (RS) event is faster than the coastdown for a shaft break (SB) event. For a shaft break, the rotor is still capable of rotating, thereby offering less resistance to flow during the rapid flow decrease. This results in the less severe coastdown for the shaft break ev'nt than for the rotor seizure event. The SB trip time is 1.74 seconds; the RS trip time is 1.18 seconds. At their respective rod drop times and loss of offsite power (LOP) times both RS and SB have identical flow rates. Because RS continues to have a faster ficw degradation than SB after their respective rod drop and LCP times, the RS with a LOP is conservative with respect to the SB with a LOP event. Therefore, RS with a LOP reruits in a lar6er amount of calculated fuel cladding failure.
C.15.3.4.3 Analysis of Effects and Consequences C.15.3.4.3.1 Core and System Performance The analysis of effects and consequences for this event is similar to that for the reactor coolant pump rotor seizure event, Section C.15.3.3. The SB coastdown is slightly slower and trip is later than those of the rotor seizure event. Although SB with a concurrent loss of offsite power has a later time of minimum DNBR than ES with a LOP, the R S with a LOP is more adverse.
C.15.3.4.3.2 Radiological Consequences The radiological consequences due to steam release from the secondary system are less revere than the consequences of the RS event as described in Section C.15.3.3 C.15.3.4.4 Conclusions The conclusion for the SB event is that this event would be no more adverse than the RS event. For both events the total number of fuel pins calculated in DNB, and which are conservatively assured to fail, is 5.0%. The rasultant radiological consequences are a small fraction of ~0CFR100 guidelines.
C - 47
/
C.15.4 REACTIVITY AND POWER DISTRIBUTION ANOMALIES C.15.4.1 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITION C.15.4.1.1 Identification of Event and Causes The identification of the event end causes for an uncontrolled sequential CEA withdrawal at low power are the same as that described in Section 15.4.1.1 of CESSAR.
C.15.4.1.2 Sequence of Events and System Operation The sequence of events and system operation are similar to that described in Section 15.4.1.2 of CESSAF. The different tirnings and parameter values of the sequence of events for the limiting CEA withdrawal tran:ient at low power (1 MWt) are snown in Table C.15.4.1-1.
C.15.4.1.3 Analysis of Effects and Consequences A. Mathematical Model The Nuclear Steam Supply System (NSSS) response to a CEA sequential withdrawal from subcritical or low power conditions was simulated using the CESEC computer program described in Section C.15.0. The thermal margin to DNB in the reactor core was calculated using the CETOP-D computer program described in Section C.15.0.
B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analyze the NSSS response to a CEA sequential withdrawal from subcritical or low power conditions are discussed in Section C.15.0. Those parameters whicn were unique to this analysis are listed in Tabl e C.15.4.1-2.
The input parameters were cnosen to produce the closest approach to the fuel design limits for a CEA withdrawal from low power level as described in Section 15.4.1:3B of CESSAR. The most adverse initial conditions allowed by the LCO's during part loop
- opegation were found to be zero power, core inlet temperature of l 569 F core inlet flow of 50% of four-pump design flow, and minimum RCS pressure of 2200 psia.
The initial core average axial power distribution assumed in the analysis corresponds to an axial shape index (ASI) of -0.64. A t
one pin radial peaking factor of 1.77, including uncertainties, was assumed.
l C - 48 i
The reactivity insertion maximum rate of 5.0 x 10 gate used in the analysis was theap /sec, i.e. ,
motion. This reactivity insertion rate is calculated based on an initial insertion of only the lead regulating bank.
C. Results ,
The dynamic behavior of important NSSS parameters following a CEA withdravel from low power conditions is presented in Figures C.15.4.1-1 through C.15.4.1-7.
The withdrawal of CEA's from low power conditions (1 MWt power) adds reactivity to the reactor core, causing both the core power level and the core heat flux to increase. The power transient causes increasing temperature and pressure transients which, togather with a top peaked axial power distribution, produce the closest approach to the specified acceptable fuel design limit on DNBR. At 81.0 seconds into the transient, a variable overpower trip is actuated. The CEA's begin dropping into the core at 81.9 seconds. If the maximum rod radial peaking factor occurs in the region of the axial power peak, the peak linear heat generation rate during the transient would reach 4.4 KW/ft.
C.15.4.1.4 Concl usions The uncontrolled CEA withdrawal from a subcritical or low power condition event meets general design criteria 25 and 20. These criteria require that the specified acceptable fuel design limits are not exceeded and the protection system action is initiated automatically. The withdrawal of-CEA's from low power conditions meets the following fuel design limits which serve as the acceptance criteria for this event: the transient terminates with a hot channel minimum DNBR greater than or equal to 1.19 and the peak linear heat generation rate during the transient is less than 21 KW/ft.
C - 49
l
- TABLE C.15.4.1-1 SEQUENCE OF EVENTS FOR THE '
SEQUENTIAL CEA WITHDRAWAL FROM LOW POWER Setpoint . Success Time Event or Value Path 0.0 Withdrawal of CEA's -- Reactivity Control 1
Initiating Event
- 81.0 Variable Overpower Trip, 17.0 Reactivity Control
% of Design Power +
81.5 CEDM Power Supply -- Reactivity Control Breakers Open 81.9 CEA's Begin to-Drop -- Reactivity Control l
83.3 Maximum Core Power, 22%
% of Design Power
! 84.0 Maximum Core Average 18%
' Heat Flux, % of Full Power Heat Flux 84.0 Minimum DNBR 4.6 85.8 Maximum Pressurizer 2358 Pressure, psia f
4 4
l .
i 4
I i
1 4
I
7, i
TABLE C.15.4.1-2 ASSUMPTIONS AND INITIAL CONDITION FOR THE LOW POWER CEA WITHDRAWAL ANALYSIS Parameter Val ue Initial core power level, MWt 1 Core inlet coolant temperature, OF 569 6
Core mass flowrate,10 lb,/h 78.7
' Reactor coolant system pressure, psia 2200 One pin radial peaking factor, with uncertainty 1.77 Steam generator pressure, psia 1217.
