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{{#Wiki_filter:Shearon Harris Nuclear Power Plant                                                                      UFSAR Chapter: 15 15.0        ACCIDENT ANALYSIS ........................................................................................... 1 15.0.1.1  ANS Condition I - Normal Operation and Operational Transients .................... 3 15.0.1.2  ANS Condition II - Faults of Moderate Frequency ........................................... 4 15.0.1.3  ANS Condition III - Infrequent Faults ............................................................... 4 15.0.1.4  ANS Condition IV - Limiting Faults ................................................................... 4 15.0.2    OPTIMIZATION OF CONTROL SYSTEMS ............................................................ 5 15.0.3    PLANT CHARACTERISTICS AND INITIAL CONDITIONS ASSUMED IN THE ACCIDENT ANALYSES.......................................................................................... 5 15.0.3.1  Statistical Core Design ..................................................................................... 5 15.0.3.2  Design Plant Conditions. .................................................................................. 5 15.0.3.3  Initial Conditions ............................................................................................... 6 15.0.3.4  Power Distribution ............................................................................................ 6 15.0.4    REACTIVITY COEFFICIENTS ASSUMED IN THE ACCIDENT ANALYSES ......... 6 15.0.5    ROD CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTICS* ....... 6 15.0.6    TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES ............................................................................................................. 7 15.0.7    DELETED ................................................................................................................ 7 15.0.8    PLANT SYSTEMS AND COMPONENTS AVAILABLE FOR MITIGATION OF ACCIDENT EFFECTS* ........................................................................................... 7 15.0.9    FISSION PRODUCT INVENTORIES AND OTHER ISOTOPE SPECIFIC PARAMETERS ....................................................................................................... 8 15.0.9.1  Fission Product Inventories .............................................................................. 8 15.0.9.2  Dose Conversion Factors ................................................................................. 9 15.0.9.3  Nuclide Decay Constants ................................................................................. 9 15.0.9.4  Fuel Handling Accident Fission Product Source Term ................................... 10 15.0.10 RESIDUAL DECAY HEAT .................................................................................... 10 15.0.10.1 Small Break LOCA Decay Heat ..................................................................... 10 15.0.10.2 Large Break LOCA Decay Heat ..................................................................... 10 15.0.10.3 Non-LOCA Decay Heat .................................................................................. 10 15.0.11 COMPUTER CODES UTILIZED ........................................................................... 11 15.0.11.1 Deleted by Amendment No. 48. ..................................................................... 11 Amendment 63                                                                                                          Page i of vii
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Shearon Harris Nuclear Power Plant                                                                              UFSAR Chapter: 15 15.0.11.2 Computer Codes Utilized in Framatome Analyses ........................................ 11 15.0.11.3 Computer Codes Utilized in Duke Energy Analyses ........................................ 12 15.0.12 LONG TERM EFFECTS AND EVENTS FOLLOWING CHAPTER 15 ACCIDENTS
              .............................................................................................................................. 13 15.0.13 SINGLE FAILURES ASSUMED IN THE ANALYSES OF CHAPTER 15 ACCIDENTS ......................................................................................................... 13
 
==REFERENCES:==
SECTION 15.0 ............................................................................................. 17 APPENDIX 15.0A .................................................................................................................... 21 15.0A.1 OFFSITE DOSE CALCULATION MODELS .......................................................... 21 15.0A.1.1 Computer Code and Model Description ......................................................... 21 15.0A.1.2 Containment Leakage Pathway ..................................................................... 22 15.0A.1.3 Mixing Between Sprayed and Unsprayed Regions ........................................ 22 APPENDIX 15.0 - DELETED BY AMENDMENT NO. 41. ...................................................... 22 15.1        INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM .................... 22 15.1.1    FEEDWATER SYSTEM MALFUNCTIONS THAT RESULT IN A DECREASE IN FEEDWATER TEMPERATURE ........................................................................... 23 15.1.1.1      Identification of Causes and Accident Description ......................................... 23 15.1.1.2      Analysis of Effects and Consequences .......................................................... 23 15.1.1.3      Conclusions .................................................................................................... 23 15.1.2    FEEDWATER SYSTEM MALFUNCTIONS THAT RESULT IN AN INCREASE IN FEEDWATER FLOW ............................................................................................ 24 15.1.2.1      Identification of Causes and Accident Description ......................................... 24 15.1.2.2      Analysis of Effects and Consequences .......................................................... 25 15.1.2.3      Conclusions .................................................................................................... 26 15.1.3    EXCESSIVE INCREASE IN SECONDARY STEAM FLOW.................................. 26 15.1.3.1      Identification of Causes and Accident Description ......................................... 26 15.1.3.2      Analysis of Effects and Consequences .......................................................... 27 15.1.3.3      Conclusions .................................................................................................... 28 15.1.4    INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE .................................................................................................................. 28 15.1.4.1      Identification of Causes and Accident Description ......................................... 28 15.1.4.2      Deleted by Amendment No. 48. ..................................................................... 30 Amendment 63                                                                                                                  Page ii of vii
 
Shearon Harris Nuclear Power Plant                                                                      UFSAR Chapter: 15 15.1.4.3  Deleted by Amendment No. 48. ..................................................................... 30 15.1.4.4  Event Disposition for AREVA Analysis ........................................................... 30 15.1.5    STEAM SYSTEM PIPING FAILURE ..................................................................... 30 15.1.5.1  Identification of Causes and Accident Description ......................................... 30 15.1.5.2  Analysis of Effects and Consequences .......................................................... 32 15.1.5.3  Radiological Consequences of a Postulated Steamline Break ...................... 36 15.1.5.4  Conclusions .................................................................................................... 38 15.2        DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM ................... 39 15.2.1    STEAM PRESSURE REGULATOR MALFUNCTION OR FAILURE THAT RESULTS IN DECREASING STEAM FLOW ....................................................... 40 15.2.2    LOSS OF EXTERNAL ELECTRICAL LOAD ......................................................... 40 15.2.2.1  Identification of Causes and Accident Description ......................................... 40 15.2.2.2  Analysis of Effects and Consequences .......................................................... 41 15.2.2.3  Conclusions .................................................................................................... 41 15.2.3    TURBINE TRIP ..................................................................................................... 42 15.2.3.1  Identification of Causes and Accident Description ......................................... 42 15.2.3.2  Analysis of Effects and Consequences .......................................................... 43 15.2.3.3  Conclusions .................................................................................................... 45 15.2.4    INADVERTENT CLOSURE OF MAIN STEAM ISOLATION VALVES .................. 45 15.2.5    LOSS OF CONDENSER VACUUM AND OTHER EVENTS RESULTING IN TURBINE TRIP ..................................................................................................... 46 15.2.6    LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES .... 46 15.2.6.1  Identification of Causes and Accident Description ......................................... 46 15.2.6.2  Analysis of Effects and Consequences .......................................................... 48 15.2.6.3  Radiological Consequences of a Loss of Non-Emergency AC Power to Plant Auxiliaries ...................................................................................................... 50 15.2.6.4  Conclusions .................................................................................................... 52 15.2.7    LOSS OF NORMAL FEEDWATER FLOW ........................................................... 52 15.2.7.1  Identification of Causes and Accident Description ......................................... 52 15.2.7.2  Analysis of Effects and Consequences .......................................................... 54 15.2.7.3  Conclusions .................................................................................................... 55 Amendment 63                                                                                                        Page iii of vii
 
Shearon Harris Nuclear Power Plant                                                                    UFSAR Chapter: 15 15.2.8    FEEDWATER SYSTEM PIPE BREAK .................................................................. 56 15.2.8.1  Identification of Causes and Accident Description ......................................... 56 15.2.8.2  Analysis of Effects and Consequences .......................................................... 58 15.2.8.3  Conclusions .................................................................................................... 62
 
==REFERENCES:==
SECTION 15.2 ............................................................................................. 62 15.3        DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE ........................... 63 15.3.1    PARTIAL LOSS OF FORCED REACTOR COOLANT FLOW .............................. 63 15.3.1.1  Identification of Causes and Accident Description ......................................... 63 15.3.1.2  Deleted by Amendment No. 48 ...................................................................... 64 15.3.1.3  Deleted by Amendment No. 48 ...................................................................... 64 15.3.1.4  Event Disposition .......................................................................................... 64 15.3.2    COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW.......................... 64 15.3.2.1  Identification of Causes and Accident Description ......................................... 64 15.3.2.2  Analysis of Effects and Consequences .......................................................... 66 15.3.2.3  Conclusions .................................................................................................... 66 15.3.3    REACTOR COOLANT PUMP SHAFT SEIZURE (LOCKED ROTOR).................. 66 15.3.3.1  Identification of Causes and Accident Description ......................................... 66 15.3.3.2  Analysis of Effects and Consequences .......................................................... 67 15.3.3.3  Radiological Consequences of a Locked Rotor ............................................. 68 15.3.3.4  Conclusions .................................................................................................... 70 15.3.4    REACTOR COOLANT PUMP SHAFT BREAK ..................................................... 70 15.3.4.1  Identification of Causes and Accident Description ......................................... 70 15.3.4.2  Conclusions .................................................................................................... 70
 
==REFERENCES:==
SECTION 15.3 ............................................................................................. 71 15.4        REACTIVITY AND POWER DISTRIBUTION ANOMALIES ................................. 71 15.4.1    UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER STARTUP CONDITION .......................................................................................................... 72 15.4.1.1  Identification of Causes and Accident Description ......................................... 72 15.4.1.2  Analysis of Effects and Consequences .......................................................... 74 15.4.1.3  Conclusions .................................................................................................... 76 Amendment 63                                                                                                      Page iv of vii
 
Shearon Harris Nuclear Power Plant                                                                              UFSAR Chapter: 15 15.4.2    UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL AT POWER ................................................................................. 76 15.4.2.1      Identification of Causes and Accident Description ......................................... 76 15.4.2.2      Analysis of Effects and Consequences .......................................................... 77 15.4.2.3      Conclusions .................................................................................................... 79 15.4.3    ROD CLUSTER CONTROL ASSEMBLY MISOPERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR) ......................................................... 79 15.4.3.1      Dropped Full Length RCCA or RCCA Bank ................................................... 80 15.4.3.2      Withdrawal of a Single Full Length RCCA ..................................................... 82 15.4.3.3      Statically Misaligned RCCA or Bank .............................................................. 85 15.4.4    STARTUP OF AN INACTIVE REACTOR COOLANT PUMP AT AN INCORRECT TEMPERATURE ................................................................................................... 87 15.4.4.1      Identification of Causes and Accident Description ......................................... 87 15.4.4.2      DELETED ....................................................................................................... 87 15.4.4.3      DELETED ....................................................................................................... 87 15.4.4.4      Event Disposition ........................................................................................... 87 15.4.5    A MALFUNCTION OR FAILURE OF THE FLOW CONTROLLER IN A BWR LOOP THAT RESULTS IN AN INCREASED REACTOR COOLANT FLOW RATE
              .............................................................................................................................. 88 15.4.6    CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION THAT RESULTS IN A DECREASE IN THE BORON CONCENTRATION IN THE REACTOR COOLANT .......................................................................................... 88 15.4.6.1      Identification of Causes and Accident Description ......................................... 88 15.4.6.2      Analysis of Effects and Consequences .......................................................... 89 15.4.6.3      Results and Conclusions ................................................................................ 91 15.4.7    INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITION......................................................................................... 91 15.4.7.1      Identification of Causes and Accident Description ......................................... 91 15.4.7.2      Analysis of Effects and Consequences .......................................................... 92 15.4.7.3      Conclusions .................................................................................................... 93 15.4.8    SPECTRUM OF ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENTS ......................................................................................................... 93 15.4.8.1      Identification of Causes and Accident Description ......................................... 93 15.4.8.2      Analysis of Effects and Consequences .......................................................... 96 15.4.8.3      Radiological Consequences of a Postulated Rod Ejection Accident ............. 97 Amendment 63                                                                                                                Page v of vii
 
Shearon Harris Nuclear Power Plant                                                                      UFSAR Chapter: 15 15.4.8.4  Conclusions .................................................................................................... 99 15.4.9    SPECTRUM OF ROD DROP ACCIDENTS IN A BWR ......................................... 99
 
==REFERENCES:==
SECTION 15.4 ............................................................................................. 99 15.5        INCREASE IN REACTOR COOLANT INVENTORY .......................................... 101 15.5.1    INADVERTENT OPERATION OF THE EMERGENCY CORE COOLING SYSTEM DURING POWER OPERATION ......................................................................... 101 15.5.1.1  Identification of Causes and Accident Description ....................................... 101 15.5.1.2  Description of Analysis ................................................................................. 101 15.5.1.3  Acceptance Criteria ...................................................................................... 102 15.5.1.4  Results ......................................................................................................... 102 15.5.1.5  Conclusions .................................................................................................. 104 15.5.2    CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION THAT INCREASES REACTOR COOLANT INVENTORY ............................................ 104 15.5.3    A NUMBER OF BWR TRANSIENTS .................................................................. 104
 
==REFERENCES:==
SECTION 15.5 ........................................................................................... 104 15.6        DECREASE IN REACTOR COOLANT INVENTORY ......................................... 105 15.6.1    INADVERTENT OPENING OF A PRESSURIZER SAFETY OR POWER OPERATED RELIEF VALVE .............................................................................. 105 15.6.1.1  Identification of Causes and Accident Description ....................................... 105 15.6.1.2  Analysis of Effects and Consequences ........................................................ 106 15.6.1.3  Conclusions .................................................................................................. 107 15.6.2    BREAK IN INSTRUMENT LINE OR OTHER LINE FROM REACTOR COOLANT PRESSURE BOUNDARY THAT PENETRATE CONTAINMENT ...................... 107 15.6.2.1  Identification of Causes and Frequency Classification ................................. 107 15.6.2.2  Sequence of Events and Systems Operation .............................................. 107 15.6.2.3  Analysis of Radiological Consequences ...................................................... 108 15.6.3    STEAM GENERATOR TUBE RUPTURE ........................................................... 109 15.6.3.1  Identification of Cause and Accident Description ......................................... 111 15.6.3.2  Westinghouse Thermal Hydraulic Analysis .................................................. 114 15.6.3.3  Westinghouse Thermal Hydraulic Analysis - Offsite Dose Case.................. 117 15.6.3.4  Radiological Consequences Analysis .......................................................... 122 Amendment 63                                                                                                        Page vi of vii
 
Shearon Harris Nuclear Power Plant                                                                      UFSAR Chapter: 15 15.6.4    SPECTRUM OF BWR STEAM SYSTEM PIPING FAILURES OUTSIDE CONTAINMENT.................................................................................................. 126 15.6.5    LOSS OF COOLANT ACCIDENTS ..................................................................... 126 15.6.5.1  Identification of Causes and Frequency Classification ................................. 126 15.6.5.2  Large Break LOCA Transient ....................................................................... 127 15.6.5.3  Small Break LOCA Transient ....................................................................... 129 15.6.5.4  Radiological Consequences Analysis of a Postulated Large Break Loss of Coolant Accident ......................................................................................... 133 15.6.5.5  Radiological Consequences Analysis of a Postulated Small Break Loss-of-Coolant Accident ......................................................................................... 139
 
==REFERENCES:==
SECTION 15.6 ........................................................................................... 139 15.7        RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT ............. 143 15.7.1    RADIOACTIVE WASTE GAS SYSTEM LEAK OR FAILURE ............................. 143 15.7.1.1  Identification of Causes ................................................................................ 143 15.7.1.2  Analysis of Events and Consequences ........................................................ 144 15.7.1.3  Radiological Consequences Analysis .......................................................... 144 15.7.2    LIQUID WASTE SYSTEM LEAK OR FAILURE .................................................. 145 15.7.3    POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID TANK FAILURE . 145 15.7.3.1  Results and Conclusions .............................................................................. 145 15.7.4    DESIGN BASIS FUEL HANDLING ACCIDENTS................................................ 145 15.7.4.1  Identification of Causes and Accident Description ....................................... 145 15.7.4.2  Radiological Consequences Analysis .......................................................... 146 15.7.4.3  DELETED ..................................................................................................... 148 15.7.4.4  Deleted ......................................................................................................... 148 15.7.4.5  Other Fuel Handling Accidents .................................................................... 148 15.7.5    SPENT FUEL CASK DROP ACCIDENTS .......................................................... 149 15.7.5.1  Cask Drop Into the New or Spent Fuel Pool ................................................ 149 15.7.5.2  Cask Drop to Flat Surface ............................................................................ 149
 
==REFERENCES:==
SECTION 15.7 ........................................................................................... 149 15.8        ANTICIPATED TRANSIENTS WITHOUT SCRAM ............................................. 150
 
==REFERENCES:==
SECTION 15.8 ........................................................................................... 150 Amendment 63                                                                                                        Page vii of vii
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.0    ACCIDENT ANALYSIS*
This chapter addresses the representative initiating events listed on pages 15-10, 15-11, and 15-12 of Regulatory Guide 1.70, Revision 3 as they apply to the Shearon Harris Nuclear Power Plant (SHNPP).
Certain items in the guide warrant comment, as follows:
Items 1.3 and 2.1 - There are no pressure regulators in the Nuclear Steam Supply System (NSSS) pressurized water reactor (PWR) design whose malfunction or failure could cause a steam flow transient.
Item 6.2 - No instrument lines from the reactor coolant system boundary in the NSSS PWR design penetrate the Containment. For the definition of the reactor coolant system boundary, refer to ANSI-N 18.2, "Nuclear Safety Criteria for the Design of Stationary PWR Plants," Section 5, 1973.
The original Chapter 15 analyses for SHNPP were performed by Westinghouse. Since then, first Framatome then Duke Energy have replaced the various analyses with their own.
Consequently, all analyses in Chapter 15 are performed with Duke Energy methods except for the following:
The Westinghouse analyses in Chapter 15 are:
* Steam Generator Tube Rupture in Section 15.6.3 - Analysis of this event is not sensitive to specific fuel design, for example, the approach to minimum DNBR and peak linear heat generation rate limits are not close enough for values to be calculated for this event.
* Fission product inventory as source term for off-site and control room doses in Section 15.0.9, and the accident specific dose calculations.
* Anticipated Transient Without Scram in Section 15.8 The Framatome analyses in Chapter 15 are:
* Loss of Nonemergency AC Power to the Station Auxiliaries in Section 15.2.6
* Loss of Normal Feedwater in Section 15.2.7
* Inadvertent Operation of the ECCS During Power Operation in Section 15.5.1 None of these Framatome analyses provide core operating limits. However, with approval of Reference 15.0-2, these transients, along with the Westinghouse SGTR analysis, will be replaced with Duke Energy analyses as soon as they are complete. The ATWS analysis in Section 15.8 will remain a Westinghouse analysis for the foreseeable future.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 Changes to Support +/-3% MSSV Setpoint Tolerance In Reference 15.0-1, HNP requested a change to the Technical Specifications to increase the as-found lift setting tolerance for the main steam line code safety valves (MSSVs) from +/-1% to
+/-3%. A consequence of the increased MSSV setpoint tolerance is a reduction of the minimum AFW flow credited from one pump in the safety analyses at the lowest lifting MSSV setpoint plus tolerance. The change to AFW flow in certain accident and transient analyses is a reduction from 390 gpm to 374 gpm. To support the requested MSSV setpoint tolerance change, Reference 15.0-1 also requested a reduction to the pressurizer water level-high reactor trip setpoint from 92% to 87% of indicated span, as well as a change to the maximum pressurizer water level Technical Specification limiting condition of operation (LCO) to 75% of indicated span. The 75% pressurizer water level LCO is tied to the initial level assumed in the Section 15.2.3 turbine trip overpressure evaluation performed by Duke Energy, which assesses the impact of the requested changes to the transient and accident analyses in this chapter. The initial pressurizer levels assumed for all other events in this chapter are not required to be associated with the pressurizer water level LCO.
Reference 15.0-1 examines the impact of the requested changes to the transient and accident analyses in this chapter and confirms that, with the exception of SBLOCA, all other events are either 1) bounded by current analysis of record, 2) bounded by the 15.2.3 turbine trip event regarding the overpressure, or 3) the fuel centerline melt and/or MDNBR limits are not affected by the requested changes. Subsequent to issuance of Reference 15.0-1, the Chapter 15 analyses performed by Duke Energy capture the MSSV tolerance, pressurizer level trip setpoint, and AFW flow changes. However, the non-Duke Energy transient and accident analyses in this chapter are not revised to include the changes to the MSSV tolerance, pressurizer level trip setpoint, and AFW flow rate. Where applicable, these analyses continue to assume an MSSV setpoint tolerance of +/-1% and an AFW flow rate of 390 gpm. For SBLOCA, the MSSV setpoint tolerance increase and AFW flow rate reduction result in an estimated peak cladding temperature penalty, which is discussed in Section 15.6.5.3.4.
15.0.1 CLASSIFICATION OF PLANT CONDITIONS**
Since 1970 the American Nuclear Society (ANS) classification of plant conditions has been used which divides plant conditions into four categories in accordance with anticipated frequency of occurrence and potential radiological consequences to the public. The four categories are as follows:
: 1) Condition I: Normal Operation and Operational Transients.
: 2) Condition II: Faults of Moderate Frequency.
: 3) Condition III: Infrequent Faults.
: 4) Condition IV: Limiting Faults.
The basic principle applied in relating design requirements to each of the conditions is that the most probable occurrences should yield the least radiological risk to the public and those extreme situations having the potential for the greatest risk to the public shall be those least likely to occur. Where applicable, functioning of the Reactor Trip System and engineered Amendment 63                                                                          Page 2 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 safeguards is assumed in fulfilling this principle to the extent allowed by considerations such as the single failure criterion.
15.0.1.1    ANS Condition I - Normal Operation and Operational Transients ANS Condition I occurrences are those which are expected frequently or regularly in the course of power operation, refueling, maintenance, or maneuvering of the plant. As such, ANS Condition I occurrences are accommodated with margin between any plant parameter and the value of that parameter which would require either automatic or manual protective action. Since ANS Condition I occurrences occur frequently or regularly, they must be considered from the point of view of affecting the consequences of fault conditions (ANS Conditions II, III and IV). In this regard, analysis of each fault condition described is generally based on a conservative set of initial conditions corresponding to adverse conditions which can occur during ANS Condition I operation.
A typical list of ANS Condition I events is listed below:
: 1) Steady state and shutdown operations
: a. Power operation (>5 to 100 percent of rated thermal power).
: b. Startup (Keff  0.99,  5 percent of rated thermal power).
: c. Hot standby (subcritical, Residual Heat Removal System isolated).
: d. Hot shutdown (subcritical, Residual Heat Removal System in operation).
: e. Cold shutdown (subcritical, Residual Heat Removal System in operation).
: f. Refueling.
: 2) Operation with permissible deviations Various deviations, which may occur during continued operation as permitted by the plant Technical Specifications, must be considered in conjunction with other operational modes. These include:
a) Operation with components or systems out of service (such as power operation with a reactor coolant pump out of service).
b) Leakage from fuel with clad defects.
c) Radioactivity in the reactor coolant
: 1) Fission products.
: 2) Corrosion products.
: 3) Tritium.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 d) Operation with steam generator leaks up to the maximum allowed by the Technical Specifications.
e) Testing as allowed by the Technical Specifications.
: 4) Operational transients a) Plant heatup and cooldown (up to 100F/hour for the Reactor Coolant System; 200F/hour for the pressurizer during cooldown and 100F/hour for the pressurizer during heatup).
b) Step load changes (up to +/-10 percent).
c) Ramp load changes (up to 5 percent/minute).
d) Load rejection up to and including design full load rejection transient.
15.0.1.2      ANS Condition II - Faults of Moderate Frequency ANS Condition II occurrences are those which are expected to occur, in general, no more than once per year. These faults, at worst, result in a reactor trip with the plant being capable of returning to operation. By definition, these faults (or events) do not cause a more serious fault, i.e., ANS Condition III or IV events. In addition, ANS Condition II events are not expected to result in fuel rod failures or Reactor Coolant System or secondary system overpressure.
Table 15.0.1-1 lists the accident category used for each of the Chapter 15 events 15.0.1.3      ANS Condition III - Infrequent Faults By definition, ANS Condition III occurrences are faults which may occur very infrequently during the life of the plant. They will be accommodated with the failure of only a small fraction of the fuel rods although sufficient fuel damage might occur to preclude resumption of the operation for a considerable outage time. An ANS Condition III fault will not, by itself, generate an ANS Condition IV fault or result in a consequential loss of function of the Reactor Coolant System or containment barriers. Table 15.0.1-1 lists the accident category used for each of the Chapter 15 events.
15.0.1.4      ANS Condition IV - Limiting Faults ANS Condition IV occurrences are faults which are not expected to take place, but are postulated because their consequences would include the potential of the release of significant amounts of radioactive material. They are the most drastic events which must be designed against and represent limiting design cases.
ANS Condition IV faults are not to cause a fission product release to the environment resulting in an undue risk to public health and safety in excess of applicable limits. A single ANS Condition IV fault is not to cause a consequential loss of required functions of systems needed to cope with the fault including those of the Emergency Core Cooling System and the Containment. Table 15.0.1-1 lists the accident category used for each of the Chapter 15 events.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.0.2    OPTIMIZATION OF CONTROL SYSTEMS A control system setpoint study is performed in order to simulate performance of the Reactor Control and Protection Systems. In this study, emphasis is placed on the development of a control system which will automatically maintain prescribed conditions in the plant even under a conservative set of reactivity parameters with respect to both system stability and transient performance.
Nominal protection system setpoints on which the accident analysis is based are also used in the controls system setpoint study. Instrumentation errors are calculated consistent with the method used in the accident analysis. These errors are applied in an adverse direction with respect to maintaining system stability and transient performance. The accident analysis and setpoint study combine to show that the plant can be operated and meet both safety and operability requirements.
For each mode of plant operation, a group of optimum controller setpoints is determined. In areas where the resultant setpoints are different, compromises based on the optimum overall performance are made and verified. A consistent set of control system parameters is derived satisfying plant operational requirements throughout the core life and for various levels of power operation.
The study contains an analysis of the following control systems: rod cluster control assembly, steam dump, steam generator level, pressurizer pressure and pressurizer level.
15.0.3    PLANT CHARACTERISTICS AND INITIAL CONDITIONS ASSUMED IN THE ACCIDENT ANALYSES*
15.0.3.1    Statistical Core Design The Duke Energy analyses are separated into two categories - Statistical Core Design (SCD) and non-SCD analyses. The SCD methodology reassigns the instrument uncertainty on core power, pressurizer pressure, RCS flow, and RCS average temperature from the thermal-hydraulic analysis and incorporates those uncertainties into the DNBR limit per the methodology provided in Reference 15.0.3-6. The resultant DNBR limit is called the statistical design limit, or SDL. Consequently, SCD analyses are those DNB analyses that use the SDL as the DNBR acceptance limit and the corresponding thermal-hydraulic analyses assume nominal conditions for power, pressure, flow and temperature. Non-SCD analyses are those thermal-hydraulic analyses that are performed for the non-DNB related acceptance criteria or are DNB analyses that do not utilize the SCD method and therefore explicitly account for the uncertainties in power, pressure, flow and temperature.
15.0.3.2    Design Plant Conditions.
The accidents presented in Chapter 15 are analyzed at limiting conditions consistent with the Technical Specifications. The initial core power level for Chapter 15 analyses can be anywhere between 0% Rated Thermal Power (RTP) and 100.34% RTP where 100% RTP is defined in the Facility Operating License as 2948 MWt. 100.34% RTP is 2958 MWt and represents the maximum thermal power allowed by the MUR as 2948 MWt + 0.34% uncertainty. Pre-MUR methodology presented the RTP as 2900 MWt with 2% added for uncertainty (2958 MWt).
Since the SCD analyses account for the 0.34% uncertainty in the DNB limit, the maximum initial Amendment 63                                                                        Page 5 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 power assumed in SCD thermal-hydraulic analyses is 2948 MWt. The maximum initial power assumed for non-SCD thermal-hydraulic analyses is 2958 MW (rated + 0.34%). The values of pertinent plant parameters utilized in the accident analyses are given in Sections 15.1 through 15.7.
15.0.3.3    Initial Conditions The initial conditions assumed in the accident analyses are given in Sections 15.1 through 15.7.
The following nominal values (used in SCD analyses) and measurement uncertainties (used in the calculation of the SDL and in non-SCD analyses) were considered:
Parameter                          Nominal (SCD)          Uncertainty Core Power                        2948 MWt              +/- 0.34% RTP Pressurizer Pressure              2235 psig              +/- 50 psi RCS Flow                          296,380 gpm            +/- 2.2%
RCS Average Temperature            588.8 °F              +6.0/-6.8 °F (non-Duke Energy analyses)
                                                          +/- 5.1 °F (Duke Energy analyses)
The component response times, setpoints, and capacities supported in the accident analyses are presented in Table 15.0.3-5. The vessel average temperature assumed in the accident analyses are presented in Sections 15.1 through 15.6.
15.0.3.4    Power Distribution The transient response of the reactor system is dependent on the initial power distribution. The nuclear design of the reactor core minimizes adverse power distribution through the placement of control rods. Power distribution may be characterized by the radial peaking factor (FH) and the total peaking factor (FQ). The power distribution factor limits are given in the Technical Specifications.
15.0.4    REACTIVITY COEFFICIENTS ASSUMED IN THE ACCIDENT ANALYSES*
The reactivity coefficients assumed in the accident analyses are given in Sections 15.1 through 15.6.
15.0.5    ROD CLUSTER CONTROL ASSEMBLY INSERTION CHARACTERISTICS*
The negative reactivity insertion following a reactor trip is a function of the acceleration of the RCCA and the variation in rod worth as a function of rod position. With respect to accident analyses, the critical parameter is the time of insertion up to the dashpot entry, or approximately 80 to 85 percent of the rod cluster travel. For accident analyses, the insertion time to dashpot entry is conservatively taken as 2.7 seconds.
The rod cluster control assembly position versus time assumed for Framatome fuel in accident analyses is shown in Figure 4.3.2-38.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 Figure 4.3.2-39 shows the fraction of total negative reactivity insertion (normalized worth) versus normalized rod position (percent inserted) for a core where the axial power distribution is skewed to the lower region of the core. An axial distribution which is skewed to the lower region of the core can arise from an unbalanced xenon distribution. This curve is used to compute the negative reactivity insertion versus time following a reactor trip. This negative reactivity insertion curve is input to all point kinetics core models used in transient analyses. The bottom skewed power distribution itself is not an input into the point kinetics core model.
There is inherent conservatism in the use of Figure 4.3.2-39 in that it is based on a skewed flux distribution which would exist relatively infrequently. For cases other than those associated with unbalanced xenon distributions, significantly greater negative reactivity would have been inserted due to the more favorable axial distribution existing prior to the trip.
The normalized rod cluster control assembly negative reactivity insertion versus time is shown in Figure 15.0.5-6. The curve shown in this figure was obtained by combining Figures 4.3.2-38 and 4.3.2-39. The insertion worth used in the analysis of each event is indicated in the event description. The insertion worth has been decreased to account for the most reactive rod stuck out. The rod cluster control assembly position versus time is normalized to 2.7 seconds drop time to provide a bounding analysis for all rod cluster control assemblies to be used in the SHNPP cores, as previously stated.
15.0.6    TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES*
A reactor trip signal acts to open two trip breakers connected in series feeding power to the control rod drive mechanisms. The loss of power to the mechanism coils causes the mechanisms to release the rod cluster control assemblies, which then fall by gravity into the core. There are various instrumentation delays associated with each trip function, including delays in signal actuation, in opening the trip breakers, and in the release of the rods by the mechanisms. The total delay to trip is the difference between the time that trip conditions are reached and the time the rods are free and begin to fall.
Limiting trip setpoints assumed in accident analyses and the time delay assumed for each trip function are given in Table 15.0.6-2.
The difference between the limiting trip point assumed for the analysis and the nominal trip point represents an allowance for instrumentation channel error and setpoint error. Nominal trip setpoints are specified in the plant Technical Specifications. During plant startup tests it will be demonstrated that actual instrument time delays are equal to or less than the assumed values.
Additionally, protection system channels are calibrated and instrument response times determined periodically in accordance with the plant Technical Specifications.
15.0.7    Deleted 15.0.8    PLANT SYSTEMS AND COMPONENTS AVAILABLE FOR MITIGATION OF ACCIDENT EFFECTS*
The NSSS is protected by design from the possible effects of natural phenomena, postulated environmental conditions and dynamic effects of the postulated accidents. In addition, the design incorporates features which minimize the probability and effects of fires and explosions.
The incorporation of these features in the NSSS, coupled with the reliability of the design, Amendment 63                                                                          Page 7 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 ensures that the normally operating systems and components listed in Table 15.0.8-1 will be available for mitigation of the events discussed in Chapter 15. In determining which systems are necessary to mitigate the effects of these postulated events, the classification system of ANSI-N18.2-1973 is utilized. The design of "systems important to safety" (including protection systems) is consistent with IEEE Standard 379-1972 and Regulatory Guide 1.53 in the application of the single failure criterion.
In the analysis of the Chapter 15 events, control system action is considered only if that action results in more severe accident results. No credit is taken for control system operation if that operation mitigates the results of an accident. For some accidents, the analysis is performed both with and without control system operation to determine the worst case.
For a dropped RCCA event in the automatic rod control mode, the Rod Control System detects the drop in power and initiates control bank withdrawal. Power overshoot may occur due to this action by the automatic rod controller after which the control system will insert the control bank to restore nominal power. Following a dropped rod event in manual rod control, the plant will establish a new equilibrium condition. The equilibrium process without control system interaction is monotonic, thus removing power overshoot as a concern and establishing the automatic rod control mode of operation as the limiting case.
Operation of the pressurizer heaters as a result of normal control action or a single failure will, depending on the transient, be less conservative or have negligible effects. See Section 5.4.10 for a discussion of the pressurizer heaters. See Sections 15.1 through 15.6 for identification of whether the pressurizer heaters were assumed available.
The principal effect of the pressurizer heaters on transients would be to maintain higher RCS pressures.
During cooldown transients or DNB limited transients higher RCS pressure is less conservative.
It could be conservative for these types of transients to not assume the pressurizer heaters are working.
The pressurizer heaters will not cause a more severe pressure transient for those accidents which have the potential to overpressurize the RCS. For overpower transients such as rod ejection, the pressurizer heaters would have no effect on the pressure response since the pressure response during these types of transients is characterized by a rapid increase and then a decrease in pressurizer pressure. This increase and decrease of pressure occurs over a period less than 20 seconds, which is much faster than the response of the heaters.
For pressurizer heater effects during accident analyses of steam generator tube rupture, refer to Section 15.6.3.
15.0.9    FISSION PRODUCT INVENTORIES AND OTHER ISOTOPE SPECIFIC PARAMETERS 15.0.9.1    Fission Product Inventories The calculation of the core fission product inventory is based on a safety analysis core power level of 2958 MWt. The fission product inventories are calculated using the ORIGEN Code (Reference 15.0.9-2) using the data library based on ENDF/B-IV and ENDF/B-V (Reference Amendment 63                                                                          Page 8 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 15.0.9-3). These inventories are given in Table 15.0.9-1. The isotopes included in Table 15.0.9-1 are the isotopes controlling from considerations of inhalation dose and from direct dose due to immersion. Design basis primary and secondary coolant activities are shown in Tables 15.0.9-2 and 15.0.9-7. These activities are used to evaluate potential releases for events that do not directly result in fuel cladding failure or fuel centerline melt conditions.
The conservative iodine spiking model used for those events which must consider spiking calculates the equilibrium iodine appearance rates based on a nominal letdown flow of 120 gpm with perfect cleanup. The nominal RCS volume used in this equilibrium Iodine spiking model is provided in Tables 15.1.5-5, 15.2.6-5, 15.3.3-6, and 15.6.3-6 for the Steam Line break, Loss of AC Power, Locked Rotor and Steam Generator Tube Rupture events, respectively. The nominal 120 gpm letdown flow with perfect cleanup is increased by 10 percent to 132 gpm (to cover uncertainty), by 10 gpm for identified leakage from the RCS, by 1 gpm for unidentified leakage from the RCS, and by 31 gpm controlled leakage. The effective letdown flow is therefore 174 gpm. Inclusion of the controlled leakage in the effective removal flow is conservative, since this flow does not remove activity from the RCS.
The iodine spike appearance rates of iodines in the Reactor Coolant System assumed in the analysis are shown in Table 15.0.9-6.
Inventory in the gap between the fuel pellet and fuel rod cladding For some design basis accidents, fuel damage as predicted in bounding analyses is limited to fuel cladding failure. For these design basis accidents, the radioactive source term is limited to the activity initially in the gap between the fuel pellets and the fuel rod cladding. The percentage of the core fission product inventory in this gap - the gap factor - for these non-LOCA accidents is consistent with the values from Table 3 in Regulatory Guide 1.183, except for the fuel handling accidents in the Fuel Handling Building and in Containment. For the fuel handling accidents, gap fractions for Kr-85, Cs-134, and Cs-137 are increasing by a factor of 3 from the Regulatory Guide 1.183, Table 3 values (Reference 15.7.4.11). The fuel cycle design ensures that no fuel rod predicted to experience DNB [departure from nucleate boiling] in any other non-LOCA accidents will have operated beyond the power/burnup criteria of Footnote 11 in Regulatory Guide 1.183 15.0.9.2    Dose Conversion Factors The total effective dose equivalent (TEDE) dose is equivalent to the committed effective dose equivalent (CEDE) or inhalation dose plus the acute dose (EDE) dose for the duration of exposure to the cloud. The dose conversion factors (DCFs) used in determining the CEDE dose are from Reference 15.0.9-9 and are given in Table 15.0.9-3. The DCFs used in determining the EDE dose are from Reference 15.0.9-10 and are given in Table 15.0.9-4.
These are the DCFs suggested by Regulatory Guide 1.183 (Reference 15.0.9-8).
15.0.9.3    Nuclide Decay Constants Decay constants for each nuclide are from Reference 15.0.9-11 and provided in Table 15.0.9-5.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.0.9.4    Fuel Handling Accident Fission Product Source Term The fission product inventory for the PWR and BWR fuel used in the Fuel Handling Accident Analysis is shown in Tables 15.7.4-1 and 15.7.4-3. For the FHA analysis in the Fuel Handling Building, the activity shown in Table 15.7.4-1 is the combined total activity of PWR fuel damaged plus the activity of the BWR fuel that is also assumed to be damaged. The gap release fractions used in the FHA analysis are listed in Tables 15.7.4-1 and 15.7.4-3.
15.0.10 RESIDUAL DECAY HEAT*
15.0.10.1 Small Break LOCA Decay Heat The decay heat for the Small Break LOCA (SBLOCA) analysis is based upon 1.2 times the draft 1971 ANS standard fission product decay heat (Reference 15.0.10-6) as required by 10CFR Part 50.46 (Reference 15.0.10-7) and Appendix K (Reference 15.0.10-8). Refer to Reference 15.0.11-20 (SBLOCA Methodology) 15.0.10.2 Large Break LOCA Decay Heat The Large Break LOCA (LBLOCA) decay heat is based on 1979 ANS Standard (Reference 15.0.10-3). The HNP LBLOCA Methodology (Reference 15.6.5-50) contains description on the application of the ANS standard.
15.0.10.3 Non-LOCA Decay Heat Duke Energy Analyses - Decay heat for the Duke Energy analyses is based on the 1979 ANS Standard (Reference 15.0.10-3).
Framatome Analyses - Decay heat for the non-LOCA analyses is based upon the 1973 ANS decay heat curve (Reference 15.0.10-4). Due to the rapidity of the majority of the non-LOCA transients analyzed, the decay heat has a negligible impact on the analysis results. For the longer non-LOCA transients, such as the Loss of Feedwater event, a high value for the decay heat versus time will be conservative. The 1973 ANS decay heat curve provides a higher decay heat than the 1979 ANS decay heat curve, and is therefore conservative (Reference 15.0.10-5).
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Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 15.0.11 COMPUTER CODES UTILIZED*
15.0.11.1 Deleted by Amendment No. 48.
15.0.11.1.1    Deleted by Amendment No. 48.
15.0.11.1.2    Deleted by Amendment No. 48.
15.0.11.1.3    Deleted by Amendment No. 48.
15.0.11.1.4    Deleted by Amendment No. 48.
15.0.11.1.5    Deleted by Amendment No. 48.
15.0.11.1.6    Deleted by Amendment No. 48.
15.0.11.2 Computer Codes Utilized in Framatome Analyses Summaries of some of the principal computer codes used in non-LOCA and LOCA analyses performed by Framatome are given below.
15.0.11.2.1    Non-LOCA Transients The codes used for non-LOCA transients are described below.
ANF-RELAP - ANF-RELAP is the PWR system transient analysis code used for simulation of the system response for the non-LOCA transients. Control volumes and junctions are defined which describe all major components in the primary and secondary systems which are important for the event being analyzed. The ANF-RELAP hydrodynamic model is a one-dimensional, transient, two-fluid model for flow of a two-phase steam-water mixture. ANF-RELAP uses a six equation model for the hydraulic solutions. These equations include two phasic continuity equations, two phasic momentum equations, and two phasic internal energy equations. The six equation model also allows both non-homogeneous and non-equilibrium situations encountered in reactor problems to be modeled. They have been generically approved by the NRC for use in non-LOCA transient analyses.
The methodology for using the ANF-RELAP code is described in Reference 15.0.11-6.
15.0.11.2.2    Large break LOCA analyses AREVA uses a series of codes for LBLOCA analyses. All of these codes have been reviewed and generically accepted by the NRC as part of the AREVA LBLOCA evaluation models to be in compliance with the requirements of 10 CFR 50.46 and Appendix K. The AREVA PWR ECCS evaluation model is described in Reference 15.0.11-11.
The codes used for LBLOCA analyses are described below.
RODEX3A - RODEX3A calculates fuel, cladding and fuel-cladding gap properties as a function of exposure. The code models a fuel rod over the power history. The code is used to determine fuel rod temperature and gap conditions in the LBLOCA analysis.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 S-RELAP5- S-RELAP5 is used for calculation of the system response. The field equations are basically the same form as RELAP5/MOD2 with the addition of full two-dimensional momentum equations. This two-dimensional capability is only applied within the reactor vessel in the Realistic Large Break LOCA methodology, but can be applied anywhere in the reactor coolant system through input. Initial fuel conditions are supplied by the realistic fuel performance code, RODEX3A. Capability for a concurrent calculation of containment backpressure based on the ICECON code was added.
15.0.11.2.3    Small break LOCA The codes used to perform the SBLOCA analysis are described below:
RODEX2 - The RODEX2 code is utilized to determine the initial fuel stored energy and gap conditions for the initialization of the system blowdown and hot rod response calculations.
S-RELAP5 - The AREVA version of S-RELAP5 is used to model the primary system and secondary side of the steam generators during the entire transient. The governing conservation equations for mass, energy, and momentum transfer are used along with appropriate correlations consistent with Appendix K of 10 CFR 50.
The methodologies for applying the SBLOCA codes described above can be found in Reference 15.0.11-20.
15.0.11.2.4    Deleted 15.0.11.3 Computer Codes Utilized in Duke Energy Analyses CASMO CASMO-5 (Reference 15.0.11-25) is a multi-group two-dimensional characteristics based transport theory code for burnup calculations on Pressurized Water Reactor (PWR) fuel assemblies or simple pin cells. The code accommodates a geometry consisting of cylindrical fuel rods of varying composition in a square pitch array with allowances for absorber-loaded fuel rods (i.e. gadolinia or erbium), integral fuel burnable absorbers (IFBA), discrete burnable absorber rods, control rods, in-core instrument channels, and water gaps. Reflector and baffle calculations can also be performed with CASMO-5. Multi-group calculations can be performed using as many as 586 neutron and 18 gamma group data from the ENDF/B-VII.1 library.
SIMULATE SIMULATE-3 (Reference 15.0.11-26) is a two-group three-dimensional coarse mesh diffusion theory code based on the QPANDA neutronics model. Fuel and moderator temperature feedback effects are accounted for using a closed-channel nodal thermal-hydraulics model. The program explicitly models the baffle and reflector region. Cross sections and nuclear constants from CASMO are used for each pin in the fuel assembly, along with inter-assembly and intra-assembly data obtained from the coarse mesh solution to reconstruct the power distribution for each pin.
SIMULATE-3K - SIMULATE-3K (Reference 15.0.11-27) is a two-group nodal code that models both neutronics and thermal-hydraulics for transient reactor core analyses. It is a transient version of SIMULATE-3 that uses the same nuclear data library but adds the time-dependent equations needed to solve transient, three-dimensional, coupled neutronics and thermal-hydraulic problems.
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Shearon Harris Nuclear Power Plant                                                  UFSAR Chapter: 15 VIPRE VIPRE-01 (Reference 15.0.11-28) is a subchannel thermal-hydraulics analysis code, based on a homogeneous formulation of the mass, energy, and momentum equations with additional models for subcooled boiling and liquid/vapor slip. It has been reviewed and approved for generic use by the NRC staff in Reference 15.0.11-29 and more specifically for use at SHNPP in Reference 15.0-2.
RETRAN 3-D - RETRAN 3-D (Reference 15.0.11-23) is a flexible, general-purpose, thermal-hydraulic computer code that can be used to represent light-water reactor systems. The code solves the governing conservation equations of mass, energy, and momentum, as applied to a network of fluid volumes and flow junctions. Conductive heat structures can be modeled, including the fuel elements in the reactor core. Changes in reactor power from neutron kinetics and decay heat are calculated to occur with time. The name, RETRAN-3D, refers to ability of the code to perform three-dimensional neutronic calculations in the core, as opposed to three-dimensional fluid dynamic capability. RETRAN-3D was generically approved by the NRC staff in Reference 15.0.11-24 with 45 limitations and conditions of use and specifically for SHNPP in References 15.0-2 and 15.0.11-30.
15.0.12 LONG TERM EFFECTS AND EVENTS FOLLOWING CHAPTER 15 ACCIDENTS For most of the events analyzed in Chapter 15, the plant will be in a safe and stable hot standby condition following the automatic actuation of reactor trip. This condition will in fact be similar to plant conditions following any normal, orderly shutdown of the reactor. At this point, the actions taken by the operator would be no different than normal operating procedures. The exact actions taken, and the time these actions would occur, will depend on what systems are available (e.g. steam dump system, main feedwater system, etc.) and the plans for further plant operation. As a minimum, to maintain the hot stabilized condition, decay heat must be removed via the steam generators. The main feedwater system and the steam dump or atmospheric relief system could be used for this purpose. Alternatively, the auxiliary feedwater system and the steam generator safety valves may be used, both of which are safety grade systems.
Although the auxiliary feed system may be started manually, it will be automatically actuated if needed by one of the signals shown on Figure 7.2.1-1 sheet 14, such as low-low steam generator water level. If hot standby conditions are maintained for an extended period of time, operator action may be required to transfer to the auxiliary feedwater source. The time at which such action is required will be sufficiently long after initiation of the event to permit operator action. Also, if the hot standby condition is maintained for an extended period of time (greater than approximately 18 hours), operator action may be required to add boric acid via the CVCS to compensate for xenon decay and maintain shutdown margin. Again, the actions taken by the operator would be no different than during normal plant shutdown.
Many Chapter 15 events result in a stable condition being reached automatically following a reactor trip and only actions typical of normal operation are required from the operator. For several events involving breaks in the reactor coolant system or secondary system piping, additional requirements for operator action can be identified (see Sections 15.1.5.2 and 15.2.8.2). (Additional information about the impact of equipment failures or erroneous operator actions may be found in WCAP-9691 "NUREG-0578.2.1.9.C, Transient and Accident Analysis".)
15.0.13 SINGLE FAILURES ASSUMED IN THE ANALYSES OF CHAPTER 15 ACCIDENTS All of the transients analyzed in Chapter 15 were analyzed assuming the most limiting single failure (e.g., loss of one protection signal of safety injection (SI) train failure). Table 15.0.13-1 Amendment 63                                                                            Page 13 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 lists the limiting single failures for each American Nuclear Society (ANS) Conditions II event (faults of moderate frequency) as provided in Reference 15.0-2. Table 15.0.13-2 lists the limiting single failures for Non-Condition II events as provided in Reference 15.0-2.
The single failures listed in Tables 15.0.13-1 and 15.0.13-2 are the limiting failures for their respective event. Though operator errors do not appear to be explicitly addressed, they have been considered and found to be bounded by the most limiting single failures listed.
: 1) Single Failures Assumed for Accidents of Moderate Frequency - The incidents of moderate frequency were analyzed consistent with the acceptance criteria given in the Standard Review Plan (SRP) concerning peak pressure (less than 110 percent of design), fuel integrity (departure from nucleate boiling ratio (DNBR) limit), generation of more serious plant conditions, and single active failures.
Pressure transients for each event are provided in the FSAR and demonstrate that the pressure remains below 110 percent of design pressure. Fuel cladding integrity is demonstrated for each case by showing that the DNBR remains above the limit value. This is discussed in the results and conclusions sections for each event.
For each transient, its associated worst single failure within the protection system assumed in the FSAR analyses is given in Table 15.0.13-1. The protection system is defined as those safety functions required to mitigate the consequences of the event. This includes not only the Solid State Protection System (SSPS), but also the Engineered Safety Features (ESF) and pressurizer and steam generator safety valves.
These single failures were selected based on the requirements of 10 CFR 50 Appendix A, the SRP, and Reg. Guide 1.53 (which addresses IEEE-279 and IEEE-379). A single failure is ". . .
an occurrence which results in the loss of capability of a component to perform its intended safety functions." (10 CFR 50 Appendix A). The single failure criterion states that a "single failure within the protection system shall not prevent proper protective action at the system level when required" (IEEE-279). The worst single failure for each event are provided in Reference 15.0-2.
The single failures which are considered are active failures, consistent with the SRP acceptance criteria. Failures in the protection system which are not required to mitigate the consequences of an accident are not considered. These are failures of systems which are not challenged during the transient and are not active failures. Such failures are independent failures and are, therefore, not within the scope of the evaluation.
For each event listed in Table 15.0.13-1, a brief discussion of the assumed single failure is provided in the respective FSAR section. These failures are failures at the system level and consider the failure of a protective function. The cause or mechanical nature of the failure which causes the system failure is not discussed, since these are addressed in the failure modes and effects analyses (FMEA's) of the SSPS and ESF and in Chapters 6, 7, and 9 of the FSAR.
Therefore, further detail beyond the systems level single failure of loss of one protection train is not provided.
The steam generator safety valves may be required to prevent a pressurization of the secondary system. Except where it is already stated in the FSAR, the steam generator valves are not challenged or required to mitigate the consequences of the event. Failures of these Amendment 63                                                                          Page 14 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 valves are not considered since they are not active failures. These independent failures are not applicable. Therefore, failure of these valves is not discussed below unless they are actuated as stated in the FSAR.
Finally, a loss of offsite power is not considered as a single failure for these events. The SRP does not require consideration of a loss of offsite power for the accidents listed in Table 15.0.13-1 (loss of AC power, FSAR Section 15.2.6, is by definition an exception). Furthermore, no single active failure will cause a loss of offsite power to the emergency buses. Therefore, consideration of this failure is not applicable for the Framatome analyses. The Duke Energy analyses for which a loss of offsite power is considered are identified in Reference 15.0-2.
The Auxiliary Feedwater System (AFS) warrants additional discussion since the only credible single failure results in the loss of 1 auxiliary feedwater pump. However, the Framatome analyses in Chapters 15.2.6 and 15.2.7 assume the loss of 2 auxiliary feedwater pumps. In the Duke Energy methodology, all pumps called upon by the ESF system are available and the single failure analysis simply determines which of the pumps is/are unavailable.
Although Table 10.4.9-2 presents the worst single failure affecting AFS flow as loss of one pump, Framatome has analyzed the Loss of Feedwater and Loss of AC Power Events using only one AFW pump in response to TMI Action Item II.E.1.1. As documented in Chapter 10.4.9 App A of the SHNPP FSAR, taking credit for this capability enabled SHNPP to achieve the AFW System reliability criteria specified in the TMI Action Item. Therefore, the Framatome analysis was performed using the flow from only one MDAFW pump in order to demonstrate that capability. It should be noted that the corresponding loss of two of three AFS pumps is more conservative than any postulated single active failure (Reference 15.0.13-1).
Separate from pump or system flow, an additional consideration presented in Table 10.4.9-2 is the number of steam generators that receive auxiliary feedwater. An alternative potential failure is temporary isolation of one steam generator. Instead of dividing the minimum design (i.e., one pump) flow among three steam generators, inadvertent isolation of a steam generator would split the full (three pump) AFS flow between two steam generators.
For the cases in which forced RCS flow is maintained to sustain the normal operating heat transfer coefficient between the primary coolant and the inside of the steam generator tubes, the alternative of delivering three or four times as much auxiliary feedwater to two of three steam generators represents a much less restrictive limitation on cooling capability. However, when loss of offsite power is the initiating event, the relative cooling capability of AFW flow rate vs number of steam generators receiving flow is not immediately obvious. Under RCP coastdown and natural circulation conditions, it is not clear whether the limiting factor for RCS cooling is steam generator secondary side inventory vs degraded heat transfer from the primary coolant to the SGs. As such, analysis of loss of AC power is based on conservatively combining these two alternative failure scenarios. With flow from only a single Auxiliary Feedwater pump, the analysis is performed with and without an additional failure of an auxiliary feedwater isolation valve in the closed position. Although postulation of multiple independent failures is not required, doing so is conservative. Therefore, the analysis basis listed in Table 15.0.13-1 is one motor driven AFS pump delivering flow to two or three steam generators (as separate cases).
In this case, a unique distinction exists where the AFW system configuration and performance assumed in the analysis (one AFW pump delivering to two steam generators) does not represent a legitimate performance requirement for the system. The analysis uses a "beyond Amendment 63                                                                          Page 15 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 licensing basis" system configuration that combines the independent failures in the AFW pump capability and the inadvertent isolation of one steam generator. The acceptable results obtained with the overly-conservative, beyond-licensing-basis model demonstrates that either of the two independent legitimate failure scenarios would also yield acceptable results if analyzed separately.
However, the analysis model itself (with its combination of independent failures) does not accurately represent a legitimate licensing basis configuration for the system and there is no commitment or requirement that the system must be able to perform as modeled in the analysis.
Therefore, the AFW system does not have to be capable of delivering 390a gpm of flow from only one AFW pump with the coincident isolation of one steam generator. If only one AFW pump is assumed operational, then the required 390a gpm of flow can be assumed to be delivered to all three steam generators. If a failure has resulted in isolation of AFW flow to one steam generator, then it may be assumed that all three AFW pumps would be available to deliver the required 390a gpm of flow. The minimum capacity of a motor driven auxiliary feedwater (MDAFW) pump is 430 gpm. For conservatism, the flow capacity of any pump was assumed to be 390a gpm in the FSAR analysis.
For the case where the single active failure is the failure of the pressurizer PORV or safety valve to close, credit can be taken for complete auxiliary feedwater capability. This would reduce the peak pressure and cause the time at which decay heat equals heat removal capability to be sooner. As stated in FSAR Section 15.2.6.1, the steam generator safety and relief valves are used to dissipate decay heat during long term cooling. Since it is desirable to have these valves open, failure to close has no impact, especially since the emergency feedwater supplies sufficient heat removal capability. Single failures which result in loss of signals which actuate auxiliary feedwater, reactor trip, or valve openings have no impact due to their redundancy, diversity and independence.
: 2) Single Failures for Non-Condition II Events - Table 15.0.13-2 addresses events other than those of moderate frequency. Fuel failures (if any) which occur are discussed in the appropriate section of the FSAR.
The reliability criteria specified in TMI Action Item II.E.1.1 (FSAR Section 10.4.9 App A) that led to Framatome modeling multiple AFW pump failures for the Loss of Feedwater and Loss of AC power Events is not applicable to the Feedline Break Event. Therefore, since no credible single-active-failure has been identified that leads to loss of two AFW pumps, credit for two AFW pumps may be assumed in order to provide the AFW flow modeled to mitigate this event.
Although operator error is not explicitly considered in the accidents analyzed in Chapter 15, it should be noted that cognitive operator errors may induce the transient or cause the limiting single failures listed. With the exception of the steam generator tube rupture (SGTR),
Inadvertent Operation of the ECCS, and the boron dilution, no credit for operator action or non-action is considered during the transient.
For the boron dilution event, operator action to terminate the dilution is taken within the time frames specified in the FSAR.
a The safety analyses support an AFW flow rate of 374 gpm from 1 MDAWF pump at the lowest lifting MSSV setpoint plus 3% tolerance. See Section 15.0 for more details.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 Operator action (subsequent to the action of the protection system) is also considered to mitigate the consequences for an SGTR event. The operator actions and the operator action times assumed for the SGTR analysis are discussed in Section 15.6.3 Operator action for the Inadvertent Operation of the ECCS is assumed to occur to terminate safety injection. See Chapter 15.5.1 for details.
 