Moderator temperature coefficient, 10-4Ap/ F +0.5 Doppler coefficient multiplier .85 CEA reactivity addition rate, 10-5 A /sec. 5.0-CEA Worth on trip, % Ap -3.6 Steam bypass control system Automatic O
4 60 i i i i 50 40 -
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0 0 25 50 75 100 125 l
TIME, SECONDS C-E f Fi vre SEQUENTIAL CEA WITHDRAWAL AT LOW POWER Eff,8PLF /j CORE POWER vs TIME C.1 4.1 i
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60 , , , ,
50 40 -
5 M ..
y
>( 30 -
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20 8
10
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0 0 25 50 75 100 125 TIME, SECONDS
/ Fi ure 9- SEQUENTIAL CEA WITHURAWAL AT LOW POWER SfSfc,./ / CORE AVERAGE HEAT FLUX vs 'iiME C.1 .,4. 1
~
2500 , , , ,
2400 2300 - _
{
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1900 0 25 50 75 100 125 TIME, SECONDS C-E Figure
/ SEQUENTIAL CEA WITHURAWAL AT LOW POWER C.15 4 1 f[898 / REACTOR COOLANT SYSTEM PRESSURE vs TIME
I 70 i i i i 4
60 - -
4 50 5 -
5 40 - -
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3 5
1 30 1
20 10 -
0 0 25 50 75 100 125 TIME, SECONDS C-E / F'gure SEQUENTIAL CEA WITHDRAWAL AT LOW POWER C. .4.1 gggPLE /[ MINIMUM DNBR vs TIME
- . - e s ,
610 i i i i 6W - -
O g 590 - -
5 3 CORE OUTLET w
E 580 - - -
W CORE AVERAGE 570 - -
\
CORE INLET l
560 - -
1 i
550 0 25 50 75 100 125 TIME, SECONDS
~
I Figure C-E / SEQUENTIAL CEA WITHDRAWAL AT LOW POWER C.1 .4.1
'l Ef25P8 / _ CORE AVERAGE COOLANT TEMPERATURES vs TIME
1500 i i i i 5
m c- 1400 - _
ui 5
YA u.i 8 1300 - -
a:
c:
$ 1200 1
s W
m 1100 1000 0 25 50 75 100 125 TIME, SECONDS l
c-e ""*
/ SEQUENTIAL CEA WITHDRAWAL AT LOW POWER C.15 4.1 Blff8 / / STEAM GENERATOR PRESSURE vs TIME L
l 20 i i i i 16 tr '
E x
N g 12 5
P f5 '
$8 - -
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E W 4 - -
a 0
0 25 50 75 100 125 TIME, SECONDS f
C-E / "S" EfMfif /! SEQUENTIAL CEA WITHDR AWAL AT LOW POWER LINEAR HEAT GENER ATION RATE vs TIME C.154.1
C.15.4.2 UNCONTROLLED CONTROL ELEMENT ASSEMBLY WITHDRAWAL AT POWER C.15.4.2.1 Identification of Event and Causes The identification of the event and causes of an uncontrolled CEA withdrawal at power are the same as that described in Section 15.4.2.1 of CESSAR.
C.15.4.2.2 Sequence of Events and System Operation The sequence of events and system operation are similar to that described in Section 15.4.2.2 of the CESSAR. The di fferent timings and parameter values of the sequence of events for the limiting CEA withdrawal transients at power are shown in Table C.15.4.2-1.
C.15.4.2.3 Analysis of Effects and Consequences A. Mathematical Model The Nuclear Steam Supply System (NSSS) response to a CEA sequential withdravel from power conditions was simulated using the CESEC computer program described in Section C.15.0. The thermal margin on DNBR in the reactor core was calculated using the CETOP-D computer program described in Section C.15.0.
B. Input Parameters and Initial Conditions The input parameters and initia'l conditions used to analyze the NSSS response to a CEA sequential withdrawal from power conditions are discussed in Section C.15.0. Those parameters which were unique to this analysis are listed in Table C.15.4.1-2.
The input parameters were chosen to produce the closest approach to the fuel design limits for a CEA withdrawal from power as described in Section 15.4.2.3 of CESSAR. The most adverse initial conditions alloked by the LC0's during part loop operation vere found to ge a core power of 50% of 3800 MWt, core inlet temperature of 569 F core inlet flow of 50% of four-pump design flow, and minimum RCS pressure of 2200 psia.
The initial crre average axial power distribution assumed in the analysis corresponds to an axial shape index (ASI) of -0.2. A one pin radial peaking factor of 1.63, including uncertainties, was assumed.
The reactivity insertion i
maximum rate of 5.v x 10~ gate used ao/sec, in the.01%
i.e., analysis acperwas inchthe of rod
( motion. This reactivity insertion rate is calculated based on an l
initial insertion of only the lead regulating bank.
l l
l 1
C - 50
C. Results The dynamic behavior of important NSSS parameters following an uncontrolled CEA group withdravel are presented in Figures C.15.4.2-1 to C.15.4.2-7.
The withdrawal of CEA's causes a positive reactivity change,_
resulting in an increase in the core power and heat flux. As a consequence, the reactor coolant temperature and pressurizer pressure increase. At 14.8 seconds after initiation of the transient, a variable overpower trip is actuated. The CEA's begin dropping into the core at 15.7 seconds which terminates the transient. The minimum DNBR reached during the transient is 1.58 at 17.3 seconds. If the maximum rod radial peaking factor occurs in the region of the axial power peak, the peak linear generation rate during the transient reaches 8.4 KW/ft.
C.15.4.2.4 Concl usions The uncontrolled CEA withdrawal from power event meets general design criteria 25 and .20. These criteria require that the specified acceptable fuel design limits are not exceeded and the protection system action is initiated automatically. The withdrawal of CEA's from power conditions meets the following fuel design limits which serve as the acceptance criteria for this event: the transient terminates with a hot channel minimum DNBR greater that or equal to 1.19 and the peak linear heat generation rate during the transient is less than 21 KW/ft.
C - 51
l l
TABLE C.15.4.2-1 SEQUENCE OF EVENTS FOR THE SEQUENTIAL CEA WITHDRAWAL EVENT Setpoint success Time Event or Value Path 0.0 Withdrawal of CEA's - -- Reactivity Control Initiating Event 14.8 Variable High Power Trip 58% of 3800 Reactivity Control Signal Generated MWt 15.4 CEDM Power Supply -- Reactivity Control Breakers Open 15.7 CEA's Begin to Drop -- Reactivity Control 16.6 Maximum Core Power, 59%
% of Design Power 17.3 Minimum DNBR 1.58 17.3 Maximum Core Average 58%
Heat Flux, % of Full Power Heat Flux 18.3 Maximum Pressurizer 2308 Pressure, psia
TABLE C.15.4.2-2 ASSUMPTIONS AND INITIAL CONDITIONS FOR THE SEQUENTIAL CEA WITHDRAWAL Parameter Val ue Core Power Level , $Mt 1908.7 Core Inlet Coolant Temperature, F -569 Reactor Coolant System Pressure, psia 2200 Steam Generator Pressure, psia 1182.7 Core Mass Flow,106 lbm/hr 78.7 Maximum Radial Power Peaking Factor 1.63 Maximum Axial Power Peak 1.34 Initial Minimum DNBR 2.16 Doppler Coefficient Multiplier 0.85 CEA Worth at Trip, % AP - 10.0 Reactivity Insertion Rate, 10-4ap/sec. 0.5 CEA Withdrawal Speed, inches / min 30.0 i
60 , , , ,
~
50 n
d 40 - -
5 o_
ci g 30 - -
u 8
20 - -
10 l
l l l 1 0
0 10 20 30 40 50 TIME, SECONDS l
l l
- C-E / Figure SEQUENTIAL CEA WITHDRAWAL AT POWER gfggp[# /[ CORE POWER vs TIME C.15.4.2
/
60 , , , ,
50 40 -
m o_
x' 3
[ 30 -
5 x
8 20 -
10 ,
l I ' ' '
0 0 10 20 30 40 50 TIME, SECONDS 4
C-E Figure SEQUENTIAL CEA WITHDR AWAL AT POWER CORE AVER AGE HEAT FLUX vs TIME C.15.4.2 E l -2
/
2500 I I g 2400 -
E 2300 us 5
y 2200 - -
m 5$
2100 - _
2000 I i ,
1900 0 10 20 30 40 50 TIME, SECONDS C-E SEQUENTIAL CEA WITHDRAWAL AT POWER Figure BE REACTOR COOLANT SYSTEM PRESSURE vs TIME C.15.4. 2
-3 i-
1 2.4 i i i
~
2.2 -
2.0 - -
5 E 1.8
- s 5
s s
1.6 - -
1.4 - -
- 1. 2 1.0 O 10 20 30 40 50 TIME, SECONDS C-E SEQUENTIAL CEA WITHDRAWAL AT POWER Fi re g MINIMUM DNBR vs TIME C. .1 s
/
640 i i i i 630 OUTET
, 620 - -
610 - -
O uf 5
{E 600 -f AVERAGE
- E W
590 -
580 -
570 -
INLET 560 0 10 20 30 40 50 TIME, SECONDS C-E Figure SEQUENTIAL CEA WITHDRAWAL AT POWER g CORE AVERAGE COOLANT TEMPER ATURES vs TIME C.15.4. 2
-5
e m wpumi- mV-e W*e=w -
/
l l l l 1800 a.
LL.
E h
1600
!E 5
ti g 1400 E
o
$ 1200 - -
1000 I I I '
800 0 10 20 30 40 50 TIME, SECONDS SEP8 // SEQUENTIAL CEA WITHDRAWAL AT POWER STEAM GENERATOR PRESSURE vs TIME C 15 k2
_g
28 , , , ,
24 t- 20 -
E x
uf E
x z 16 o
p b
g 12 W
E u
~ ~
8 E
a.
4 - _
' i ' '
0 0 10 20 30 40 50 TIME, SECONDS i
C-E S " *
/ SEQUENTIAL CEA WITHDRAWAL AT POWER C.15.'4.2 Ef26P8 / I PEAK LINEAR HEAT GENERATION R ATE t _..
i C.15.4.3 SINGLE FULL LENGTH CONTROL ELEMENT ASSEMBLY DROP C.15.4.3.1 Identification of Event and Causes The identification of event and causes for the single full length CEA drop are the same as that described in Section 15.4.3.1 of tne CESSAR. The CEA drop for part loop operation will not approach tne DNBR criterion of 1.19.
C.15.4.3.2 Sequence of Events and Subsystems The sequence of events and system operation are the same as that described in Section 15.4.3.2 of the CESSAR. However, the minimum DNBR during the CEA drop for part loop operation will not approach the criterion of 1.19. Rather, it will be held to a substantially higher value even withcut a reactor trip.
C.15.4.3.3 Analysis of Effects and Consequences The dynamic behavior of important NSSS parameters following the drop of a single full length CEA from part loop conditions is similar to that presented in Figures 15.4.3-1 to 15.4.3-12 of the CESSAR.
The selection of the power distribution penalty factors upplied t
by the CEA calculators for part loop operation will provide substantially greater conservatism than do those for full loop operation. Thus, the majority of CEA drop events will result in immediate trips and no decrease in the initial margin to the SAFDLs on DNBR and fuel centerline melt. In addition, the larger values provided for these penalty factors will assure that even those CEA drops that do result in an immediate trip will not approach as close to the SAFDLs as did tne CEA drop described in Section 15.4.3 of CESSAR. Thus, the CEA drop for part loop operation is less limiting than that presented in Section 15.4.3 of CESSAR.
C.15.4.3.4 Concl usions The drop of a single full length CEA me ts general design criteria 25 and 20. These criteria require that the specified acceptable fuel design limits are not exceeded and tne protection system action is initiated automatically. The drop of a CEA meets the following fuel design limits which serve as the acceptance criteria for this event: the transient terminates with a hot channel minimum DNBR greater than or equal to 1.19 and the peak linear heat generation rate during the transient is less than 21 kw/ft.
C.15.4.4 STARTUP 0F AN INACTIVE REACTOR COOLANT PUMP The Startup of an Inactive Reactor Coolant Pump event was not ar.elyzed since measures will be taken to assure that the two inactive reactor coolant pumps cannot be restarted.
C - 52
~
C.15.4.3 SINGLE FULL LENG E CONTROL ELEMENT ASSEFELY DROP C.15.4.3.1 Identification of Event and Causes i
The identification of event and causes for the single full length:
CEA drop are .the same as that described -in section 15.4.3 1 of the CESSAR. The CEA drop for part loop operation will not approach the DNBR criterion of 1.19
, C.15.h.3.2 Sequence of Events and Systems Operation The sequence of events and system operation are the'same as that described in section 15.4.3.2 of the CESSAR. However, th?
minimum DNBR during the CEA drop for part-loop operationLaill not approach the criterion of 1.19. Rather, it vill be held to a i substantially higher value even without a reacter trip.
C.15.4.3.3 Analysis of Effects and Consequences I- The dynamic behavior of important NSSS parameters following the
] drop of a single full length CEA from part loop conditions is similar to that presented in Figures 15.4.3-1 to 15.4.3-12 of the CESSAR.