==REFERENCES:==
SECTION 15.0 15.0            Deleted by Amendment No. 48.
15.0-1          Letter from B.C. Waldrep (Duke Energy) to NRC (Serial HNP-15-038) dated December 17, 2015, "License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change." (Safety Evaluation Report received by {{letter dated|date=July 25, 2016|text=letter dated July 25, 2016}}).
15.0-2          DPC-NE-3009, Revision 0, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, April 2018.
15.0.3-1        Deleted by Amendment 42.
15.0.3-2        Deleted by Amendment No. 48.
15.0.3-3        Deleted by Amendment No. 48.
15.0.3-4        Deleted by Amendment No. 48.
15.0.3-5        Deleted by Amendment No. 48.
15.0.3-6        DPC-NE-2005-PA, Revision 5, Thermal-Hydraulic Statistical Core Design Methodology, March 2016.
15.0.9-1        Deleted by Amendment No. 44.
15.0.9-2        ORNL-4628 "ORIGEN - The ORNL Isotope Generation and Depletion Code," M.
J. Bell, May 1973.
15.0.9-3        RSIC-OLC-38, "ORIGEN" yields and cross sections - Nuclear Transmutation and Decay Data from ENDF/B-IV, "Radiation Shielding Information Center, Oak Ridge National Laboratory, September 1975.
15.0.9-4        Baker, D. A., et. al., "Assessment of the Use of Extended Burnup Fuel in Light Water Reactors," NUREG/CR-5009, February 1988.
15.0.9-5        Kersting, P. J., et. al., "Extended Burnup Evaluation of Westinghouse Fuel,"
WCAP-10125-P-A, December 1985.
15.0.9-6        Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 7 to Facility Operating License No. NPF 63, Carolina Power &
Amendment 63                                                                      Page 17 of 151
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 Light Company, et. al., Shearon Harris Nuclear Power Plant, Unit 1, Docket No.
50-400.
15.0.9-7      J.J. DiNunno, F.D. Anderson, R.E. Baker, R.L. Waterfield, Calculation of Distance Factors for Power and Test Reactor Sites, Technical Information Document TID-14844, Division of Licensing and Regulation, U.S. Atomic Energy Commission, March 23, 1962.
15.0.9-8      Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
15.0.9-9      EPA Federal Guidance Report No. 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," EPA-520/1-88--0202, September 1988.
15.0.9-10    EPA Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water and Soil," EPA 402-R-93-081, September 1993.
15.0.9-11    International Commission on Radiological Protection, "Radionuclide Transformations, Energy and Intensity of Emissions," ICRP 38, 1983.
15.0.9-12    International Commission on Radiological Protection, "Limits for Intakes or Radionuclides by Workers," ICRP Publication 30, 1979.
15.0.9-13    Bechtel Corporation, LOCADOSE NE-319 Users Manual, Latest Revision to Date.
15.0.9-14    Bechtel Corporation, LOCADOSE NE-319 Theoretical Manual, Latest Revision to Date.
15.0.9-15    Bechtel Corporation, LOCADOSE NE-319 Validation Manual, Latest Revision to Date.
15.0.9-16    U.S. Nuclear Regulatory Commission, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Plants, Regulatory Guide 1.195, May 2003.
15.0.9-17    Murphy, K. G., Campe, K. M., Nuclear Power Plant Control Room Ventilation System Design for Meeting General Design Criterion 19, 13th AEC Air Cleaning Conference, August 1974.
15.0.10-1    Deleted by Amendment No. 48.
15.0.10-2    Deleted by Amendment No. 48.
15.0.10-3    ANSI/ANS-5.1-1979, "Decay Heat Power in Light Water Reactors," August 29, 1979.
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Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 15.0.10-4    American Nuclear Society Proposed Standard, ANS 5.1, Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors, October 1971, Revised October 1973.
15.0.10-5    NUREG-1230, "Compendium of ECCS Research of Realistic LOCA" Analysis, Section 6.12, published December 1988.
15.0.10-6    Proposed ANS Standard, Decay Energy Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors, ANS-5.1, October 1971.
15.0.10-7    10CFR Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Reactors," Federal Register, Volume 39, No.3, January 4, 1974.
15.0.10-8    Appendix K, "ECCS Evaluation Models," Federal Register, Volume 39, No.3, January 4, 1974.
15.0.11-1    Deleted by Amendment No. 48.
15.0.11-2    Deleted by Amendment No. 48.
15.0.11-3    Deleted by Amendment No. 48.
15.0.11-4    Deleted by Amendment No. 48.
15.0.11-5    Deleted by Amendment No. 48.
15.0.11-6    ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events, ANF-89-151(P)(A) and Correspondence, Advanced Nuclear Fuels Corporation, Richland, WA 99352, May 1992.
15.0.11-7    Steamline Break Methodology for PWRs, EMF-84-093(P)(A) Revision 1, Advanced Nuclear Fuels Corporation, Richland, WA 99352, February 1999.
15.0.11-8    Deleted by Amendment No. 63.
15.0.11-9    Deleted by Amendment No. 63.
15.0.11-10    Deleted by Amendment No. 59 15.0.11-11    EMF-2103(P)(A) Revision 0, Realistic Large Break LOCA Methodology, AREVA NP, April 2003 15.0.11-12    Deleted by Amendment No. 63.
15.0.11-13    Deleted by Amendment No. 63.
15.0.11-14    Deleted by Amendment No. 50 15.0.11-15    Deleted by Amendment No. 50 Amendment 63                                                                  Page 19 of 151
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 15.0.11-16    Deleted by Amendment No. 63.
15.0.11-17    Deleted by Amendment No. 63.
15.0.11-18    Deleted by Amendment No. 54 15.0.11-19    Deleted by Amendment No. 63.
15.0.11-20    EMF-2328(P)(A) Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, AREVA NP, March 2001.
15.0.11-21    Deleted by Amendment No. 59 15.0.11-22    Deleted by Amendment No. 63.
15.0.11-23    RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid FLow Systems, EPRI NP-7450(A), Revision 10, September 2014.
15.0.11-24    Letter from S.A. Richards (NRC) to G.L. Vine (EPRI) dated January 2001, "Safety Evaluation Report on EPRI Topical Report NP-7450(P), Revision 4,
              'RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems' (TAC No. MA4311)."
15.0.11-25    J. D. Rhodes III, R. M. Ferrer, and J. M. Hykes, CASMO5 A Fuel Assembly Burnup gram Methodology Manual, Studsvik Proprietary, SSP-08/405 Revision 3, August 2014 15.0.11-26    STUDSVIK/SOA-95/18, Revision 0, SIMULATE-3 Methodology Advanced Three-Dimensional Two-Group Reactor Analysis Code, 1995.
15.0.11-27    Studsvik Scandpower, SIMULATE-3K Models and Methodology, SSP-93/13, Revision 6, January 2009.
15.0.11-28    EPRI NP-2511-CCM-A, Revision 4, VIPRE A Thermal-Hydraulic Code for Reactor Cores, June 2007.
15.0.11-29    Rossi, C.F., U.S. Nuclear Regulatory Commission, letter to J.A. Blaisdell, Utility Group for Regulatory Applications, Acceptance for Referencing of Licensing CM, VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, Volumes 1, 2, 3, and 4, May 31, 1986.
15.0.11-30    DPC-NE-3008, Revision 0, Thermal-Hydraulic Models for Transient Analysis, April 2018.
15.0.13-1    NLS-87-232 letter from R. A. Watson, CP&L, to Dr. J. N. Grace, NRC, dated October 29, 1987, 10 CFR 21 Notification Follow-Up Report.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 APPENDIX 15.0A 15.0A.1 OFFSITE DOSE CALCULATION MODELS 15.0A.1.1 Computer Code and Model Description The Bechtel computer code LOCADOSE is used in the analyses of radiological consequences of all design basis accidents with postulated releases of fission products to the environment.
The code has been verified, documented, and controlled within Bechtel, but has not been submitted to the NRC for generic approval under a topical report (Reference 15.0.9-13, 15.0.9-14, 15.0.9-15). The code calculates transport and release of radioactivity for a network of nodes and flow paths defined by the user. The generalized transport equations used in the code are the same as those listed by NRC staff in applicable regulatory positions (Reference 15.0.9-16).
In particular, the code solves the time dependent Murphy-Campe Equation to calculate the activity in the control room (Reference 15.0.9-17). LOCADOSE is used to calculate offsite and control room personnel doses for the Alternative Source Terms implementation under Regulatory Guide 1.183 (Reference 15.0.9-8). Figure 15.0A.1-1 shows a simplified containment diagram containing elements of the model LOCADOSE uses to evaluate the effects of containment sprays and containment leakage.
Using the appropriate event specific assumptions, the time dependent release of nuclides is calculated. No credit is taken for cloud depletion by ground deposition or by radioactive decay during transport to the exclusion area boundary or to the outer boundary of the low population zone. The time dependent release of nuclides is used to calculate TEDE doses to the individuals at specific offsite locations. The TEDE dose is equivalent to the committed effective dose equivalent (CEDE) or inhalation dose plus the acute dose (EDE) dose for the duration of exposure to the cloud. The offsite TEDE doses are calculated using the following equations:
Offsite inhalation doses (CEDE) are calculated using the following equation:
DCEDE =            (IAR) (BR) (/Q) where:
DCEDE = CEDE dose via inhalation (rem).
DCFi = CEDE dose conversion factor via inhalation for isotope i (rem/Ci)
(Table 15.0.9-3)
(IAR)ij = integrated activity of isotope i released during the time interval j (Ci)
(BR)j = breathing rate during time interval j (m3/sec) (Table 15.6.3-10)
(/Q)j = atmospheric dispersion factor during time interval j (sec/m3) (Table 15.6.3-10)
Offsite external exposure (EDE) doses are calculated using the following equation:
DEDE =            (IAR) (/Q)
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Shearon Harris Nuclear Power Plant                                                  UFSAR Chapter: 15 where:
DEDE= external exposure dose via cloud immersion (rem)
DCFi = EDE dose conversion factor via external exposure for isotope i (rem x m3/Ci x sec) (Table 15.0.9-4)
(IAR)ij = integrated activity of isotope i released during the time interval j (Ci)
(/Q)j =atmospheric dispersion factor during time interval j (sec/m3)(Table 15.6.3-10) 15.0A.1.2 Containment Leakage Pathway After a major LOCA, containment leakage is assumed to be equal to 0.1 percent of the containment volume per day for the first 24 hours and to 50 percent of this value for the duration of this accident. For the purpose of dose calculations, this leakage has been assumed to reach the environs unfiltered, bypassing those portions of the Reactor Auxiliary Building which are maintained under negative pressure with engineered safety feature-grade filtered exhaust systems.
15.0A.1.3 Mixing Between Sprayed and Unsprayed Regions As shown in Figure 15.0A.1-1, a portion of the containment does not receive direct coverage by the containment spray system (sprayed region is conservatively modeled as 85.9% of the maximum net free volume of containment). To determine the effects of mixing between the sprayed and unsprayed regions, the two regions were modeled separately in LOCADOSE, with air and radionuclide transfer between the two regions determined by the minimum (one train, two units, half speed) containment fan cooler flow rate. No iodine removal or particulate filtration credit is applied to this transfer mechanism.
APPENDIX 15.0 - DELETED BY AMENDMENT NO. 41.
15.1      INCREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM A number of events have been postulated which could result in an increase in heat removal from the Reactor Coolant System (RCS) by the Secondary System. Detailed analyses are presented for several such events which have been identified as limiting cases.
Discussions of the following RCS cooldown events are presented in this section:
: 1) Feedwater system malfunctions that result in a decrease in feedwater temperature (ANS Condition II event).
: 2) Feedwater system malfunctions that result in an increase in feedwater flow (ANS Condition II event).
: 3) Excessive increase in secondary steam flow (ANS Condition II event).
: 4) Inadvertent opening of a steam generator relief or safety valve (ANS Condition II event).
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15
: 5) Steam system piping failure (ANS Condition III and IV events).
Section 15.0.1 contains a discussion of ANS classifications.
15.1.1    FEEDWATER SYSTEM MALFUNCTIONS THAT RESULT IN A DECREASE IN FEEDWATER TEMPERATURE 15.1.1.1    Identification of Causes and Accident Description Reductions in feedwater temperature will cause an increase in core power by decreasing reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary plant and of the RCS. The overpower/ overtemperature protection (overtemperature and overpower T trips) prevents any power increase which could lead to a departure from nucleate boiling ratio (DNBR) less than the safety analysis limit.
A reduction in feedwater temperature may be caused by the accidental opening of a feedwater bypass valve, which diverts flow around a portion of the feedwater heaters. In the event of an accidental opening of the bypass valve, there is a sudden reduction in feedwater inlet temperature to the steam generators. At power, this increased subcooling will create a greater load demand on the RCS.
With the plant at no-load conditions the addition of cold feedwater may cause a decrease in RCS temperature and thus an effective reactivity insertion due to the effects of the negative moderator coefficient of reactivity. However, the rate of energy change is reduced as load and feedwater flow decrease, so the no-load transient is less severe than the full power case.
The net effect on the RCS due to a reduction in feedwater temperature is similar to the effect of increasing secondary steam flow, i.e., the reactor will reach a new equilibrium condition at a power level corresponding to the new steam generator T.
A decrease in normal feedwater temperature is classified as an ANS Condition II event, fault of moderate frequency. See Section 15.0.1 for a discussion of ANS Condition II events.
The protection available to mitigate the consequences of a decrease in feedwater temperature is the same as that for an excessive steam flow increase, as discussed in Section 15.0.8 and listed in Table 15.0.8-1.
15.1.1.2    Analysis of Effects and Consequences Per Reference 15.1.5-13, the decrease in feedwater temperature event is bounded by the increase in secondary steam flow event (Section 15.1.3). Consequently, no analysis is presented here.
15.1.1.3    Conclusions The decrease in feedwater temperature transient is less severe than the increase in secondary steam flow event (Section 15.1.3). Based on results presented in Section 15.1.3, the applicable acceptance criteria for the decrease in feedwater temperature event have been met.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.1.2    FEEDWATER SYSTEM MALFUNCTIONS THAT RESULT IN AN INCREASE IN FEEDWATER FLOW 15.1.2.1    Identification of Causes and Accident Description Addition of excessive feedwater will cause an increase in core power by decreasing reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary plant and of the RCS. The overpower/ overtemperature protection (overtemperature and overpower T trips) prevents any power increase which could lead to a DNBR less than the safety analysis limit.
An example of excessive feedwater flow would be a full opening of a feedwater control valve due to a feedwater control system malfunction or an operator error. At power, this excess flow causes a greater load demand on the RCS due to increased subcooling in the steam generator.
With the plant at no-load conditions, the addition of cold feedwater may cause a decrease in RCS temperature and thus an effective reactivity insertion due to the effects of the negative moderator coefficient of reactivity.
Continuous addition of excessive feedwater is prevented by the steam generator high-high level trip, which closes the feedwater valves.
An increase in normal feedwater flow is classified as an ANS Condition II event, fault of moderate frequency. See Section 15.0.1 for a discussion of ANS Condition II events. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur.
: 3) The event should not generate a more serious plant condition without other faults occurring independently.
: 4) For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel element cladding, shall occur.
Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod insertability is maintained.
No single active failure will prevent operation of the RPS.
Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.1.2.2    Analysis of Effects and Consequences
: 1) Method of Analysis - The system is analyzed to demonstrate plant behavior in the event that excessive feedwater addition, due to a control system malfunction or operator error which allows a feedwater control valve to open fully, occurs.
This event is analyzed at both HFP conditions and HZP conditions. The HFP and HZP cases analyzed are:
a) HFP conditions, minimum (BOC) reactivity feedback, automatic rod control, full open main feedwater valve used to initiate the transient b) HFP conditions, maximum (EOC) reactivity feedback, manual rod control, full open main feedwater valve used to initiate the transient c) HFP conditions, maximum reactivity feedback, automatic rod control, full open main feedwater valve used to initiate the transient d) HZP conditions, maximum reactivity feedback, manual rod control, step change in main feedwater flow from 0 to 120% nominal used to initiate the transient The feedwater flow resulting from a fully open control valve is terminated by a steam generator high-high level trip signal which closes all feedwater control and isolation valves, trips the main feedwater pumps, and trips the turbine.
This event is analyzed using the SGR/Uprating values for Tavg (Tavg=588.8°F at 100% power and Tavg=557°F at HZP).
The Increase in Feedwater event is asymmetric, affecting only one of the three steam generators. Therefore, the moderator reactivity feedback is conservatively computed using the moderator temperature in the cold leg of the affected loop rather than the core average moderator temperature. This method of computing moderator feedback is conservative for EOC cases because it maximizes the reactivity insertion rate. It is also conservative for the BOC case because it results in the most aggressive rod pull.
Conservative conditions established for the analysis of this event are presented in Table 15.1.2-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. For conservatism, only the high power range and turbine trips were active in the analysis of this event. Key operating parameters used in the analysis of this event are presented in Table 15.1.2-2. The range of neutronics parameters supported by this analysis are presented in Table 15.1.2-3.
The transient response of the reactor system is calculated using the ANF RELAP (Reference 15.1.2-3) computer program. The core thermal hydraulic boundary conditions from the ANF-RELAP calculation are used as input to the XCOBRA-IIIC code (References 15.1.2-4 and 15.1.2-5) to predict the MDNBR for the event.
Results - The limiting case is the HZP case with maximum (EOC) reactivity feedback and manual rod control. This transient tripped on a high neutron flux trip. The sequence of events for the limiting case is given in Table 15.1.2-4. The responses to key system variables are given in Figures 15.1.2-1 to 15.1.2-5.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 15.1.2.3    Conclusions The results of the analysis demonstrate that the event acceptance criteria are met. Both the maximum reactor primary system pressure and the maximum secondary system pressure are less than the design limits of 2750 psia and 1320 psia, respectively. The predicted MDNBR is greater than the safety limit. The critical heat flux correlation limit ensures that, with 95%
probability and 95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. Fuel centerline melt threshold is not penetrated during this event.
The analysis for this event supports full power operation at nominal primary average temperature less than, or equal to, 588.8°F.
15.1.3    EXCESSIVE INCREASE IN SECONDARY STEAM FLOW 15.1.3.1    Identification of Causes and Accident Description An excessive increase in secondary system steam flow (excessive load increase incident) is defined as a rapid increase in steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. The Reactor Control System is designed to accommodate a 10 percent step load increase or a 5 percent per minute ramp load increase in the range of 15 to 100 percent of full power. Any loading rate in excess of these values may cause a reactor trip actuated by the Reactor Protection System. Steam flow increases greater than 10 percent are analyzed in Sections 15.1.4 and 15.1.5. Steam flow increases due to a major steam line rupture are equivalent to a step increase in steam flow of over 200 percent.
The event initiator is a 10% step increase in steam flow. The feedwater regulating valves open to increase the feedwater flow in an attempt to match the increased steam demand and maintain steam generator water level. In response to the increased steam flow, the secondary system pressure decreases, resulting in an increase in the primary-to-secondary heat transfer rate. The primary side steam generator outlet temperature decreases due to the enhanced heat removal. As a consequence, the primary system core average temperature decreases and the primary system fluid contracts, resulting in an outsurge of fluid from the pressurizer. The pressurizer level and pressure decrease as fluid is expelled from the pressurizer. If the moderator temperature coefficient (MTC) is negative, the reactor core power increases as the moderator temperature decreases due to the mismatch between the power being removed by the steam generators and the power being generated in the core.
The reactor responds to the mismatch between the power being removed by the steam generators and the power being generated in the core. The rod control system is conservatively assumed to be in the automatic state.
This accident could result from either an administrative violation such as excessive loading by the operator, or an equipment malfunction in the steam dump control or turbine speed control.
During power operation, steam dump to the condenser is controlled by reactor coolant condition signals, i.e., high reactor coolant temperature indicates a need for steam dump. A single controller malfunction does not cause steam dump; an interlock is provided which blocks the opening of the valves unless a large turbine load decrease or a turbine trip has occurred.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 Protection against an excessive load increase accident is provided by the following reactor protection system signals:
: 1) Overpower T.
: 2) Overtemperature T.
: 3) Power range high neutron flux.
Depending on the magnitude of the increase in steam demand, a reactor trip may not be activated. Instead, the reactor system will reach a new steady-state condition at a power level greater than the initial power level which is consistent with the increased heat removal rate. The final steady-state conditions which are achieved will depend upon the capacity of the turbine control valves, the magnitude of the MTC, and whether or not the rod control system is in automatic. If the MTC is positive, the reactor power would decrease as the core average coolant temperature decreased, and this event would not produce a challenge to the acceptance criteria.
An excessive load increase incident is considered to be an ANS Condition II event, fault of moderate frequency. See Section 15.0.1 for a discussion of Condition II events. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur.
: 3) The event should not generate a more serious plant condition without other faults occurring independently.
: 4) For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel element cladding, shall occur.
Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod insertability is maintained.
No single active failure has been identified for this event.
15.1.3.2    Analysis of Effects and Consequences Method of Analysis - This event is predominantly a cooldown event and is evaluated at full power conditions. At full power, the margin to limits is the smallest and, therefore, bounds operation at lower power levels. The reactor control system is designed to accommodate a 10% increase in load (step increase) or a 5% per minute load ramp for power levels between 15% and 100% of full power. The 10% step increase in load is analyzed because it is the highest expected increase that would occur. Two cases are analyzed: one for minimum neutronics feedback (BOC conditions) and the other for maximum neutronics feedback (EOC conditions). Both cases are evaluated with automatic rod control.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 Conservative conditions established for the analysis of this event are presented in Table 15.1.3-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. All trips listed in this table were modeled for the analysis of this event. Key operating parameters used in the analysis of this event are presented in Table 15.1.3-2. The transient is evaluated for operation at a reduced feedwater temperature of 375 °F from the nominal 440 °F. Operation at 375 °F provides a more limiting MDNBR that remains well above the DNB threshold. Since this is infrequent operation, the analysis presented uses 440 °F. The range of neutronics parameters supported by this analysis are presented in Table 15.1.3-3.
The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.1.2-3) computer program following the method described in Reference 15.1.5-12 and 15.1.5-
: 13. The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.1.2-5) to predict the MDNBR for the event using the Statistical Core Design (SCD) methodology (Reference 15.1.5-14) and the HTP CHF correlation (Reference 15.1.5-15) for the Framatome 17x17 Advanced W-HTP fuel.
Results - The event is initiated by a rapid opening of the turbine control valves, the atmospheric dump valves and/or the turbine bypass valves resulting in a 10% step increase in steam flow.
The maximum increased steam flow rate at full power is 110% of rated. The limiting case is the minimum (BOC) neutronics feedback case. There was no reactor trip for this transient.
The sequence of events is given in Table 15.1.3-4. The responses to key system variables are given in Figures 15.1.3-1 to 15.1.3-4.
15.1.3.3    Conclusions The results of the analysis demonstrate that the event acceptance criteria are met. Both the maximum reactor primary system pressure and the maximum secondary system pressure are less than the design limits of 2750 psia and 1320 psia, respectively. The predicted MDNBR is greater than the safety limit. The critical heat flux correlation limit ensures that, with 95%
probability and 95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. The fuel centerline melt threshold is not penetrated during this event.
15.1.4    INADVERTENT OPENING OF A STEAM GENERATOR RELIEF OR SAFETY VALVE 15.1.4.1    Identification of Causes and Accident Description The most severe core conditions resulting from an accidental depressurization of the Main Steam System are associated with an inadvertent opening of a single steam dump, power operated relief or safety valve. The analyses performed, assuming a rupture of a main steam line, are given in Section 15.1.5.
The steam release as a consequence of this accident results in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of reactor coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity.
The analysis is performed to demonstrate that the following criterion is satisfied:
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 Assuming a stuck rod cluster control assembly, and a single failure in the Engineered Safety Features System, the DNB will remain above the safety analysis limit following a steam release equivalent to the spurious opening, with failure to close, of the largest of any single steam dump, power operated relief, or safety valve. Accidental depressurization of the secondary system is classified as an ANS Condition II event. See Section 15.0.1 for a discussion of Condition II events.
The following systems provide the necessary protection against an accidental depressurization of the Main Steam Supply System.
: 1) Safety Injection System actuation from any of the following:
a) Two out of three pressurizer pressure signals.
b) Two out of three High-1 containment pressure signals.
c) Two out of three low steamline pressure signals in any one main steam line.
: 2) The overpower reactor trips (neutron flux and T), low pressurizer pressure reactor trip, and the reactor trip occurring in conjunction with receipt of the safety injection signal.
: 3) Redundant isolation of the main feedwater lines.
Sustained high feedwater flow would cause additional cooldown. Therefore, in addition to the normal control action which will close the main feedwater valves following reactor trip, a safety injection signal will rapidly close all feedwater control valves and back up feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves.
: 4) Trip of the fast-acting main steam line isolation valves (designed to close in less than 7 seconds) on:
a) High-2 containment pressure.
b) Safety injection system actuation derived from two out of three low steam line pressure signal in any one main steam line (above Permissive P-11).
c) High negative steam pressure rate indication from two out of three signals in any one main steam line (below Permissive P-11).
Plant systems and equipment which are available to mitigate the effects of the accident are also discussed in Section 15.0.8 and listed in Table 15.0.8-1.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.1.4.2    Deleted by Amendment No. 48.
15.1.4.3    Deleted by Amendment No. 48.
15.1.4.4    Event Disposition for AREVA Analysis Mode 1: The maximum steam flow through a single steam dump, power operated relief, or safety valve drives a thermal load increase less than that considered in Event 15.1.3. Ultimate reactor power level and the potential challenge to SAFDLs is greater for Event 15.1.3. The event is therefore bounded by Event 15.1.3 before trip and by Event 15.1.5 after trip, since the Condition II criteria are met by the more challenging Condition IV event.
Mode 2: The reactor will achieve a steady state power level equal to its initial power plus the additional load imposed by the steam flow through the failed valve. Because the initial power level is less than 5% of rated, and the additional load is less than 10% of rated power, the reactor power will not rise to a level at which a significant challenge to SAFDLs is posed. Event 15.1.4 is bounded in Mode 2 by the "Off-site Power Available" Case analyzed for Mode 2 of Event 15.1.5, "Steamline Break". The thermal load increase associated with the latter event is significantly greater than encountered here and the "Off-site Power Available" Steamline Break Cases continue to meet the Condition II acceptance criteria.
Mode 3: The event will proceed as described for Mode 2, except that the reactor is subcritical by at least 1000 pcm at event initiation for Mode 3. The initial margin to criticality for Mode 3 will slow the evolution of the event relative to Mode 2. Time available for operator response is greater than in Mode 2. Event 15.1.4 is bounded in Mode 3 by the "Off-site Power Available" Case analyzed for Mode 2 of Event 15.1.5, "Steamline Break". The thermal load increase associated with the latter event is significantly greater than encountered here and the "Off-site Power Available" Steamline Break Cases continue to meet the Condition II acceptance criteria.
Modes 4, 5, and 6: The reactor coolant temperature and secondary pressure are significantly reduced relative to Mode 3 for these modes. Thus, the potential cooldown is smaller than that in Mode 3 due to the decreased steam flow through the affected valve in these modes. The challenge to SAFDLs is reduced directly by the lower coolant temperatures. The Mode 3 event, therefore, bounds the event in Modes 4, 5, and 6. In Modes 5 and 6, the event cannot occur with consequences because the primary and secondary temperatures are below saturation at atmospheric pressure.
15.1.5    STEAM SYSTEM PIPING FAILURE 15.1.5.1    Identification of Causes and Accident Description The steam release arising from a rupture of a main steam line would result in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the Reactor Coolant System (RCS) causes a reduction of reactor coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. If the most reactive rod cluster control assembly (RCCA) is assumed stuck in its fully withdrawn position after reactor trip, there is an increased possibility that the core will become critical and return to power. A return to power following a steam line rupture is a potential problem mainly because of the high power peaking factors which exist assuming the most reactive RCCA to be stuck in its fully withdrawn position.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 The core is ultimately shut down by the boric acid injection delivered by the Safety Injection System.
A major steam line rupture is classified as an ANS Condition IV event. See Section 15.0.1 for a discussion of ANS Condition IV events. The acceptance criteria for this event are:
: 1) Pressure in the reactor coolant and main steam systems should be maintained below acceptable design limits.
: 2) Any fuel failure calculated to occur must be of sufficiently limited extent that the core will remain in place and intact with no loss of core cooling capability.
: 3) The integrity of the reactor coolant pumps should be maintained, such that a loss of AC power and containment isolation will not result in pump seal damage.
: 4) The auxiliary feedwater system must be safety grade and, when required, automatically initiated.
: 5) Tripping of the reactor coolant pumps should be consistent with the resolution of TMI Action Plan item II.K.3.5.
: 6) The radiological analysis and acceptance criteria are described in the Alternative Source Term Regulatory Guide 1.183, Reference 15.0.9-8.
a) For a Main Steam Line Break (MSLB) with an assumed pre-accident iodine spike or with accident induced fuel failure, the calculated doses should be less than the 10CFR50.67 limits of 25 rem TEDE offsite and 5 rem TEDE in the Control Room.
b) For a Main Steam Line Break (MSLB) with an accident initiated iodine spike, the calculated offsite doses should be less than 10% of the 10 CFR 50.67 limits (i.e.,
less than 2.5 rem TEDE) and the control room doses should be less than the 10CFR50.67 limit of 5 rem TEDE.
Effects of minor secondary system pipe breaks are bounded by the analysis presented in this section. Minor secondary system pipe breaks are classified as ANS Condition III events, as described in Section 15.0.1.3.
The following functions provide the protection for a steam line rupture:
: 1) Safety Injection System actuation from any of the following:
a) Two out of three low pressurizer pressure signals.
b) Two out of three Hi-1 Containment pressure signals.
c) Two out of three low steam line pressure signals in any one main steam line.
: 2) The overpower reactor trips (neutron flux and T), low pressurizer pressure reactor trip, and the reactor trip occurring in conjunction with receipt of the safety injection signal.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15
: 3) Redundant isolation of the main feedwater lines - Sustained high feedwater flow would cause additional cooldown. Therefore, in addition to the normal control action which will close the main feedwater valves after a reactor trip, a safety injection signal will rapidly close all feedwater control valves and backup feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves.
: 4) Trip of the fast acting main steam line isolation valves (designed to close in less than 7 seconds, which represents the signal delay plus valve stroke time) on:
a) Hi-2 Containment pressure.
b) Safety Injection System actuation derived from two out of three low steam line pressure signals in any one main steam line (above P-11).
c) High negative steam pressure rate indication from two signals in any main steam line (below Permissive P-11).
Fast-acting isolation valves are provided in each main steam line; these valves will fully close following a large break in the steam line. For breaks downstream of the isolation valves, closure of all valves would completely terminate the blowdown. For any break, in any location, no more than one steam generator would experience an uncontrolled blowdown even if one of the isolation valves fails to close. A description of steam line isolation is included in Chapter 10.
Flow restrictors are installed in the steam generator outlet nozzle and are an integral part of the steam generator. The effective throat area of the nozzles is 1.4 square ft., which is considerably less than the main steam pipe area; thus, the nozzles also serve to limit the maximum steam flow for a break at any location.
The AFW system is designed to detect conditions indicative of a steam line or feedwater line break and automatically isolate AFW flow to the affected Steam Generator. For the MSLB Event, this helps to reduce the heat removal via the ruptured Steam generator loop and thus reduces the reactivity feedback to the "affected" sector of the core.
Table 15.1.5-1 lists the equipment required in the recovery from a high energy line rupture. Not all equipment is required for any one particular break, since the requirements will vary depending upon postulated break location and details of balance of plant design and pipe rupture criteria as discussed elsewhere in the FSAR. Design criteria and methods of protection of safety-related equipment from the dynamic effects of postulated piping ruptures are provided in Section 3.6.
15.1.5.2      Analysis of Effects and Consequences Method of Analysis - Four base scenarios were considered. The reactor was assumed to be initially operating at either hot full power (HFP) or hot zero power (HZP) conditions. From both of these initial conditions, the transient was assumed to occur either with or without offsite power. With offsite power available, reactor coolant pumps remained operating throughout the entire transient. With loss of offsite power, the pumps were tripped coincident with turbine trip.
For these base scenarios, the most reactive control rod was assumed to be stuck out of the core.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 The HFP cases were initiated at a nominal power level of 2948 MWt. The results of HFP cases are driven by the maximum rate of positive moderator reactivity insertion, which is predominantly a function of the largest break flow rate (i.e. largest break size) and the most negative moderator temperature coefficient (MTC).
The methodology for analyzing MSLB events is documented in Reference 15.1.5-13. Three computer codes are used in the analysis of this event: RETRAN-3D (Reference 15.1.2-3),
SIMULATE-3 (Reference 15.1.5-4), and VIPRE-01 (Reference 15.1.2-5). First, RETRAN-3D is used to calculate the general system thermal-hydraulic responses during an MSLB. RETRAN-3D produces a set of core boundary conditions for each case analyzed. The boundary conditions include core inlet flows and temperatures, core exit pressure, and core averaged fuel surface heat flux. The time of peak core averaged fuel surface heat flux as calculated by RETRAN-3D is used as a time for transferring boundary conditions to SIMULATE-3 and VIPRE-01.
Second, SIMULATE-3 is used to calculate the axial and radial power distributions and reactivity at the time of MDNBR and peak core averaged fuel surface heat flux as calculated by RETRAN-3D. The calculated power distributions are inputs to VIPRE-01. The reactivity calculated by SIMULATE-3 is compared with the reactivity calculated by RETRAN-3D to ensure that conservatism exists in the RETRAN-3D representation of reactivity feedback.
Fuel failure analysis is based on DNB and fuel centerline melt criteria. DNB criterion is evaluated using the VIPRE-01 code and is based on the CHF correlations given in Reference 15.1.5-13 for the HZP analysis and the HTP (Reference 15.1.5-8) correlation for the HFP analysis. DNB propagation is considered. Fuel centerline melt criterion is evaluated using the maximum post-scram core average fuel temperature from RETRAN-3D and the nuclear heat flux hot channel factor from SIMULATE-3.
The assumptions used in the analysis of this event are as follows:
: 1) A double ended guillotine break is assumed to occur in the loop 1 steam line, downstream of the flow restrictor and at the main steam isolation valve (MSIV). A double ended guillotine break will lead to maximum steam flow out of the break which results in maximum cooldown and maximum return to power. The flow is choked at the flow restrictor, having an area of 1.4 ft2. Only steam is allowed to flow out the break.
The break flows are calculated based on the Moody critical flow model. On the steam generator side of the break, steam flows out the break throughout the entire transient.
On the MSIV side of the break, break flow terminates after the MSIVs are fully closed.
The MSIVs are fully closed 7 seconds after reaching the low steam line pressure point.
: 2) For conservatism, no steam generator tube plugging is assumed for the MSLB analysis.
: 3) The single failure assumed in this analysis is loss of one of two Charging/Safety Injection Pumps (CSIP's). Although three CSIP's are physically present, only two of the three are electrically connected and considered operational at any given time. Therefore, in the analysis, only one CSIP is assumed to be available to mitigate the event. The pump is assumed to take suction from the refueling water storage tank (RWST) at 40°F with a boron concentration of 2400 ppm and discharge through the boron injection tank (BIT).
Initially, the BIT is assumed to be filled with unborated water. The water in the injection lines between the RWST and cold legs is also assumed to be unborated. The time Amendment 63                                                                          Page 33 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 required to flush this unborated water from the BIT and the injection lines is included in the RETRAN-3D model. The single failure assumed in this analysis plays an important role in ensuring long term subcriticality in the core.
: 4) At break initiation, maximum main feed water (MFW) is assumed, directed entirely to the faulted steam generator. MFW is isolated 8 seconds after receiving the isolation signal.
Auxiliary feed water (AFW) is assumed to start at SI signal without delay. The AFW pumps are assumed to take suction from the condensate storage tank (CST) at 40°F.
All the AFW is assumed to be delivered to the broken steam generator. Although isolation of AFW flow to the faulted steam generator will occur within about 60 seconds, the analysis conservatively delayed AFW isolation until 600 seconds.
: 5) The reactor kinetics in RETRAN-3D is calculated using a point kinetics model. The total reactivity calculated by RETRAN-3D is compared with the reactivity calculated by SIMULATE-3 at selected conditions during the transient to ensure that conservatism exists in the RETRAN-3D representation of reactivity feedback. SIMULATE-3 provides a three-dimensional, two-group diffusion calculation of the reactivity feedback.
: 6) The most negative technical specification value of moderator temperature coefficient (MTC), -50 pcm/°F is used in the MSLB analysis. Other neutronic parameters, are based on values consistent with end-of-cycle (EOC) conditions.
This analysis considers only the response of the primary and secondary systems and does not include containment response for structural or equipment qualification purposes.
Results -
Significant margin with respect to fuel centerline melting exists for all MSLB cases (both HFP and HZP). The most limiting case with respect to the approach to DNBR is the HZP MSLB with offsite power available and with the stuck rod. The sequence of events for this case is summarized in Table 15.1.5-3. In addition, some key system parameters describing the transient are illustrated in Figures 15.1.5-1 through 15.1.5-6. With offsite power available, the primary reactor coolant pumps (RCP) operate throughout the event. In all cases, HHSI is actuated 37 seconds after the low steamline pressure setpoint is reached to account for diesel startup time and the time to switch suction from the volume control tank to the refueling water storage tank. The transient is terminated at 600 seconds.
As shown in Figure 15.1.5-1, the pressure in all three steam generators decreases immediately after transient initiation due to steam flow out of the break (Figure 15.1.5-2). After MSIV closure, the flow from the unaffected steam generators is terminated and the pressure recovers.
The pressure of the affected and unaffected steam generators decrease slowly thereafter to reflect the continuing break flow of the affected steam generator (Figure 15.1.5-2).
The break flow (Figure 15.1.5-2) from down stream of MSIV is terminated when the MSIVs close. After reaching an early maximum, the break flow from the upstream of MSIV gradually decreases throughout the remainder of the transient.
The mass inventory (Figure 15.1.5-3) increases for the affected generator but decreases immediately for the unaffected steam generators before the MSIVs close to reflect the MFW and AFW assumptions. Thereafter, the mass in the unaffected steam generators stabilizes and Amendment 63                                                                        Page 34 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 remains essentially constant. After the MSIV closure and the AFW actuation, the mass in the affected steam generator decreases until the end of the simulation.
The pressurizer pressure (Figure 15.1.5-4) decreases rapidly during the initial phase of the transient due to thermal contraction of the primary system coolant. The pressurizer pressure becomes stable at approximately 40 seconds due to the addition of inventory via the SIS. After 300 seconds the pressurizer pressure increases until the end of the simulation.
The reactor power is initially at 2.948E-6 MW. The cooldown results in power increasing significantly at about 30 seconds (Figure 15.1.5-5), reaching a peak of approximately 739 MW at 244 seconds, and then begins to decline due to boron injection.
Initially, the total core reactivity (Figure 15.1.5-6) increases due to moderator and Doppler feedback associated with the primary cooldown. Once the reactor power begins to increase, the Doppler feedback changes from positive to negative and the reactor is brought to a quasi steady-state. The boron component of reactivity begins to show an effect at approximately 252 seconds as boron from the HHSI system begins to reach the core.
Long Term Effects and Events - Following the hypothetical steamline break incident, a steamline isolation signal will be generated almost immediately, causing the steamline isolation valves to close within a few seconds. If the break is downstream of the isolation valves, all of which subsequently close, the break will be isolated. If the break is upstream of the isolation valves, or if one valve fails to close, the break will be isolated to two steam generators while the faulted one will continue to blow down. Only the case in which the steam generator continues to blow down is discussed here since the downstream break followed by isolation of all steam generators will terminate the transient.
An excessive cooldown protection signal will cause main feedwater isolation to occur. The only source of water available to the faulted steam generator is then the auxiliary feedwater system.
The first required operator action is to identify the faulted steam generator and then isolate the auxiliary feedwater flow to that steam generator if this has not been accomplished by the Automatic AFW Isolation System. Table 15.1.5-7 provides a complete assessment of the operator's role. Following steamline isolation, steam pressure in the steamline with the faulted steam generator will continue to fall rapidly, while the pressure stabilizes in the remaining two steam lines. The indication of the different steam pressures will be available to the operator within a few seconds of the steamline isolation. This will provide the necessary information to identify the faulted steam generator so that auxiliary feedwater to it can be isolated. The operator is instructed by Emergency Operating Procedures to isolate the affected steam generator by shutting the steam generator AFW isolation valve. Manual controls are provided in the control room for start and stop of the AFW pumps and for the control and isolation valves associated with the AFW system. The means for detecting the faulted steam generator and isolating auxiliary feedwater to it requires only the use of safety grade equipment available following the break.
Following the automatic safety injection actuation and after the faulted steam generator is completely isolated, the continued operation of the safety injection system will repressurize the reactor coolant system and continue to increase the RCS volume inventory. The second required operator action is to manually control the repressurization of the reactor coolant system and stop one Charging/Safety Injection Pump (CSIP) to control pressurizer level. The operator may then establish normal charging and letdown to restore pressure and level control. The Amendment 63                                                                          Page 35 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 operator has available, in the control room, an indication of pressurizer level from the instrumentation in the reactor protection system. After the indicated water level returns to the pressurizer, RCS pressure has returned to the normal range and the RCS is sufficiently subcooled, the operator is instructed to stop the CSIP's, reestablish normal charging and letdown flows, and reestablish operation of the pressurizer heaters to maintain a steam bubble in the pressurizer to limit system repressurization. The pressurizer level instrumentation and manual controls for operation of the CSIP's meet the required standards for safety systems.
The removal of decay heat in the long-term (following the initial cooldown) using the remaining intact steam generators requires only the auxiliary feedwater system as a water source and the secondary system safety valves to relieve steam.
The requirements to terminate auxiliary feedwater flow to the faulted steam generator, reestablishing normal charging and letdown flows, and reestablishing operation of the pressurizer heaters can be met by simple switch actions by the operator. Thus the required actions to limit the cooldown and repressurization can be easily recognized, planned and performed within the necessary time. For decay heat removal and plant cooldown the operator has a considerably longer time period in which to respond because of the large initial cooldown associated with a steamline break transient.
15.1.5.3    Radiological Consequences of a Postulated Steamline Break The analysis of the steam line break (SLB) radiological consequences uses the analytical methods and assumptions outlined in the RG 1.183, Appendix E. A summary of input parameters and assumptions is provided in Table 15.1.5-5. Additional clarification is provided as follows:
a) The noble gas activity concentration in the RCS at the time the accident occurs is based on a one-percent fuel defect level. This is approximately equal to the Technical Specification value of 100/E bar Ci/gm for gross radioactivity. The noble gas concentrations in the RCS are given in Table 15.0.9-2. The iodine activity concentration of the secondary coolant at the time the SLB occurs is assumed to be equivalent to the Technical Specification limit of 0.1 Ci/gm of DE I-131. The iodine secondary coolant activity concentration is given in Table 15.0.9-7.
b) The amount of primary to secondary SG tube leakage is assumed to be 1 gpm total.
The primary to secondary SG tube leakage is apportioned between the affected and unaffected SGs to provide a conservative result. Leakage to the affected (faulted) SG is directly released to the atmosphere thus using 0.35 gpm to the affected SG and 0.65 gpm to the two unaffected SGs would maximize the dose.
c) The SG connected to the broken steam line is assumed to boil dry within the initial two minutes following the SLB. The entire liquid inventory of this SG is assumed to be steamed off and all of the iodine initially in this SG is released to the environment. In addition, iodine carried over to the faulted SG by tube leaks is assumed to be released directly to the environment with no credit taken for iodine retention in the SG.
d) In the intact SGs an iodine partition factor of 0.01 (curies iodine/gm steam)/ (curies iodine/gm water) is used.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 e) All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.
f)  Eight hours after the accident, the RHR System is assumed to be placed into service for heat removal. After eight hours there are no further steam releases to the atmosphere from the intact steam generators.
g) Within 40 hours after the accident, the reactor coolant system has been cooled to below 212°F, and there are no further steam releases to atmosphere from the faulted steam generator.
h) No fuel damage is postulated to occur for the limiting SLB event. Therefore, per RG 1.183, Appendix E, two cases of iodine spiking are assumed.
15.1.5.3.1 Pre-accident Iodine Spike Case It is assumed that a reactor transient has occurred prior to the SLB and has raised the RCS iodine concentration to the Technical Specification limit for a transient of 60 Ci/gm of dose equivalent (DE) I-131. The pre-accident spike iodine concentrations are given in Table 15.0.9-7.
15.1.5.3.2 Accident-Initiated Iodine Spike Case The reactor trip associated with the SLB creates an iodine spike in the RCS which increases the iodine release rate from the fuel to the RCS to a value 500 times greater than the release rate corresponding to a maximum equilibrium RCS concentration of 1.0 Ci/gm of DE I-131. The iodine spike appearance rates are given in Table 15.0.9-6. The duration of the accident-initiated iodine spike is limited by the amount of activity available in the fuel-clad gap. Based on having 12 percent of the iodine in the fuel-clad gap, the gap inventory would be depleted within 5.0 hours and the spike is terminated at that time.
15.1.5.3.3 Fuel Failure Case Deleted by Amendment No. 63.
15.1.5.3.4 Offsite Doses The offsite doses are calculated using the assumptions and equations in Section 15.0A.1 15.1.5.3.5 Control Room Doses The control room assumptions are provided in Section 15.6.5.4.3 and Table 15.6.5-15.
In the event of SLB, the low steamline pressure SI setpoint will be reached shortly after event initiation. The SI signal causes the control room HVAC to switch from the normal operation mode to the post-accident recirculation mode of operation. It is assumed that the Si signal is generated at 10 seconds. The control room HVAC switches from normal operation to post-accident recirculation mode of operation at 25 seconds (10 seconds for SI signal plus 15 second delay time). Two hours after the control room HVAC is in post-accident recirculation mode an operator action switches the control room to the pressurization mode.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.1.5.3.6 Results The potential radiological consequences resulting from the occurrence of a postulated main steam line break have been conservatively analyzed, using assumptions and modes in the previous sections. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The radiological analysis results for this event are listed in Table 15.1.5-6. The resultant doses are within the applicable limits.
15.1.5.4    Conclusions The methodology for analyzing the MSLB event provided a conservative method of calculating the system and core responses during an MSLB. Significant margin from the fuel centerline melt standpoint exists for all MSLB cases (both HFP and HZP). The most limiting scenario from the MDNBR standpoint was the case initiated from HZP with offsite power available and with the stuck rod. Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable and not precluded by the acceptance criteria presented in Section 15.1.5.1, the analysis shows that no fuel failure is predicted to occur as a result of this accident.
The radiological consequences presented in Section 15.1.5.3 are within the applicable limits.
 