The selection of the. power distribution penalty factors supplied by the CEA calculators for part-loop operation will provide substantially greater conservatism than do those for full loop operation. Thus the majority of CEA drop events will result in 3
immediate trips and no decrease in the initial margin to the SAFDL's on DNBR and fuel centerline melt. In addition, the larger values provided for these penalty factors will assure that even those CEA drops that do not result in an immediate trip will-not approach as close to the SAFDL's as did the CEA drop 3
! described in Section 15.4.3 of CESSAP. Thus the CEA drop for part-loop operation is less limiting than that presented in section 15.4.3 of CESSAR.
! C.15.4.3.4 Conclusions l The drop of a single full length CEA meets general design -
l criteria 25 and 20. These criteria require that the specified i acceptable fuel design limits are not exceeded and the protection system action is initiated automatically. The drop of a CEA
- meets the following fuel design limits which serve as the
! acceptance criteria for this event: the transient terminates with a hot channel minimum DMDR greater than or equal to 1.19 and the peak linear heat generation rate during the transient is less i than 21 KW/ft.
f C.15.4.4 STARTUP OF AN INACTIVE REACTOR COOLANT PUMP l The Startup of an Inactive Reactor Coolant Pump event was not
- analyzed since the two inactive reactive coolant pumps will be disconnected from elactrical power supplies by redundant methods.
C - 53
s; -
/
C.15.4.5 FLOW CONTRCLLER MALFUNCTION CAUSING AN INCREASE IN BWR CORE FLOW RATE This event is not applicable to Pressurized Water Reactor and therefore is not included here.
C.15.4.6 INADVERTENT DEBORATION C.15.4.6.1 Identification of Event and Causes The identification and causes of an Inadvertent Deboration (ID) are the same as that described in the section 15.4.6.1 of the CESSAR.
C.15.4.6.2 Sequence of Events and System Coeration The sequence of events and system operation are the same as that described in the section 15.4.6.2 of the CESSAR.
C.15.4.6.3 Analysis of Effects and Consequences The analysis of effects and consequences is the same as that presented in the section 15.4.6.3 of the CESSAR because the factors or parameters affecting the boron concentration in the RCS remain unchanged during part loop operation.
C.15.4.6.4 Conclusions The inadvertent deboration event during part loop operation will be no more limiting than the same event discussed in the section 15.4.6 of the CESSAR. The ID event during part loop operation will result in acceptable consequences with sufficient time available for the operator to detect and to terminate the event if it occurs.
C.15.4.7 INADVERTENT LOADING CF A FUEL ASSEVELY INTO THE IMPROPER PCSITION C.15.4.7.1 Identification of Events and Causes The identification and causes of'an Inadvertent loading of a Fuel Assembly into the Improper Position are the same as described in section 15.4.7.1 of the CESSAR.
C.15.4.7.2 Sequence of Events and System Oneration The sequence of events and system operation are the same es described in section 15.4.7.2 of the CESSAR.
C.15.4.7.3 Analysis of Effects and Consequences The analysis of effects and consequences are no more adverse than those presented in section 15.4.7.3 of the CESSAR. The minimon i DNBR of 1.19 will not be approached with no clad failure expected l to occur.
l C - 54
C.15.4.7.4 Conclusion Those Inadvertent Loading of a Fuel Assembly into the Improper Position Events which are not detected during part-loop operation do not result in fuel cladding failure and hence radiological releases. Therefore the radiological consequences are within 10CFR100 guidelines.
C.15.4.8 CCUTROL ELEMENT ASSEMBLY (CEA) EJECTION C. 15.4.8.1 Identification of Event and Causes The identification of event and causes for CEA Ejection is the same as that described in Section 15.4.8.1 of the CESSAR.
C.15.4.S.2 Sequence of Events and Systems Cperation The sequence of events and systems operation for the part loop CEA ejection with loss of offsite power event is similar to that described in Section 15.4.8.2 of the CESSAR.
C.15.4.3.3 Analysis of Effects and Consequences A. Mathematical Model The mathematical model used to evaluate the CEA ejection with lcss of offsite power is the same as that described in Section 15.4.G.3.A of the CESSAR.
B. Input Paraneters and Initial Conditions The ranges of initial conditions considered are given in Section C.15.0. Table C.15.4.8-1 contains assumptions regnrding the initial reactor states analyzed for this event. The initial conditions and process varinbles were varied as described in Section 15.4.8.3 of the CESSAR. The parameters were varied in an attempt to reach a COLSS power operating limit. The initial conditions chosen were those that resulted in the mininum margin to DNB.
C. Results The spectrum of initial reactor states contained in Table C.15.4.8-1 was analyzed to show that each case met the crite~ia established in Regulatory Guide 1.77. All cases resulted na radial average fuel enthalpy less than 280 cal /gran at the hottest axial location of the hot fuel pin. The case that resulted in the greatest potential for offsite dose consequences (i.e., the case resulting in tha largest number of postulated fuel failures) was identified as the case initiated from maximum power (MP) beginning-of-cycle (DCC) initial conditions. Refer to Table C.15.4.8-2 for the initial conditions and assumptions used for this analysis.
C - 55
The dynamic behavior of the important core parameters resulting from the limiting CEA Ejection with Loss of Offsite Power case is illustrated in Figures C15.4.B-1 to Figures C15.4.8-6. Tcble C.15.4.3.2-3 summarizes the maxirum values that occur for core power, clad surface temperature, and fuel centerline
-temperature.
The maximum RCS and pressurizer pressures are less than in Section 15.4.8 3 of the CESSAR. The pressure is less for the following reasons. One, the increase in core average heat flux is less than in Section 15.4.8.3 of the CESSAR. Two, the heat rejection that occurs following turbine' trip from part-loop operation is less than the heat rejection that would occur following turbine trip for full power operating conditions as in Section 15.4.8.3 of the CESSAR.
Following a postulated CEA ' Ejection Event, 8.8" of the fuel-is calculated to experience DNB. Regulatory Guide 1.77 recommends that the onset of DMB be used as the basis for predicting clad failure. C-E does not equate onset of DN3 with cladding failure. Nevertheless, this criterion was used to determine.the percentage of pins that suffer clad failure.
The radiological consequences for this event are less than the CEA Ejection event described in SectionL15.4.8 of the CESSAR.
This is primarily due to the lower amount of failed fuel.
C.15.4.8.4 Conclusions The rupture of a CEDM nozzle or housing and the subsequent ejection of a CEA will not result in a radial average fuel enthalpy greater than 230 cal /gm at any axial location in any fuel rod.
The peak RCS pressure is less than the peak RCS pressure calculated for the CEA Ejection event in Section 15.4.8 of the CESSAR. Therefere the peak RCS pressure is less than the Service Limit C value as defined in the ASME code.