==REFERENCES:==
SECTION 15.1 15.1.2-1        Deleted by Amendment No. 48.
15.1.2-2        Deleted by Amendment No. 45.
15.1.2-3        EPRI NP-7450(A), Revision 10, RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, September 2014.
15.1.2-4        Deleted by Amendment No. 63.
15.1.2-5        EPRI NP-2511-CCM-A, Revision 4, VIPRE A Thermal-Hydraulic Code for Reactor Cores, June 2007.
15.1.4-1        Deleted by Amendment No. 48.
15.1.5-1        Deleted by Amendment No. 48.
15.1.5-2        Deleted by Amendment No. 63.
15.1.5-3        Deleted by Amendment No. 63.
15.1.5-4        STUDSVIK/SOA-95/18, Revision 0, SIMULATE-3 Methodology Advanced Three-Dimensional Two-Group Reactor Analysis Code, 1995.
15.1.5-5        Deleted by Amendment No. 51 15.1.5-6        Deleted by Amendment No. 48.
15.1.5-7        Deleted by Amendment No. 51 Amendment 63                                                                        Page 38 of 151
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 15.1.5-8      EMF-92-153(P)(A) Revision 1, HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, January 2005.
15.1.5-9      Deleted by Amendment No. 63.
15.1.5-10      Deleted by Amendment No. 63.
15.1.5-11      Deleted by Amendment No. 63.
15.1.5-12      DPC-NE-3008, Revision 0, Thermal-Hydraulic Models for Transient Analysis, April 2018.
15.1.5-13      DPC-NE-3009, Revision 0, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, April 2018.
15.1.5-14      DPC-NE-2005-PA, Revision 5, Thermal-Hydraulic Statistical Core Design Methodology, March 2016.
15.1.5-15      EMF-92-153-PA, Revision 1, HTTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, January 2005.
15.2  DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM A number of transients and accidents have been postulated which could result in a reduction of the capacity of the secondary system to remove heat generated in the Reactor Coolant System (RCS). Detailed analyses are presented in this section for several such events which have been identified as more limiting than the others.
Discussion of the following RCS coolant heatup events are presented in Section 15.2:
: 1) Steam pressure regulator malfunction or failure that results in decreasing steam flow (ANS Condition II event).
: 2) Loss of external electrical load (ANS Condition II event).
: 3) Turbine trip (ANS Condition II event).
: 4) Inadvertent closure of main steam isolation valves (ANS Condition II event.
: 5) Loss of condenser vacuum and other events resulting in turbine trip (ANS Condition II event).
: 6) Loss of nonemergency AC power to the station auxiliaries (ANS Condition II event).
: 7) Loss of normal feedwater flow (ANS Condition II event).
: 8) Feedwater system pipe break (ANS Condition IV event).
Section 15.0.1 contains a discussion of ANS classifications.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.2.1    STEAM PRESSURE REGULATOR MALFUNCTION OR FAILURE THAT RESULTS IN DECREASING STEAM FLOW There are no steam pressure regulators in the Shearon Harris Nuclear Power Plant (SHNPP) whose failure or malfunction could cause a steam flow transient.
15.2.2    LOSS OF EXTERNAL ELECTRICAL LOAD 15.2.2.1    Identification of Causes and Accident Description A major load loss on the plant can result from loss of external electrical load due to some electrical system disturbance. Offsite alternating current (AC) power remains available to operate plant components such as the reactor coolant pumps; as a result, the onsite emergency diesel generators are not required to function for this event. Following the loss of generator load, an immediate fast closure of the turbine control valves will occur. This will cause a sudden reduction in steam flow, resulting in an increase in pressure and temperature in the steam generator shell. As a result, the heat transfer rate in the steam generator is reduced, causing the reactor coolant temperature to rise, which in turn causes reactor coolant expansion, pressurizer insurge, and RCS pressure rise.
For a loss of external electrical load without subsequent turbine trip, no direct reactor trip signal would be generated. The plant would be expected to trip from the Reactor Protection System if a safety limit were approached. A continued steam load of approximately 5 percent would exist after total loss of external electrical load because of the steam demand of plant auxiliaries.
In the event that a safety limit is approached, protection would be provided by the high pressurizer pressure and overtemperature T trips. Voltage and frequency relays associated with the reactor coolant pump provide no additional safety function for this event. Following a complete loss of external load, the maximum turbine overspeed would be approximately 8 to 9 percent, resulting in an overfrequency of less than 6 hertz (hz). Any degradation in their performance could be ascertained at that time. Any increased frequency to the reactor coolant pump motors will result in slightly increased flow rate and subsequent additional margin to safety limits. For postulated loss of load and subsequent turbine generator overspeed, any overfrequency condition is not seen by other Reactor Protection System equipment. Reactor Protection System equipment is supplied from the 118 volt AC instrument power supply system, which in turn is supplied from the inverters; the inverters are supplied from a direct current (DC) bus energized from batteries or by a rectified AC voltage from safeguards buses. Safeguards loads are transferred to offsite power or, alternately, to standby diesel generators upon loss of the turbine.
In the event the steam dump valves fail to open following a large loss of load, the steam generator safety valves may lift and the reactor may be tripped by the high pressurizer pressure signal, the high pressurizer water level signal, or the overtemperature T signal. The steam generator shell side pressure and reactor coolant temperatures will increase rapidly. The pressurizer safety valves and steam generator safety valves are, however, sized to protect the RCS and steam generator against overpressure for all load losses without assuming the operation of the Steam Dump System, pressurizer spray, pressurizer power operated relief valves, automatic rod cluster control assembly control, or direct reactor trip on turbine trip.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 The steam generator safety valve capacity is sized to remove the steam flow from the steam generator without exceeding 110 percent of the steam system design pressure. The pressurizer safety valves capacity are sized based on a complete loss of heat sink with the plant initially operating at the maximum calculated turbine load, along with operation of the steam generator safety valves. The pressurizer safety valves are then able to relieve sufficient steam to maintain the RCS pressure within 110 percent of the RCS design pressure.
A more complete discussion of overpressure protection can be found in Section 5.2.2.
A loss of external load is classified as an ANS Condition II event, fault of moderate frequency.
See Section 15.0.1 for a discussion of Condition II events.
The primary side transient is caused by a decrease in heat transfer capability from primary to secondary due to a rapid termination of steam flow to the turbine, accompanied by an automatic reduction of feedwater flow (should feed flow not be reduced, a larger heat sink would be available and the transient would be less severe). Termination of steam flow to the turbine following a loss of external load occurs due to automatic fast closure of the turbine control valves in approximately 0.3 seconds. Following a turbine trip event, termination of steam flow occurs via turbine stop valve closure, which occurs in approximately 0.1 seconds. Therefore, the transient in primary pressure, temperature, and water volume will be less severe for the loss of external load than for the turbine trip due to a slightly slower loss of heat transfer capability.
Therefore, a detailed transient analysis is not presented for the loss of external load.
The protection available to mitigate the consequences of a loss of external load is the same as that for a turbine trip, as listed in Table 15.0.8-1.
15.2.2.2    Analysis of Effects and Consequences Refer to Section 15.2.3.2 for the method used to analyze the limiting transient (turbine trip) in this grouping of events. The results of the turbine trip event analysis are more severe than those expected for the loss of external load, as discussed in Section 15.2.2.1.
Normal reactor control systems are assumed to function only if their operation results in more severe accident conditions. The engineered safety systems are not required to function. The Auxiliary Feedwater System may, however, be automatically actuated following a loss of main feedwater; this will further mitigate the effects of the transient.
The Reactor Protection System may be required to function following a complete loss of external load to terminate core heat input and prevent departure from nucleate boiling (DNB).
Depending on the magnitude of the load loss, pressurizer safety valves and/or steam generator safety valves may be required to open to maintain system pressure below allowable limits. No single active failure will prevent operation of any system required to function. Refer to Reference 15.8.0-1 for a discussion of anticipated transients without scram (ATWS) considerations.
15.2.2.3    Conclusions Based on results obtained for the turbine trip event (Section 15.2.3) and considerations described in Section 15.2.2.1, the applicable acceptance criteria for a loss of external load event are met.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.2.3      TURBINE TRIP 15.2.3.1      Identification of Causes and Accident Description For a turbine trip event, the reactor would be tripped directly (unless below approximately 10 percent power) from a signal derived from the turbine stop emergency trip fluid pressure and turbine stop valves. The turbine stop valves close rapidly (typically 0.1 seconds) on loss of trip fluid pressure actuated by one of a number of possible turbine trip signals. Mechanical turbine trip initiation signals include:
: 1) Generator trip
: 2) Low condenser vacuum
: 3) Loss of lubricating oil
: 4) Turbine thrust bearing failure
: 5) Turbine overspeed
: 6) Manual trip
: 7) Loss of power to TCS Controllers
: 8) Steam Generator HI-HI Level
: 9) ATWS Mitigating System initiated
: 10) Reactor trip
: 11) Safety injection
: 12) Closure of 4 throttle valves at 10% power or greater Upon initiation of stop valve closure, steam flow to the turbine stops abruptly. Sensors on the stop valves detect the turbine trip and initiate steam dump and, if above 10 percent power, a reactor trip. The loss of steam flow results in an almost immediate rise in secondary system temperature and pressure with a resultant primary system transient as described in Section 15.2.2.1 for the loss of external load event. A slightly more severe transient occurs for the turbine trip event due to the more rapid loss of steam flow caused by the more rapid valve closure.
The automatic Steam Dump System would normally accommodate the excess steam generation. Reactor coolant temperatures and pressure do not significantly increase if the Steam Dump System and Pressurizer Pressure Control System are functioning properly. If the turbine condenser was not available, the excess steam generation would be dumped to the atmosphere and main feedwater flow would be lost. For this situation feedwater flow would be maintained by the Auxiliary Feedwater System to ensure adequate residual and decay heat removal capability. Should the Steam Dump System fail to operate, the steam generator safety Amendment 63                                                                      Page 42 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 valves may lift to provide pressure control. See Section 15.2.2.1 for a further discussion of the transient.
A turbine trip is classified as an ANS Condition II event, fault of moderate frequency. See Section 15.0.1 for a discussion of ANS Condition II events. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur.
: 3) The event should not generate a more serious plant condition without other faults occurring independently.
: 4) For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel element cladding, shall occur.
Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod insertability is maintained.
No single active failure will prevent operation of any system required to function.
Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
Reactor Trip and OTT. Trip response times are presented in FSAR Section 15.0.6. The primary system is protected against overpressurization by the pressurizer safety and relief valves. Pressure relief on the secondary side is afforded by the steam line safety/relief valves.
A turbine trip event is more limiting than loss of external load, loss of condenser vacuum, and other decrease in heat removal events. As such, this event has been analyzed in detail.
Results and discussion of the analysis are presented in Section 15.2.3.2.
15.2.3.2    Analysis of Effects and Consequences Method of Analysis - The purpose of analyzing this event is to demonstrate that the primary and secondary pressure relief capability is sufficient to limit the pressures to less than 110% of their respective design values. This event is also analyzed to ensure that the reactor protection system is properly set to prevent penetration of the SAFDLs under the limiting assumptions of no credit for a direct reactor trip on turbine trip.
Three cases are analyzed for this event: one challenging the primary overpressurization criterion, one challenging the secondary overpressurization criterion, and one challenging the fuel design limits. In all cases, the input parameters are biased (BOC kinetics) to maximize the increase in reactor power during the transient. However, in the first case, the parameters and equipment operational states are selected to maximize the primary system overpressurization, in the second case the parameters and equipment states have been selected to maximize the Amendment 63                                                                          Page 43 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 secondary side overpressurization, and in the final case the parameters and equipment states have been selected to reduce the primary system pressurization which provides a conservative estimate of the minimum DNBR during the transient.
In this analysis, the behavior of the Unit is evaluated for a complete loss of steam load from the Measurement Uncertainty Recapture (MUR) power plus uncertainty without direct reactor trip primarily to show the adequacy of the pressure relieving devices and also to demonstrate core protection margins; that is, the turbine is assumed to trip without actuating any of the sensors for reactor trip on the turbine stop valves. This assumption delays reactor trip until conditions in the RCS result in a trip due to other signals (high pressurizer pressure, OTT, high neutron flux, high pressurizer water level, and low-low steam generator water level). Thus, the analysis assumes a worst transient. In addition, no credit is taken for steam dump. Main feedwater flow is terminated at the time of turbine trip, with no credit taken for auxiliary feedwater to mitigate the consequences of the transient.
Conservative conditions established for the analysis of this event are presented in Table 15.2.3-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. The trip setpoints and time delays assumed in the analysis of this event are also listed in Table 15.0.6-2. All relevant trips listed in this table except reactor trip OPT were modeled for the analysis of this event. For all cases analyzed, the reactor trip on turbine trip is also disabled. Key operating parameters used in the analysis of this event are presented in Table 15.2.3-2. The range of neutronics parameters supported by this analysis are presented in Table 15.2.3-3. Part power cases were analyzed for the DNB case to assure the Technical Specification limits on MTC are supported.
See Section 15.0 for additional discussion on the turbine trip analysis.
Major assumptions are summarized below:
: 1) Reactor Control - From the standpoint of the maximum pressures attained it is conservative to assume that the reactor is in manual control. If the reactor were in automatic control, the control rod banks would move prior to trip and reduce the severity of the transient.
: 2) Steam Release - No credit is taken for the operation of the Steam Dump System or steam generator power operated relief valves.
: 3) Pressurizer Spray and Power Operated Relief Valves:
a) For the secondary side overpressurization and the DNB cases, the pressurizer spray and power operated relief valves are conservatively assumed to operate in reducing or limiting the reactor coolant pressure. Safety valves are also available.
b) For the primary side overpressurization case, no credit is taken for the effect of pressurizer spray and power operated relief valves in reducing or limiting the reactor coolant pressure. Safety valves are operable.
: 4) Feedwater Flow - Main feedwater flow to the steam generators is assumed to be lost at the time of turbine trip. No credit is taken for auxiliary feedwater flow since a stabilized plant condition will be reached before auxiliary feedwater initiation is normally assumed Amendment 63                                                                          Page 44 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 to occur; however, the auxiliary feedwater pumps would be expected to start on a trip of the main feedwater pumps. The auxiliary feedwater flow would remove core decay heat following plant stabilization.
: 5) Reactor trip is actuated by the first reactor protection system trip setpoint reached with no credit taken for the direct reactor trip on the turbine trip. Trip signals are expected due to high pressurizer pressure, overtemperature T, high neutron flux, high pressurizer water level, and low-low steam generator water level.
The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.2.3-5) computer program following the method described in Reference 15.2.8-4 and Reference 15.2.8-5. The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.2.8-1) to predict the MDNBR for the event using the Statistical Core Design (SCD) Methodology (Reference 15.2.8-2) and the HTP CHF correlation (Reference 15.2.8-3) for the Framatome 17x17 Advanced W-HTP fuel.
Results - For the overpressurization cases, the peak reactor primary system pressure and the peak secondary system pressures are less than 110% of design limits. The predicted MDNBR is greater than the safety limit. The primary and secondary overpressurization transients tripped on high pressurizer pressure and OTT, respectively. The DNB transient tripped on OTT.
The sequences of events for the overpressurization cases are given in Table 15.2.3-4 and Table 15.2.3-5. The responses to key system variables for the primary overpressurization case are given in Figures 15.2.3-1 to 15.2.3-4 and for the secondary side overpressurization, these responses are presented in Figures 15.2.3-9 to 15.2.3-12.
The sequence of events for the DNB case is given in Table 15.2.3-6. The responses to key system variables are given in Figures 15.2.3-5 to 15.2.3-7. Part power cases indicated that the full power cases with 0.0 MTC was the limiting DNB case.
15.2.3.3    Conclusions The results of the analysis demonstrate that the event acceptance criteria are met. For the overpressurization cases, both the maximum reactor primary system pressure and the maximum secondary system pressure are less than the design limits of 2750 psia and 1320 psia, respectively. For the DNB case, the predicted MDNBR is greater than the safety limit.
The critical heat flux correlation limit ensures that, with 95% probability and 95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. This event does not pose a credible challenge to FCM safety criteria.
15.2.4    INADVERTENT CLOSURE OF MAIN STEAM ISOLATION VALVES Inadvertent closure of the main steam isolation valves (MSIVs) would result in a complete loss of steam flow similar to but less severe than the turbine trip event analyzed in Section 15.2.3.
The main steam line isolation valves close more slowly than the turbine stop valves resulting in less severe transient. Therefore, this event is bounded by the results of the analysis for Event 15.2.3.
This event is classified as a Condition II event. See Section 15.0.1 for a discussion of ANS Condition II events.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 15.2.5    LOSS OF CONDENSER VACUUM AND OTHER EVENTS RESULTING IN TURBINE TRIP Loss of condenser vacuum is one of the events that can cause a turbine trip. Turbine trip initiating events are described in Section 15.2.3. A loss of condenser vacuum would preclude the use of steam dump to the condenser; however, since steam dump is assumed not to be available in the turbine trip analysis, no additional adverse effects would result if the turbine trip were caused by loss of condenser vacuum. Therefore, the analysis results and conclusions contained in Section 15.2.3 apply to loss of condenser vacuum.
This event is classified as a Condition II event. See Section 15.0.1 for a discussion of ANS Condition II events.
15.2.6    LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 15.2.6.1      Identification of Causes and Accident Description A complete loss of nonemergency AC power may result in the loss of all power to the station auxiliaries, i.e., the reactor coolant pumps, condensate pumps, etc. The loss of power may be caused by a complete loss of the offsite grid accompanied by a turbine generator trip at the station, or by a loss of the onsite AC distribution system. Loss of main feedwater occurs on loss of nonemergency AC power. The combination of the decrease in primary coolant flow rate, the cessation of main feedwater flow and trip of the turbine generator compounds the event consequences. The decrease of main feedwater to the steam generators decreases the primary-to-secondary system heat transfer rate resulting in heatup of the primary coolant system. The increase in primary coolant temperature results in overpressurization of the RCS.
The reactor will trip: 1) due to turbine trip; 2) upon reaching one of the trip setpoints in the primary and secondary systems, as a result of the flow coastdown and decrease in secondary heat removal; or 3) due to loss of power to the control rod drive mechanisms, as a result of the loss of power to the plant.
Following a loss of AC power with turbine and reactor trips, the sequence described below will occur:
: 1) Plant vital instruments are supplied from emergency DC power sources.
: 2) As the steam system pressure rises following the trip, the steam generator power operated relief valves may be automatically opened to the atmosphere. Steam dump to the condenser is assumed not to be available. If power is available, power operated relief valves will open and control the steam pressure eliminating the pressure rise from opening the self-actuated safety valves. The self-actuated safety valves will lift to dissipate the sensible heat of the fuel and reactor coolant plus the residual decay heat produced in the reactor.
: 3) As the no-load temperature is approached, the steam generator power operated relief valves (or the self-actuated safety valves, if the power operated relief valves are not available) are used to dissipate the residual decay heat and to maintain the plant at the hot standby condition.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15
: 4) The standby diesel generators, started on loss of voltage on the plant emergency buses, begin to supply plant vital loads.
The Auxiliary Feedwater System is started automatically as follows:
Two motor driven auxiliary feedwater pumps are started on any of the following:
: 1) Low-low level in any steam generator.
: 2) Trip of all main feedwater pumps.
: 3) A safety injection signal.
: 4) Loss of offsite power.
: 5) Manual actuation.
One turbine driven auxiliary feedwater pump is started on any of the following:
: 1) Low-low level in any two steam generators.
: 2) Loss of offsite power.
: 3) Manual actuation.
Refer to Section 10.4.9 for a description of the Auxiliary Feedwater System.
The motor driven auxiliary feedwater pumps are supplied power by the diesel generators and the turbine driven pump utilizes steam from the Main Steam System. Both type pumps are designed to start within one minute even if a loss of all AC power occurs simultaneously with loss of normal feedwater. The turbine exhausts the secondary steam to the atmosphere. The auxiliary feedwater pumps take suction from the condensate storage tank for delivery to the steam generators.
Upon the loss of power to the reactor coolant pumps, reactor coolant flow necessary for core cooling and the removal of residual heat is maintained by natural circulation in the reactor coolant loops.
A loss of nonemergency AC power to the station auxiliaries is classified as an ANS Condition II event, a fault of moderate frequency. See Section 15.0.1 for a discussion of ANS Condition II events. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur. In addition, the AFW system response maintains a sufficient steam generator secondary side water inventory Amendment 63                                                                        Page 47 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 to remove long term decay heat from the primary side (i.e., steam generator dryout is avoided).
: 3) The event should not generate a more serious plant condition without other faults occurring independently.
: 4) For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel element cladding, shall occur.
Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod insertability is maintained.
An actual loss of nonemergency AC power would cause a loss of all feedwater, loss of reactor coolant flow (flow coastdown) and a reactor trip due to either loss of power to the RCCA's or any of the primary coolant flow trips within 1.5 seconds of the loss of AC power. The analysis presented in 15.2.6.2 is more conservative than an actual loss of AC in that it does not assume that these events occur simultaneously.
At 0 seconds, the loss of AC power and the resulting loss of feedwater occurs. However, the reactor trip and loss of RCS flow, which would normally occur, is not assumed to happen at this time. This causes the primary side coolant to heat up and the steam generator inventory to decrease. The reactor is finally tripped on an OTT signal, and at this time, the loss of primary flow due to the loss of AC is assumed to occur.
The above assumptions are more conservative than an actual loss of nonemergency AC because the reactor power is maintained following the loss of AC/loss of feedwater at 0 seconds. This minimizes the steam generator heat transfer capability and increases the amount of RCS stored energy at the time of reactor trip and loss of primary coolant flow.
Following the reactor coolant pump coastdown caused by the loss of AC power, the natural circulation capability of the RCS will remove residual and decay heat from the core, aided by auxiliary feedwater in the secondary system. An analysis is presented here to show that the natural circulation flow in the RCS following a loss of AC power event is sufficient to remove residual heat from the core.
The plant systems and equipment available to mitigate the consequences of a loss of AC power event are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
The worst single failure for this event is either the failure of an available auxiliary feedwater pump or the failure of an auxiliary feedwater isolation valve in the closed position. However, in compliance with SHNPP's TMI Action Plan Item II.E.1.1, this analysis conservatively accommodates failure of two of the three pumps. Furthermore, the analysis is performed with and without an additional failure of an auxiliary feedwater isolation valve in the closed position.
These two scenarios are analyzed as separate cases. The case with a single AFW pump delivering to only two Steam Generators represents an especially conservative combination of multiple failures.
15.2.6.2    Analysis of Effects and Consequences Method of Analysis - The transient response of the reactor system is calculated using the ANF-RELAP (Reference 15.2.3-1) computer program. Initial conditions, trip setpoints, plant systems Amendment 63                                                                            Page 48 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 availability, and boundary conditions are conservatively adjusted to maximize the calculated pressurizer water level.
The assumptions used in the analysis are as follows:
: 1) The plant is initially operating at 102 percent of the engineered safety features design rating.
: 2) Reactor trip occurs on steam generator low-low level or OTT. No credit is taken for immediate release of the control rod drive mechanisms caused by a loss of offsite power.
: 3) Auxiliary feedwater system delivers flow from one or two auxiliary feedwater pumps (see Section 15.0.13 for explanation of failure mode assumptions). For conservatism in this analysis the flow capacity of one motor driven auxiliary feedwater pump is assumed to be 374 gpm.
: 4) Secondary system steam relief is achieved through the steam generator safety valves.
: 5) The pressurizer power operated relief valves and pressurizer spray are conservatively assumed to function. This maximizes the peak transient pressurizer water volume.
: 6) RCS pump coastdown is conservatively assumed to occur at reactor scram to maximize primary system heatup and minimize steam generator inventory.
: 7) The worst single failure for this event is either the failure of the available turbine driven auxiliary feedwater pump or the failure of an auxiliary feedwater isolation valve in the closed position. In addition, to be in compliance with SHNPP's TMI Action Plan Item II.E.1.1, this analysis conservatively does not take credit for the operation of 1 available auxiliary feedwater pump.
The assumptions used in the analysis are similar to the loss of normal feedwater flow incident (Subsection 15.2.7), except that power is assumed to be lost to the reactor coolant pumps at the time of reactor trip.
This event is analyzed at the SGR/Uprating value for Tavg (588.8°), which bounds operation at 580.8°F.
Conservative conditions established for the analysis of this event are presented in Table 15.2.6-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. The trip setpoints and time delays assumed in the analysis of this event are also listed in Table 15.0.6-2. Key operating parameters used in the analysis of this event are presented in Table 15.2.6-2. The range of neutronics parameters supported by the analysis are presented in Table 15.2.6-3.
Results - The case without failure of an AFW isolation valve (with one AFW pump delivering to all three Steam Generators) was determined to be more limiting than the corresponding case with one AFW pump delivering to (only) two Steam Generators. The sequence of events is given in Table 15.2.6-4. The responses to key system variables are given in Figures 15.2.6-1 to 15.2.6-5.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 The transient is initiated from 2958 MW (rated + 0.34%) power initial conditions by a sudden and complete loss of main feedwater at time zero. The loss of feedwater results in degraded heat transfer capacity by the steam generators. Due to degraded heat transfer, the primary system begins to heat up. With a conservatively modeled positive moderator temperature coefficient, the increasing primary system temperature results in an increasing reactor core power. The pressurizer sprays and PORVs operate to mitigate the pressurizer pressure increase; however, the operation of these components also maximizes the pressurizer water level.
Without feedwater, the steam generator water level decreases until a reactor trip and subsequent turbine trip occur at approximately 41 seconds due to an OTT trip. The reactor scram terminates the short term primary system power increase at 110% and short term pressurizer water level increases to approximately 92% span.
The auxiliary feedwater system is also initiated by the low-low steam generator water level trip and delivers auxiliary feedwater flow to the steam generators by 109 seconds into the transient.
By approximately 500 seconds, the reactor coolant system is in a natural circulation flow mode.
After the turbine trip occurs, the main steam safety valves (MSSVs) begin cycling to remove the decay heat load. At 1320 seconds, the steam generator inventories reach a minimum value and then the steam generators begin to refill.
The transient simulation is terminated at 10,000 seconds with the primary system being cooled and the steam generators being refilled. No operator actions are credited in this event simulation.
15.2.6.3    Radiological Consequences of a Loss of Non-Emergency AC Power to Plant Auxiliaries A loss of non-emergency AC power to plant auxiliaries would result in a turbine and reactor trip on loss of condenser vacuum. Heat removal from the secondary system would occur through the steam generator power-operated relief valves or safety valves. Due to the pressure differential between the primary and secondary systems and assumed steam generator tube leakage, fission products are discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere through either the atmospheric relief valves or safety valves.
In addition, iodine activity is contained in the secondary coolant before the accident and some of this activity is released to the atmosphere as a result of steaming from the steam generators following the accident.
The analysis of the loss of offsite power (LOOP) radiological consequences uses the analytical methods and assumptions outlined in RG 1.183, Appendix G (Locked Rotor) for secondary system leakage release path modeling and RG 1.183 Appendix E (Main Steam Line Break) for iodine spiking.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 A summary of input parameters and assumptions is provided in Table 15.2.6-5. Additional clarification is provided as follows:
a) The noble gas and alkali metal activity concentrations in the RCS at the time the accident occurs are based on a one percent fuel defect level. The noble gas and alkali metal concentrations in the RCS are given in Table 15.0.9-2. The iodine activity concentration of the secondary coolant at the time the LOOP occurs is assumed to be equivalent to the Technical Specification limit of 0.1 Ci/gm of DE I-131. The secondary coolant iodine activity concentration is given in Table 15.0.9-7. The alkali metal activity concentration of the secondary coolant at the time the LOOP occurs is assumed to be 10% of the primary side concentration.
b) The amount of primary to secondary SG tube leakage is assumed to be equal to the Technical Specification limit of 1 gpm total.
c) An iodine partition factor in the SGs of 0.01 (curies iodine/gm steam)/(curies iodine/gm water) is used. This partition factor is also applied to alkali metals.
d) All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.
e) At 8 hours after the accident, the RHR System is assumed to be placed into service for heat removal and there is no further steam release to the atmosphere from the secondary system.
15.2.6.3.1 Pre-Accident Iodine Spike Case It is assumed that a reactor transient had occurred prior to the LOOP and had raised the RCS iodine concentration to the Technical Specification limit for a transient of 60 Ci/gm of dose equivalent (DE) I-131. The pre-accident spike iodine concentrations are given in Table 15.0.9-7.
15.2.6.3.2 Accident Initiated Iodine Spike Case The reactor trip associated with the LOOP creates an iodine spike in the RCS which increases the iodine release rate from the fuel to the RCS to a value 500 times greater than the release rate corresponding to a maximum equilibrium RCS concentration of 1.0 Ci/gm of DE I-131.
The iodine spike appearance rates are given in Table 15.0.9-6. The duration of the accident-initiated iodine spike is limited by the amount of activity available in the fuel-clad gap. Based on having 12 percent of the iodine in the fuel-clad gap, the gap inventory would be conservatively depleted within 5.0 hours and the spike is terminated at that time.
15.2.6.3.3 Offsite Doses The offsite doses are calculated using the assumptions and equations in Section 15.0A.1.
15.2.6.3.4 Control Room Doses The control room assumptions are provided in Section 15.6.5.4.3 and Table 15.6.5-15. The control room HVAC is switched to the emergency post-accident recirculation mode after Amendment 63                                                                          Page 51 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 receiving a high radiation signal. The high radiation signal is reached at 3 seconds into the event. The control room HVAC is switched over to the emergency post-recirculation mode at 18 seconds (3 second signal initiation plus 15 second delay time for switching between modes).
An operator action switches the control room from the post-accident recirculation mode to the pressurization mode at 2 hours after event initiation. The 15-second delay to switch between modes was also assumed with the operator action. Thus 2 hours and 33 seconds was actually modeled from the time of operator action switchover to the pressurization mode.
15.2.6.3.5 Results The potential radiological consequences resulting from the occurrence of a loss of non-emergency AC power have been conservatively analyzed, using assumptions and models described in previous sections. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The radiological analysis results for this event are listed in Table 15.2.6-6. The resultant doses are within the guideline values. For the case with an assumed pre-accident iodine spike, the calculated doses are less than the 10CFR50.67 limits of 25 rem TEDE offsite and 5 rem TEDE in the control room. For the case with an accident initiated iodine spike, the offsite doses are less than 10%
of the 10CFR50.67 limits (i.e., less than 2.5 rem TEDE) and the control room doses are less than the 10CFR50.67 limit of 5 rem TEDE.
15.2.6.4    Conclusions The results of the analysis demonstrate that the event acceptance criteria are met since overpressurization of the primary and secondary system is avoided. Pressurizer filling is avoided and solid liquid discharge through the pressurizer relief or safety valves does not occur.
Additionally, it is demonstrated that the assumed minimum motor driven auxiliary feedwater pump capacity of 374 gpm is sufficient to prevent steam generator dryout and to accomplish long term decay heat removal.
The analysis for this event supports full power operation at a nominal primary Tavg between 580.8°F and 588.8°F, inclusive.
15.2.7    LOSS OF NORMAL FEEDWATER FLOW 15.2.7.1    Identification of Causes and Accident Description A Loss of Normal Feedwater Flow transient is initiated by main feedwater pump failure or a malfunction in the feedwater control valves. The loss of main feedwater flow decreases the amount of subcooling in the secondary-side downcomer which diminishes the primary-to-secondary system heat transfer and leads to an increase in the primary system coolant temperature. The increase in primary coolant temperature results in overpressurization of the RCS.
The opening of the secondary-side safety valves acts to remove decay heat load and to mitigate the primary system heatup. The long-term cooling of the primary system is governed by the heat removal capacity of the auxiliary feedwater flow. The auxiliary feedwater pumps are automatically started upon a steam generator low-low water level trip.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 If an alternative supply of feedwater were not supplied to the plant, core residual heat following reactor trip would heat the Reactor Coolant System water to the point where water relief from the pressurizer would occur, resulting in a substantial loss of water from the RCS. Since the plant is tripped well before the steam generator heat transfer capability is reduced, the reactor coolant system variables never approach a DNB condition.
The following events occur upon loss of normal feedwater (assuming main feedwater pump failures or valve malfunctions).
: 1) As the steam system pressure rises following reactor and turbine trips, the steam generator power operated relief valves are automatically opened to the atmosphere.
Steam dump to the condenser is assumed not to be available. If the steam flow rate through the power operated relief valves is not available, the steam generator self-actuated safety valves will lift to dissipate the sensible heat of the fuel and reactor coolant plus the residual decay heat produced in the reactor.
: 2) As the no-load temperature is approached, the steam generator power operated relief valves (or the self-actuated safety valves, if the power operated relief valves are not available) are used to dissipate the residual decay heat and to maintain the plant at the hot standby condition.
A loss of normal feedwater is classified as an ANS Condition II event, fault of moderate frequency. See Section 15.0.1 for a discussion of ANS Condition II events. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur. In addition, fission product decay heat must be transferred from the reactor coolant system following a loss of normal feedwater flow (i.e., steam generator dryout is avoided).
: 3) The event should not generate a more serious plant condition without other faults occurring independently.
: 4) For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel element cladding, shall occur.
Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod insertability is maintained.
Reactor trip on low-low water level in any steam generator provides protection for a loss of normal feedwater.
The Auxiliary Feedwater System is started automatically as discussed in Section 15.2.6.1. The steam driven auxiliary feedwater pump utilizes steam from the Main Steam System and exhausts to the atmosphere. The motor driven auxiliary feedwater pumps are supplied by Amendment 63                                                                            Page 53 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 power from the diesel generators. The pumps take suction directly from the condensate storage tank for delivery to the steam generators.
An analysis of the system transient is presented below to show that following a loss of normal feedwater, the Auxiliary Feedwater System is capable of removing the stored and residual heat, thus preventing either overpressurization of the RCS or loss of water from the reactor core, and returning the plant to a safe condition.
The worst single failure for this event is the failure of the available turbine driven auxiliary feedwater pump. In compliance with SHNPP's TMI Action Plan Item II.E.1.1, this analysis conservatively does not credit the operation of 1 of 2 available motor driven auxiliary feedwater pumps.
Plant systems and equipment which are available to mitigate the effects of a loss of normal feedwater accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
15.2.7.2    Analysis of Effects and Consequences Method of Analysis - The transient response of the reactor system is calculated using the ANF-RELAP (Reference 15.2.3-1) computer program. The primary concern in analyzing the loss of normal feedwater event is to avoid filling the pressurizer, resulting in liquid discharge through the pressurizer relief valves and loss of primary system water inventory. Therefore, the initial conditions, trip setpoints, plant systems availability, and boundary conditions are conservatively adjusted to maximize the calculated pressurizer water level.
Assumptions made in the analysis are:
: 1) The plant is initially operating at 102 percent of the engineered safety features design rating.
: 2) Reactor trip occurs on steam generator low-low level or OTT.
: 3) Auxiliary feedwater system delivers flow to the three steam generators from one auxiliary feedwater pump (see Section 15.0.13 for explanation of failure mode assumptions). For conservatism in this analysis, the flow capacity of the motor driven auxiliary feedwater pump is assumed to be 374 gpm.
: 4) Secondary system steam relief is achieved through the self-actuated safety valves.
Note that steam relief will, in fact, be through the power operated relief valves or condenser dump valves for most cases of loss of normal feedwater. However, for the sake of analysis these have been assumed unavailable.
: 5) The pressurizer power operated relief valves and pressurizer spray are assumed to function, this maximizes the peak transient pressurizer water volume.
The assumptions used in the analysis are similar to the loss of AC power incident (Subsection 15.2.6), except that the reactor coolant pumps are assumed to continue to operate.
This event is analyzed at the SGR/Uprating value for Tavg (588.8°F), which bounds operation at 580.8°F.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 Conservative conditions established for the analysis of this event are presented in Table 15.2.7-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. Key operating parameters used in the analysis of this event are presented in Table 15.2.7-2. The range of neutronics parameters supported by this analysis are presented in Table 15.2.7-3.
Results - The sequence of events is given in Table 15.2.7-4. The responses to key system variables are given in Figures 15.2.7-1 to 15.2.7-5.
The transient is initiated from full power initial conditions by a sudden and complete loss of main feedwater at time zero. The loss of feedwater results in degraded heat transfer capacity by the steam generators. Due to degraded heat transfer, the primary system begins to heat up. With a conservatively modeled positive moderator temperature coefficient, the increasing primary system temperature results in an increasing reactor core power. The pressurizer sprays and PORVs operate to mitigate the pressurizer pressure increase. However, their operation also maximizes the pressurizer water inventory and level.
A reactor trip and subsequent turbine trip occur at approximately 41 seconds due to the overtemperature T (OTT) trip. This is approximately 11 seconds before the low low steam generator water level trip.
The auxiliary feedwater system is also initiated by the low steam generator water level signal and delivers auxiliary feedwater flow to the steam generators by 110 seconds into the transient.
After the turbine trip occurs, the MSSVs begin cycling to remove the decay heat load. At 2965 seconds, the steam generator water levels reach a minimum and then the steam generators begin to refill. The transient simulation is terminated at 10,000 seconds with the primary system being cooled and all three steam generators being refilled. No operator actions are credited in this event simulation.
15.2.7.3    Conclusions Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the RCS, or the Main Steam System since the auxiliary feedwater capacity is such that Reactor Coolant System does not overpressurize and solid liquid is not relieved from the pressurizer power operated relief or safety valves.
The results of the analysis demonstrate that the event acceptance criteria are met. Additionally, it is demonstrated that the assumed minimum motor driven auxiliary feedwater pump capacity of 374 gpm is sufficient to prevent steam generator dryout and to accomplish long term decay heat removal. The radiological consequences of this event would be less severe than the steamline break accident analyzed in Section 15.1.5.
The analysis for this event supports nominal primary Tavg operation at full power between 588.8 and 580.8°F, inclusive.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.2.8    FEEDWATER SYSTEM PIPE BREAK 15.2.8.1    Identification of Causes and Accident Description A major feedwater line rupture is defined as a break in a feedwater line large enough to prevent the addition of sufficient feedwater to the steam generators to maintain shell side fluid inventory in the steam generators. If the break is postulated in a feedline between the check valve and the steam generator, liquid followed by steam from the steam generator will be discharged through the break. (A break upstream of the feedline check valve would affect the Nuclear Steam Supply System only as a loss of feedwater. This case is covered by the evaluation in Section 15.2.7). Depending upon the size of the break and the plant operating conditions at the time of the break, the break could cause either a RCS cooldown (by excessive energy discharge through the break) or a RCS heatup. Potential RCS cooldown resulting from a secondary pipe rupture is evaluated in Section 15.1.5. Therefore, only the RCS heatup effects are evaluated for a feedwater line rupture.
The event follows three phases. A mild primary system heatup occurs due to the loss of feedwater to the steam generators prior to a reactor scram. (This phase may not occur under certain kinetics assumptions.) This is followed by a cooldown phase of the primary-side coolant due to the energy removal during the steam generator blowdown stage. Finally, the eventual depletion of secondary-side inventory in the ruptured steam generator and lack of main feedwater to the intact steam generators results in a long term, primary system heatup much like a Loss of Normal Feedwater Flow event.
A feedwater line rupture reduces the ability to remove heat generated by the core from the RCS for the following reasons:
: 1) Feedwater flow to the steam generators is reduced. Since feedwater is subcooled, its loss may cause reactor coolant temperatures to increase prior to reactor trip.
: 2) Fluid in the steam generator may be discharged through the break, and would then not be available for decay heat removal after trip.
: 3) The break may be large enough to prevent the addition of any main feedwater after trip.
Since the auxiliary feedwater (AFW) flow is injected into the steam generators via a separate piping network than the main feedwater, the delivery of auxiliary feedwater will not be interrupted by the pipe rupture. An Auxiliary Feedwater System is provided to assure that adequate feedwater will be available such that:
: 1) No substantial overpressurization of the RCS shall occur.
: 2) Sufficient liquid in the RCS shall be maintained so that the core remains in place and geometrically intact with no loss of core cooling capability.
A major feedwater line rupture is classified as an ANS Condition IV event. See Section 15.0.1 for a discussion of ANS Condition IV events. The acceptance criteria for this event are:
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15
: 1) Pressure in the reactor and main steam systems should be maintained below 110% of design pressures for low probability events and below 120% of design pressures for very low probability events such as double-ended guillotine breaks.
: 2) The potential for core damage is evaluated on the basis that it is acceptable if the minimum DNBR remains above the 95/95 DNBR limit. If DNBR falls below the 95/95 DNBR limit, fuel failure must be assumed for all rods that do not meet this criteria unless it can be shown, based on an acceptable fuel damage model, which includes the potential adverse effects of hydraulic instabilities, that fewer failures occur. Any fuel damage calculated to occur must be of sufficiently limited extent that the core will remain in place and intact with no loss of core cooling capacity.
: 3) Any radioactivity release must be such that the calculated doses are within applicable guidelines. For a case with an assumed pre-accident iodine spike or accident induced fuel failure, the calculated doses should be less than the 10CFR50.67 limits of 25 rem TEDE offsite and 5 rem TEDE in the control room. For a case with an accident initiated iodine spike, the offsite doses should be less than 10% of the 10CFR50.67 limits (i.e.,
less than 2.5 rem TEDE) and the control room doses should be less than the 10CFR50.67 limit of 5 rem TEDE.
: 4) The integrity of the reactor coolant pumps should be maintained, such that loss of AC power and containment isolation will not result in seal damage.
: 5) The auxiliary feedwater system must be safety grade and automatically initiated when required.
: 6) Tripping of the reactor coolant pumps should be consistent with the resolution to TMI Action Plan Item II.K.3.5.
Analyses are performed for three different cases: short-term core cooling (DNB), long-term core cooling, and primary system pressurization. Secondary system pressurization is bounded by the turbine trip event and is not analyzed. The worst single active failure could be either the failure of an auxiliary feedwater pump or the failure of an auxiliary feedwater isolation valve in the closed position.
Plant systems and equipment which are available to mitigate the effects of a feedwater line break accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
The following provides the necessary protection for a main feedwater rupture:
: 1) A reactor trip on any of the following conditions.
a) High pressurizer pressure.
b) Overtemperature T.
c) Low-low steam generator water level in any steam generator.
d) Safety injection signals from any of the following:
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 1) two out of three low steam line pressure in any one main steam line.
: 2) two out of three high containment pressure (Hi-1).
: 3) low pressurizer pressure.
: 2) An Auxiliary Feedwater System to provide an assured source of feedwater to the steam generators for decay heat removal. (Refer to Section 10.4.9 for a description of the Auxiliary Feedwater System). Automatic AFS isolation is actuated on high steam line differential pressure coincident with main steam line isolation to divert all available cooling flow to the intact steam generators. (Refer to Sections 7.3.1.3.3 and 10.4.9.3 and Figure 7.2.1-1, Sheet 7 of 15.)
: 3) Automatic main steam line isolation to terminate steam flow from the intact steam generators to the break (through the affected steam generator).
The analysis assumed operator control of safety injection (and auxiliary feedwater) 30 minutes after event initiation.
15.2.8.2    Analysis of Effects and Consequences Method of Analysis - The availability of offsite power affects the results of the Feedwater Line Break event. Therefore, cases are evaluated with offsite power available and with offsite power lost coincident with turbine trip.
The Feedwater Line Break event is postulated as a double ended rupture of the largest feedwater line that could result in steam generator blowdown. By assuming the largest possible break area, the steam generator blowdown rate is maximized, which maximizes the resulting reactor coolant system heatup.
The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.2.3-5) computer program following the method described in Reference 15.2.8-4 and Reference 15.2.8-5. The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.2.8-1) to predict the MDNBR for the event using the Statistical Core Design (SCD) Methodology (Reference 15.2.8-2) and the HTP CHF correlation (Reference 15.2.8-3) for the Framatome 17x17 Advanced W-HTP fuel.
The primary concerns in analyzing the Feedwater Line Break event are to avoid primary side over pressurization, ensure that fuel design limits are met, and provide long term decay heat removal. The initial conditions, trip setpoints, plant systems availability, and boundary conditions are adjusted to provide a conservative result for each acceptance criterion analyzed.
Major assumptions made in the analyses are as follows:
: 1) The plant is initially operating at 2948 MW in the short-term core cooling case. In the peak primary pressure and long-term core cooling cases, the plant is initially operating at 2958 MW (rated + 0.34%).
: 2) No credit is taken for the pressurizer power operated relief valves or pressurizer spray in the peak primary pressure case.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 3) No credit is taken for safety injection on high containment pressure.
: 4) Main feedwater flow to all steam generators is assumed to stop at the time the break occurs (all main feedwater spills out through the break).
: 5) The worst possible break area is assumed. This maximizes the blowdown discharge rate, which maximizes the resultant heatup of the reactor coolant.
: 6) Reactor trip is assumed to be initiated on the safety injection signal when the low steam line pressure setpoint is reached.
: 7) The Auxiliary Feedwater System is actuated by the safety injection signal or loss of offsite power. A 61.5 second delay was assumed to allow time for startup of the standby diesel generators and the auxiliary feedwater pumps.
: 8) No credit is taken for charging or letdown.
: 9) Receipt of a low-low steam generator water level signal in at least one steam generator starts both the motor driven auxiliary feedwater pumps, and in two out of three steam generators starts the turbine driven auxiliary feed pump, which in turn initiates auxiliary feedwater flow to the steam generators. Receipt of a safety injection signal will also start the motor driven auxiliary feedwater pumps. Both motor driven auxiliary feedwater pumps and the turbine driven auxiliary feedwater pump start on a loss of offsite power.
Similarly receipt of a low steam line pressure signal in at least one steam line initiates a steam line isolation signal which closes the main steam line isolation valves in all steam lines. This signal also gives a safety injection signal which initiates flow of borated water into the RCS. The amount of safety injection flow is a function of RCS pressure.
: 10) For the feedwater line break case with loss of offsite power, the reactor coolant pumps trip upon the loss of offsite power.
: 11) The flow coastdown associated with a loss of offsite power minimizes the DNBR.
Therefore, only the offsite power lost case is analyzed for short-term core cooling. There is no single failure that adversely impacts the minimum DNBR.
: 12) Cases are analyzed with both offsite power maintained and offsite power lost for long-term core cooling.
: 13) Continued forced flow from the reactor coolant pumps maximizes the pressure difference between the pressurizer and bottom of the reactor vessel lower plenum where the peak primary pressure occurs. Therefore, the offsite power lost case is bounded by the offsite power maintained case for peak primary pressure and is not analyzed.
: 14) For the long-term core cooling cases, early termination of HHSI flow slows the repressurization and minimizes the addition of cooler water to the RCS. Therefore, it is assumed that no HHSI flow is delivered since the termination criteria are met early in the event.
: 15) For the peak primary pressure case, two pump maximum HHSI flow is assumed in order to maximize RCS pressurization.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 16) The auxiliary feedwater system delivers flow from one or two auxiliary feedwater pumps.
For conservatism in this analysis, the minimum flow capacity is assumed to be 374 gpm for one motor driven pump and 700 gpm for two motor driven pumps.
Following the trip of the reactor coolant pumps, there will be a flow coastdown until reactor coolant loop flow reaches the natural circulation value. The natural circulation capability of the RCS has been shown in Section 15.2.6, for the loss of AC power transient, to be sufficient to remove core decay heat following reactor trip. Pump coastdown characteristics are demonstrated in Sections 15.3.1 and 15.3.2 for single and multiple reactor coolant pump trips, respectively.
Emergency operating procedures following a feedwater line break will call for the following actions to be taken by the reactor operator:
: 1) Isolation of feedwater flow spilling out the break of faulted steam generator and align system so that the level in the intact steam generators is recovered.
: 2) Stop Charging/Safety Injection Pumps (CSIP's) if: 1) wide range reactor coolant pressure is stable or increasing and subcooling exists, 2) the pressurizer level is on span plus errors and, 3) steam generator narrow range level indication exists in at least one steam generator or minimum auxiliary feedwater flow exists.
Subsequent to recovery of level in the intact steam generators, safety injection flow will be isolated and plant operating procedures will be followed in cooling the plant to hot shutdown conditions.
Conservative conditions established for the analysis of this event are presented in Table 15.2.8-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. Key operating parameters used in the analysis of this event are presented in Table 15.2.8-2. The range of neutronics parameters supported by this analysis are presented in Table 15.2.8-3.
No reactor control systems are assumed to function. The Reactor Protection System is required to function following a feedwater line rupture as analyzed here. No single active failure will prevent operation of this system.
The engineered safety systems assumed to function are the Auxiliary Feedwater System, Main Steamline Isolation, and the Safety Injection System. The worst single active failures considered are the failure of an auxiliary feedwater pump and the failure of an auxiliary feedwater isolation valve in the closed position (Table 10.4.9-2). Separate AFW isolation valves exist for motor-driven and turbine-driven AFW, and thus the failure of one of these valves reduces AFW flow delivery to the associated steam generator but does not totally prevent it.
Since each intact steam generator receives some AFW flow regardless of the single failure assumed, the fraction of primary side heat removed by each steam generator is not as important as the total amount of AFW flow removing that heat.
For the cases with offsite power maintained, both motor driven AFW pumps receive a start signal from a safety injection signal on low steam line pressure. Since the minimum flow from either one or two motor driven AFW pumps is sufficient to avoid reaching the low-low steam generator level setpoint in the intact steam generators, the turbine driven AFW pump will not start. Therefore, the worst single failure for the cases with offsite power maintained is the failure Amendment 63                                                                        Page 60 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 of a motor driven AFW pump, which results in one motor driven AFW pump feeding two intact steam generators.
For the case with offsite power lost, both motor driven AFW pumps and the turbine driven AFW pump receive a start signal on the loss of offsite power. Therefore, the worst single failure for the case with offsite power lost is the failure of the turbine driven AFW pump, which results in two motor driven AFW pumps feeding two intact steam generators.
A detailed description and analysis of the Safety Injection System is provided in Section 6.3.
The Auxiliary Feedwater System is described in Section 10.4.9.
Results The sequence of events for the short-term core cooling analysis is presented in Table 15.2.8-4.
The limiting case includes BOC kinetics and is terminated prior to the delivery of HHSI or AFW flow. The reactor coolant pumps trip 3 seconds after the loss of offsite power coincident with turbine trip. The responses to key system variables for the short-term core cooling analysis are very similar to those for the long-term core cooling analysis with offsite power lost and are presented in Figures 15.2.8-21 through 15.2.8-24. The transient simulation is terminated at 30 seconds after the minimum DNBR is reached, and the predicted MDNBR is greater than the safety limit.
The sequence of events for the long-term core cooling analysis with offsite power maintained is presented in Table 15.2.8-5. The limiting case includes EOC kinetics to maximize decay heat, no HHSI flow, and the single failure of a motor driven AFW pump. The responses to key system variables are given in Figures 15.2.8-1 through 15.2.8-9. The transient simulation is terminated at 14400 seconds. The MSSVs and AFW system are removing the decay heat load. The primary system is in a stable, slow cooldown mode. The operator is available to control water level in the intact steam generators with AFW beginning at 30 minutes.
The sequence of events for the long-term core cooling analysis with offsite power lost is presented in Table 15.2.8-7. The limiting case includes EOC kinetics to maximize decay heat, no HHSI flow, and the single failure of the turbine driven AFW pump. The reactor coolant pumps trip 3 seconds after the loss of offsite power coincident with turbine trip. The responses to key system variables are given in Figures 15.2.8-11 through 15.2.8-19. The transient simulation is terminated at 3600 seconds. The MSSVs and AFW system are removing the decay heat load.
The primary system is in a stable, slow cooldown mode. The operator is available to control water level in the intact steam generators with AFW beginning at 30 minutes.
The sequence of events for the peak primary pressure analysis is presented in Table 15.2.8-8.
The limiting case includes BOC kinetics, maximum HHSI flow, offsite power maintained, and the single failure of a motor driven AFW pump. The responses to key system variables are given in Figures 15.2.8-25 through 15.2.8-32. The transient simulation is terminated at 3600 seconds.
The MSSVs and AFW system are removing the decay heat load. The primary system is in a stable, slow cooldown mode. HHSI is terminated by the operator at 30 minutes, but the primary system remains water solid. The operator is available to control water level in the intact steam generators with AFW beginning at 30 minutes. The maximum reactor vessel pressure does not exceed 120% of the primary system design pressure of 3000 psig.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 Long Term Effects and Events - For a feedwater line break, auxiliary feedwater is initiated automatically as is Safety Injection. For the feedline break downstream of the main feedwater isolation valves the required operator actions are similar in nature to the required actions for the steamline break. Table 15.2.8-6 provides a complete assessment of the operator's role.
Where possible, the operator should increase the auxiliary feedwater flow to the intact steam generators in order to shorten the time until primary temperatures begin to decrease. As a minimum, the operator must provide for decay heat removal through the intact steam generators by maintaining steam generator water level using auxiliary feedwater as a makeup supply. The operator can use the steam dump system or the steam generator PORV's to begin a controlled cooldown, or the unit may be maintained in hot standby using the steam side safety valves for decay heat removal.
Once the Safety Injection termination criteria are met, the Charging/Safety Injection Pumps will be turned off and plant operating procedures will be followed in cooling the plant to shutdown conditions. The operator must observe the primary steam pressure-temperature relationship to ensure that voiding does not occur in the reactor coolant system. The operator uses safety grade instrumentation and controls to manually control the primary system pressure and pressurizer level.
Operator action in the postulated feedwater line rupture to terminate the safety injection and cycle AFW flow within 30 minutes has been credited in the analysis. Refer to Section 5.2.2.2 for a discussion of pressurizer SRV qualification for this event.
15.2.8.3    Conclusions The results of the analysis demonstrate that the event acceptance criteria are met since over pressurization of the primary system is avoided. In addition, the analysis indicates that long term decay heat removal is adequate. The minimum DNBR remains above the safety limit. The radiological consequences of this event would be less severe than the steamline break accident analyzed in Section 15.1.5.
 
==REFERENCES:==
SECTION 15.2 15.2.3-1        ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events, ANF-89-151(P)(A), Advanced Nuclear Fuels Corporation, Richland, WA 99352, May 1992.
15.2.3-2        Deleted by Amendment No. 63.
15.2.3-3        Deleted by Amendment No. 63.
15.2.3-4        Letter from B.C. Waldrep (Duke Energy) to NRC (Serial HNP-15-038) dated December 17,2015, "License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change." (Safety Evaluation Report received by {{letter dated|date=July 25, 2016|text=letter dated July 25, 2016}}).
15.2.3-5        EPRI NP-7450(A), Revision 10, RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, September 2014.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.2.8-1        EPRI NP-2511-CCM-A, Revision 4, VIPRE A Thermal-Hydraulic Code for Reactor Cores, June 2007.
15.2.8-2        DPC-NE-2005-PA, Revision 5, "Thermal-Hydraulic Statistical Core Design Methodology," March 2016.
15.2.8-3        EMF-92-153-PA, Revision 1 HTTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, January 2005.
15.2.8-4        DPC-NE-3009, Revision 0, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, April 2018.
15.2.8-5        DPC-NE-3008, Revision 0, Thermal-Hydraulic Models for Transient Analysis, April 2018.
15.3    DECREASE IN REACTOR COOLANT SYSTEM FLOW RATE A number of faults are postulated which could result in a decrease in Reactor Coolant System flow rate. These events are discussed in this section. Detailed analyses are presented for the most limiting of these events.
Discussions of the following flow decrease events are presented in Section 15.3:
: 1) Partial loss of forced reactor coolant flow. (ANS Condition II event).
: 2) Complete loss of forced reactor coolant flow. (ANS Condition III event, however, is analyzed as a Condition II event).
: 3) Reactor coolant pump shaft seizure (locked rotor). (ANS Condition IV event).
: 4) Reactor coolant pump shaft break. (ANS Condition IV event).
Section 15.0.1 contains a discussion of ANS classifications.
15.3.1    PARTIAL LOSS OF FORCED REACTOR COOLANT FLOW 15.3.1.1    Identification of Causes and Accident Description A partial loss of forced reactor coolant flow accident can result from a mechanical or electrical failure in a reactor coolant pump, or from a fault in a reactor coolant pump bus. If the reactor is at power at the time of the accident, the immediate effect of loss of forced reactor coolant flow is a rapid increase in the reactor coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor is not tripped promptly.
Normal power for the reactor coolant pumps is supplied through individual buses connected to the turbine generator. When a generator trip occurs, the buses are automatically transferred to a transformer supplied from external power lines, and the reactor coolant pumps will continue to supply coolant flow to the core. Following any turbine trip where there are no electrical faults which require tripping the generator from the network, the generator remains connected to the network for approximately 30 seconds. The reactor coolant pumps remain connected to the Amendment 63                                                                        Page 63 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 generator, thus ensuring full flow for approximately 30 seconds after the reactor trip before any transfer is made.
This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Section 15.0.1.
The necessary protection against a partial loss of forced reactor coolant flow accident is provided by the low reactor coolant flow reactor trip signal which is actuated in any reactor coolant loop by two out of three low flow signals. Above Permissive 8, low flow in any loop will actuate a reactor trip. Between approximately 10 percent power (Permissive 7) and the power level corresponding to Permissive 8, low flow in any two reactor coolant loops will actuate a reactor trip. Reactor Trip on Low flow is blocked below Permissive 7. Above Permissive 7, two or more reactor coolant pump circuit breakers opening will actuate the corresponding undervoltage relays. This results in a reactor trip which serves as a backup to the low flow trip.
Plant systems and equipment which are necessary to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1. No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.
15.3.1.2    Deleted by Amendment No. 48 15.3.1.3    Deleted by Amendment No. 48 15.3.1.4    Event Disposition Mode 1 (Above P-7) - The amount of flow reduction and the rate of flow decrease is less than that in Mode 1 of Event 15.3.2, because only 1 reactor coolant pump is affected in Event 15.3.1.
The challenge to Specified Acceptable Fuel Design Limits (SAFDLs) is therefore bounded for this event by Event 15.3.2, as long as Event 15.3.2 satisfies the SAFDLs since Event 15.3.2 is classified as a Condition III event. Mode 1 (Above P-7) bounds all other modes of operation for this event.
In the analysis of Event 15.3.2 (Section 15.3.2), the Condition II acceptance criteria were met.
Therefore, this event does not require reanalysis.
15.3.2    COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2.1    Identification of Causes and Accident Description A complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to all reactor coolant pumps. If the reactor is at power at the time of the accident, the immediate effect of loss of forced reactor coolant flow is a rapid increase in the reactor coolant temperature. This increase could result in DNB with subsequent fuel damage if the reactor were not tripped promptly.
Normal power for the reactor coolant pumps is supplied through buses from a transformer connected to the turbine generator. When a generator trip occurs, the buses are automatically transferred to a transformer supplied from external power lines, and the pumps will continue to supply reactor coolant flow to the core. Following any turbine trip where there are no electrical faults which require tripping the generator from the network, the generator remains connected to Amendment 63                                                                        Page 64 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 the network for approximately 30 seconds. The reactor coolant pumps remain connected to the generator, thus ensuring full flow for 30 seconds after the reactor trip before any transfer is made.
This event is classified as an ANS Condition III incident (an infrequent incident) as defined in Section 15.0.1. This transient has been analyzed against Condition II acceptance criteria in order to bound the Chapter 15.3.1 "Partial Loss of Forced Reactor Coolant Flow" Event. The Condition II acceptance criteria are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur.
: 3) The event should not generate a more serious plant condition without other faults occurring independently.
: 4) For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel element cladding, shall occur.
Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod insertability is maintained.
The following signals provide the necessary protection against a complete loss of flow accident:
: 1) Reactor coolant pump power supply undervoltage or underfrequency.
: 2) Low reactor coolant loop flow.
The reactor trip on reactor coolant pump undervoltage is provided to protect against conditions which can cause a loss of voltage to all reactor coolant pumps, i.e., station blackout. This function is blocked below approximately 10 percent power (Permissive 7).
The reactor trip on reactor coolant pump underfrequency is provided to trip the reactor for an underfrequency condition, resulting from frequency disturbances on the power grid. If the maximum grid frequency decay rate is less than approximately 5 Hz/sec., this trip function will protect the core from underfrequency events without requiring tripping of the RCP breakers.
Refer to Chapter 7 for interface requirements concerning tripping of the RCP breakers for underfrequency events.
The reactor trip on low reactor coolant loop flow is provided to protect against loss of flow conditions which affect only one reactor coolant loop. This function is generated by two out of three low flow signals per reactor coolant loop. Above Permissive 8 power level, low flow in any reactor coolant loop will actuate a reactor trip. Between approximately 10 percent power (Permissive 7) and the power level corresponding to Permissive 8, low flow in any two reactor coolant loops will actuate a reactor trip. If the maximum grid frequency decay rate is less than approximately 2.5 Hz/sec., the low flow trip function will protect the core from underfrequency events.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 Plant systems and equipment which are necessary to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1. No single active failure will adversely affect the consequences of the event.
15.3.2.2    Analysis of Effects and Consequences Method of Analysis - A conservative reactor trip actuation that bounds both the pump power supply undervoltage trip and the pump power supply underfrequency trip is analyzed.
Conservative conditions established for the analysis of this event are presented in Table 15.3.2-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. Key operating parameters used in the analysis of this event are presented in Table 15.3.2-2. The range of neutronics parameters supported by this analysis are presented in Table 15.3.2-3.
The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.3.2-2) computer program following the method decribed in Reference 15.3.2-5 and 15.3.2-7.
The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.3.2-4) to predict the MDNBR for the event using the Statistical Core Design (SCD) methodology (Reference 15.3.2-6) and the HTP CHF correlation (Reference 15.3.2-8) for the Framatome 17x17 Advanced W-HTP fuel.
Results - The complete loss of forced reactor coolant flow event was analyzed using a conservative reactor trip actuation that bounds both the underfrequency and undervoltage events. The sequence of events is given in Table 15.3.2-4. The responses to key system variables are given in Figures 15.3.2-1 to 15.3.2-7.
15.3.2.3    Conclusions The results of the analysis indicate that the predicted MDNBR is greater than the safety limit.
The critical heat flux correlation limit ensures that, with 95% probability and 95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. This event does not pose a credible challenge to FCM safety criteria. Thus, Condition II acceptance criteria are met for this event. Since Condition II acceptance criteria are met for this event, Event 15.3.1 is bounded.
15.3.3    REACTOR COOLANT PUMP SHAFT SEIZURE (LOCKED ROTOR) 15.3.3.1    Identification of Causes and Accident Description The accident postulated is an instantaneous seizure of a reactor coolant pump rotor such as is discussed in Section 5.4.1. Flow through the affected reactor coolant loop is rapidly reduced, leading to an initiation of a reactor trip on a low flow signal.
Following initiation of the reactor trip, heat stored in the fuel rods continues to be transferred to the reactor coolant causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generators is reduced, first because the reduced flow results in a decreased tube side film coefficient and then because the reactor coolant in the tubes cool down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the reactor coolant in the reactor core, combined with reduced heat transfer in the steam generators, causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steam volume, actuates Amendment 63                                                                          Page 66 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 the automatic pressurizer spray system, opens the pressurizer power operated relief valves, and potentially opens the pressurizer safety valves, in that sequence. The three pressurizer power operated relief valves are designed for reliable operation and would be expected to function properly during the accident. However, for conservatism, their pressure reducing effect as well as the pressure reducing effect of the spray is not included in the overpressurization case of the analysis.
This event is classified as an ANS Condition IV incident (a limiting fault) as defined in Section 15.0.1. The acceptance criteria for this event are:
: 1) The pressure in the reactor coolant and main steam systems should be maintained below acceptable design limits (i.e., the pressure in the reactor coolant and main steam systems should be maintained below 120% of the design pressures).
: 2) The potential for core damage is evaluated on the basis that it is acceptable if the minimum DNBR remains above the 95/95 DNBR limit. If DNBR falls below the 95/95 limit, fuel failure must be assumed for all rods that do not meet this criteria unless it can be shown, based on an acceptable fuel damage model, which includes the potential adverse effects of hydraulic instabilities, that fewer failures occur. Any fuel damage calculated to occur must be of sufficiently limited extent that the core will remain in place and intact with no loss of core cooling capacity.
: 3) Any reactivity release must be such that the calculated offsite doses should be less than 10% of the 10CFR50.67 limits (i.e., 2.5 rem TEDE) and the control room doses should be less than the 10CFR50.67 limit of 5 rem TEDE.
: 4) A rotor seizure or shaft break in a reactor coolant pump should not, by itself, generate a more serious condition or result in a loss of function of the reactor coolant system or containment barriers.
No single active failure will adversely affect the consequences of the event.
Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
15.3.3.2    Analysis of Effects and Consequences Methods of Analysis -The analysis of the Locked Rotor event has considered the effect of a coincident loss of offsite power which causes the remaining two pumps to coastdown.
Both an overpressurization case and a MDNBR case are analyzed for this event.
The analysis is performed with BOC kinetics parameters.
Normal power for the reactor coolant pumps is supplied through individual buses connected to the turbine generator. When a generator trip occurs, the buses are automatically transferred to a transformer supplied from external power lines, and the reactor coolant pumps will continue to supply coolant flow to the core. Following any turbine trip where there are no electrical faults which require tripping the generator from the network, the generator remains connected to the network for approximately 30 seconds. The reactor coolant pumps remain connected to the Amendment 63                                                                          Page 67 of 151
 