The radiological consequences for this event are less than that presented in Section 15.4.8 of the CESSAR. Therefore the radiological consequences are less than the guidelines set forth in 10CFR Part 100.
C - 56
Table C.15.4.8-1 INITIAL REACTOR STATES CONSIDERED FOR THE CEA EJECTION EVENT Initial Reactor Power Bank 5 Insertion (Mwt) Depth (%)
1900 15 1425 24 950 33 475 42 0 60
/
Table C15.4.8-2 ASSUMPTIONS USED FOR THE CEA EJECTION ANALYSIS MAXIMUM POWER BEGINNING 0F CYCLE INITIAL CONDITIONS Parameters Assumption Initial Core Power Level, Mwt 1900 Delayed Neutron Fraction, S .00730
" Moderator Temperature Coefficient Section 15.0 Most Positive value Ejected CEA Worth, % ap 0.15 Doppler Weighting Factor, 1.0 Initial Three-Dimensional Fuel 2.37 Pin Peaking Factor Ejected Three-Dimensional Fuel 2.95 Pin Peaking Factor Total CEA Worth Available for -2.94 Insertion on Reactor Trip. % ap Postulated CEA Ejection Time, sec 0.05.
Core Inlet Coolant Temperature,*F 569 Core Mass Flow Rate,106 lbm/hr 78.1 ,
Reactor Coolant System Pressure, psia 2200
Table C.15.4,8 3
SUMMARY
OF RESULTS FOR THE CEA EJECTION EVENT Parameter Val ue leximum Core 64.8 Power, % of
[esign Power Maximum Clad Surface 852 Temperature in the Hot Node, F Maximum Fuel Centerline 1857 Temperature in the Hot Node, F
% of Fuel Pins 8.8 Calculated to Experience DNB 4
~
70 , , , ,
60 - _
5 E
2
--j 50 -
5 g 40 - -
u 5
o.
_ _ 30 - -
5 s
2 g 20 - -
o o
i 10 -
l 0 0 1 2 3 4 5 TIME, SECONDS C-E F* ' S "
'/ CEA EJECTION SfEfd> / / CORE POWER vs TIME C.15.4.8
-1
'j S
f f
8C P-y.8 E &E 8u 235 SMD boE~ B 2g 235
/
R Ew 235 uz7C
/
/ -
1 1 1 1 1 2 2 4 6 8 0 2 4 6 8 0 0 0 0 0 0 0 0 0 0 0 0 - - - - - - -
P E
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o 0 ' ' ' '
- 0 1 2 3 4 5 l TIME, SECONDS C-E / Figure
/ CEA EJECTION SfMP8v / / CORE AVERAGE HEAT FLUX vs TIME C.15.4.8
-3
180 , , , ,
s h 160 -
E d
2 140 8
ox 120 -
53 o_ u-I--
'6 3I u.w 100 EO um dQ 80 -
z z
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o - -
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z 40 - -
20 ' ' ' '
0 1 2 3 4 5 l TIME, SECONDS C-E i CEA EJECTION Figure gggg,,f [ PEAK HOT CHANNEL HEAT FLUX vs TIME C.15.4.8
-4
r_
MAXIMUM HOT CHANNEL
--- MAXIMUM AVERAGE CHANNEL 2000 , , , ,
5 1800 5 n a:
E 1600 s
W FUEL CENTERLINE a
5 1400 o e a a 5
d 1200 - -
g ' ,
J, -
~ s a
w N s 5 1000 _ FUEL RADIAL AVERAGE _
s - -
a _ -- s U "
h
< 800 - -
CLAD SURFACE Q
l 5 600 - -
! Ei t
x .
400 0 1 2 3 4 5 TIME, SECONDS
~~
C-E / CEA EJECTION Figure ggggt, f [ HOT AND AVERAGE CHANNEL FUEL AND CLAD TEMPERATURES vs TIME C 154.8
-5
i
- 0. 2 , , , ,
EJECTED CEA 0.1 1 TOTAL I
I l /
0
! MODERATOR .
./
g '%.w __
N
,_. DOPPLER z s b
e
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N
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E SHUTDOWN CEA's \
\
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x
-1.0 -
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\
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\
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-2.0 \
1
-3. 0 0 1 2 3 4 5 i TIME, SECONDS l
l C-E I / Figure CEA EJECTION Efl5FJ., / / REACTIVITY vs TIME C.15.4.8
-6
.]
l C.15.5 IUCREASE IU RCS I!!VENTORY
- j. C.15.5.1 INADVERTE!!T CPERATIOM OF THE ECCS
{ C.15.5.1.1 Identification of Event and Causes The inadvertent operation of the emergency core coding system (ECCS) identification and causes are the same as that described in section 15.5.1.1 of the CESSAR.
. C.15.5.1.2 Sequence of Events and Systems 0;eration The sequence of events and. systems operation are the same as that described in section 15.5.1.2 of the CESSAR.
C.15.5.1.3 Analysis of Effects and Consequences ,
t
~
i The analysis of effects and consequences is the same as that 4 ~
presented in section 15.5.1.3 of the CESSAR. 1 C.15.5.1.4 Conc 1usion
_ The peak pressurizer pressure reached during the inadvertent operation of the ECCS is well within 1107, cf design pressure. .
Additionally, the pressure-temperature limits for brittle fracture of the RCS are not violated by this transient. The fuel i
' ~
integrity is not challenced by this event.
C.15:5.2 CVCS MALFUNCTION - PRESSURIZER LEVEL C0:ITROL SYSTEM MALFU!!CTICN WITH LCSS OF OFFSITE PC'ER
- C.15.5.2.1 Adentification of Event and causes i
The. identification and causes of a Pressurizer Level Control
- System (PLCS) malfunction are the same as described in the i - section 15.5.2.1 of the CESSAR.
- j. .
l C.15.5.2.2 Sequence of Events and System Cperation l .,Thesequenceofeventsandsystcboperationaresimilartothose i
described in section 15.5.2.2 of the CESSAR. Due to the 1 difference in initial conditions defined for four pump operation
! and two pump operation, a reactor trip will occur earlier in the i
s event for part-loop operation. Section C.15.5.2.3 discusses the j affect of the initial conditions on the transient response.