Shearon Harris Nuclear Power Plant                                                  UFSAR Chapter: 15 generator ensuring full flow for approximately 30 seconds after the reactor trip before any supply power transfer is made. This analysis conservatively assumes the remaining two pumps are connected to the generator for 3 seconds following the turbine trip.
Conservative conditions established for the analysis of this event are presented in Table 15.3.3-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. For the analysis of this event, all trips listed in this table were disabled except low primary coolant flow. Key operating parameters used in the analysis of this event are presented in Table 15.3.3-2. The range of neutronics parameters supported by this analysis are presented in Table 15.3.3-3.
The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.3.2-2) computer program following the method described in Reference 15.3.2-5 and 15.3.2-
: 7. The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.3.2-4) to predict the MDNBR for the event using the Statistical Core Design (SCD) methodology (Reference 15.3.2-6) and the HTP CHF correlation (Reference 15.3.2-8) for the Framatome 17x17 Advanced W-HTP fuel..
Results - This transient tripped on low primary coolant flow. For the overpressurization case, the maximum primary system pressure is less than the design limits. The maximum secondary pressure is bounded by Section 15.2.3 Turbine Trip. The sequence of events for the bounding overpressurization case is given in Table 15.3.3-4. The responses to key system variables are given in Figures 15.3.3-1 to 15.3.3-6.
The sequence of events for the MDNBR case is given in Table 15.3.3-5. This transient tripped on low primary coolant flow. The responses to key system variables are given in Figures 15.3.3-7 to 15.3.3-13.
15.3.3.3    Radiological Consequences of a Locked Rotor An instantaneous seizure of a reactor coolant pump rotor is assumed to occur which rapidly reduces flow through the affected reactor coolant loop. Fuel clad damage may be predicted to occur as a result of this accident. No fuel centerline melt is predicted to occur. Due to the pressure differential between the primary and secondary systems and assumed steam generator tube leakage, fission products are discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere through either the atmospheric relief valves or safety valves. In addition, iodine activity is contained in the secondary coolant before the accident and some of this activity is released to the atmosphere as a result of steaming from the steam generators following the accident.
The analysis of the locked rotor radiological consequences uses the analytical methods and assumptions outlined in RG 1.183, Appendix G (Locked Rotor).
A summary of input parameters and assumptions is provided in Table 15.3.3-6. Additional clarification is provided as follows:
a) It is assumed that 8% of the fuel rods in the core suffer damage as a result of the locked rotor sufficient that all of their gap activity is released to the reactor coolant system.
Eight percent of the total I-131 core activity is in the fuel-cladding gap. Ten percent of the total Kr-85 core activity is in the fuel-cladding gap. Five percent of other iodine isotopes and other noble gases and 12 percent of the total core activity for alkali metals Amendment 63                                                                              Page 68 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 are assumed to be in the fuel-cladding gap. In the calculation of activity releases from the failed fuel and maximum radial peaking factor of 1.73 was applied. The core fission product inventory is given in Table 15.0.9-1.
b) The iodine activity concentration of the secondary coolant at the time the locked rotor occurs is assumed to be equivalent to the Technical Specification limit of 0.1 Ci/gm of DE I-131. The iodine activity concentration of the secondary coolant is given in Table 15.0.9-7. The alkali metal activity concentration of the secondary coolant at the time the locked rotor occurs is assumed to be 10% of the primary side concentration.
c) The amount of primary to secondary SG tube leakage is assumed to be 1 gpm total.
d) An iodine partition factor in the SGs of 0.01 (curies iodine/gm steam) / (curies iodine/gm water) is used. This partition factor is also applied to alkali metals.
e) All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.
f)  At 8 hours after the accident, the RHR system is assumed to be placed into service for heat removal and there is no further steam release to the atmosphere from the secondary system.
15.3.3.3.1 Offsite Doses The offsite doses are calculated using the assumptions and equations in Section 15.0A.1.
15.3.3.3.2 Control Room Doses The control room assumptions are provided in Section 15.6.5.4.3 and Table 15.6.5-15.
The control room HVAC is switched to the emergency post-accident recirculation mode after receiving a high radiation signal. The high radiation signal is reached at 3 seconds into the event. The control room HVAC is switched over to the emergency post-recirculation mode at 18 seconds (3 seconds signal initiation plus 15 seconds delay time for switching between modes).
An operator action switches the control room from the post-accident recirculation mode to the pressurization mode at 2 hours after event initiation. The 15-second delay to switch between modes was also assumed with the operator action. Thus 2 hours and 33 seconds actually modeled for the time of operator action switchover to the pressurization mode.
15.3.3.3.3 Results The potential radiological consequences resulting from a locked rotor have been conservatively analyzed, using assumptions and models described in previous sections. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The radiological analysis results for this event are listed in Table 15.3.3-7. The resultant doses are within the applicable limits.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 15.3.3.4      Conclusions For the overpressurization case, the maximum reactor primary system pressure is less than the 120% design limit (3000 psia). The maximum secondary pressure is bounded by Section 15.2.3 Turbine Trip.
For the MDNBR case, the predicted MDNBR is less than the 95/95 safety limit. Less than 8% of the fuel is predicted to fail based on DNB criteria. This event does not pose a credible challenge to FCM safety criteria. The radiological doses have been calculated based on 8%
assumed cladding failure. The offsite radiological doses are less that 10% of the 10CFR50.67 limits (i.e. 2.5 REM TEDE) and the control room doses are less than the 10CFR50.67 limit of 5 REM TEDE. Thus, the event acceptance criteria are met.
15.3.4    REACTOR COOLANT PUMP SHAFT BREAK 15.3.4.1      Identification of Causes and Accident Description The accident is postulated as an instantaneous failure of a reactor coolant pump shaft. Flow through the affected reactor coolant loop is rapidly reduced, though the initial rate of reduction of coolant flow is greater for the reactor coolant pump rotor seizure event. Reactor trip is initiated on a low flow signal in the affected reactor coolant loop.
Following initiation of the reactor trip, heat stored in the fuel rods continues to be transferred to the reactor coolant causing the coolant to expand. At the same time, heat transfer to the shell side of the steam generators is reduced, first because the reduced flow results in a decreased tube side film coefficient and then because the reactor coolant in the tubes cools down while the shell side temperature increases (turbine steam flow is reduced to zero upon plant trip). The rapid expansion of the reactor coolant in the reactor core, combined with reduced heat transfer in the steam generators causes an insurge into the pressurizer and a pressure increase throughout the RCS. The insurge into the pressurizer compresses the steam volume, actuates the automatic pressurizer spray system, opens the power operated relief valves, and opens the pressurizer safety valves, in that sequence. The three power operated relief valves are designed for reliable operation and would be expected to function properly during the accident.
However, for conservatism, their pressure reducing effect as well as the pressure reducing effect of the spray is not included in the analysis. The peak RCS pressure is bounded by that for a locked rotor (see Sections 15.3.3.2). This event is classified as an ANS Condition IV incident (a limiting fault) as defined in Section 15.0.1.
No single active failure will adversely affect the consequences of the event.
Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
15.3.4.2      Conclusions The consequences of a reactor coolant pump shaft break would be no worse than those calculated for the locked rotor incident (see Section 15.3.3). With a failed shaft, the impeller could conceivably be free to spin in a reverse direction as opposed to being fixed in position as assumed in the locked rotor analysis. However, the net effect on core flow is negligible, resulting in only a slight decrease in the end point (steady state) core flow. For both the shaft break and Amendment 63                                                                          Page 70 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 locked rotor incidents, reactor trip occurs very early in the transient. In addition, the locked rotor analysis conservatively assumes that DNB occurs at the beginning of the transient. The calculations performed for this event showed that the MDNBR, fuel centerline melt, and overpressurization consequences are bounded by those of the Pump Rotor Seizure/Locked Rotor event. The Condition IV acceptance criteria are met for the Pump Rotor Seizure/Locked Rotor event, as noted in Section 15.3.3.3.
 
==REFERENCES:==
SECTION 15.3 15.3.1-1        Deleted by Amendment No. 48 15.3.1-2        Deleted by Amendment No. 48 15.3.2-1        Deleted by Amendment No. 45 15.3.2-2        EPRI NF-7450(A), Revision 10, RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, September 2014.
15.3.2-3        Deleted by Amendment No. 63.
15.3.2-4        EPRI NP-2511-CCM-A, Revision 4, VIPRE A Thermal-Hydraulic Code for Reactor Cores, June 2007.
15.3.2-5        DPC-NE-3009, Revision 0, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, April 2018.
15.3.2-6        DPC-NE-2005-PA, Revision 5, "Thermal-Hydraulic Statistical Core Design Methodology," March 2016.
15.3.2-7        DPC-NE-3008, Revision 0, Thermal-Hydraulic Models for Transient Analysis, April 2018.
15.3.2-8        EMF-92-153-PA, Revision 1 HTTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, January 2005.
15.4    REACTIVITY AND POWER DISTRIBUTION ANOMALIES A number of faults have been postulated which could result in reactivity and power distribution anomalies. Reactivity changes could be caused by control rod motion or ejection, boron concentration changes, or addition of cold water to the Reactor Coolant System (RCS). Power distribution changes could be caused by control rod motion, misalignment, or ejection, or by static means such as fuel assembly mislocation. These events are discussed in this section.
Detailed analyses are presented for the most limiting of these events.
Discussions of the following incidents are presented in Section 15.4:
a) Uncontrolled rod cluster control assembly bank withdrawal from a subcritical or low power startup condition. (ANS Condition II event).
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 b) Uncontrolled rod cluster control assembly bank withdrawal at power. (ANS Condition II event).
c) Rod cluster control assembly misalignment. (ANS Condition II and III events).
d) Startup of an inactive reactor coolant pump at an incorrect temperature. (ANS Condition II event).
e) Chemical and Volume Control System malfunction that results in a decrease in the boron concentration in the reactor coolant. (ANS Condition II event).
f)  Inadvertent loading and operation of a fuel assembly in an improper position. (ANS Condition III event).
g) Spectrum of rod cluster control assembly ejection accidents. (ANS Condition IV event).
Section 15.0.1 contains a discussion of ANS classifications.
15.4.1    UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER STARTUP CONDITION 15.4.1.1    Identification of Causes and Accident Description A rod cluster control assembly (RCCA) withdrawal accident is defined as an uncontrolled addition of reactivity to the reactor core caused by withdrawal of RCCA's resulting in a power excursion. Such a transient could be caused by a malfunction of the reactor control or rod control systems. This could occur with the reactor subcritical or during startup. An "at power" case is discussed in Section 15.4.2.
Although the reactor is normally brought to power from a subcritical condition by means of RCCA withdrawal, initial startup procedures with a clean core calls for boron dilution to a condition with minimal RCCA insertion. The maximum rate of reactivity increase in the case of boron dilution is less than that assumed in this analysis (see Section 15.4.6).
The RCCA drive mechanisms are wired into preselected bank configurations which are not altered during reactor life. These circuits prevent the RCCAs from being automatically withdrawn in other than their respective banks. Power supplied to the banks is controlled such that no more than two banks can be withdrawn at the same time and in their proper withdrawal sequence. The RCCA drive mechanisms are of the magnetic latch type and coil actuation is sequenced to provide variable speed travel. The maximum reactivity insertion rate analyzed in the detailed plant analysis is that occurring with the simultaneous withdrawal of the combination of two sequential control banks having the maximum combined worth at maximum speed.
This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Section 15.0.1. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
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: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur.
No single active failure will adversely affect the consequences of the event.
Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
The neutron flux response to a continuous, rapid reactivity increase is characterized by a very fast rise terminated by the reactivity feedback effect of the negative Doppler coefficient. This self-limitation of the power excursion is of primary importance since it limits the power to a tolerable level during the delay time for protective action. Should a continuous RCCA withdrawal accident occur, the transient will be terminated by the following automatic features of the Reactor Protection System:
: 1) Source range high neutron flux reactor trip. Actuated when either of two independent source range channels indicates a neutron flux level above a preselected manually adjustable setpoint. This trip function may be manually bypassed only after an intermediate range flux channel indicates a flux level above a specified level. It is automatically reinstated when both intermediate range channels indicate a flux level below a specified level.
: 2) Intermediate range high neutron flux reactor trip. Actuated when either of two independent intermediate range channels indicates a flux level above a preselected manually adjustable setpoint. This trip function may be manually bypassed only after two out of the four power range channels are reading above approximately 10 percent of full power and is automatically reinstated when three out of the four channels indicate a power level below this value.
: 3) Power range high neutron flux reactor trip (low setting). Actuated when two out of the four power range channels indicate a power level above approximately 25 percent of full power. This trip function may be manually bypassed when two out of the four power range channels indicate a power level above approximately 10 percent of full power and is automatically reinstated only after three out of the four channels indicate a power level below this value.
: 4) Power range high neutron flux reactor trip (high setting). Actuated when two out of the four power range channels indicate a power level above a preset setpoint. This trip function is always active.
: 5) High nuclear flux rate reactor trip. Actuated when the positive rate of change of neutron flux on two out of four nuclear power range channels indicate a rate above the preset setpoint. This trip function is always active.
In addition, control rod stops on high intermediate range flux level (one out of two) and high power range flux level (one out of four) serve to discontinue rod withdrawal and prevent the need to actuate the intermediate range flux level trip and the power range flux level trip, Amendment 63                                                                          Page 73 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 respectively. This event is precluded by plant procedures in Mode 3 when Tavg <551&deg;F and in Modes 4-6.
15.4.1.2    Analysis of Effects and Consequences Method of Analysis - The objective of this analysis is to bound plant operational modes below approximately 10% of rated power to Mode 3 when Tavg >551&deg;F. The analysis examines the possible operational modes and state conditions between these two limits to develop a bounding case.
This event is driven by the magnitude and rate of reactivity insertion.
The maximum positive reactivity insertion rate is rapid enough that very high neutron powers are calculated, but of short enough duration that excessive energy deposition does not occur.
Because the event is very rapid, an increase in primary coolant temperature lags behind power.
The low coolant flow rate in the core accompanied by a rapid surge of power makes this event a challenge to both the Specified Acceptable Fuel Design Limits (SAFDLs) and system pressurization. Lower reactivity insertion rates produce extensive reactor coolant system heatup which may also challenge the SAFDLs. The challenge to the SAFDLs is controlled by the rate of energy dissipation from the fuel rod. The challenge to the system pressurization is due to the large and rapid thermal expansion of the coolant in the core.
A low initial power yields the maximum margin to trip and, hence, maximum time for withdrawal to trip. This will yield the largest prompt multiplication which maximizes overshoot past trip. The initial power selected conservatively bounds the shutdown condition. Therefore, this event is analyzed at HZP conditions. Two reactor coolant pumps are assumed operational to minimize the coolant flow. The event is analyzed using BOC neutronics conditions.
The overpressurization scenario was analyzed using the methodology described in Reference 15.4.2-3 and determined to be bounded by Event 15.2.3, Turbine Trip. Furthermore, peak secondary pressure is also bounded by 15.2.3 and is not presented here. An MDNBR case is analyzed to evaluate the challenge to the SAFDLs. The MDNBR case bounds operation in Mode 3 with Tavg >551&deg;F and operation in Mode 1 or 2 while reactor power is below the power range flux trip (low setting) reset setpoint, approximately 10% rated power.
Conservative conditions established for the analysis of this event are presented in Table 15.4.1-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. For the MDNBR analysis of this event, all trips listed in this table were disabled except power range neutron flux, low setting. This approach bounds Mode 3 conditions as the power range neutron flux reactor trip setpoint is higher than that of the source range reactor trip, resulting in a later reactor trip at a higher reactor power. Key operating parameters used in the analysis of this event are presented in Table 15.4.1-2. The range of neutronics parameters supported by this analysis are presented in Table 15.4.1-3.
In order to give conservative DNBR results for a startup accident, the following assumptions are made:
: 1) Since the magnitude of the power peak reached during the initial part of the transient for any given rate of reactivity insertion is strongly dependent on the Doppler coefficient, conservatively low values as a function of temperature are used.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 2) Contribution of the moderator reactivity coefficient is secondary during the initial part of the transient because the heat transfer time between the fuel and the moderator is much longer than the neutron flux response time. However, after the initial neutron flux peak, the succeeding rate of power increase is affected by the moderator reactivity coefficient.
A conservative value of +5 pcm/&deg;F at hot zero power is used in the analysis to yield the maximum peak heat flux.
: 3) The reactor is assumed to be at hot zero power. This assumption is more conservative than that of a lower initial system temperature. The higher initial system temperature yields a less negative (smaller absolute magnitude) Doppler coefficient which tends to reduce the Doppler feedback effect thereby increasing the neutron flux peak.
: 4) Reactor trip is assumed to be initiated by power range high neutron flux (low setting)
The most adverse combination of instrument and setpoint errors, as well as delays for trip signal actuation and RCCA release, is taken into account. Per Table 15.0.6-2, the analytical trip setpoint assumed is 34.4% of rated thermal power. The neutron flux signal input to the reactor trip accounts for the effects of rod shadowing, detector miscalibration, and downcomer attenuation. In addition, the reactor trip insertion characteristic is based on the assumption that the highest worth RCCA is stuck in its fully withdrawn position. See Section 15.0.5 for RCCA insertion characteristics.
: 5) The maximum positive reactivity insertion rate assumed is greater than that for the simultaneous withdrawal of the combination of the two sequential control banks having the greatest combined worth at maximum speed (45 in./ min.). Control rod drive mechanism design is discussed in Section 4.6. A spectrum of reactivity insertion rates up to the maximum positive reactivity insertion rate are analyzed to determine the limiting case.
: 6) The initial power level was assumed to be below the power level expected for any shutdown condition. The combination of highest reactivity insertion rate and lowest initial power produces the highest peak heat flux.
: 7) Two reactor coolant pumps are assumed to be in operation.
The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.4.1-3) computer program following the method described in Reference 15.4.1-4 and Reference 15.4.2-3. The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.4.1-5) to predict the MDNBR for the event using the statistical core design methodology (Reference 15.4.3-3) and the HTP CHF correlation (Reference 15.4.1-6) for the Framatome 17x17 Advanced W-HTP fuel.
Results The event tripped on power range neutron flux, low setting. The sequence of events is given in Table 15.4.1-4. The transient response from the limiting MDNBR case is given in Figures 15.4.1-1 through 15.4.1-4. The VIPRE-01 analysis showed that the limiting MDNBR case occurred at a lower reactivity insertion rate, due to extensive RCS heatup. The limiting centerline fuel melt case occurs at the maximum positive reactivity insertion rate, as the highest reactivity insertion rate produces the highest peak heat flux.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 The results of this analysis also demonstrate that the hot spot centerline temperature is below the centerline melt temperature.
15.4.1.3      Conclusions The results of the analysis indicate that the predicted MDNBR is greater than the safety limit.
The statistical design limit (Reference 15.4.3-3) ensures that, with 95% probability and 95%
confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. The fuel centerline melt threshold temperature is not penetrated during this event. Thus, the acceptance criteria for this event are met.
15.4.2    UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL AT POWER 15.4.2.1      Identification of Causes and Accident Description Uncontrolled RCCA bank withdrawal at power results in an increase in the core heat flux. Since the heat extraction from the steam generator lags behind the core power generation until the steam generator pressure reaches the power operated relief or safety valve setpoint, there is a net increase in the reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant reactor coolant temperature rise could eventually result in DNB. Therefore, in order to avert damage to the fuel clad, the Reactor Protection System is designed to terminate any such transient before the DNBR falls below the safety analysis value.
This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Section 15.0.1. The acceptance criteria for this event are:
: 1) The pressure in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur.
The automatic features of the Reactor Protection System which prevent core damage following the postulated accident include the following:
: 1) Power range neutron flux instrumentation actuates a reactor trip if two out of four channels exceed an overpower setpoint.
: 2) Reactor trip is actuated if any two out of three T channels exceed an overtemperature T setpoint. This setpoint is automatically varied with axial power imbalance, reactor coolant temperature and pressure to protect against DNB.
: 3) Reactor trip is actuated if any two out of three T channels exceed an overpower T setpoint. This setpoint is automatically varied with axial power imbalance to ensure that the allowable heat generation rate (kW/ft.) is not exceeded.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 4) A high pressurizer pressure reactor trip actuated from any two out of three pressure channels, which is set at a fixed point. This set pressure is less than the set pressure for the pressurizer safety valves.
: 5) A high pressurizer water level reactor trip actuated from any two out of three level channels when the reactor power is above approximately 10 percent (Permissive-7).
In addition to the above listed reactor trips, there are the following RCCA withdrawal blocks:
: 1) High neutron flux (one out of four power range).
: 2) Overpower T (two out of three).
: 3) Overtemperature T (two out of three).
The manner in which the combination of overpower and overtemperature T trips provide protection over the full range of RCS conditions is described in Chapter 7.
The area of permissible operation (power, pressure, and temperature) is bounded by the combination of reactor trips: high neutron flux (fixed setpoint); high pressure (fixed setpoint);
low pressure (fixed setpoint); overpower and overtemperature T (variable setpoints).
Plant systems and equipment which are available to mitigate the effects of the accident are discussed in Section 15.0.8 and listed in Table 15.0.8-1. No single active failure in any of these systems or equipment will adversely affect the consequences of the accident.
15.4.2.2    Analysis of Effects and Consequences The uncontrolled RCCA bank withdrawal at power analysis consists of the system thermal-hydraulic for DNB analysis, and the primary side overpressurization analysis. Secondary system pressurization is bounded by the turbine trip event and is not analyzed.
Method for DNB Analysis - The power range to be considered in this analysis is from the power range high flux (low setting) trip reset point (approximately 10% of rated power), up to full power. Uncontrolled RCCA bank withdrawal at power levels below 10% rated power are considered in Event 15.4.1. This analysis considers a spectrum of reactivity insertion rates at initial power levels of 10%, 50% and 100%. Since neutronics feedback as a function of cycle exposure and design also influences the results, these effects are also included in the analysis.
A broad range of reactivity insertion rates are possible. Therefore, a spectrum of reactivity insertion rates were evaluated in order to bound events ranging from a slow dilution of the primary system boron concentration to the maximum possible RCCA bank withdrawal rate at maximum bank worth.
Reactivity feedback effects are bounded by analyzing a series of BOC and EOC cases using the bounding least negative moderator temperature coefficient and Doppler temperature coefficient.
The reactivity insertion rate due to the bank withdrawal is varied until the maximum withdrawal rate is met.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 Conservative conditions established for the system thermal-hydraulic calculation for DNB analysis of this event are presented in Table 15.4.2-1. Available reactor protection system trips are presented in Table 15.0.6-2. Key operating parameters used in the analysis of this event are presented in Table 15.4.2-2. The range of neutronics parameters supported by this analysis are presented in Table 15.4.2-3.
In order to obtain conservative results for an uncontrolled rod withdrawal at power accident, the following assumptions are made:
: 1) The reactor trip on high neutron flux is assumed to be actuated at a conservative value of 113.5 percent of nominal full power. The calculated neutron flux is conservatively modified to account for transient changes in the flux incident on the excore detectors.
The T trips include all adverse instrumentation and setpoint errors; the delays for trip actuation are assumed to be the maximum values.
: 2) The RCCA trip insertion characteristic is based on the assumption that the highest worth assembly is stuck in its fully withdrawn position.
: 3) The maximum positive reactivity insertion rate is greater than that for the simultaneous withdrawal of the combinations of the two control banks in normal overlap having the maximum combined worth at maximum speed.
The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.4.1-3) computer program following the method described in Reference 15.4.1-4 and Reference 15.4.2-3. The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.4.1-5) to predict the MDNBR for the event using the Statistical Core Design (SCD) Methodology (Reference 15.4.3-3) and the HTP correlation (Reference 15.4.1-6) for the Framatome 17x17 advanced W-HTP fuel.
Method for Primary Side Overpressurization Analysis - The maximum primary side overpressurization analysis is performed using RETRAN-3D computer program (Reference 15.4.1-3) with the non-SCD methodology (Reference 15.4.2-3) that accounts for uncertainties directly in initial power, RCS pressure, pressurizer level, and RCS Tavg. The maximum reactivity insertion rate at initial power levels of 8% and 12% are analyzed, which is 10% power plus or minus uncertainty. Conservative conditions established for the primary side overpressurization analysis of this event are presented in Table 15.4.2-1. Available reactor protection system trips are presented in Table 15.0.6-2 with the following assumed conservative setpoints.
: 1) The high pressurizer pressure trip setpoint is identified in Table 15.0.6-1.
: 2) The high pressurizer level trip is identified in Table 15.0.6-1.
Key operating parameters used in the analysis of this event are presented in Table 15.4.2-2a.
The range of neutronics parameters supported by this analysis are presented in Table 15.4.2-3.
Assumptions 1), 2), and 3) in the DNB analysis method are also applied.
DNB Analysis Results - The uncontrolled RCCA bank withdrawal for DNB transients are analyzed for a spectrum of reactivity insertion rates at initial power levels of 10%, 50% and HFP.
The limiting uncontrolled RCCA bank withdrawal for DNB transient occurred at 10% power with BOC kinetics and the most limiting reactivity insertion rate among a spectrum of analyzed Amendment 63                                                                        Page 78 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 values and the most limiting power peakings. Figure 15.4.2-1 presents the MDNBR results for the range of reactivity addition rates analyzed for the 10% with BOC kinetics. The predicted MDNBR is greater than the safety limit.
The sequence of events for the limiting Uncontrolled RCCA Bank Withdrawal transient is given in Table 15.4.2-4. The responses to key system variables for the limiting transient are given in Figures 15.4.2-2 to 15.4.2-9.
Primary Side Overpressurization Analysis Results - The limiting primary side overpressurization transient occurs at 8% power level and for a maximum reactivity insertion rate. The sequence of events of this limiting transient is given in Table 15.4.2-5. The maximum primary side pressure, which occurs at the bottom of the reactor vessel and is shown in Figure 15.4.2-10, is below the acceptance criterion. The primary side overpressurization of uncontrolled RCCA bank withdrawal is bounded by the Turbine Trip event (Event 15.2.3).
15.4.2.3    Conclusions Reactivity insertion transient calculations demonstrate that the DNB correlation limit will not be penetrated during any credible reactivity insertion transient at full power. Analysis results demonstrate that transients initiated at power levels below full power are more limiting for DNBR. The critical heat flux correlation limit ensures that, with 95% probability and 95%
confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. The fuel centerline melt threshold is not penetrated during this event. The maximum primary pressurization result is below the acceptance criterion and is bounded by the Turbine Trip event.
15.4.3    ROD CLUSTER CONTROL ASSEMBLY MISOPERATION (SYSTEM MALFUNCTION OR OPERATOR ERROR)
Rod cluster control assembly (RCCA) misoperation accidents include:
: 1) One or more dropped assemblies within the same group.
: 2) A dropped full length assembly bank.
: 3) Statically misaligned full length assembly.
: 4) Withdrawal of a single full length assembly.
Each RCCA has a position indicator channel which displays position of the assembly. The displays of assembly positions are grouped for the operator's convenience. Fully inserted assemblies are further indicated by a rod at bottom signal, which actuates a local alarm and a Control Room annunciator. Group demand position is also indicated.
Full length RCCAs are always moved in preselected banks, and the banks are always moved in the same preselected sequence. Each bank of RCCAs is divided into two groups. The rods comprising a group operate in parallel through multiplexing thyristors. The two groups in a bank move sequentially such that the first group is always within one step of the second group in the bank. A definite schedule of actuation (or deactuation of the stationary gripper, movable gripper, and lift coils of a mechanism) is required to withdraw the RCCA attached to the mechanism. Since the stationary gripper, movable gripper, and lift coils associated with the Amendment 63                                                                        Page 79 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 RCCAs of a rod group are driven in parallel, any single failure which would cause rod withdrawal would affect a minimum of one group. Mechanical failures are in the direction of insertion, or immobility.
Plant systems and equipment which are available to mitigate the effects of the various control rod misoperations are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
15.4.3.1      Dropped Full Length RCCA or RCCA Bank 15.4.3.1.1 Identification of causes and accident description The dropped RCCA and dropped RCCA bank events are initiated by a de-energized control rod drive mechanism or by a malfunction associated with a RCCA bank during power operation.
The result is that a single RCCA or RCCA bank falls into the core. The dropped RCCA promptly inserts negative reactivity which reduces reactor power and disturbs the power distribution, resulting in an increase (augmentation) of local power peaking. Two operational states are available: Manual and Automatic Rod Control (ARC).
In the automatic rod control mode, the ARC system receives signals from the excore detectors and the turbine to indicate a primary/secondary side power mismatch. In an attempt to eliminate a mismatch, the ARC system initiates the movement of a partially inserted control bank. The ARC system detects the drop in power and initiates control bank withdrawal. Power overshoot may occur, after which the control rod system will insert the control bank and return the plant to nominal power. The magnitude of the power overshoot is a function of:
* Core reactivity coefficients;
* Dropped rod worths;
* Differential bank worths; and
* Rod shadowing factors (RSF).
An automatic and redundantly actuated reduction in turbine load demand (turbine runback) is provided as protection for this event. When turbine runback occurs, the automatic rod controller setpoint reference temperature is also set back to an average primary coolant temperature corresponding to the new turbine load demand. With the turbine runback the reduction in load initially results in a load mismatch if the dropped rod (bank) reactivity does not match that required for the setback power level. If the dropped rod worth is not equal to the reactivity to match the power runback, a power mismatch between primary and secondary occurs and is detected by either coolant temperature or neutron power above or below setpoint. The controller output signal is sent to the control rod driver controller, which acts to minimize the mismatch.
The ARC system uses the highest NI channel reading of the four available signals for rod control. To incorporate a single failure, the analysis uses the second highest NI channel reading to drive the ARC system.
In the manual mode, a decrease in moderator temperature results from the initial power reduction. At EOC conditions, automatic action taken by the turbine control system to open the Turbine Governor Valves in combination with a strongly negative moderator temperature coefficient can return the reactor to a new equilibrium condition with an elevated radial power Amendment 63                                                                          Page 80 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 peaking factor consequent to the dropped RCCA. Elevated clad heat flux in the hot assembly may result in an approach to the DNBR SAFDL.
An automatic and redundantly actuated reduction in turbine load demand (turbine runback) is also provided as protection for this event in the manual mode. When turbine runback occurs, the turbine load reduction reduces secondary steam flow, causing a tendency for the secondary side temperature and pressure to increase. Thus the primary coolant temperature decrease characteristic of the event is mitigated, reducing the reactivity insertion contingent on cooldown and reducing the ultimate power level at which the reactor stabilizes.
A dropped assembly or assembly bank is detected by:
: 1) Sudden drop in the core power level as seen by the Nuclear Instrumentation System.
: 2) Asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples.
: 3) Rod at bottom signal.
: 4) Rod deviation alarm.
: 5) Rod position indication.
The Dropped Full Length RCCA or RCCA Bank event is classified as a Condition II event. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur.
The most adverse credible single failure assumption in the Dropped Full Length RCCA event is the loss of one NIS channel. This results in a reduced ability for the NIS and the ARC system to detect the core power redistribution characteristic of the event. If cases exist where a power range high neutron flux trip intercedes, the trip is delayed until two of the three remaining channels reach the trip setpoint. In cases where the plant stabilizes at a new equilibrium condition without a reactor trip, no further protective action is required and this assumption has no impact on the RPS response. Therefore, consideration of other single failures within the protection system is not applicable.
15.4.3.1.2 Analysis of effects and consequences Method of Analysis - The characteristic system response for this event is strongly dependent on the neutron kinetics feedback, the worth of the dropped RCCA or RCCA bank, and on the availability of the ARC system. The Dropped RCCA transients are analyzed with automatic rod control. Manual rod control cases are bounded by the automatic rod control cases and are not analyzed. This event is evaluated at BOC, MOC, and EOC conditions for a spectrum of drop rod Amendment 63                                                                          Page 81 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 worths, differential worths, and rod shadowing factors. The calculated neutron flux is conservatively modified to account for transient changes in the flux incident on the same detectors.
Conservative conditions established for the analysis of this event are presented in Table 15.4.3-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. Key operating parameters used in the analysis of this event are presented in Table 15.4.3-2. The range of neutronics parameters supported by this analysis are presented in Table 15.4.3-3.
The analysis evaluates the impact of the dropped rod event against the DNBR and fuel centerline melt acceptance criteria. The transient system response is calculated using the RETRAN-3D computer code (Reference 15.4.1-3). The minimum DNBR analysis is performed using the VIPRE-01 computer code (Reference 15.4.1-5). The minimum DNBR is determined using SCD methodology and the W-HTP correlation (Reference 15.4.3-3 and Reference 15.4.1-
: 6) for the Framatome 17x17 advanced W-HTP fuel. Peak primary and secondary pressures are bounded by other events and are not analyzed per Reference 15.4.2-3.
The RETRAN-3D analysis generates the core state point conditions corresponding to the time of minimum DNBR. The core thermal-hydraulic boundary conditions from the RETRAN-3D analysis and the core power distribution are then input to the VIPRE-01 model to calculate peaking limits that preclude DNB.
The dropped RCCA bank is distinguished from the dropped RCCA in the RETRAN-3D analysis only by the greater magnitude of rod bank worth and radial peaking associated with the dropped RCCA bank.
Results - The dropped RCCA transients are analyzed with automatic rod control. The ARC cases are more limiting than the manual rod control cases. The limiting ARC dropped RCCA transient occurred with EOC kinetics, a Control Bank D worth of 680 pcm, and a 400 pcm dropped rod worth.
The sequence of events for the limiting Dropped RCCA transient is given in Table 15.4.3-4a.
The responses to key system variables for the limiting transient are given in Figures 15.4.3-1 to 15.4.3-11.
15.4.3.1.3 Conclusions The results of the analysis of the Dropped RCCA/Bank event demonstrate that the acceptance criteria are met. The predicted MDNBR is greater than the safety limit. The critical heat flux correlation limit ensures that, with 95% probability and 95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. The fuel centerline melt threshold is not penetrated during this event.
15.4.3.2    Withdrawal of a Single Full Length RCCA 15.4.3.2.1 Identification of causes and accident description The rod withdrawal event is initiated by an electrical or mechanical failure in the Rod control System that causes the inadvertent withdrawal of a single RCCA. A rod is withdrawn from the reactor core causing an insertion of positive reactivity which results in a power excursion Amendment 63                                                                          Page 82 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 transient, increasing the core heat flux and creating a challenge to DNB margin. The DNB margin is further reduced by the mismatch between the constant energy removal rate of the steam generators and the increased energy generation rate in the core which increases the primary system temperature.
The system response is essentially the same as that occurring in the Uncontrolled Bank Withdrawal at Power event (Event 15.4.2). The single RCCA withdrawal is distinguished from the withdrawal of an RCCA bank by the severe radial power redistribution. High radial peaking is localized in the region of the single withdrawn RCCA and may, in severe cases, surpass the design limits.
Automatic protection for this event is afforded by the OTT and high pressurizer level reactor trips. Because of the localized power peaking, there is the possibility of violating the SAFDLs.
This is acceptable since this event is classified as a Condition III event, in which a small fraction of fuel failure is permitted. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) A small fraction of fuel failures may occur, but these fuel failures should not hinder core coolability.
: 3) Radiological consequences should be less than 10% 10 CFR 50.67 limits (i.e.,2.5 rem TEDE) offsite and less than the 10CFR50.67 limit of 5 rem TEDE in the control room.
: 4) The event should not generate a limiting fault or result in the consequential loss of the reactor coolant or containment barriers.
No single electrical or mechanical failure in the Rod Control System could cause the accidental withdrawal of a single RCCA from the inserted bank at full power operation. The operator could deliberately withdraw a single RCCA in the control bank since this feature is necessary in order to retrieve an assembly should one be accidentally dropped. The event analyzed must result from multiple wiring failures (probability for single random failure is on the order of 10-4/year (refer to Section 7.7.2.2) or multiple significant operator actions and subsequent and repeated operator disregard of event indication. The probability of such a combination of conditions is low enough that the limiting consequences may include slight fuel damage.
In the extremely unlikely event of simultaneous electrical failures which could result in single RCCA withdrawal, rod deviation and rod control urgent failure would both be displayed on the plant annunciator, and the rod position indicators would indicate the relative positions of the assemblies in the bank. The urgent failure alarm also inhibits automatic rod motion in the group in which it occurs. Withdrawal of a single RCCA by operator action, whether deliberate or by a combination of errors, would result in activation of the same alarm and the same visual indications. Withdrawal of a single RCCA results in both positive reactivity insertion tending to increase core power, and an increase in local power density in the core area associated with the RCCA. Automatic protection for this event is provided by the overtemperature T or high pressurizer level reactor trip, although due to the increase in local power density it is not possible in all cases to provide assurance that the core safety limits will not be violated.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 This event is analyzed assuming the worst single failure is the loss of an excore detector adjacent to the withdrawn RCCA.
15.4.3.2.2 Analysis of effects and consequences Method of Analysis - The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.4.1-3) computer program following the method described in Reference 15.4.2-3 and Reference 15.4.1-4. The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.4.1-5) to predict the MDNBR for the event using the Statistical Core Design (SCD) Methodology (Reference 15.4.3-3) and the HTP CHF correlation (Reference 15.4.1-6) for the Framatome 17x17 Advanced W-HTP fuel.
The result of this calculation is a power peaking limit which, at the statepoint, yields a DNBR equal to the limit value. All fuel pins exceeding this peaking limit are assumed to undergo DNB and subsequently fail.
Plant characteristics and initial conditions are listed in Table 15.4.3-6. In order to obtain conservative results for a single RCCA withdrawal accident, the following assumptions are made:
: 1.      The values assumed for initial reactor power, pressurizer pressure, RCS average temperature, and RCS flow include no allowance for uncertainties.
Uncertainties in these initial conditions are included in the SCD DNBR limit as described in Reference 15.4.3-3. The key operating parameters for the single RCCA withdrawal analyses are listed in Table 15.4.3-7.
: 2.      A least negative moderator density coefficient of reactivity is assumed corresponding to the beginning of core life. A least negative variable fuel temperature coefficient is assumed corresponding to the beginning of core life. These feedback assumptions lead to the most limiting statepoint with respect to thermal-hydraulic conditions. Peaking limits derived from this statepoint are compared to peaking results from beginning of core life to ensure that the percent of fuel rods in DNB is smaller than the acceptance criterion. The range of neutronics parameters is identified in Table 15.4.3-8.
: 3.      The T and high pressurizer level trips include all adverse instrumentation and setpoint errors. The delays for trip actuation are assumed to be maximum values.
: 4.      The RCCA trip insertion characteristic is based on the assumption that the highest worth assembly is stuck in its fully withdrawn position.
: 5.      The maximum positive insertion rate is greater than that for the maximum speed withdrawal of the most reactive single Control Bank D RCCA from at or above its insertion limit, accounting for uncertainties in the indicated RCCA position.
: 6.      The case presented assumes normal pressurizer heater operation and the pressurizer PORVs operable. Sensitivity studies are performed to ensure that Amendment 63                                                                          Page 84 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 the minimum DNBR for this case bounds other pressurizer pressure control availability assumptions.
: 7.      The rod withdrawal in the case presented is initiated at 100% RTP. Sensitivity studies are performed to ensure that the minimum DNBR for this case bounds other initial power level assumptions of 10% and 50% RTP. Reload specific power distributions are compared to DNB and fuel centerline melt limits to account for localized effects from the withdrawal of the control rod.
Results - Table 15.4.3-9 shows the sequence of events for the 100% single uncontrolled rod withdrawal transient. Figures 15.4.3-12, 15.4.3-13, 15.4.3-14, and 15.4.3-15 show the transient response for the single uncontrolled rod withdrawal from an initial power of 100% RTP. System temperature and system pressure increase until reactor trip occurs, which is after the RCCA is completely withdrawn.
For the case presented, the RCCAs are in manual control mode, where continuous withdrawal of a single RCCA results in both an increase in core power and coolant temperature, and an increase in the local hot channel factor in the area of the withdrawing RCCA. In terms of the overall system response, this case is similar to those presented in Section 15.4.2. Depending on initial bank insertion and location of the withdrawn RCCA, automatic reactor trip on OTT or high pressurizer pressure may not occur fast enough to prevent the minimum core DNBR from falling below the design limit (Reference 15.4.3-3); however an upper limit for the number of rods with a DNBR less than the limit value is determined in the radiological analysis.
15.4.3.2.3 Conclusions The results of the analysis of the Single Control Rod Withdrawal event demonstrate that Condition III acceptance criteria are met. The predicted MDNBR is less than the 95/95 safety limit. Less than 9% of the fuel is predicted to fail based on DNB criteria. No fuel is predicted to fail based on fuel centerline melt criteria. Enveloping, conservative fuel damage assumptions were used to determine radiological consequences. The radiological doses have been calculated based on a 9% assumed cladding failure. A summary of input parameters and assumptions is provided in Table 15.4.3-5. The radiological analysis results for this event are presented in Table 15.4.3-6a.
15.4.3.3    Statically Misaligned RCCA or Bank 15.4.3.3.1 Identification of causes and accident description The static misalignment events occur when a malfunction of the Control Rod Drive (CRD) mechanism causes a control rod to be out of alignment with its bank. Misalignment occurs when the rod is either higher or lower than any of the other control rods in the same bank.
During this event, the reactor is at steady-state rated full power conditions, and no excursion of core temperature, pressure, flow, or power occurs. For extreme RCCA misalignments, the core radial power distribution may be characterized by peaking factors in excess of design limits.
Highly localized increases in clad heat flux, coolant temperature, and flow diversions may occur.
In severe cases, the SAFDL on DNB may be approached.
This event is classified as a Condition II event as defined in Section 15.0.1.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 Misaligned assemblies are detected by:
: 1) Asymmetric power distribution as seen on out-of-core neutron detectors or core exit thermocouples.
: 2) Rod deviation alarm.
: 3) Rod position indicators.
The deviation alarm alerts the operator to rod-to-rod deviations within the same bank in excess of 12 steps. If the rod deviation alarm is not operable, the operator is required to take action as required by the Technical Specifications.
If one or more rod position indicator channels should be out of service, operating instructions shall be followed to assure the alignment of the non-indicated assemblies. The operator is also required to take action as required by the Technical Specifications. Indirect means of checking control rod position presented in plant operating instructions include review of temperature indications from core outlet thermocouples and review of data provided by movable in-core neutron detectors. Review of in-core detector data is specifically required following any significant movement of the non-indicated control rod assembly.
No single failure will prevent operation of any system required to function.
15.4.3.3.2 Analysis of effects and consequences Cycle operation with a statically misaligned RCCA or Bank could result in core power distributions which are significantly more peaked than predicted. Steady state power distributions are calculated in three dimensions for several misaligned cases. Full power operation with the most severe peaking at any core location resulting from undetected misalignments will be analyzed.
Key operating parameters used in the analysis of this event are the nominal rated-thermal power conditions used for the Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position accident presented in Table 15.4.7-1.
The VIPRE-01 code (Reference 15.4.1-5) is used to verify that DNB does not occur.
15.4.3.3.3 Conclusions The results of the analysis demonstrate that Condition II event acceptance criteria are met. The predicted MDNBR is greater than the safety limit. The critical heat flux correlation limit ensures that, with 95% probability and 95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. The fuel centerline melt threshold is not penetrated during this event.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.4.3.3.4 Deleted by Amendment No. 50 15.4.4    STARTUP OF AN INACTIVE REACTOR COOLANT PUMP AT AN INCORRECT TEMPERATURE 15.4.4.1    Identification of Causes and Accident Description If the plant is operating with one pump out of service, there is reverse flow through the inactive loop due to the pressure difference across the reactor vessel. The cold leg temperature in an inactive loop is identical to the cold leg temperature of the active loops (the reactor core inlet temperature). If the reactor is operated at power, and assuming the secondary side of the steam generator in the inactive loop is not isolated, there is a temperature drop across the steam generator in the inactive loop and, with the reverse flow, the hot leg temperature of the inactive loop is lower than the reactor core inlet temperature.
Administrative procedures require that the Unit be brought to a load of less than 25 percent of full power prior to starting the pump in an inactive loop in order to bring the inactive loop hot leg temperature closer to the core inlet temperature. Starting of an idle reactor coolant pump without bringing the inactive loop hot leg temperature close to the core inlet temperature would result in the injection of cold water into the core, which would cause a reactivity insertion and subsequent power increase.
This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Section 15.0.1.
Should the startup of an inactive reactor coolant pump accident occur, the transient will be terminated automatically by a reactor trip when the power range neutron flux (two out of four channels) exceeds the P-8 setpoint, which has been previously reset for two loop operation.
No single active failure will adversely affect the consequences of the event.
15.4.4.2    DELETED 15.4.4.3    DELETED 15.4.4.4    Event Disposition Modes 1 and 2: The event is incredible in these modes because the Plant Technical Specifications require that three reactor coolant pumps operate in Modes 1 and 2.
Modes 3 through 6: In these modes, the reactor is subcritical and there is no significant load on the plant. The potential for a significant reactivity excursion is nil. Even low levels of backflow through the inactive loop will preclude a static condition in which significant cooling of the inactive loop water inventory might occur, and the primary system will remain essentially isothermal. The consequences of the event in these modes are bounded by those of Event 15.4.1.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 15.4.5    A MALFUNCTION OR FAILURE OF THE FLOW CONTROLLER IN A BWR LOOP THAT RESULTS IN AN INCREASED REACTOR COOLANT FLOW RATE This section is not applicable to the Shearon Harris Nuclear Power Plant.
15.4.6    CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION THAT RESULTS IN A DECREASE IN THE BORON CONCENTRATION IN THE REACTOR COOLANT 15.4.6.1    Identification of Causes and Accident Description Reactivity can be added to the core by feeding primary grade water into the RCS via the reactor makeup portion of the Chemical and Volume Control System (CVCS), resulting in decreasing boron concentration in the reactor coolant system. The dilution of primary system boron adds positive reactivity to the core. This event can lead to an erosion of shutdown margin for subcritical initial conditions, or a slow power excursion for at-power conditions. A boron dilution for at-power conditions behaves in a similar manner to a slow Uncontrolled RCCA Bank Withdrawal event (Event 15.4.2).
Boron dilution is a manual operation under strict administrative controls with procedures calling for a limit on the rate and duration of dilution. A boric acid blend system is provided to permit the operator to match the boron concentration of reactor coolant makeup water during normal charging to that in the RCS. The CVCS is designed to limit, even under various postulated failure modes, the potential rate of dilution to a value which, after indication through alarms and instrumentation, provides the operator sufficient time to correct the situation in a safe and orderly manner.
The opening of the reactor makeup water control valve provides makeup to the RCS which can dilute the reactor coolant. Inadvertent dilution from this source can be readily terminated by closing the control valve. In order for makeup water to be added to the RCS at pressure, at least one charging pump must be running in addition to a reactor makeup water pump.
The rate of addition of unborated makeup water to the RCS is limited by operator response to alarm setpoints for reactor makeup water flow, charging flow, and letdown flow.
The boric acid from the boric acid tank is blended with primary grade water in the blender and the composition is determined by the preset flow rates of boric acid and primary grade water on the main control board.
In order to dilute two separate operations are required:
: 1) The operator must switch from the automatic makeup mode to the dilute mode.
: 2) The Stop-Start switch must be turned to the start position.
Omitting either step would prevent dilution.
Information on the status of the reactor coolant makeup is continuously available to the operator. Lights are provided on the main control board to indicate the operating condition of the pumps in the CVCS. Alarms are actuated to warn the operator if boric acid or demineralized water flow rates deviate from preset values as a result of system malfunction.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 This event is classified as an ANS Condition II incident (a fault of moderate frequency) as defined in Section 15.0.1. The acceptance criteria for this event are:
: 1) During cold shutdown (Mode 5):
If operator action is required to terminate the transient, a minimum time interval of 15 minutes must be available between the time when an alarm announces an unplanned moderator dilution and the time of loss of shutdown margin.
: 2) During hot shutdown, hot standby, startup, and power operation (Modes 4, 3, 2, and 1):
If operator action is required to terminate the transient, a minimum time interval of 15 minutes must be available from the time of initiation of the dilution to the time of loss of shutdown margin.
A boron dilution event can result from any of the following:
: 1) Resin sluice connections to the CVCS and BTRS demineralizers.
: 2) Reactor makeup water connection to the BTRS.
: 3) Pumping water of unknown boron concentration from the recycle holdup tanks to the charging pump suction.
: 4) Reactor makeup water connection to the boric acid batching tank(s).
: 5) Operation of the BTRS in the dilution mode.
: 6) Malfunction of the CVCS reactor makeup control system.
All of the above except items 5 and 6 require the opening of normally closed local manual valves, some of which are normally locked closed. Measures required to prevent a dilution from these sources during refueling are addressed in Section 15.4.6.2.
No dilution accident can originate from safeguard systems based on the premise that the technical specifications require periodic verification of the boron concentration in the RWST or the accumulators.
The containment spray system has been evaluated and determined not to be a boron dilution source during shutdown.
The effect of increasing boron worth with dilution was taken into account for all cases analyzed.
A conservatively low (most negative pcm/ppm) value was used.
For all boron dilution events analyzed, all fuel assemblies are in the core.
15.4.6.2    Analysis of Effects and Consequences
: 1) Method of Analysis - To cover all phases of the plant operation, boron dilution during refueling, startup, cold shutdown, hot shutdown, hot standby and power operation are Amendment 63                                                                          Page 89 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 considered in this analysis. Per the applicable methodology (Reference 15.4.2-3) an instantaneous mixing model is used for modes where at least one reactor coolant pump is operating.
: 2) Dilution During Refueling (Mode 6) - An uncontrolled boron dilution accident cannot occur during refueling as a result of a reactor coolant makeup system malfunction. This accident is prevented by administrative controls which isolate the RCS from the potential source of unborated water.
During refueling operations, potential boron dilution is prevented by administrative control of the valves listed on Table 15.4.6-2. Any makeup which is required during refueling will be from a borated water source.
: 3) Dilution During Cold Shutdown (Mode 5) - The boron dilution analysis during cold shutdown must ensure that the operator has at least 15 minutes after the actuation of the high flux at shutdown alarm to terminate the dilution before losing shutdown margin.
The shutdown margin requirement for cold shutdown is dependent on the reactor coolant system (RCS) temperature and cycle burnup, with a minimum requirement of 1000 pcm.
Two minimum RCS water volumes are assumed in the analysis. The first volume corresponds to the minimum RCS volume for reduced inventory operation with residual heat removal (RHR) cooling and a single train of CVCS. The second corresponds to the minimum RCS volume with at least one RCP in operation with a single train of RHR and a single train of CVCS.
The dilution flow rate is conservatively high corresponding to the maximum capacity of the reactor makeup water system with one makeup water pump locked out.
: 4) Dilution During Hot Shutdown (Mode 4) - The boron dilution analysis during hot shutdown must ensure that the operator has at least 15 minutes to terminate the dilution before losing shutdown margin. The shutdown margin requirement is dependent on the RCS temperature and cycle burnup, and whether a RCP is in operation or cooling is being provided by RHR. The shutdown margin requirement is at least 1770 pcm for all hot shutdown conditions.
Two minimum RCS water volumes are assumed in the analysis. The first volume corresponds to the minimum RCS volume with residual heat removal (RHR) cooling and a single train of CVCS. The second corresponds to the minimum RCS volume with at least one RCP in operation and a single train of CVCS.
The dilution flow rate is limited by the combined capacity of the two reactor makeup water pumps. This flow rate is conservatively high corresponding to the maximum capacity of the reactor makeup water system.
: 5) Dilution During Hot Standby (Mode 3) - The boron dilution analysis during hot standby must ensure that the operator has at least 15 minutes to terminate the dilution before losing shutdown margin. The shutdown margin requirement is temperature and exposure dependent, with a minimum requirement of 1770 pcm for all hot standby conditions.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 A minimum RCS water volume is assumed in the analysis. This is the active RCS volume, excluding the pressurizer and assuming at least one RCP in operation, one train of CVCS, and 3% SGTP. The dilution flow rate is limited by the combined capacity of the two reactor makeup water pumps. This flow rate is conservatively the maximum capacity of the reactor makeup water system.
: 6) Dilution During Startup (Mode 2) - The boron dilution during startup is analyzed to ensure that the operator has at least 15 minutes to terminate the dilution before losing shutdown margin. The shutdown margin requirement is 1770 pcm.
A minimum RCS water volume is assumed in the analysis. This is the active RCS volume, excluding the pressurizer and assuming at least one RCP in operation, one train of CVCS, and 3% SGTP.
The dilution flow rate is limited by the combined capacity of the two reactor makeup water pumps. This flow rate is conservatively the maximum capacity of the reactor makeup water system.
: 7) Dilution During Full Power Operation (Mode 1) - The boron dilution during startup is analyzed to ensure that the operator has at least 15 minutes to terminate the dilution before losing shutdown margin. The shutdown margin requirement is 1770 pcm. The analysis is performed with control rods in manual and control rods in automatic. While control rods are in manual, the alarm function is provided by the earliest reactor trip setpoint reached. With the reactor in automatic control, the rod insertion limit alarms provide the operator with adequate time to isolate the cause of the dilution and initiate reboration.
The dilution flow rate is limited by the combined capacity of the two reactor makeup water pumps. This flow rate is conservatively the maximum capacity of the reactor makeup water system.
15.4.6.3    Results and Conclusions The results of the boron dilution analysis show that the current Technical Specification shutdown margin requirements provide the operator with adequate time to manually terminate the source of dilution flow during operational Modes 1 to 5. No analysis was performed for a boron dilution event in Mode 6, since administrative controls are in place to prevent an uncontrolled boron dilution while the unit is in the refueling mode.
15.4.7  INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITION 15.4.7.1    Identification of Causes and Accident Description Fuel and core loading errors can arise from the inadvertent loading of one or more fuel assemblies into improper positions or the improper addition or removal of discrete burnable absorber rod assemblies, when applicable. These loading errors can result in severe changes in the core power distribution which may be undetectable by the incore instrumentation.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 The in-core system of moveable flux detectors which is used to verify power shapes at the start of life, is capable of revealing an assembly loading error which causes power shapes to be peaked in excess of the design value.
Fuel assembly loading errors are prevented by administrative procedures and controls implemented during fabrication. To reduce the probability of core loading errors, each fuel assembly is marked with an identification number and loaded in accordance with a core loading diagram. During core loading, the identification number will be checked before each assembly is moved into the core. After core loading is completed, a core map will be performed as a further check on proper placing of the fuel in the core.
The power distortion due to any combination of misplaced fuel assemblies would significantly raise peaking factors and would be potentially observable with in-core flux monitors. In addition to the flux monitors, thermocouples are located at the outlet of about one third of the fuel assemblies in the core. There is a high probability that these thermocouples would also indicate any abnormally high coolant enthalpy rise. In-core flux measurements are taken during the startup subsequent to every refueling operation.
This event is classified as an ANS Condition III incident (an infrequent fault) as defined in Section 15.0.1. The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110%
of design values.
: 2) A small fraction of fuel failures may occur, but these fuel failures should not hinder core coolability.
: 3) Radiological consequences should be less than 10% of 10 CFR 50.67 limits (i.e., 2.5 rem TEDE) offsite and less than 10 CFR 50.67 limit of 5 rem TEDE in the control room.
15.4.7.2      Analysis of Effects and Consequences Method of Analysis - Cycle operation with an improperly loaded core could result in core power distributions which are significantly more peaked than predicted. Steady state power distributions are calculated in three dimensions for a spectrum of postulated misloading events.
For each case analyzed, the assembly powers in instrumented core locations are compared to a normally loaded core to determine if the case would be detected at the time of the initial low-power flux map used in verifying that the core is properly loaded. Full power operation with the most severe peaking at any core location resulting from undetected misloadings will be analyzed.
Misloadings which exceed the criteria from the low-power flux map are detectable with the incore instrumentation system. Misloadings which are undetectable at the time of the low-power flux map are analyzed to ensure fuel failures in excess of allowed limits will not occur as a result of this event.
The misload assembly detection criteria are a function of the number of operable thimble locations during the initial low-power flux map. The misload detection criteria are based on combinations of acceptable differences in measured-to-predicted reaction rate (or instrumented assembly power) errors, core-wide (root-mean-square) errors for instrumented assemblies, Amendment 63                                                                        Page 92 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 differences in reaction rate (or power) in symmetric instrumented core locations, and differences in core quadrant tilt. A flux map which exceeds the misload detection criteria does not necessarily indicate a misload, but requires further examination to ensure power distribution limits will not be exceeded.
Key operating parameters used in the analysis of this event are presented in Table 15.4.7-1.
The VIPRE-01 code (Reference 15.4.1-5) is used to predict peaking that would preclude DNB from occurring.
15.4.7.3    Conclusions The results of the analysis demonstrate that Condition III event acceptance criteria are met.
Fuel failures due to DNB and fuel centerline melt criteria are less than the values assumed in the radiological dose analysis. Enveloping, conservative fuel damage assumptions were used to determine the predicted radiological consequences. The radiological doses have been calculated based on 9% assumed cladding failure. These are the same assumptions as were made in Section 15.4.3.2.3. The radiological doses are the same as those referenced in Section 15.4.3.2.3.
15.4.8    SPECTRUM OF ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENTS 15.4.8.1    Identification of Causes and Accident Description This accident is defined as the mechanical failure of a control rod mechanism pressure housing resulting in the ejection of a RCCA and drive shaft. The consequence of this mechanical failure is a rapid positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage. The transient is terminated by the Doppler reactivity effects of increased fuel temperature, and by an automatic reactor trip. This event challenges deposited enthalpy, radiological consequences and pressurization acceptance criteria.
15.4.8.1.1 Design precautions and protection Certain features in the SHNPP pressurized water reactors are intended to preclude the possibility of a rod ejection accident, or to limit the consequences if the accident were to occur.
These include a sound, conservative mechanical design of the rod housings, together with a thorough quality control (testing) program during assembly, and a nuclear design which lessens the potential ejection worth of RCCA's, and minimizes the number of assemblies inserted at high power levels.
Mechanical Design - The mechanical design is discussed in Section 4.6. Mechanical design and quality control procedures intended to preclude the possibility of a RCCA drive mechanism housing failure are listed below:
: 1) Each full length control rod drive mechanism housing is completely assembled and shop tested at 3107 psig.
: 2) The latch mechanism housings are attached to the closure head penetration nozzles and are hydrotested with the completed closure head.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 3) Stress levels in the mechanism are not affected by anticipated system transients at power, or by the thermal movement of the reactor coolant loops. Moments induced by the design earthquake can be accepted within the allowable primary working stress range specified by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, for Class 1 components.
: 4) The latch mechanism housing and rod travel housing are each a single length of forged Type 304 stainless steel. This material exhibits excellent notch toughness at all temperatures which will be encountered.
: 5) The CRDM housing plug is an integral part of the rod travel housing.
A significant margin of strength in the elastic range together with the large energy absorption capability in the plastic range gives additional assurance that gross failure of the housing will not occur. The joints between the latch mechanism housing and rod travel housing, are threaded joints reinforced by canopy type seal welds. Administrative regulations require periodic inspections of these (and other) welds.
Nuclear Design - Even if a rupture of a RCCA drive mechanism housing is postulated, the operation utilizing chemical shim is such that the severity of an ejected RCCA is inherently limited. In general, the reactor is operated with the RCCA's inserted only far enough to permit load follow. Reactivity changes caused by core depletion and xenon transients are compensated by boron changes. Further, the location and grouping of control RCCA banks are selected during the nuclear design to lessen the severity of a RCCA ejection accident.
Therefore, should a RCCA be ejected from its normal position during full power operation, only a minor reactivity excursion, at worst, could be expected to occur.
However, it may be occasionally desirable to operate with larger than normal insertions. For this reason, a rod insertion limit is defined as a function of power level. Operation with the RCCA's above this limit guarantees adequate shutdown capability and acceptable power distribution. The position of all RCCAs is continuously indicated in the Control Room. An alarm will occur if a bank of RCCAs approaches its insertion limit or if one RCCA deviates from its bank. A low alarm requires boron addition by following normal procedures with the CVCS. A low-low alarm requires boron addition by following the emergency boration procedure.
Reactor Protection - The reactor protection in the event of a rod ejection accident is provided by high neutron flux trip (high and low setting) and high rate of neutron flux increase trip. If a rapid flux trip does not occur, further protection is provided by the OTT These protection functions are described in detail in Section 7.2.
Effects on Adjacent Housings - Disregarding the remote possibility of the occurrence of a RCCA mechanism housing failure, investigations have shown that failure of a housing due to either longitudinal or circumferential cracking would not cause damage to adjacent housings. The full length control rod drive mechanism is described in Section 3.9.4.
Effects of Rod Travel Housing Longitudinal Failures - If a longitudinal failure of the rod travel housing should occur, the region of the position indicator assembly opposite the break would be stressed by the reactor coolant pressure of 2250 psia. The most probable leakage path would be provided by the radial deformation of the position indicator coil assembly, resulting in the Amendment 63                                                                        Page 94 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 growth of axial flow passages between the rod travel housing and the hollow tube along which the coil assemblies are mounted.
If failure of the position indicator coil assembly should occur, the resulting free radial jet from the failed housing could cause it to bend and contact adjacent rod housings. If the adjacent housings were on the periphery, they might bend outward from their bases. The housing material is quite ductile; plastic hinging without cracking would be expected. Housings adjacent to a failed housing, in locations other than the periphery, would not be bent because of the rigidity of multiple adjacent housings.
Effect of Rod Travel Housing Circumferential Failures - If circumferential failure of a rod travel housing should occur, the broken-off section of the housing would be ejected vertically because the driving force is vertical and the position indicator coil assembly and the drive shaft would tend to guide the broken-off piece upwards during its travel. Travel is limited by the missile shield, thereby limiting the projectile acceleration. When the projectile reached the missile shield it would partially penetrate the shield and dissipate its kinetic energy. The water jet from the break would continue to push the broken-off piece against the missile shield.
If the broken-off piece of the rod travel housing were short enough to clear the break when fully ejected, it would rebound after impact with the missile shield. The top end plates of the position indicator coil assemblies would prevent the broken piece from directly hitting the rod travel housing of a second drive mechanism. Even if a direct hit by the rebounding piece were to occur, the low kinetic energy of the rebounding projectile would not cause significant damage.
Possible Consequences - From the above discussion, the probability of damage to an adjacent housing must be considered remote. However, even if damage is postulated, it would not be expected to lead to a more severe transient since RCCAs are inserted in the core in symmetric patterns, and control rods immediately adjacent to worst ejected rods are not in the core when the reactor is critical. Damage to an adjacent housing could, at worst, cause that RCCA not to fall on receiving a trip signal; however this is already taken into account in the analysis by assuming the limiting stuck rod in addition to the ejected rod.
Summary - The considerations given above lead to the conclusion that failure of a control rod housing, due either to longitudinal or circumferential cracking, would not cause damage to adjacent housings that would increase severity of the initial accident.
15.4.8.1.2 Limiting criteria This event is classified as an ANS Condition IV incident. See Section 15.0.1 for a discussion of ANS classifications. Due to the extremely low probability of a RCCA ejection accident, some fuel damage could be considered an acceptable consequence. The acceptance criteria for this event are:
: 1) To ensure core coolability, the peak radially averaged enthalpy must not exceed 230 cal/gm at any axial location in any fuel rod.
: 2) The peak fuel temperature must remain below melting conditions.
: 3) Offsite Radiological consequences should be less than ~25% of the 10 CFR 50.67 limits (i.e., 6.3 rem TEDE) and less than the 10 CFR 50.67 limit of 5 rem TEDE in the control Amendment 63                                                                          Page 95 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 room.
For HFP and part-power cases, cladding failure is evaluated via the DNB 95/95 safety limit. For HZP cases, cladding failure is evaluated via a high temperature cladding failure fuel pellet enthalpy limit of 150 cal/g or 170 cal/g, depending on rod internal pressure.
: 4) The maximum reactor pressure during any portion of the assumed excursion should be less than the value that will cause stresses to exceed the "Service Limit C" as defined in Section III of the ASME Boiler and Pressure Vessel Code (i.e., maximum reactor pressure should be less than 120% of design values).
The worst single failure is the failure of the NI detector with the highest indication of power.
15.4.8.2    Analysis of Effects and Consequences Method of Analysis This analysis is performed using the methodology in Reference 15.4.2-3. The analysis requires a suite of computer codes to verify the acceptance criteria on fuel pellet enthalpy, fuel temperature, and reactor pressure above are met, and to ensure the number of pins which fail due to DNB does not exceed the assumptions in the dose analysis. The radiological consequences themselves are analyzed separately as described in Section 15.4.8.3.
SIMULATE-3K (Reference 15.4.8-8) is used to calculate the core power level and nodal power distributions during the transient. This information is used by VIPRE-01 (Reference 15.4.1-5).
One calculation with VIPRE-01 evaluates the transient with respect to fuel temperature and fuel pellet enthalpy, and a separate VIPRE-01 calculation evaluates the transient with respect to DNB. If the SIMULATE-3K model does not predict a trip on high flux or high positive flux rate, the core power level from SIMULATE-3K is used as input to a RETRAN-3D (Reference 15.4.1-
: 3) analysis, which determines the intervention of other Reactor Protection System trips, such as OTT, OPT, and low pressurizer pressure.
The analysis is performed at a variety of power levels and burnups. Initial conditions for the analysis are shown in Table 15.4.8-1. The range of neutronics parameters for the limiting fuel temperature and enthalpy cases are shown in Table 15.4.8-3.
The ejected rod worths are calculated using the SIMULATE-3 code (Reference 15.4.8-10) and methodology described in Reference 15.4.8-9. No credit is taken for the power flattening effects of Doppler or moderator feedback in the calculation of ejected rod worths. The hot rod and hot pellet peaking factors are calculated using either the SIMULATE-3 or SIMULATE-3K code.
The peak primary pressure for this event is bounded by the Uncontrolled Bank Withdrawal from Subcritical or Low Power event presented in Section 15.4.1 and is therefore not analyzed here.
Results Fuel failure based on the 95/95 DNB safety limit is less than that assumed in the dose analyses.
No fuel is predicted to melt. Enveloping conservative fuel damage assumptions were used to determine the predicted radiological consequences. The radiological doses have been calculated based on the failure percentage provided in Table 15.4.8-5. The peak fuel pellet Amendment 63                                                                          Page 96 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 enthalpy is less than the 230 cal/g limit. In addition, no pins at HZP exceed the high temperature cladding failure fuel pellet enthalpy limit.
The limiting peak fuel temperature and peak fuel enthalpy case is the EOC 50% case. The sequence of events for this case is given in Table 15.4.8-4b. This case tripped on power range high positive neutron flux rate. The core power and total reactivity versus time for this case are shown in Figures 15.4.8-9 and 15.4.8-11.
15.4.8.3      Radiological Consequences of a Postulated Rod Ejection Accident It is assumed that a mechanical failure of a control rod mechanism pressure housing has occurred, resulting in the ejection of a rod cluster control assembly and drive shaft. As a result of the accident, some fuel clad damage is assumed to occur. Due to the pressure differential between the primary and secondary systems, radioactive reactor coolant is discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere through either the atmospheric relief valves or the main steam safety valves.
Iodine and alkali metals group activity is contained in the secondary coolant prior to the accident, and some of this activity is released to the atmosphere as a result of steaming the steam generators following the accident. Finally, radioactive reactor coolant is discharged to the containment via the spill from the opening in the reactor vessel head. A portion of this radioactivity is released through containment leakage to the environment.
The analysis of the rod ejection radiological consequences uses the analytical methods and assumptions outlined in RG 1.183, Appendix H. Separate calculations are performed to calculate the dose resulting from the release of activity to containment and subsequent leakage to the environment and the dose resulting from the leakage to activity to the secondary system and subsequent release to the environment. The total offsite and control room doses are the sum of the doses resulting from each of the postulated release paths and nuclides considered.
A summary of input parameters and assumptions is provided in Table 15.4.8-5. Additional clarification is provided as follows:
Source Term a) In determining the offsite doses following a rod ejection accident, it is assumed that 10%
of the fuel rods in the core suffer sufficient damage that all of their gap activity is released. Ten percent of the total core activity of iodine and noble gases and 12 percent of the total core activity for alkali metals are assumed to be in the fuel-cladding gap. In the calculation of activity releases from the failed/melted fuel the maximum radial peaking factor of 1.73 was applied. The core fission product inventory is given in Table 15.0.9-1.
b) For both the containment leakage release path and the primary to secondary leakage release path all noble gas and alkali metal activity contained in the failed fuel gap is available for release.
c) For the containment leakage release path all of the iodine contained in the failed fuel gap is available for release.
d) For the primary to secondary leakage release path all of the iodine contained in the failed fuel gap is available for release from the reactor coolant system.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 e) Prior to the accident the iodine activity concentration of the primary coolant is 1.0 Ci/gm of DE I-131. The iodine activity concentration in the primary coolant is given in Table 15.0.9-7. The noble gas and alkali metal activity concentration in the RCS at the time the accident occurs is based on a one percent fuel defect level. The noble gas and alkali metal activity concentration in the RCS is given in Table 15.0.9-2. The iodine activity of the secondary coolant at the time the rod ejection occurs is assumed to be equivalent to the Technical Specification limit of 0.1  Ci/gm of DE I-131. The iodine activity concentration of the secondary coolant is given in Table 15.0.9-7. The alkali metal activity concentration of the secondary coolant at the time the rod ejection occurs is assumed to be 10% of the primary side concentration.
f)  Iodine in containment is assumed to be 4.85% elemental, 0.15% organic and 95%
particulate.
g) Iodine released from the secondary system is assumed to be 97% elemental and 3%
organic.
Containment Release Pathway a) The containment is assumed to leak at the design leak rate of 0.1% per day for the first 24 hours of the accident and then to leak at half that rate (0.05% per day) for the remainder of the 30 day period following the accident considered in the analysis.
b) For the containment leakage pathway, no credit is taken for plateout onto containment surfaces or for containment spray operation which would remove airborne particulate and elemental iodine. Sedimentation of alkali metal particulate in containment is credited.
Primary to secondary Leakage Release Pathway a) When determining doses due to the primary to secondary steam generator tube leakage, all the iodine, alkali metals group and noble gas activity (from prior to the accident and resulting from the accident) is assumed to be in the primary coolant (and not in the containment). The primary to secondary tube leakage and steaming from the steam generators continue until the reactor coolant system pressure drops below the secondary pressure. A conservative time of 2 hours was used for this analysis, although analyses of the small break LOCA pressure transient have shown that this would occur well before that time. The rod ejection pressure transient is similar to that of a small break LOCA.
b) The amount of primary to secondary SG tube leakage is assumed to be 1 gpm total.
Although the primary to secondary pressure differential drops throughout the event, the constant flow rate is conservatively maintained.
c) An iodine partition factor in the SGs of 0.01 (curies iodine/gm steam) / (curies iodine/gm water) is used. This partition factor is also applied to alkali metals.
d) All noble gas activity carried over to the secondary side through SG tube leakage is assumed to be immediately released to the outside atmosphere.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.4.8.3.1 Offsite Doses The Offsite doses are calculated using the assumptions and equations in Section 15.0A.1.
15.4.8.3.2 Control Room Doses The control room assumptions are provided in Section 15.6.5.4.3 and Table 15.6.5-15.
In the rod ejection accident, the SI setpoint will be reached within 30 seconds from event initiation. The SI signal causes the control room HVAC to switch from the normal operation mode to the post-accident recirculation mode of operation. A 15-second delay for the control room to switch between normal and post-accident recirculation modes is modeled. An operator action switches the control room from the post-accident recirculation mode to the pressurization mode at 2 hours after event initiation.
15.4.8.3.3 Results The potential radiological consequences resulting from a rod ejection have been conservatively analyzed, using assumptions and models described in previous sections. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The radiological analysis results for this event are listed in Table 15.4.8-6. The resultant doses are within the guideline values. The doses reported in Table 15.4.8-6 are for the combined release pathways of containment leakage and steaming from the secondary system for 2 hours. These doses are bounding for the postulated situation in which the steaming from the secondary system is the only release pathway even though in that situation steaming would continue for 8 hours (as described in Reg Guide 1.183, Appendix H, Section 7).
15.4.8.4    Conclusions The results of the analysis demonstrate that the event acceptance criteria are met. The DNB fuel failure percentages is less than that assumed in the radiological dose analyses. No fuel is predicted to melt. The peak fuel pellet enthalpy is less than the 230 cal/g limit.
15.4.9    SPECTRUM OF ROD DROP ACCIDENTS IN A BWR This section is not applicable to the Shearon Harris Nuclear Power Plant.
 