C.15.5.2.3 Analysis of Effects and Consequences i
,- The analysis of effects and consequences will be similar to that given in the section 15.5.2.3 of the CESSAR with the consequences for part-loop operation less severe. For part-loop operation the initial. conditions are restricted to n pressurizer pressure of l
,, 2200 psia and an initial press 2rizer water volume of 55% of the 3 .otal volume (see Table C.15.0-5). Where as, for the analysis in C - 57
r -- - .~
section 15.5.2.3 of the CESSAR the value of initial pressurizer pressure is 1785 psia and the value of initial pressurizer water volume is 60% of total volume. The higher initial pressure for part-loop operation results in smaller time to reactor trip on high pressurizer pressure, and lower pressurizer watervolume results in less RCS inventory in the pressurizer prior to trip.
Therefore, the overall effect on the results of the analysis for the part loop operation will be less severe than in section 15.5.2 of CESSAR.
C.15.5.2.4 conclusion The PLCS malfunction event during part-loop operation will be less limiting than the same event discussed in the section 15.5.2 of the CESSAR. The PLCS malfunction event during part-loor operation will result in acceptable consequences and meets the acceptance criteria for system pressure and fuel performance.
9 C - 58
r -
1 C.15.6 DECREASE IM REACTOR COCLAMT SYSTEi INVEMTORY C.15.6.1 IMADVERTENT OPEMIMG CF A PRESSURIZER SAFETY / RELIEF VALVE The Inadvertent Opening of a Pressurizer Safety Valve Event as described in SRP 15.6.1 is evaluated in the EmerCency Core Cooling Systems analyses (section C.6.3).
C.15.6.2 DOUELE ENDED BRE!.K CF A LETDC*al LINE OUTSIDE CONTAIUMENT C.15.6.2.1 Identification of Event and Causes The identification of event and causes for a double ended break of a letdown line outside containment, upstream of the letdown control valve (DELLOCUS) during part-loop operation are the same as those described in the section 15.6.2.1 of the CESSAR.
C.15.6.2.2 Sequence of Events and Systems Operation The sequence of events and systems operation are the same as those described in Section 15.6.2.2 of the CESSAR with the following exception. Because of the restrictions en the initial conditions for part-loop operation (Table C.15.6.2-1), the time at which the third charging pump starts and the pressurizer backup heaters are turned on will not be exactly the same. For example, the pressurizer backup heaters, for the part-loop analysis, will be turned on at 2260 psia instead of 2360 psia as was shown in Table 15.5.2.2 of CESSAR. This will result in a different time of actuation of the backup pressurizer heaters for part-loop operation since the lower break flow resulting from the lower initial RCS pressure will cause a slower decrease in pressurizer pressure relative to the analysis of Section 15.6.2 of CESSAR.
C.15.6.2.3 Annlysis of Effects and Consequences C.15.6.2.3.1 Core and Systen Performance A. Mathematical Model The mathematical model is the same as that described in Section 15.6.2.3.1 of CESSAR.
B. Input Paramters and Initial Conditions Table C.15.6.2-1 compares the initial conditions for the analysis presented in CESSAR section 15.6.2 with those for part-loop operation. The major differences are in the initial core . cower level, core mass flow, and pressurizer pressure.
C. Results The dynanic behavior of important USSS parameters following a DELLOCUS during part-loop operation is similar to the behavior presented in Figures 15.6.2-2 to 15.6.2-14 of CESSAR. The lower C - 59
1 power level and core mass flow rate have very little effect on the amount of primary fluid released through the break in the
- letdown line. The lower initial .RCS pressure will result in less integrated mass release from the primary system than was reported in Section 15.6.2 of CESSAR.
Ten minutes into the transient, the operator isolates the letdown
, line. This action terminates the loss of primary system mass.
C.15.6.2.3.2 Radiological Consequences A. Phthematical Model The mathematical model is tne same as that used in Section 15.6.2.3.2 of the CESSAR.
! B. Assumptions and Parameters The assumptions and parameters are identical to those presented in Section 15.6.2.3.2 of the CESSAR.
C. Results 1
The radiological consequences are no more severe than those presented in Section 15.6.2.3.2 of the CESSAR.
C.15.6.2.4 Concl usions The double-ended break of a letdown line outside containment upstream of the letdown line control valve results in gradual depressurization of the reactor coolant system. The departure from nucleate boiling ratio (DNBR) does not approach the value of 1.19 throughout the transient. Hence, no fuel pins are calculated to experience DNB for this event.
During the 600 second duration of the transient, no more than 30,766 pounds of primary system coolant is released outside the containment. This results in no more than 23.7 rem for the two hour thyroid inhalation dose at the exclusion area boundary.
This value is significantly less than the 10 CFR 100 guideline dose value of 300 rems.
i C - 60
. . - - . . ~ , , . , . . - - . - _ - . . - , . - - -_ - -
r-l TABLE C.15.6.2-1 ASSUMED INPUT PARAMETERS AND INITIAL CONDITIONS FOR THE DOUBLE-ENDED BREAK OF THE LETDOWN LINE OUTSIDE CONTAINMENT UPSTREAM OF THE LEIDOWN LINE CONTROL VALVE-CESSAR Part Loop Parameter Assumed Val ue, Assumed Value Core Power Level, MWt 3876 1909 Core Inlet Temperature,'F 580 570 Pressurizer Pressure, psia 2400 2300 Core Mass Flow,106 lbm/hr 153 78.9 Pressurizer Liquid Volume, ft 3 1116 1023 Steam Generator Pressure, psia 1206 1206 Doppler coefficient Multiplier 1.15 1.15 CEA Worth at Trip, % Ap (most -10.0 -10.0 reactive CEA fully withdrawn)
Break Size (double-ended), ft 2 0.01556 0.01556 2
w f
n i
4 1
r C.15.6.3 STEAM GENERATOR TUBE RUPTURE C.15.6.3.1 Steam Generator Tube Rupture Witnout a Concurrent Loss of Offsite Power C.15.6.3.1.1 Identification of Event and Causes The identification of event and causes for a steam generator tube rupture (SGTR) without a concurrent loss of offsite power during part loop operation are the same as that described in Section 15.6.3.1.1 of the CESSAR.
C.15.6.3.1.2 Sequence of Events and Systems Operation The sequence of events and systems operation are the same as those described in Section 15.6.3.1.2 of the CESSAR with the following exception. Because of the restrictions on the initial conditions and setpoints (see Table C.15.6.3-1) for part loop operation, the events presented in Table 15.6.3-1 of the CESSAR will be shifted with respect to the times shown in the table. In particular,the lower initial RCS pressure and the higher low pressurizer pressure trip '
setpoint during part loop operation will result in the reactor tripping at an earlier time than for the case presented in CESSAR Section 15.6.3.1.
C.15.6.3.1.3 Analysis of Effects and Consequences C.15.6.3.1.3.1 Core and System Performance A. Mathematical Model The mathematical model is the same as that described in Section 15.6.3.1.3.1 of the CESSAR.