==REFERENCES:==
SECTION 15.4 15.4.1-1        Deleted by Amendment No. 45.
15.4.1-2        Deleted by Amendment No. 63.
15.4.1-3        EPRI NP-7450(A), Revision 10, RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, September 2014.
15.4.1-4        DPC-NE-3008, Revision 0, Thermal-Hydraulic Models for Transient Analysis, April 2018.
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Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 15.4.1-5      EPRI NP-2511-CCM-A, Revision 4, VIPRE A Thermal-Hydraulic Code for Reactor Cores, June 2007.
15.4.1-6      EMF-92-153-PA, Revision 1, HTTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, January 2005.
15.4.2-1      Deleted by Amendment No. 51 15.4.2-2      Deleted by Amendment No. 45.
15.4.2-3      DPC-NE-3009, Revision 0, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, April 2018.
15.4.3-1      Deleted by Amendment No. 63..
15.4.3-2      Deleted by Amendment No. 45.
15.4.3-3      DPC-NE-2005-PA, Revision 5, Thermal-Hydraulic Statistical Core Design Methodology, March 2016.
15.4.4-1      Deleted by Amendment No. 51 15.4.4-2      Deleted by Amendment No. 51 15.4.4-3      Deleted by Amendment No. 51 15.4.4-4      Deleted by Amendment No. 51 15.4.6-1      Deleted by Amendment No. 63.
15.4.8-1      Deleted by Amendment No. 45.
15.4.8-2      Deleted by Amendment No. 45.
15.4.8-3      Deleted by Amendment No. 45.
15.4.8-4      Deleted by Amendment No. 45.
15.4.8-5      Deleted by Amendment No. 45.
15.4.8-6      Deleted by Amendment No. 45.
15.4.8-7      Deleted by Amendment No. 45.
15.4.8-8      Studsvik Scandpower, SIMULATE-3K Models and Methodology, SSP-98/13, Revision 6, January 2009.
15.4.8-9      DPC-NF-2010-A, Revision 3, Nuclear Physics Methodology for Reload Design, May 2017.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.4.8-10      STUDSVIK/SOA-95/18, Revision 0, SIMULATE-3 Methodology Advanced Three-Dimensional Two-Group Reactor Analysis Code, 1995.
15.5    INCREASE IN REACTOR COOLANT INVENTORY Discussion and analysis of the following events are presented in this section:
: 1) Inadvertent operation of the Emergency Core Cooling System (IOECCS) during power operation (ANS Condition II event).
: 2) Chemical and Volume Control system malfunction that increases reactor coolant inventory (ANS Condition II event).
15.5.1    INADVERTENT OPERATION OF THE EMERGENCY CORE COOLING SYSTEM DURING POWER OPERATION 15.5.1.1    Identification of Causes and Accident Description Spurious Emergency Core Cooling System (ECCS) operation at power could be caused by operator error or a false electrical actuation signal. A spurious signal may originate from any of the safety injection system actuation channels as described in Section 7.3.
Following the actuation signal, the suction of the charging pumps is re-aligned to the Refueling Water Storage Tank (RWST) from the Volume Control Tank (VCT). The valves isolating the Boron Injection Tank (BIT) from the charging pumps and the valves isolating the BIT from the injection header automatically open. The charging pumps then force concentrated boric acid from the RWST into the Reactor Coolant System (RCS). If a reactor trip does not occur coincident with safety injection actuation, the turbine throttle valves will open to offset the addition of negative reactivity from the Safety Injection System (SIS). The transient is eventually terminated by the reactor protection system due to low pressurizer pressure or manual trip. The time to trip is affected by the initial operating conditions including core burnup history that affects boron concentration, rate of change of boron concentration, and Doppler and moderator coefficients.
The operator will determine if Safety Injection should be terminated. For spurious occurrence, the operator would stop the safety injection after ensuring satisfactory plant conditions per operating procedures and maintain the plant in hot standby conditions.
15.5.1.2    Description of Analysis Per Reference 15.5.1-5, the minimum departure from nucleate boiling ratio (MDNBR) for this event is bounded by the inadvertent opening of a pressurizer relief or safety valve (Event 15.6.1). An inadvertent actuation of the ECCS results in a pressure decrease similar to, but less severe than, an inadvertent opening of a pressurizer relief or safety valve. Neither event involves a reduction in the RCS flow rate since the reactor coolant pumps are not tripped.
Therefore, a quantitative core cooling capability analysis is not required.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 Pressurizer Overfill Case: The intent of this case is to examine the fluid thermal-hydraulic conditions at the inlet to the pressurizer PORVs and SRVs. The event was biased to conservatively ensure that the fluid temperatures seen by the pressurizer valves will be as low as realistically achievable.
The transient response of the reactor system is calculated using the ANF-RELAP computer program (Reference 15.5.1-4). The Reference 15.5.1-4 methodology contains the following safety evaluation report (SER) restriction that relates to this event:
The [ANF-RELAP] methodology cannot be used in situations where... the boron tracking model is needed without further justification. If... the steam line break boron tracking model is used with the [ANF-RELAP] methodology, then the applicability of the steam line break methods to the non-LOCA event under considerations should be justified.
The boron reactivity feedback model from the NRC approved MSLB methodology was conservatively implemented in this analysis. This model is justified as conservative for use in this application because the negative reactivity associated with the boron injection is delayed until the borated water is near the top of the core. Any positive feedback effects due to colder water entering the core is seen before the boron effects. Addition of boron in the pressurizer overfill cases does not impact the results.
15.5.1.3    Acceptance Criteria
: 1) The pressure in the reactor coolant and main steam systems should be maintained below 110% of design values.
: 2) Fuel cladding integrity shall be maintained by ensuring that the minimum DNBR remains above the 95/95 limit for PWRs.
: 3) An incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently.
15.5.1.4    Results An analysis was performed to support plant operation following steam generator replacement and power uprate (SGR/PUR). The analysis bounds plant operation for up to 2948 MWt rated power plus uncertainty (0.34% for a total analyzed power of 2958 MWt), and up to 3% steam generator tube plugging, and with RWST boron concentrations between the Technical Specification limits of 2400 ppm and 2600 ppm. The analysis was performed at RCS Tavg of 588.8&deg;F and bounds operation at a reduced RCS Tavg of 580.8&deg;F.
The results of the analysis are discussed with respect to each of the acceptance criteria as follows:
Acceptance Criteria 1 Per Reference 15.5.1-5, peak primary and secondary pressures are bounded by the turbine trip transient (Event 15.3.2).
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 Acceptance Criteria 2 Per Reference 15.5.1-5, the minimum DNBR is bounded by the inadvertent opening of a pressurizer relief or safety valve transient (Event 15.6.1).
Acceptance Criteria 3 An incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently. If the Reactor Protection System (RPS) initiates a reactor trip on a SIS signal, the plant would be brought to hot standby or cold shutdown condition after ensuring satisfactory plant conditions per operating procedures and Technical Specifications.
For this condition, the potential exists for the pressurizer to overfill from continued ECCS injection. The liquid flow capacity of the pressurizer PORVs and/or SRVs greatly exceeds the capacity of the high-head safety injection system; thus, RCS overpressurization is not a concern.
The pressurizer PORVs and SRVs were previously evaluated to determine if the valves remain operable during the discharge if subcooled water in accordance with NUREG 0737 II.D.1 as outlined in FSAR - TMI Appendix. The PORVs were shown to remain operable at inlet pressures of 2532 to 2545 psia and temperatures of 446 to 670&deg;F. The SRVs were shown to remain operable at an inlet pressure of 2475 psia and an inlet temperature of 635&deg;F.
An analysis of the IOECCS event was performed to evaluate the thermal hydraulic conditions at the inlet to the pressurizer SRVs and PORVs subsequent to SGR/PUR. The event analysis was biased to conservatively ensure that the fluid temperatures seen by the pressurizer PORVs and SRVs are as low as realistically achievable. The assumed state of plant systems and input biases for this case is given in the pressurizer overfill columns of Tables 15.5.1-1 and 15.5.1-2.
Two of the three pressurizer PORVs are safety-related. The associated pneumatic power and controls are designed to function by remote manual operation. The controls associated with manual operation of the valve are safety-related and the accumulator and piping leading from the accumulator to the valve operators are safety-related. The remaining portions of the pneumatic supply and the automatic actuation/control system are not safety-related but are very reliable for the following reasons:
: 1) Reference 15.5.1-2 concludes that the PORV circuitry meets the requirements of NUREG-0737, Item II.D.1 stating that "the PORVs were qualified under the pump and valve operability program (PVORT), the actuation transmitters are environmentally qualified, the cable is qualified (although not run as 1E), and the PIC cabinets are essentially the same hardware as the class 1E cabinets.
: 2) The PORVs have two diverse pneumatic supplies. One supply is the nitrogen system and the other is the instrument air system. The instrument air system 1A and 1B air compressors can be manually loaded onto the "A" and "B" train emergency diesel generators, respectively, in the event of a loss of offsite power.
The PORVs were assumed to initially open to increase the rate of pressurizer fill. The PORVs were then conservatively modeled as having exhausted their motive air supply. At this point, the PORVs are modeled in the closed position, in order to allow pressure increases that would potentially challenge the operation of the SRVs. The sequence of events and results of the Amendment 63                                                                          Page 103 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 analysis are given in Table 15.5.1-5. The responses of key system variables are given in Figures 15.5.1-15 through 15.5.1-21. Figure 15.5.1-15 shows that the pressurizer level is at 100% span at approximately 600 seconds (10 minutes) and the pressurizer is filled solid at 918.5 sec's. The liquid inlet pressures and temperatures for the pressurizer PORVs and SRVs remain above 2250 psia and 635&deg;F for approximately 950 seconds or 15.8 minutes (see Figures 15.5.1-20 and 15.5.1-21). The liquid temperature from 950 seconds to event termination at 1200 seconds remains above 564&deg;F. For additional discussion of valve operability, see Section 5.2.2.2.
15.5.1.5    Conclusions Consistent with the current licensing basis, the results demonstrate that the acceptance criteria for this event are met for SGR/Uprating conditions. The analysis for this event supports operation with the Model Delta 75 replacement steam generators at core power of 2948 MWt (with 0.34% uncertainty or a total power of 2958 MWt) with nominal primary Tavg at full power from 580.8&deg;F to 588.8&deg;F for steam generator tube plugging from 0% to 3%. The analysis bounds plant operation with RWST boron concentrations between the Technical Specification limits of 2400 ppm and 2600 ppm.
Conclusions with respect to each of the identified acceptance criteria are provided as follows:
Peak primary and secondary pressures are bounded by the turbine trip transient (Event 15.2.3).
Minimum DNBR is bounded by the inadvertent opening of a pressurizer relied or safety valve transient (Event 15.6.1).
The results of the pressurizer overfill analysis performed for SGR/PUR indicate that conditions at the inlet to the pressurizer SRVs are well within the range of conditions previously evaluated as acceptable for compliance with NUREG-0737 II.D.1 for almost 16 minutes after event initiation. Conditions remain well within those previously evaluated for the PORVs up to event termination at 20 minutes after event initiation. The PORVs are expected to mitigate the event, thereby eliminating the challenge to SRVs.
15.5.2    CHEMICAL AND VOLUME CONTROL SYSTEM MALFUNCTION THAT INCREASES REACTOR COOLANT INVENTORY An increase in reactor coolant inventory which results from the addition of cold, unborated water to the RCS is analyzed in Section 15.4.6. An increase in reactor coolant inventory which results from the injection of highly borated water into the RCS is analyzed in Section 15.5.1.
15.5.3    A NUMBER OF BWR TRANSIENTS This section is not applicable to the SHNPP.
 