B. Input Parameters and Initial Conditions Table C.15.6.3-1 compares the assumptions and initial conditions for part loop operation with those presented in Section 15.6.3.1 of the CESSAR. The major differences are in the initial core power levdl, initial core mass flow rate, the initial pressurizer pressure and the low pressurizer pressure trip setpoint. Additionally, tne secondary side pressure in the steam generators for part loop operation is, in general, nigher than that for full power operation. This is a direct result of the lower core power (i.e., lower neat transfer rate in the steam generators) and the lower RCS flow rate for part loop operation. However, in the present analysis, the assumption made is that the steam generator pressures for the two cases are equal.
Tnis is a conservative assumption because a lower steam generator pressure results in a higher primary to secondary leak rate tnrough the break.
C. Results The dynamic benavior of important NSSS parameters following a SGTR during part loop operation will be similar to the behavior presented in Figures 15.6.3-2 througn 15.6.3-17 of C_-JU
i the CESSAR. The lower RCS pressure results in a lower primary to secondary leakage through the break for the part loop case. This pressure, canbined with tile earlier expected trip time (see Section C.15.6.3.1.2), will result in a smaller integrated leakage for the part loop SGTR event. The lower initial power level will result in a lower integrated energy release from tne core after the reactor trip. This will result in a lower main steam safety valve (MSSV) release for the part loop case.
C.15.6.3.1.3.2 Radiological Consequences
, A. Physical Modei The physical model used in the evaluation of radiological consequences is the same as that used in Section 15.6.3.1.3.2 of the CESSAR.
B. Assumptions and Conditions
+ The assumptions and conditions used in tne evaluation of radiological consequences are the same as those given in Section 15.6.3.1.3.2 of the CESSAR with the following 4
exceptions.
j 1. The total amount of primary to secondary leakage for the part loop SGTR case is smaller than the amount shown in condition 11 of Section 15.6.3.1.3.2 of the CESSAR. The justification for this is presented in Section C.15.6.3.1.3.1.
- 2. The amount of steam flow to the cond'enser for the part loop operation case during the first two hours and during the two-to-eignt hour period are smaller than those shown in condition 12 of Section 15.6.3.1.3.2 of the CESSAR. This is the result of the lower initial power level during part loop operation and the earlier time of reactor trip.
C. Matnematical Model The mathematical model used in the evaluation of radiological consequences is the same as the model used in Section 15.6.3.1.3.2 of the CESSAR.
D. Results The two nour exclusion area boundary (EAB) inhalation dose and the eight hour low population zone (LPZ) boundary inhalation dose for SGTR during part loop operation are no larger than those for the SGTR event presented in Section 15.6.3.1 of the CESSAR. This is due to the amounts of integrated primary to secondary leakage, the MSSV steam release and tne steam release to the condenser for the SGTR event during part loop operation being lower than the corresponaing amounts presented in Section 15.6.3.1 of the CESSAR.
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F i
C.15.6.3.1.4 Conclusions The radiological releases for the SGTR event without a concurrent loss of offsite power during part loop operation are well withfn the 10 CFR 100 guidelines. The RCS and secondary system pressures are well below 110% of the design pressure limits, thus, assuring the integrity of these systems. Additionally, no violation of the fuel thermal limits occurs. The minimum DNBR remains above the 1.19 value throughout the duration of the event.
Tne plant is maintained in a stable condition due to automatic actions and, after thirty minutes, the operator employs the plant emergency procedure for the steam generator tube rupture event to cool down the plant to shutdown cooling entry conditions.
C.15.6.3.2 Steam Generator Tube Rupture With a Concurrent Loss of Of fsite Power C.15.6.3.2.1 Identification of Event and Causes The identification of event and causes for a steam generator tube rupture (SGTR) with a concurrent loss of offsite power during part loop operation are the same as that described in Section 15.6.3.2.1 of the CESSAR, C.15.6.3.2.2 Sequence of Events and Systems Operation
. The sequence of events and systems operation are the same as those described in Section 15.6.3.2.2 of the CESSAR with the following exception. Because of the restrictions on the initial conditions and setpoints (see Table C.15.6.3-2) for part loop operation, the events presented in Table 15.6.3-6 of tne CESSAR will be snifted witn respect to the times snown. In particular, the lower initial RCS pressure and the higher low pressurizer pressure trip setpoint during part loop operation will result in the reactor tripping at an earlier time than for tne case presented in tne CESSAR Section 15.6.3.2.
C.15.6.3.2.3 Analysis of Effects and Consequences C.15.6.3.2.3.1 Core and System Performance A. Mathematical Model The mathematical model is the same as that described in Section 15.6.3.1.3.1 of the CESSAR.
B. Input Parameters and Initial Conditions Table C.15.6.3-2 compares the assumptions and initial conditions for part loop operation with tnose presented in Section 15.6.3.2 of the CESSAR. The major differences are in the initial core power level, the initial core mass flow rate, the initial pressurizer pressure and the low pressurizer pressure trip setpoint. Additionally, the secondary side pressure in the steam generator for part loop C - 63
f ,
operation is,'in general, higher tnan tnat for full power operation. This is a direct result of the lower core power-(i.e., lower heat transfer rate in the steam generators) and tne lower RCS flow rate for part loop operation. However,
. in the present analysis, the assumption made is that the steam generator pressures for the two cases are equal . ;
This is a conservative assumption since a lower steam .
1 generator pressure results in a higher primary to secondary i leak rate through the break.
C. Results The dynamic behavior of important NSSS parameters following a SGTR during part loop operation will be similar to the behavior presented in Figures 15.6.3-19 tnrougn 15.6.3-34 of ;
the CESSAR. The lower RCS pressure results in a, lower '
primary to secondary leakage througn the break for the part loop case. Tris, combined with the earlier expected trip time (see Section C.15.6.3.2.2), will result in a smaller integrated leakage for the part loop SGTR event. The lower initial power level will result in a lower integrated energy release from the core after the reactor trip. As a result, the amount of steam released tnrough the main steam safety valves will be no greater than that presented in Section 15.6.3.2.3.1 of tne CESSAR. Also, tne amount of steam flow to the condenser during tne first two hours of the event will be smaller for part loop operation because of the lower initial power level and the earlier time of reactor trip. For the same reason, the amount of atmospheric dump valva ( ADV) release for the first two hours and during the two to eight hour period will be lower for part loop operation than that presented in Section 15.6.3.2.3.1 of the CESSAR.
C.15.6.3.2.3.2 Radiological Consequences A. Physical Model The physical model used in the evaluation of radiological consequences is the same as that used in Section 15.5.3.2.3.2 of the CESSAR.
B. Assumptions and Conditions
! The assumptions and conditions used in the evaluation of i radiological consequences are the same as those given in i Section 15.6.3.2.3.2 of tne CESSAR witn the following exceptions.
- 1. The total amount of primary to secondary leakage for the part loop SGTR case is smaller than tne amount snown in condition 12 of Section 15.6.3.2.3.2 of tne CESSAR. The justification for this is stated in Section C.15.6.3.2.3.1.
- 2. The total amount of steam flow through the condenser, main steam safety valves, and ADVs during the first two hours and during the two to eight hour period for the C - 64 i
/
part loop case are no greater than those presented in condition 2 of the CESSAR Section 15.6.3.2.3.2. This is the result of the lower initial power level during part loop operation and the earlier time of reactor trip.
C. Mathematical Model The mathematical model used in the evaluation of radiological consequences is the same as the model used in Section 15.6.3.2.3.2 of the CESSAR.
D. Results The two hour exclusion area boundary (EAB) inhalation dose and the eight hour low population zone (LPZ) boundary inhalation dose for SGTR with a concurrent loss of offsite power during part loop operation are no larger than those for the SGTR event presented in Section 15.6.3.2 of the CESSAR. This is due to the amount of integrated primary to secondary leakage, the MSSV steam release, the ADV steam release, and the steam release to the condenser for the part loop case being lower than the corresponding amount presented in Section 15.6.3.2 of the CESSAR.
C.15.6.3.2.4 Conclusions The radiological releases calculated for the SGTR event with a loss of offsite power during part loop operation are well within the 10 CFR 100 guidelines. The RCS and secondary system pressures are well below 110% of the design pressure limits, thus, assuring the integrity of these systems. Additionally, no violation of the fuel thermal limits occurs, since the minimum DNBR remains above 1.19 throughout the duration of the event.
Voids will form in the reactor vessel upper head region during the transient due to the thermal hydraulic decoupling of this region from the rest of the RCS. The upper head region liquid level will remain well above the top of the hot leg throughout the transient. Therefore, natural circulation cooldown will not be impaired during the 1 transient. Furthermore, the upper head voids will begin to collapse upon actuation of the safety injection flow, indicative of stable plant conditions. After thirty minutes, the operator employs the plant emergency procedure for the steam generator tube rupture event to cool down the plant to shutdown cooling entry conditions.
C.15.6.4 RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE BREAK OUTSIDE CONTAINMENT (BWR)
This event is not applicable to a pressurized water reactor and therefore is not included here.
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t TABLE C.15.6.3-1 ASSUMPTIONS AND INITIAL CONDITIONS FOR THE STEAM GENERATOR TUBE RUPTURE
! 1 Assumed Value Assumed Value for Parameter for CESSAR Part Loop Operation
! Core Powr Level, MWt 3876 1909 Core Inlet Coolant Temperature, UF 565 564 Reactor Coolant System Pressure, psia 2400 2300 Core Mass Flow Rate,106 lbm/hr 183.1 86.8
~
One Pin Integrated Radial Peaking Factor, 1.55 1.55 4- with Uncertainty i Steam Generator Pressure, psia 1020 1020 Modgratgr Temperature Coefficient, -3.5 -3.5 10 47 / F ,
) i Doppler Coefficient Multiplier 1.15 1.15 CEA Worth at Trip, ".op(most reactive. CEA -10.0 -10.0 fully withdrawn)
Low Pressurizer Pressure Trip Setpoint 1785 1870 i
4 1
S I'
l
+,,n .<- - ,,--..n,,-,,,-.c,-,-.-- - , - - - , , - - - - , -
l TABLE C .15. 6. 3-2 ASSUMPTIONS AND INITIAL CONDITIONS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF 0FFSITE POWER Assumed Value Assumed Value for Parameter for CESSAR Part Loop Operation 1
Core Power Level , MWt 3876 1909 Core Inlet Coolant Temperature, F 565 564 Reactor Coolant System Pressure, psia 2400 2300 Core Mass Flow Rate,100 lbn/hr 166 78.7 One Pin Integrated Radial Peaking Factor, 1.49 1.49 with Uncertainty Steam Generator Pressure, psia 1020 1020 Modegator Temperature Coefficient, -3.5 -3.5 10 of /g F Doppler Coefficient Multiplier 1.15 1.15 CEA Worth at Trip, % Ap (most reactive CEA -10.0 -10.0 fully withdraun)
Low Pressurizer Pressure Trip Setpoint 1785 1870.
e
' l l
C.15.6.5 LOSS OF COOLANT ACCIDENT (LOCA)
C.15.6.5.1 Identification of Causes The identification of causes is the same as that described in Section 15.6.5.1 of the CESSAR.
C.15.6.5.1 Analysis of Events and Consequences The analysis of events and consequences is the same as that presented in Section 15.6.5.2 of the CESSAR.
C.15.7 RADI0 ACTIVE MATERIAL RELEASE FROM A SUBSYSTEM OR COMP 0NENT I C.15.7.1 WASTE GAS SYSTEM FAILURE (see fpplicant's SAR)
C.15.7.2 RADI0 ACTIVE LIQUID WASTE SYSTEM LEAK OR FAILURE (see Applicant's SAR)
C.15.7.3 RADI0 ACTIVE RELEASE DUE TO LIQUID CONTAINING TANK FAILURE (see 3
Applicant's SAR)
C.15.7.4 FUEL HANDLING ACCIDENT
, C.15.7.4.1 Identification of Ever,t and Causes The identification and causes of a Fuel Handling Accident are the same as that described in the Section 15.7.4.1 of the CESSAR.
C.15.7.4.2 Sequence of Events and Systems Operation The sequence of events and systems operation are the same as that described in Section 15.7.4.2 of the CESSAR.
C.15.7.4.3 Analysis of Effects and Consequences The analysis of effects and consequences is the same as that given in Section 15.7.4.3 of the CESSAR.
C.15.7.4.4 Cbncl usion .
The exclusion area boundary dose resulting from the fuel assembly drop event will be determined by the Applicant.
C.16 TECHNICAL SPECIFICATIONS Tecnnical specifications will be provided for two pumps operating in opposite loops.
C.17 QUALITY ASSURANCE Analyses performed to confirm the adequacy of par' loop operation will be subject to the quality assurance of design provisions established for System 80. Therefore, tne information presented
- in Chapter 17 of CESSAR-F is applicable to part loop operation.
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