==REFERENCES:==
SECTION 15.5 15.5.1-1        FANP Letter VNG:00:292, Revision 1, "Transmittal of the Letter Report and Calculation Notebook for the Evaluation of IOECCS Event for Harris at Uprated Conditions" dated October 27, 2000.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.5.1-2        Letter from NRC's Richard A. Becker to CP&L's Lynn Eury dated May 31, 1989 "Evaluation of Carolina Power and Light Company's Shearon Harris Unit 1, Plant Specific Submittals in Response to NUREG-0737, TMI Action Plan Requirement, Item II.D.1 (TAC No. 63565)."
15.5.1-3        EPRI NP-2628-SR, "EPRI PWR Safety and Relief Valve Test Program" dated December 1982.
15.5.1-4        ANF-89-151 (P)(A), ANF-RELAP Methodology for Pressurized Water Reactors:
Analysis of Non-LOCA Chapter 15 Events, May 1992.
15.5.1-5        DPC-NE-3009, Revision 0, FSAR / UFSRA Chapter 15 Transient Analysis Methodology, April 2018.
15.6    DECREASE IN REACTOR COOLANT INVENTORYf Events which result in a decrease in reactor coolant inventory as discussed in this section are as follows:
: 1) Inadvertent opening of a pressurizer safety or relief valve (ANS Condition II event).
: 2) Break in instrument line or other lines from reactor coolant pressure boundary that penetrate containment (ANS Condition II event).
: 3) Steam generator tube failure (ANS Condition IV event).
: 4) Loss of coolant accident (LOCA) resulting from a spectrum of postulated piping breaks within the reactor coolant pressure boundary (ANS Condition IV event for large-break LOCA and condition III event for small-break LOCA).
15.6.1      INADVERTENT OPENING OF A PRESSURIZER SAFETY OR POWER OPERATED RELIEF VALVE 15.6.1.1      Identification of Causes and Accident Description An accidental depressurization of the Reactor Coolant System (RCS) could occur as a result of an inadvertent opening of a pressurizer power operated relief or safety valve.
Since a safety valve is sized to relieve approximately twice the steam flow rate of a power operated relief valve, and will, therefore, allow a much more rapid depressurization upon opening, the most severe core conditions resulting from an accidental depressurization of the RCS are associated with an inadvertent opening of a pressurizer safety valve. Initially the event results in a rapidly decreasing RCS pressure until this pressure reaches a value corresponding to the hot leg saturation pressure. At this time, the pressure decrease is slowed considerably.
The pressure continues to decrease throughout the transient.
The reactor may be tripped by the following Reactor Protection System signals:
f Further information is contained in the TMI Appendix.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 1. Overtemperature T.
: 2. Pressurizer low pressure.
An inadvertent opening of a pressurizer safety valve is classified as an ANS Condition II event, a fault of moderate frequency. See Section 15.0.1 for a discussion of ANS Condition II events.
The acceptance criteria for this event are:
: 1) The pressures in the reactor coolant and main steam systems should be less than 110% of design values.
: 2) The fuel cladding integrity should be maintained by ensuring that fuel design limits are not exceeded. This is demonstrated by assuring that the minimum calculated departure from nucleate boiling ratio (DNBR) is not less than the applicable limits of the DNBR correlation being used and that fuel centerline melt does not occur.
: 3) The event should not generate a more serious plant condition without other faults occurring independently.
: 4) For transients of moderate frequency in combination with a single failure, no loss of function of any fission product barrier, other than fuel element cladding, shall occur. Core geometry is maintained in such a way that there is no loss of core cooling capability and control rod insertability is maintained.
Plant systems and equipment which are necessary to mitigate the effects of RCS depressurization caused by an inadvertent safety valve opening are discussed in Section 15.0.8 and listed in Table 15.0.8-1.
Normal reactor control systems are not required to function. The Reactor Protection System functions to trip the reactor on the appropriate signal. No single active failure will prevent the Reactor Protection System from functioning properly.
15.6.1.2    Analysis of Effects and Consequences Method of Analysis - This event is a depressurization of the primary coolant system. No pressurization criteria need therefore be addressed. The challenge to the SAFDLs results from the depressurization which occurs prior to reactor scram. The analysis utilized a steam flow rate from the pressurizer of 475,000 lbs/hr at 2500 psia. This is in excess of the rated flow rate for one pressurizer safety valve.
Conservative conditions established for the analysis of this event are presented in Table 15.6.1-
: 1. Available reactor protection system trips are presented in Table 15.0.6-2. Key operating parameters used in the analysis of this event are presented in Table 15.6.1-2. The range of neutronics parameters supported by this analysis are presented in Table 15.6.1-3.
The transient response of the reactor system is calculated using the RETRAN-3D (Reference 15.6.1-1) computer program following the method described in Reference 15.6.1-3 and Reference 15.6.1-4). The core thermal hydraulic boundary conditions from the RETRAN-3D calculation are used as input to the VIPRE-01 code (Reference 15.6.1-2) to predict the MDNBR Amendment 63                                                                        Page 106 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 for the event using the statistical core design methodology (Reference 15.6.1-5) and the HTP CHF correlation (Reference 15.6.1-6) for the Framatome 17x17 Advanced W-HTP fuel.
Results - The sequence of events for this analysis is given in Table 15.6.1-4. This transient tripped on low pressurizer pressure. The responses to key system variables are given in Figures 15.6.1-1 to 15.6.1-7.
15.6.1.3    Conclusions The results of the analysis demonstrate that the event acceptance criteria are met. The predicted MDNBR is greater than the safety limit. The critical heat flux correlation limit ensures that, with 95% probability and 95% confidence, DNB is not expected to occur; therefore, no fuel is expected to fail. This event does not pose a credible challenge to FCM safety criteria.
Both the maximum reactor primary system pressure and the maximum secondary system pressures are less than 110% of design pressure, 2750 psia and 1320 psia, respectively.
15.6.2    BREAK IN INSTRUMENT LINE OR OTHER LINE FROM REACTOR COOLANT PRESSURE BOUNDARY THAT PENETRATE CONTAINMENT 15.6.2.1    Identification of Causes and Frequency Classification The estimated frequency of a primary sample or instrument line rupture classifies it as a limiting fault incident. A primary sample or instrument line break provides a release path for reactor coolant outside Containment. The line break selected for analysis is the letdown line which penetrates the Containment. This is the largest penetration whose failure could result in an event in this category. This failure would result in larger releases than would be the case for the smaller instrument and sample lines.
This event is classified as an ANS Condition II event, a fault of moderate frequency. See Section 15.0.1 for a discussion of ANS Condition II events.
Plant systems and equipment which are necessary to mitigate the effects of the event are discussed in Section 15.0.8 and listed in Table 15.0.8-1. No single active failure in any of these systems or equipment will adversely affect the consequences of the event.
15.6.2.2    Sequence of Events and Systems Operation The integrity of lines containing primary coolant external to the Containment is significant radiologically since a rupture of this barrier results in the release of reactor coolant outside Containment. Following such a break, the RCS pressure decreases due to the loss of reactor coolant. When the pressurizer pressure has reached the low pressure setpoint, a reactor trip is initiated. A turbine trip follows a reactor trip and results in an increase in secondary side pressure to the steam generator safety valve set pressure. The safety injection signal (SIS) on low pressurizer pressure terminates the break flow by isolating the letdown line inside Containment. The reactor coolant inventory is replenished by the charging pumps. Operation of these pumps ensures that the core will not be uncovered and prevents any significant increase in clad temperatures.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 At about 30 minutes into the transient, the ruptured line is isolated, terminating the leak flow.
Prior to isolation of the line, reactor coolant has been released from the RCS to the Reactor Auxiliary Building at the rate of 200 gpm.
15.6.2.3    Analysis of Radiological Consequences 15.6.2.3.1 Design basis 15.6.2.3.1.1    Physical model A break in fluid-bearing lines which penetrate the Containment could result in the release of radioactivity to the environment. There are no instrument lines connected to the RCS which penetrate the Containment. There are, however, other piping lines from the RCS to the Chemical and Volume Control System (CVCS) and the Process Sampling System which penetrate the Containment.
The most severe rupture with respect to radioactivity release during normal plant operation is the rupture of the letdown line outside Containment. For such a break, the reactor coolant letdown flow would have passed from the cold leg and through the regenerative heat exchanger.
15.6.2.3.1.2    Assumptions and Parameters The analysis of the small line break outside of containment (SLBOC) radiological consequences uses the analytical methods and assumptions outlined in SRP 15.6.2 since this accident is not discussed in RG 1.183. The major assumptions and parameters used in the analysis are listed in Table 15.6.2-1 and discussed below:
a) The SRP indicates that accident-initiated iodine spiking be modeled. The accident-initiated iodine spike increases the iodine release rate from the fuel to the RCS to a value 500 times greater than the release rate corresponding to a maximum equilibrium RCS concentration of 1.0 Ci/gm of DE I-131. The iodine spike appearance rates are given in Table 15.0.9-6.
b) The noble gas activity concentration in the RCS at the time the accident occurs is based on a one-percent fuel defect level. This is approximately equal to the Technical Specification value of 100/E bar Ci/gm for gross radioactivity. The noble gas concentrations in the RCS are given in Table 15.0.9-2.
c) The transfer of the primary coolant to the environment through the letdown line break is 200 gpm until 30 minutes. The iodine flashing factor for the released letdown flow is 0.4.
Therefore of the iodine contained in the water released in the letdown line break, only 40%
of the iodine is released to the auxiliary building atmosphere and of that all is released to the environment.
15.6.2.3.1.3    Offsite Doses The offsite doses are calculated using the assumptions and equations in Section 15.0A.1.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.6.2.3.1.4      Control Room Doses The control room assumptions are provided in Section 15.6.5.4.3 and Table 15.6.5-15.
It is assumed that the control room HVAC system begins in normal mode. The activity level in the intake duct causes a high radiation signal almost immediately. It is conservatively assumed that the post-accident recirculation control room HVAC mode is entered 15 seconds after event initiation. The control room is assumed to be placed in pressurization mode at 2 hours after isolation signal.
15.6.2.3.1.5      Results The potential radiological consequences resulting from a letdown line break outside of containment have been conservatively analyzed, using assumptions and models described in previous sections. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The radiological analysis results for this event are listed in Table 15.6.2-2. The resultant doses are within the guideline values. The calculated offsite doses are less than 10% of the 10CFR50.67 limits (i.e., 2.5 rem TEDE) and the calculated control room doses are less than the 10CFR50.67 limit of 5 rem TEDE.
15.6.3      STEAM GENERATOR TUBE RUPTURE A steam generator tube rupture (SGTR) results in the leakage of contaminated reactor coolant into the secondary system and the subsequent release of a portion of that activity to the atmosphere. Therefore, the analysis must demonstrate that the offsite radiological consequences resulting from a SGTR are within allowable guidelines. One of the primary assumptions of the offsite dose assessment methodology is that the steam generator with the leaking tube can eventually be isolated from the release to the atmosphere. Therefore, it is essential to demonstrate that the secondary side of the ruptured Steam Generator does not overfill from water entering the secondary side of the steam generator from primary coolant flow through the broken tube and AFW flow delivered to the Steam Generator. Such an overfill condition in the Steam Generator could force water discharge through the Main Steam PORV's or Safeties potentially causing damage that could prevent their subsequent closure to isolate the release path.
The SGTR analysis then becomes a two-step process. The first part of the analysis is to ensure that overfill of the steam generator does not occur. Having demonstrated that Margin to Overfill exists, the Offsite Dose Assessment can be performed using the standard assumptions and methodology. Since the Margin to Overfill case and the Offsite Dose analysis have different sensitivities to plant parameters, two separate analyses must be performed to ensure that conservative scenarios are analyzed for each case.
Carolina Power & Light Company (CP&L) participated as a member of the Westinghouse Owners' Group (WOG) Steam Generator Tube Rupture (SGTR) Subgroup, which was responsible for the production of WCAP-10698 P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," (December 1984) and Supplement 1 to WCAP-10698-P-A, "Evaluation of Off Site Radiation Doses for a Steam Generator Tube Rupture Accident,"
(May 1985). Carolina Power & Light Company has used the methodology documents in these reports as a basis for the plant specific analysis of an SGTR. The NRC staff provided their Amendment 63                                                                      Page 109 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 assessment of this methodology and approval for referencing in plant specific submittals in a {{letter dated|date=March 30, 1987|text=March 30, 1987 letter}} from Mr. C. E. Rossi to Mr. A. E. Ladeiu, Chairman, SGTR Subgroup.
The original design basis steam generator tube rupture (SGTR) analysis of record for the Shearon Harris Nuclear Power plant (SHNPP) is presented in WCAP-12403 (Reference 15.6.3-2). The SHNPP SGTR analysis in WCAP-12403 included an analysis of the margin to steam generator overfill, as well as an analysis of the offsite radiation doses for a design basis SGTR.
The analysis results demonstrated that there was margin to steam generator overfill with the most limiting single failure with respect to overfill, and that the calculated offsite radiation doses would be acceptable assuming the most limiting single failure for the offsite dose evaluation.
Plant response to the event was modeled using the LOFTTR2 computer code with conservative assumptions of break size and location, condenser availability and initial secondary water mass in the ruptured steam generator. The analysis methodology includes the simulation of the operator actions for recovery from a steam generator tube rupture based on the SHNPP Emergency Operating Procedures (EOPs), which were developed from the Westinghouse Owners Group Emergency Response Guidelines (ERGs). The operator action times used in the simulation of the SGTR recovery actions for the SHNPP Analysis were based on the results from plant simulator studies.
The SGTR Event was subsequently reanalyzed to account for the recognition that the Main Steam PORV capacity was greater than assumed on the original analysis of record (WCAP-12403). The reanalysis was used as an opportunity to incorporate a few other existing conditions and to investigate the potential for possible future modifications. This reanalysis uses identical methodologies and operator action times as the original analysis of WCAP-12403 but it considers revised or expanded ranges for the AFW System performance and the SI system performance as well as the Main Steam PORV capacity. This revised analysis (WCAP-12403 Supplement 1/Ref. 15.6.3-7) was the analysis of record for SHNPP as amended by Reference 15.6.3-8, prior to installation of the model 75 replacement steam generators.
In support of the Shearon Harris Nuclear Power Plant (SHNPP) model 75 replacement steam generator program, an evaluation for a design basis steam generator tube rupture (SGTR) event has been performed to demonstrate that the potential consequences are acceptable. The evaluation discussed herein considers operation with a full power average temperature (Tavg) of 588.8&deg;F and assumes that up to 10% of the steam generator tubes are plugged. The analysis supports a main feedwater temperature window of 375&deg;F to 440&deg;F. Operation at the rated power of 2948 MWt with 0.34% uncertainty (analyzed core power of 2958 MWt) was also considered. As noted in Tables 15.0.3-5 and 15.0.6-2, values used in analysis of SGTR for Reactor Protection System time constants and Auxiliary Feedwater delivery are slightly different for this particular event. This analysis is documented in WCAP-14778 (Reference 15.6.3-9).
Reference 15.6.3-9 presents the margin to overfill analysis.
The WCAP-14778 analysis was subsequently amended to address the decay heat issue presented in NSAL-07-11, (Reference 15.6.3-12). In addition to decay heat, safety injection and auxiliary feedwater temperatures were evaluated and single failure issues were addressed. The results of the amended SGTR Margin to Overfill analysis are presented in this FSAR section.
The dose analysis presented in the revised licensing submittal for Alternate Source Term implementation (Reference 15.6.3-13) contains the Regulatory Guide 1.183 results for the SGTR event. The thermal and hydraulic analysis in Reference 15.6.3-9 was used as input to Amendment 63                                                                          Page 110 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 the Regulatory Guide 1.183 dose calculations. Both of these analyses will be presented separately in this FSAR section based on credit for a measurement uncertainty recapture (MUR) power uprate being applied. The MUR uprate increased NSSS power to 2970.4 MWt.
15.6.3.1    Identification of Cause and Accident Description The accident examined is the complete severance of a single steam generator tube. The accident is assumed to take place at power with the reactor coolant contaminated with fission products corresponding to continuous operation with a limited number of defective fuel rods.
The accident leads to an increase in contamination of the secondary system due to leakage of radioactive coolant from the RCS. In the event of a coincident loss of offsite power, or failure of the Condenser Steam Dump System, discharge of radioactivity to the atmosphere takes place via the steam generator power operated relief valves (and safety valves if their setpoint is reached).
In view of the fact that the steam generator tube material is Inconel 690 and is a highly ductile material, it is considered that the assumption of a complete severance is somewhat conservative. The more probable mode of tube failure would be one or more minor leaks of undetermined origin. Activity in the steam and power conversion system is subject to continual surveillance and an accumulation of minor leaks which exceed the limits established in the Technical Specifications is not permitted during the Unit operation.
Due to a series of alarms as described below, the operator will readily determine that a steam generator tube rupture has occurred, identify and isolate the ruptured steam generator, and complete the required recovery actions to stabilize the plant and terminate the primary to secondary break flow. The recovery procedure can be completed on a time scale which ensures that break flow to the secondary system is terminated before water level in the affected steam generator rises into the main steam pipe. Sufficient indications and controls are provided to enable the operator to carry out these functions satisfactorily.
Assuming normal operation of the various plant control systems, the following sequence of events is initiated by a tube rupture:
: 1) Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side there is a steam flow/feedwater flow mismatch before trip as feedwater flow to the affected steam generator is reduced due to the break flow which is now being supplied to that steam generator from the primary side.
: 2) The condenser vacuum pump effluent radiation monitor, steam generator blowdown line radiation monitor, and main steamline radiation monitor will alarm, indicating a sharp increase in radioactivity in the secondary system.
: 3) Continued loss of reactor coolant inventory leads to a reactor trip signal generated by low pressurizer pressure or by overtemperature T. Resultant plant cooldown following reactor trip leads to a rapid decrease in RCS pressure and pressurizer level. A safety injection signal, initiated by low pressurizer pressure, follows soon after the reactor trip. The safety injection signal automatically terminates steam generator blowdown, normal feedwater supply and initiates auxiliary feedwater addition via the motor-driven AFW pumps. If the Amendment 63                                                                        Page 111 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 steam generator level decreases below the low-low level setpoint in two of the three steam generators or a loss of offsite power occurs, the turbine-driven pump will also be started.
: 4) The reactor trip automatically trips the turbine and if offsite power is available the steam dump valves open permitting steam dump to the condenser. In the event of coincident loss of offsite power, the steam dump valves would automatically close to protect the condenser.
The steam generator pressure would rapidly increase resulting in steam discharge to the atmosphere through the steam generator power-operated relief valves (and safety valves if their setpoint is reached).
: 5) Following reactor trip and safety injection actuation, the continued action of auxiliary feedwater supply and borated safety injection flow (supplied from the refueling water storage tank) provide a heat sink which absorbs some of the decay heat. This reduces the amount of steam bypass to the condenser, or in the case of loss of offsite power, steam relief to atmosphere.
: 6) Safety injection flow results in stabilization of RCS pressure and pressurizer water level, and the RCS pressure trends towards an equilibrium value where the safety injection flow rate equals the break flow rate.
In the event of an SGTR, the plant operators must diagnose the SGTR and perform the required recovery actions to stabilize the plant and terminate the primary to secondary leakage. The operator actions for SGTR recovery are provided in the plant Emergency Operating Procedures.
The major operator actions include identification and isolation of the ruptured steam generator, cooldown and depressurization of the RCS to restore inventory, and termination of SI to stop primary to secondary leakage. These operators actions are described below.
: 1) Identify the ruptured steam generator.
High secondary side activity, as indicated by the condenser vacuum pump effluent radiation monitor, steam generator blowdown line radiation monitor, or main steamline radiation monitor, typically will provide the first indication of an SGTR event. The ruptured steam generator can be identified by a mismatch between steam and feedwater flow, high activity in a steam generator water sample, or a high radiation indication on the corresponding main steamline radiation monitor. For an SGTR that results in a reactor trip at high power, the steam generator water level will decrease to near the bottom of the narrow range scale for all of the steam generators.
The AFW flow will begin to refill the steam generators, distributing flow to each of the steam generators. Since primary to secondary leakage adds additional liquid inventory to the ruptured steam generator, the water level will increase more rapidly in that steam generator. This response, as displayed by the steam generator water level instrumentation, provides confirmation of an SGTR event and also identifies the ruptured steam generator.
: 2) Isolate the ruptured steam generator from the intact steam generators and isolate feedwater to the ruptured steam generator.
Once a tube rupture has been identified, recovery actions begin by isolating steam flow from and stopping feedwater flow to the ruptured steam generator. In addition to minimizing radiological releases, this also reduces the possibility of overfilling the ruptured steam generator with water by 1) minimizing the accumulation of feedwater flow and 2) enabling the operator to Amendment 63                                                                        Page 112 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 establish a pressure differential between the ruptured and intact steam generators as a necessary step toward terminating primary to secondary leakage.
: 3) Cool down the RCS using the intact steam generators.
After isolation of the ruptured steam generator, the RCS is cooled as rapidly as possible to less than the saturation temperature corresponding to the ruptured steam generator pressure by dumping steam from only the intact steam generators. This ensures adequate subcooling will exist in the RCS after depressurization of the RCS to the ruptured steam generator pressure in subsequent actions. If offsite power is available, the normal steam dump system to the condenser can be used to perform this cooldown. However, if offsite power is lost, the RCS is cooled using the power operated relief valves (PORVs) on the intact steam generators.
: 4) Depressurize the RCS to terminate primary to secondary leakage.
When the cooldown is completed, SI flow will increase RCS pressure until break flow matches SI flow. Consequently, SI flow must be terminated to stop primary to secondary leakage.
However, adequate reactor coolant inventory must first be assured. This includes both sufficient reactor coolant subcooling and pressurizer inventory to maintain a reliable pressurizer level indication after SI flow is stopped. Since leakage from the primary side will continue after SI flow is stopped until RCS and ruptured steam generator pressures equalize, an "excess" amount of inventory is needed to ensure pressurizer level remains on span. The "excess" amount required depends on RCS pressure and reduces to zero when RCS pressure equals the pressure in the ruptured steam generator.
The RCS depressurization is performed using normal pressurizer spray if the reactor coolant pumps (RCPs) are running. However, if offsite power is lost or the RCPs are not running for some other reason, normal pressurizer spray is not available. In this event, RCS depressurization can be performed using the pressurizer PORVs or auxiliary pressurizer spray.
: 5) Terminate SI to stop primary to secondary leakage.
The previous actions will have established adequate RCS subcooling, secondary side heat sink and sufficient reactor coolant inventory to ensure that SI flow is no longer needed. When these actions have been completed, SI flow must be stopped to terminate primary to secondary leakage. Primary to secondary leakage will continue after SI flow is stopped until RCS and ruptured steam generator pressures equalize. Charging flow, letdown, and pressurizer heaters will then be controlled to prevent repressurization of the RCS and reinitiation of leakage into the ruptured steam generator.
Following SI termination, the plant conditions will be stabilized, the primary to secondary break flow will be terminated, and all immediate safety concerns will have been addressed. At this time, a series of operator actions are performed to prepare the plant for cooldown to cold shutdown conditions. Subsequently, actions are performed to cool down and depressurize the RCS to cold shutdown conditions and to depressurize the ruptured steam generator.
This event is classified as an ANS Condition IV event, a limiting fault (postulated accident). See Section 15.0.1 for a discussion of ANS Condition IV events.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 Plant systems and equipment which are necessary to mitigate the effects of the event are discussed in Section 15.0.8 and listed in Table 15.0.8-1. No single active failure in any of these systems or equipment will adversely affect the consequences of the event. The analyses presented here consider the limiting single failure.
15.6.3.2    Westinghouse Thermal Hydraulic Analysis Steam Generator Overfill. One of the primary assumptions of the offsite dose assessment methodology is that the steam generator with the leaking tube can eventually be isolated from relieving to the atmosphere. Therefore, it is essential to demonstrate that the secondary side of the ruptured steam generator does not overfill and potentially damage the Main Steam PORV's or Safeties by forcing water discharge through them. Therefore, the first part of the analysis is to ensure that overfill of the steam generator does not occur. Having demonstrated that margin to overfill exists, the Offsite Dose Assessment can be performed using the standard assumptions and methodology. Since the Margin to Overfill case and the Offsite Dose analysis have different sensitivities to plant parameters, two separate analyses must be performed to ensure that conservative scenarios are analyzed for each case. The Overfill Analysis is presented in this section.
15.6.3.2.1 Analysis assumptions A detailed discussion of the Design Basis Accident, along with the SHNPP-specific inputs and Conservative Assumptions, is provided in the base analysis presented in WCAP-14778. The analysis is further revised to address the decay heat issue presented in NSAL-07-11 (Reference 15.6.3-12) and to address potential single failures. The following discussion summarizes those assumptions.
The accident modeled is a double ended break of one steam generator tube located at the top of the tube sheet on the outlet (cold leg) side of the steam generator. It was also assumed that loss of offsite power occurs at the time of reactor trip. The limiting single failure with respect to steam generator overfill used for the analysis is a failure within the turbine-driven AFW Pump speed controller, resulting in steady-state pump operation at the upper end of the speed control band (4100 rpm) and increased delivery to the ruptured steam generator.
Most of the conservative assumptions and initial conditions used in this analysis are discussed in detail in WCAP-14778 (Reference 15.6.3-9). These assumptions include simulation of turbine runback based on the calculated reactor trip time, a conservatively high initial steam generator secondary mass, and initiation of AFW flow from both motor-driven AFW pumps and the turbine driven AFW pump immediately after reactor trip.
Key assumptions include:
: 1) A maximum AFW flowrate of 750 gpm to the ruptured steam generator, until isolation, for the overfill analysis.
: 2) Steam generator PORV capacity of 795,000 lbm/hr per valve at 1200 psig (both MSPORVs are credited in the overfill analysis).
: 3) Safety Injection flow from two high head safety injection pumps.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 To address NSAL-07-11 (Reference 15.6.3-12) and single failure effects, the following were modeled:
: 4) Low decay heat based on the 1979 ANS model -2 uncertainty.
: 5) The minimum AFW enthalpy of 8.0 Btu/lbm (based on 40&deg;F).
: 6) A failure within the turbine-driven AFW Pump speed controller, resulting in steady-state pump operation at the upper end of the speed control band (4100 rpm).
The major operator actions for SGTR recovery provided in the SHNPP Plant Operating Manual, Procedure Number EOP-E-3 (ERG E-3) were explicitly modeled in this analysis. The operator actions modeled include (1) identification and isolation of the ruptured steam generator, (2) cooldown of the RCS using the PORVs on the intact steam generators to ensure subcooling, (3) depressurization of the RCS using a pressurizer PORV to restore inventory, and (4) termination of SI to stop primary to secondary leakage. The operator action times used for the analysis are presented in Table 15.6.3-1.
15.6.3.2.2 Transient description The analysis results for the SHNPP margin to overfill analysis are described below. The sequence of events for this transient is presented in Table 15.6.3-2.
Following the tube rupture, reactor coolant flows from the primary into the secondary side of the ruptured steam generator since the primary pressure is greater than the steam generator pressure. In response to this loss of reactor coolant, pressurizer level decreases as shown in Figure 15.6.3-1. The RCS pressure also decreases as shown in Figure 15.6.3-2 as the steam bubble in the pressurizer expands. As the RCS pressure decreases due to the continued primary to secondary leakage, automatic reactor trip occurs at approximately 112 seconds on an overtemperature -T trip signal.
After reactor trip, core power rapidly decreases to decay heat levels. The turbine stop valves close and steam flow to the turbine is terminated. The steam dump system is designed to actuate following reactor trip to limit the increase in secondary pressure, but the steam dump valves remain closed due to the loss of condenser vacuum resulting from the assumed loss of offsite power at the time of reactor trip. Thus, the energy transfer from the primary system causes the secondary side pressure to increase rapidly after reactor trip until the steam generator PORVs (and safety valves if their setpoints are reached) lift to dissipate the energy, as shown in Figure 15.6.3-3. After the plant sensible heat is dissipated, the steaming rate decreases to the level required to remove the core decay heat and the secondary pressure is maintained near the PORV setpoint of 1121 psia. The main feedwater flow will be terminated and AFW flow will be automatically initiated following the loss of offsite power assumed at the time of reactor trip.
The RCS pressure decreases more rapidly after reactor trip as energy transfer to the secondary shrinks the reactor coolant and the tube rupture break flow continues to deplete primary inventory. Pressurizer level also decreases more rapidly following reactor trip. The decrease in RCS inventory results in a low pressurizer pressure SI signal. After SI actuation, the SI flow rate initially exceeds the tube rupture break flow rate and the pressurizer level begins to Amendment 63                                                                      Page 115 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 increase. This also results in an increase in the RCS pressure which trends toward the equilibrium value where the SI flow rate equals the break flow rate.
Since offsite power is assumed lost at reactor trip, the RCPs trip and a gradual transition to natural circulation flow occurs. Immediately following reactor trip the temperature differential across the core decreases as core power decays (see Figure 15.6.3-4); however, the temperature differential subsequently increases as the reactor coolant pumps coast down and natural circulation flow develops. The cold leg temperatures trend toward the steam generator temperature as the fluid residence time in the tube region increases. The hot leg temperature reaches a peak and then slowly decreases, as steady state conditions are reached, until operator actions are initiated to cool down the RCS.
15.6.3.2.3 Major operator actions
: 1) Identify and Isolate the Ruptured Steam Generator. Once a tube rupture has been identified, recovery actions begin by isolating steam flow from the ruptured steam generator and isolating the auxiliary feedwater flow to the ruptured steam generator. The ruptured steam generator is assumed to be identified and AFW flow to the ruptured steam generator isolated when the narrow range level reaches 30% on the ruptured steam generator or at 8.8 minutes after initiation of the SGTR, whichever is longer. For the Shearon Harris analysis, the time to reach 30% is less than 8.8 minutes; therefore, isolation of the AFW to the ruptured generator occurs at 8.8 minutes. The 8.8-minute time ensures a positive margin to overfill assuming a single failure within the turbine-driven AFW Pump speed controller that results in steady-state pump operation at the upper end of the speed control band (4100 rpm). This time limit ensures sufficient space for cooldown, depressurization, and SI termination following a speed-controller failure, which can also be accomplished if AFW is isolated prior to 80% narrow range level. Complete isolation of steam flow from the ruptured steam generator is verified when the narrow range level reaches 30% on the ruptured steam generator or at 12 minutes after initiation of the SGTR, whichever is longer.
For the Shearon Harris analysis the time to reach 30% is less than 12 minutes, and thus the ruptured steam generator is assumed to be completely isolated at 12 minutes.
: 2) Cool down the RCS to Establish Subcooling Margin. After isolation of the ruptured steam generator, there is a 5 minute operator action time imposed prior to initiating the cooldown.
After this time, actions are taken to cool the RCS as rapidly as possible by dumping steam from the intact steam generators. Since offsite power is lost, the RCS is cooled by dumping steam to the atmosphere using the PORVs on the intact steam generators. Both of the intact steam generator PORVs are assumed to be opened at 1020 seconds for RCS cooldown. The cooldown is continued until RCS subcooling at the ruptured steam generator pressure is 20&deg;F plus an allowance for subcooling uncertainty. When these conditions are satisfied at 1356 seconds, it is assumed that the operator closes the intact steam generator PORVs to terminate the cooldown. This cooldown ensures that there will be adequate subcooling in the RCS after the subsequent depressurization of the RCS to the ruptured steam generator pressure. The reduction in the intact steam generator pressure required to accomplish the cooldown is shown in Figure 15.6.3-3, and the effect of the cooldown on the RCS temperatures is shown in Figure 15.6.3-4. The pressurizer level and RCS pressure also decrease during this cooldown process due to shrinkage of the reactor coolant as shown in Figures 15.6.3-1 and 15.6.3-2.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 3) Depressurize RCS to Restore Inventory. After the RCS cooldown, a 4 minute operator action time is included prior to the RCS depressurization. The RCS depressurization is performed to assure adequate coolant inventory prior to terminating SI flow. With the RCPs stopped, normal pressurizer spray is not available and thus the RCS is depressurized by opening a pressurizer PORV. The RCS depressurization is initiated at 1598 seconds and continued until any of the following conditions are satisfied: RCS pressure is less than the ruptured steam generator pressure and pressurizer level is greater than 10%, or pressurizer level is greater than 75%, or RCS subcooling is less than 20&deg;F. For this case, the RCS depressurization is terminated because the RCS pressure is reduced to less than the ruptured steam generator pressure and the pressurizer level is above 10%. The RCS depressurization reduces the break flow as shown in Figure 15.6.3-5, and increases SI flow to refill the pressurizer as shown in Figure 15.6.3-1.
: 4) Terminate SI to Stop Primary to Secondary Leakage. The previous actions establish adequate RCS subcooling, a secondary side heat sink, and sufficient reactor coolant inventory to ensure that SI flow is no longer needed. When these actions have beencompleted, the SI flow must be stopped to prevent repressurization of the RCS and to terminate primary to secondary leakage. The SI flow is terminated at this time if RCS subcooling is greater than 20&deg;F, minimum AFW flow is available or at least one intact steam generator level is in the narrow range, the RCS pressure is stable or increasing, and the pressurizer level is greater than 10%. For the SHNPP analysis, SI was not terminated until the RCS pressure increased to 50 psi above the ruptured steam generator pressure to assure that RCS pressure is increasing.
After depressurization is completed, an operator action time of 3 minutes was assumed prior to initiation of SI termination. Since the above requirements are satisfied, SI termination actions were performed at this time by closing off the SI flow path. After SI termination, the RCS pressure begins to decrease as shown in Figure 15.6.3-2. The intact steam generator PORVs are also opened to dump steam to maintain the prescribed RCS temperature to ensure that subcooling is maintained. When the PORVs are opened, the increased energy transfer from primary to secondary also aids in the depressurization of the RCS to the ruptured steam generator pressure. The primary to secondary leakage continues after the SI flow is terminated until the RCS and ruptured steam generator pressures equalize.
The primary to secondary break flow rate throughout the recovery operations is presented in Figure 15.6.3-5, and the water volume in the ruptured steam generator is presented as a function of time in Figure 15.6.3-6. It is noted that the water volume in the ruptured steam generator when the break flow is terminated is approximately 66 cubic feet less than the total steam generator volume of 5545 ft3. Therefore, it is concluded that overfill of the ruptured steam generator will not occur for a design basis SGTR for Shearon Harris.
15.6.3.3      Westinghouse Thermal Hydraulic Analysis - Offsite Dose Case An analysis was also performed to determine the offsite radiological consequences, assuming the limiting single failure relative to offsite doses without steam generator overfill. Since steam generator overfill does not occur, the results of this analysis represent the limiting consequences for an SGTR for SHNPP, and the results of the offsite radiological consequences analysis are discussed below.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 This analysis was performed to determine the plant response for a design basis SGTR and to determine the integrated primary to secondary break flow and the mass releases from the ruptured and intact steam generators to the condenser and to the atmosphere. This information was then used to calculate the quantity of radioactivity released to the environment and the resulting radiological consequences.
The plant response following an SGTR was analyzed with the LOFTTR2 program until the primary to secondary break flow is terminated. The reactor protection system and the automatic actuation of the engineered safeguards systems were modeled in the analysis. The major operator actions which are required to terminate the break flow for an SGTR were also simulated in the analysis.
15.6.3.3.1 Analysis assumptions The accident modeled is a double-ended break of one steam generator tube located at the top of the tube sheet on the outlet (cold leg) side of the steam generator. It was assumed that the reactor is operating at full power at the time of the accident and the initial secondary mass was assumed to correspond to operation at the nominal steam generator mass minus an allowance for uncertainties. An initial minimum AFW flow of 390 gpm per steam generator was assumed for the offsite dose analysis of record. However, the analysis also finds that a reduction in AFW flow due to a 3% MSSV setpoint tolerance (i.e. an increase in SG pressure) will have no impact on calculated doses. It was also assumed that a loss of offsite power occurs at the time of reactor trip and the highest worth control assembly was assumed to be stuck in its fully withdrawn position at reactor trip.
The limiting single failure was assumed to be the failure of the PORV on the ruptured steam generator. Failure of this PORV in the open position will cause an uncontrolled depressurization of the ruptured steam generator which will increase primary to secondary leakage and the mass release to the atmosphere. It was assumed that the ruptured steam generator PORV fails open when the ruptured steam generator is isolated and that the PORV was subsequently isolated by locally closing the associated block valve.
The major operator actions required for the recovery from an SGTR are discussed in Section 15.6.3.1 and these operator actions were simulated in the analysis. The operator action times which were used for the analysis are presented in Table 15.6.3-3. It is noted that the PORV on the ruptured steam generator was assumed to fail open at the time the ruptured steam generator was isolated. Before proceeding with the recovery operations, the failed open PORV on the ruptured steam generator was assumed to be isolated by locally closing the associated block valve. It was assumed that the ruptured steam generator PORV is isolated at 20 minutes after the valve was assumed to fail open. After the ruptured steam generator PORV was isolated, the additional delay time of approximately five minutes (Table 15.6.3-3) was assumed for the operator action time to initiate the RCS cooldown.
15.6.3.3.2 Transient description The LOFTTR2 analysis results are described below. The sequence of events for this transient is presented in Table 15.6.3-4.
Following the tube rupture, reactor coolant flows from the primary into the secondary side of the ruptured steam generator since the primary pressure is greater than the steam generator Amendment 63                                                                        Page 118 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 pressure. In response to this loss of reactor coolant, pressurizer level decreases as shown in Figure 15.6.3-7. The RCS pressure also decreases as shown in Figure 15.6.3-8 as the steam bubble in the pressurizer expands. As the RCS pressure decreases due to the continued primary to secondary leakage, automatic reactor trip occurs on an overtemperature T trip signal.
After reactor trip, core power rapidly decreases to decay heat levels. The turbine stop valves close and steam flow to the turbine is terminated.
The steam dump system is designed to actuate following reactor trip to limit the increase in secondary pressure, but the steam dump valves remain closed due to the loss of condenser vacuum resulting from the assumed loss of offsite power at the time of reactor trip. Thus, the energy transfer from the primary system causes the secondary side pressure to increase rapidly after reactor trip until the steam generator PORVs (and safety valves if their setpoints are reached) lift to dissipate the energy, as shown in Figure 15.6.3-9. The main feedwater flow will be terminated and AFW flow will be automatically initiated on loss of offsite power assumed at the time of reactor trip.
The RCS pressure decreases more rapidly after reactor trip as energy transfer to the secondary shrinks the reactor coolant and the leak flow continues to deplete primary inventory.
Pressurizer level also decreases more rapidly following reactor trip. The decrease in RCS inventory results in a low pressurizer pressure SI signal. After SI actuation, the RCS pressure and pressurizer level tend to stabilize until the ruptured steam generator PORV is assumed to fail open.
Since offsite power is assumed lost at reactor trip, the RCPs trip and a gradual transition to natural circulation flow occurs. Immediately following reactor trip, the temperature differential across the core decreases as core power decays (see Figures 15.6.3-10 and 15.6.3-11);
however, the temperature differential subsequently increases as natural circulation flow develops. The cold leg temperatures trend toward the steam generator temperature as the fluid residence time in the tube region increases. The intact steam generator loop temperatures continue to slowly decrease due to the continued AFW flow until operator actions are taken to control the AFW flow to maintain the specified level in the intact steam generators. The ruptured steam generator loop temperatures also continue to slowly decrease until the ruptured steam generator was isolated and the PORV was assumed to fail open.
15.6.3.3.3 Major operator actions
: 1) Identify and Isolate the Ruptured Steam Generator. The ruptured steam generator was assumed to be identified and AFW flow to the ruptured steam generator isolated at 10 minutes after the initiation of the SGTR or when the narrow range level reaches 30%,
whichever time is greater. Since the time to reach 30% narrow range level is less than 10 minutes, it was assumed that the ruptured steam generator is isolated at 10 minutes.
Complete isolation of steam flow from the ruptured steam generator is verified when the narrow range level reaches 30% on the ruptured steam generator or at 12 minutes after initiation of the SGTR, whichever is longer. For the Shearon Harris analysis the time to reach 30% is less than 12 minutes, and thus the ruptured steam generator is assumed to be completely isolated at 12 minutes. The ruptured steam generator PORV was also assumed to fail open at this time and the failure was simulated at 722 seconds. The failure causes the ruptured steam generator to rapidly depressurize as shown in Figure 15.6.3-9, which Amendment 63                                                                      Page 119 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 results in an increase in primary to secondary leakage The depressurization of the ruptured steam generator increases the break flow and energy transfer from primary to secondary which results in a decrease in the ruptured loop temperatures as shown in Figure 15.6.3-11.
As noted previously, the intact steam generator loop temperatures also decrease, as shown in Figure 15.6.3-10, until the AFW flow to the intact steam generators is throttled. After this time, the heat transfer to the intact steam generators decreases and the temperature differential across the intact steam generators decreases. As the intact steam generator hot leg temperatures decrease below the steam generator water temperature, reverse heat transfer takes place for a short time period as shown in Figure 15.6.3-10. It was assumed that the time required for the operator to identify that the ruptured steam generator PORV is open and to locally close the associated block valve is 20 minutes. Thus, at 1922 seconds the depressurization of ruptured steam generator was terminated.
: 2) Cool Down the RCS to Establish Subcooling Margin. After the ruptured steam generator PORV block valve was closed, there is an approximately five minute operator action time imposed prior to initiation of cooldown. The depressurization of the ruptured steam generator affects the RCS cooldown target temperature since the temperature is dependent upon the pressure in the ruptured steam generator. Since offsite power was lost, the RCS was cooled by dumping steam to the atmosphere using the intact steam generator PORVs.
The cooldown was continued until RCS subcooling at the ruptured steam generator pressure is 20&deg;F plus an allowance for instrument uncertainty. Because of the lower pressure in the ruptured steam generator, the associated temperature the RCS must be cooled to is also lower, which has the net effect of extending the time for cooldown. The cooldown was initiated at 2224 seconds and was completed at 2996 seconds.
The reduction in the intact steam generator pressures required to accomplish the cooldown is shown in Figure 15.6.3-9 and the effect of the cooldown on the RCS temperature is shown in Figure 15.6.3-10. The pressurizer level and RCS pressure also decrease during this cooldown process due to shrinkage of the reactor coolant as shown in Figures 15.6.3-7 and 15.6.3-8.
: 3) Depressurizes RCS to Restore Inventory. After the RCS cooldown, a 4 minute operator action time is included prior to the RCS depressurization. The RCS is depressurized to assure adequate coolant inventory prior to terminating SI flow. With the RCPs stopped, normal pressurizer spray is not available and thus the RCS is depressurized by opening a safety grade pressurizer PORV. The depressurization is initiated at 3236 seconds and continued until the criteria in the Emergency Operating Procedures are satisfied. The RCS depressurization reduces the break flow as shown in Figure 15.6.3-13, and increases SI flow to refill the pressurizer as shown in Figure 15.6.3-7.
: 4) Terminate SI to Stop Primary to Secondary Leakage. The previous actions establish adequate RCS subcooling, a secondary side heat sink, and sufficient reactor coolant inventory to ensure that SI flow is no longer needed. When these actions have been completed, the SI flow must be stopped to prevent repressurization of the RCS and to terminate primary to secondary leakage. The SI flow is terminated at this time if the SI termination criteria in the Emergency Operating Procedures are satisfied.
After depressurization was complete, an operator action time of 3 minutes was assumed prior to initiation of SI termination. Since the SI termination requirements are satisfied, SI termination actions were performed at this time by closing off the SI flow path. After SI Amendment 63                                                                      Page 120 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 termination, the RCS pressure begins to decrease as shown in Figure 15.6.3-8. The intact steam generator PORVs are also opened to dump steam to maintain the prescribed RCS temperature to ensure that subcooling is maintained. When the PORVs are opened, the increased energy transfer from primary to secondary also aids in the depressurization of the RCS to the ruptured steam generator pressure. The differential pressure between the RCS and the ruptured steam generator is shown in Figure 15.6.3-12. Figure 15.6.3-13 shows that the primary to secondary leakage continues after the SI flow is stopped until the RCS and ruptured steam generator pressures equalize.
The ruptured steam generator water volume is shown in Figure 15.6.3-14. The water volume in the ruptured steam generator when the break flow is terminated is less than the volume for the margin to overfill case in Reference 15.6.3-14 and is significantly less than the total steam generator volume of 5545 ft3 for the radiological case. The mass of water in the ruptured steam generator is also shown as a function of time in Figure 15.6.3-15.
15.6.3.3.4 Mass releases The mass releases were determined for use in evaluating the exclusion area boundary and low population zone radiation exposure. The steam releases from the ruptured and intact steam generators, the feedwater flows to the ruptured and intact steam generators, and primary to secondary break flow into the ruptured steam generator were determined for the period from accident initiation until two hours after the accident and from two to eight hours after the accident. The releases for 0-2 hours were used to calculate the radiation doses at the exclusion area boundary for a two-hour exposure, and the releases for 0-8 hours were used to calculate the radiation doses at the low population zone for the duration of the accident.
The operator actions for the SGTR recovery up to the termination of primary to secondary leakage were simulated in the LOFTTR2 analysis. Thus, the steam releases from the ruptured and intact steam generators, the feedwater flows to the ruptured and intact steam generators, and the primary to secondary leakage into the ruptured steam generator were determined from the LOFTTR2 results for the period from the initiation of the accident until the leakage is terminated. Following the termination of leakage, it was assumed that the actions are taken to cool down the plant to cold shutdown conditions. The PORVs for the intact steam generators were assumed to be used to cool down the RCS to the RHR system operating temperature of 325&deg;F at the maximum allowable cooldown rate of 100&deg;F/hr. The steam releases and the feedwater flows for the intact steam generators for the period from leakage termination until two hours were determined from a mass and energy balance using the calculated RCS and intact steam generator conditions at the time of leakage termination and at two hours. The RCS cooldown was assumed to be continued after two hours until the RHR system in-service temperature of 325&deg;F is reached. Depressurization of the ruptured steam generator was then assumed to be performed to the RHR in-service pressure of 365 psia via steam release from the ruptured steam generator PORV. The RCS pressure was also assumed to be reduced concurrently as the ruptured steam generator is depressurized. It was assumed that the continuation of the RCS cooldown and depressurization to RHR operating conditions are completed within eight hours after the accident since there is ample time to complete the operations during this time period. The steam releases and feedwater flows from two to eight hours were determined for the intact and ruptured steam generators from a mass and energy balance using the conditions at two hours and at the RHR system in-service conditions.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 After eight hours, it was assumed that further plant cooldown to cold shutdown as well as long-term cooling is provided by the RHR system. Therefore, the steam releases to the atmosphere were terminated after RHR in-service conditions were assumed to be reached at eight hours.
For the time period from initiation of the accident until leakage termination, the releases were determined from the LOFTTR2 results for the time prior to reactor trip and following reactor trip.
Since the condenser is in service until reactor trip, any radioactivity released to the atmosphere prior to reactor trip would be through the condenser vacuum pump exhaust. After reactor trip, the releases to the atmosphere were assumed to be via the steam generator PORVs. The mass release rates to the atmosphere from the LOFTTR2 analysis are presented in Figures 15.6.3-16 and 15.6.3-17 for the ruptured and intact steam generators, respectively, for the time period until leakage termination. The mass releases calculated from the time of leakage termination until two hours and from 2-8 hours were also assumed to be released to the atmosphere via the steam generator PORVs. The mass releases for the SGTR event for the 0-2 hour and 2-8 hour time intervals considered are presented in Table 15.6.3-5.
15.6.3.4    Radiological Consequences Analysis The evaluation of the radiological consequences of a steam generator tube rupture (SGTR) assumes that the reactor has been operating at the Technical Specification limit for primary coolant activity and primary to secondary leakage for sufficient time to establish equilibrium concentrations of radionuclides in the reactor coolant and in the secondary coolant.
Radionuclides from the primary coolant enter the steam generator via the ruptured tube and primary to secondary leakage and are released to the atmosphere through the steam generator safety or power operated relief valves (PORVs) and via the condenser air ejector exhaust.
The quantity of radioactivity released to the environment due to an SGTR depends upon primary and secondary coolant activity, iodine spiking effects, primary to secondary break flow, break flow flashing fractions, attenuation of iodine carried by the flashed portion of the break flow, partitioning of iodine between the liquid and steam phases, the mass of fluid released from the generator, and liquid-vapor partitioning in the turbine condenser hot well. All of these parameters were conservatively evaluated for a design basis double ended rupture of a single tube.
15.6.3.4.1 Design basis analytical assumptions The major assumptions and parameters used in the analysis are itemized in Table 15.6.3-6.
15.6.3.4.2 Source term calculations The radionuclide concentrations in the SHNPP primary and secondary system prior to and following the SGTR were determined as follows:
: 1) The iodine concentrations in the reactor coolant will be based upon pre-accident and accident-initiated iodine spikes.
: a. Accident-Initiated Spike - The initial primary coolant iodine concentration is 1.0 Ci/gm of Dose Equivalent (D.E.) I-131. Following the primary system depressurization associated with the SGTR, an iodine spike is initiated in the primary system which increases the iodine release rate from the fuel to the coolant to a value 335 times greater than the Amendment 63                                                                        Page 122 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 release rate corresponding to the initial primary system iodine concentration. The iodine spike appearance rates are given in Table 15.0.9-6.
: b. Pre-accident Spike - A reactor transient has occurred prior to the SGTR and has raised the primary coolant iodine concentration from 1 to 60 Ci/gram of D.E. I-131. The pre-accident iodine concentrations are given in Table 15.0.9-7.
: 2) The initial secondary coolant iodine concentration is 0.1 Ci/gram of D.E. I-131. The initial secondary coolant iodine concentrations are given in Table 15.0.9-7.
: 3) The chemical form of iodine in the primary and secondary coolant is assumed to be 97%
elemental and 3% organic.
: 4) The initial noble gas concentration in the reactor coolant is based upon 1% fuel defects.
The noble gas concentrations in the RCS are given in Table 15.0.9-2.
15.6.3.4.3 Dose calculations The iodine transport model utilized in this analysis considers break flow flashing, steaming, and partitioning. The model assumes that a fraction of the iodine carried by the break flow becomes airborne immediately due to flashing and atomization. The fraction of primary coolant iodine which is not assumed to become airborne immediately mixes with the secondary water and is assumed to become airborne at a rate proportional to the steaming rate and the iodine partition coefficient. This analysis conservatively assumes an iodine partition coefficient of 100 between the steam generator liquid and steam phases. Droplet removal by the dryers is conservatively assumed to be negligible. The iodine transport model is illustrated in Figure 15.6.3-18. The radiological consequences analysis did not consider steam generator tube uncovery in the calculation, since the steam generator tube uncovery issue was investigated and closed by the Westinghouse Owners Group (WOG). Reference 15.6.3-10 documents the Westinghouse position on the issue, that the effects of tube uncovery on the limiting SGTR transient is essentially negligible and need not be considered in the analysis. Reference 15.6.3-11 documents the NRC agreement on the issue.
The following assumptions and parameters were used to calculate the activity released to the atmosphere and the offsite doses following an SGTR.
: 1) The mass of reactor coolant discharged into the secondary system through the rupture and the mass of steam released from the ruptured and intact steam generators to the atmosphere are presented in Table 15.6.3-5
: 2) The time dependent fraction of rupture flow that flashes to steam and is immediately released to the environment is presented in Figure 15.6.3-19. The break flow flashing fraction was conservatively calculated assuming that 100 percent of the break flow comes from the hot leg side of the steam generator, whereas the break flow actually comes from both the hot leg and cold leg sides of the steam generator.
: 3) In addition to the releases calculated in the thermal hydraulic analysis, steam released from the ruptured steam generator to the turbine driven auxiliary feedwater (TDAFW) pump is considered in the dose analysis. A flow of 41,310 lbm/hr is considered from the time of auxiliary feedwater initiation until the ruptured steam generator is isolated. All of the iodine Amendment 63                                                                        Page 123 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 contained in this steam, determined from the steam generator activity and the water/steam partition coefficient of 100, is assumed to be released directly to the atmosphere.
: 4) The total primary to secondary leak rate to the intact steam generators is assumed to be 0.70 gpm. The leakage to the intact steam generators is assumed to persist for the duration of the accident.
: 5) The iodine partition coefficient between the liquid and steam of the ruptured and intact steam generators is assumed to be 100.
: 6) No credit was taken for radioactivity decay during release and transport or for cloud depletion by ground deposition during transport to the control room, exclusion area boundary (EAB) or outer boundary of the low population zone (LPZ).
: 7) Short term atmospheric dispersion factors (x/Qs) for accident analysis and breathing rates are provided in Table 15.6.3-10. The breathing rates were obtained from NRC Regulatory Guide 1.183 (Reference 15.6.3-4).
15.6.3.4.4 Offsite Dose Calculation Model No credit is taken for cloud depletion by ground deposition or by radioactive decay during transport to the exclusion area boundary or to the outer boundary of the low population zone.
Offsite inhalation doses (CEDE) are calculated using the following equation.
DCEDE=            (IAR) (BR) (/Q) where:
DCEDE = CEDE dose via inhalation (rem).
DCFi    = CEDE dose conversion factor via inhalation for isotope i (rem/Ci)(Table 15.0.9-3)
(IAR)ij = integrated activity of isotope i released during the time interval j (Ci)
(BR)j = breathing rate during time interval j (m3/sec)(Table 15.6.3-10)
(/Q)j = atmospheric dispersion factor during time interval j (sec/m3)(Table 15.6.3-10)
Offsite external exposure (EDE) doses are calculated using the following equation:
DEDE=            (IAR) (/Q) where:
DEDE    = external exposure dose via cloud immersion (rem)
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 DCFi    = EDE dose conversion factor via external exposure for isotope I (rem*m3/Ci*sec)(Table 15.0.9.4)
(IAR)ij = integrated activity of isotope i released during the time interval j (Ci)
(/Q)j = atmospheric dispersion factor during time interval j (sec/m3)(Table 15.6.3-10) 15.6.3.4.5 Control Room Dose Calculation Models Control room inhalation doses are calculated using the following equation:
DCEDE=            Conc (BR) where:
DCEDE    = CEDE dose via inhalation (rem)
DCFi      = CEDE dose conversion factor via inhalation for isotope i (rem/Ci)(Table 15.0.9-3)
Concij = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleakage, filtered recirculation and filtered inflow (Ci-sec/m3)
(BR)j    = breathing rate during time interval j (m3/sec)(Table 15.6.5-15)
Control room external exposure doses are calculated using the following equation:
DEDE=            (  Conc )
where:
DEDE    = external exposure dose via cloud immersion in rem.
GF      = geometry factor, calculated based on Reference 15.6.5-23, using the equation
          =    . where V is the control room volume ft3 DCFI    = EDE dose conversion factor via external exposure for isotope i (rem*m3/Ci*sec)(Table 15.0.9-4)
Concij = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleakage, filtered recirculation and filtered inflow (Ci-sec/m3) 15.6.3.4.6 Control Room Doses The control room assumptions are provided in Section 15.6.5.4.3 and Table 15.6.5-15.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 The control room HVAC begins in normal mode. Once the safety injection actuation setpoint is reached at ~178 seconds and after a delay of 15 seconds the control room HVAC is switched to the post-accident recirculation mode. After 2 hours of operation in post-accident recirculation mode the operator switches the control room HVAC system to the pressurized mode.
15.6.3.4.7 Results Offsite and control room doses are calculated for the limiting thermal hydraulic analysis presented in Section 15.6.3.3. The doses at the EAB and the LPZ for an SGTR with an assumed pre-accident iodine spike must be within the 10 CFR 50.67 limit of 25 rem TEDE. The doses at the EAB and the LPZ for an SGTR with an assumed accident-initiated iodine spike must be within 10% of the 10CFR50.67 limit (i.e., less than 2.5 rem TEDE). The doses in the control room must be less than the 10CFR50.67 dose limit of 5 rem TEDE for all cases. The radiological analysis results for this event are presented in Table 15.6.3-13 and are within the applicable limits.
15.6.4    SPECTRUM OF BWR STEAM SYSTEM PIPING FAILURES OUTSIDE CONTAINMENT This section is not applicable to the Shearon Harris Nuclear Power Plant.
15.6.5    LOSS OF COOLANT ACCIDENTS 15.6.5.1    Identification of Causes and Frequency Classification A loss-of-coolant accident (LOCA) is the result of a pipe rupture of the Reactor Coolant System (RCS) pressure boundary. A major pipe break (large break) is considered a limiting fault, an ANS Condition IV event, in that it is not expected to occur during the lifetime of the plant, but is postulated as a conservative design basis. See Section 15.0.1 for a discussion of Condition IV events.
A minor pipe break (small break) is defined as a rupture of the reactor coolant pressure boundary in which the normally operating charging system flow is not sufficient to sustain pressurizer level and pressure. This is considered a ANS Condition III event in that it is an infrequent fault that may occur during the life of the plant. See Section 15.0.1 for a discussion of Condition III events.
The acceptance criteria for the loss-of-coolant accident is described in 10 CFR 50 Paragraph 46 (Reference 15.6.5-1) as follows:
: 1) The calculated peak fuel element clad temperature is below the requirement of 2200&deg;F.
: 2) The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
: 3) The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxidation limits of 17 percent are not exceeded during or after quenching.
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15
: 4) The core remains amenable to cooling during and after the break.
: 5) The core temperature is reduced and decay heat is removed for an extended period of time, as required by the longlived radioactivity remaining in the core.
These criteria were established to provide significant margin in ECCS performance following a LOCA.
15.6.5.2    Large Break LOCA Transient 15.6.5.2.1 Description of large break LOCA transient The rupture of a RCS pipe is assumed to occur on the pump discharge side of a cold leg pipe.
Loss-of-offsite power is assumed to occur co-incident with the LOCA. Primary coolant pump coastdown occurs co-incident with the loss-of-offsite power. Following the break, depressurization of the reactor coolant system, including the pressurizer, occurs. A reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached. Reactor trip and scram are conservatively neglected in the LOCA analysis. Early in the blowdown, the reactor core experiences flow reversal and stagnation which causes the fuel rods to pass through critical heat flux (CHF). Following CHF, the fuel rods dissipate heat through the transition and film boiling modes of heat transfer. Rewet is precluded during blowdown by Appendix K of 10 CFR 50.
A Safety Injection System (SIS) signal is actuated when the appropriate setpoint (high containment pressure) is reached. Due to loss-of-offsite power, a time delay for startup of the diesel generators and SIS pumps is assumed. Once the time delay criteria is met and the system pressure falls below the shutoff head of the Charging/Safety Injection Pumps (CSIPs) and Low Head Safety Injection/Residual Heat Removal (LHSI/RHR) pumps, SIS flow is injected into the cold legs. The single failure criterion is met by assuming the loss of one ECCS pumped injection train. One HHSI pump, one LHSI pump and two containment spray pumps are assumed to be operating. When the system pressure falls below the Cold Leg Accumulator pressure, flow from the Cold Leg Accumulator is injected into the cold legs. Flow from the Emergency Core Cooling System (ECCS)is assumed to bypass the core and flow to the break until the end-of-bypass (EOBY) is predicted to occur (sustained downflow in the downcomer).
Following EOBY, ECCS flow fills the downcomer and lower plenum until the liquid level reaches the bottom of the core (beginning-of-core-recovery or BOCREC time). During this downcomer and lower plenum refill period, heat is transferred from the fuel rods by radiation heat transfer.
Reflood begins at BOCREC time. ECCS fluid fills the downcomer and provides the driving head to move coolant through the core. As the mixture level moves up the core, steam is generated.
Steam binding occurs as the steam flows through the intact and broken loop steam generators and pumps. The pumps are assumed to have a locked rotor (per Appendix K of 10 CFR 50) which tends to reduce the reflood rate. The fuel rods are eventually cooled and quenched by radiation and convective heat transfer as the quench front moves up the core. The reflood heat transfer rate is predicted through experimentally determined heat transfer and carry-over rate fraction correlations.
Continued operation of the ECCS pumps supplies water during long-term cooling. Core temperatures have been reduced to long-term steady state levels associated with dissipation of residual heat generation. After the water level of the refueling water storage tank (RWST)
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Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching from the injection mode to the cold leg recirculation mode of operation in which spilled borated water is drawn from the containment sumps by the low head safety injection (RHR) pump and returned to the RCS cold legs. The Containment Spray System continues to operate to further reduce containment pressure. Hot leg recirculation is established to control boric acid concentration in the reactor vessel when the following criteria are met:
: 1) The safety injection system has previously been aligned for cold leg circulation (meaning that the Refueling Water Storage Tank level has been depleted), and
: 2) 6.5 hours have passed since the beginning of the event, and
: 3) Safety Injection has not been terminated such that a single Charging Safety Injection Pump has been realigned to the charging header (meaning that Reactor Coolant System subcooling and Pressurizer level have been established)(Reference 15.6.5-34). See Sections 6.3.2.5.2.3 and 6.3.2.8.
15.6.5.2.2 Large break LOCA evaluation model The PWR ECCS evaluation model described in Reference 15.6.5-36 was used to perform the LBLOCA analysis. Additional restrictions are applied to the implementation of EMF-2103 as described in the plant specific implementation methodology ANP-3011 (Reference 15.6.5-50).
The EMF-2103 model consists of the following computer codes:
: 1) RODEX2 3A for initial stored energy, fission gas release, and gap conductance;
: 2) S-RELAP5 for the system calculation and;
: 3) AUTORLBLOCA for generation of ranged parameter values, transient input, transient runs, and general output documentation.
The Shearon Harris nuclear reactor is a Westinghouse three-loop pressurized water reactor with a dry containment. The reactor coolant system (RCS) is divided into control volumes representing reasonably homogeneous regions, interconnected by flow paths or "junctions."
The reactor coolant pump performance characteristics are the Westinghouse pump homologous curves built into the S-RELAP5 code. Three percent of the tubes in each steam generator are assumed to be plugged.
The transient behavior was determined from the governing equations for the conservation of mass, energy, and momentum. Energy transport, flow rates, and heat transfer are determined from appropriate correlations. The reactor core model uses heat generation rates determined from reactor kinetics equations which use reactivity feedback and decay heating from the 1979 ANSI/ANS standard.
15.6.5.2.3 Input parameters and initial conditions Table 15.6.5-1 lists important input parameters, fuel design parameters and initial conditions used in the analysis. The LOCA analysis is based on a full core of AREVA fuel operating at a power of 2958 MWt (2948 MWt plus 10 MWt uncertainty), a peak rod average exposure less Amendment 63                                                                    Page 128 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 than 62,000 MWD/MTU, a total peaking factor ( ) of 2.52, a nuclear enthalpy rise factor (FH) of 1.73 (including 4% uncertainty) and no axial or burnup dependent power peaking limit.
15.6.5.2.4      Large break LOCA results The Peak Cladding Temperature (PCT) was calculated to be 1935&deg;F for the limiting case which has a 3.6168 ft2 Cold Leg Guillotine (CLG) (one sided break area shown) and models the loss of one ECCS train (i.e., one HHSI and one LHSI/RHR pump). The maximum local Zr-H2O reaction was calculated to be 4.2%. The core wide Zr-H2O reaction was less than 1.0%.
The NRC review of the HNP specific methodology (ANP-3011) identified a previously un-quantified issue with axial relocation of fuel pellet material in a rod that bursts during a LBLOCA.
Progress Energy committed (Reference 15.6.5-51) to include a penalty to 138&deg;F to the analysis PCT value (1935&deg;F) to account for the uncertainties of the effect and the potential benefits of the rupture in the flow channel. Including the impact of errors found in the analysis, the net result is a LOCA PCT of 2095&deg;F for the purposes of NRC reporting against the regulatory limit of 2200&deg;F.
This analysis supports full power operation Tavg ranging from 582&deg;F to 594.8&deg;F.
Figures 15.6.5-2 through 15.6.5-20 show transient results for the limiting case.
Break Spectrum versus Axial Shape Results - Calculations were performed with a range of DECLG break sizes from approximately 0.26 times to 1.0 times the full break area.
The limiting case has a 3.6168 ft2 CLG break (total break area of 7.2336 ft2). A constant value of K(z) = 1 was used for all core elevations.
Single Failure - The single failure evaluated was the loss of one diesel generator, namely a HHSI pump and a LHSI pump. However, all containment sprays and fans are assumed to function with minimum start time delays.
Exposure Study Results - The current AREVA methodology considers the effects of peak fuel rod exposures. The Linear Heat Generation Rate limit is therefore independent of exposure up to a peak rod average exposure of 62,000 MegaWatt Days/Metric Ton of Uranium.
Long Term Criticality Analysis Results - The long term post-LOCA analysis assures that the mixed mean boron concentration in the containment sump is higher than the corresponding critical boron concentration with no credit for control rod insertion.
15.6.5.3    Small Break LOCA Transient 15.6.5.3.1 Description of small break LOCA transient The postulated SBLOCA covers a range of break areas that encompasses small lines which penetrate the primary pressure boundary. Small breaks could involve relief and safety valves, charging and letdown lines, drain lines, and instrumentation lines. The most limiting break location is in the cold leg pipe at the discharge side of the pumps. This break location results in the largest amount of inventory loss and the largest fraction of Emergency Core Cooling System Amendment 63                                                                          Page 129 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 (ECCS) fluid being ejected out through the break. This produces the greatest degree of core uncovery and the longest fuel rod heatup time.
The SBLOCA transient is characterized by a slow depressurization of the primary system with a reactor trip occurring at a low primary pressure (1934.7 psia). The Safety Injection Actuation Signal (SIAS) occurs when the system has further depressurized (1714.7 psia for Harris). The capacity and shutoff head of Charging/Safety Injection Pumps (CSIPs) are important parameters in the SBLOCA transient.
The SBLOCA transient can be categorized into three ranges of break sizes. The scenario is different for each range of break sizes. The "small" small breaks are characterized by inventory losses that are less than the makeup capacity of the CSIPs such that core uncovery is limited or precluded. The core level is eventually recovered and hot rod heatup is limited. "Large" small breaks are characterized by a larger primary system depressurization rate such that the accumulator pressure is reached in sufficient time to limit the core uncovery and hot rod heatup.
The CSIPs have limited influence in the "large" small break transient. The break sizes between the "small" and "large" small breaks are generally the most limiting. For "medium" small breaks, the rate of inventory loss from the primary system is large enough that the CSIPs cannot preclude significant core uncovery. The primary system depressurization rate is very slow, extending the time required to reach the accumulator pressure. This tends to maximize the heatup time of the hot rod and produces the maximum Peak Cladding Temperature (PCT). It also results in the longest time with the core being at elevated temperatures, which maximizes the local cladding oxidation. Core recovery for the limiting break begins when intact loop SI flow and accumulator flow exceed primary system inventory lost out the break.
As contrasted with the large break, the blowdown phase of the small break occurs over a longer time period. Thus, for the small-break LOCA there are only three characteristic stages, i.e., a gradual blowdown in which the decrease in water level is checked, core recovery, and long-term recirculation.
15.6.5.3.2 Small break LOCA evaluation model The AREVA NP SBLOCA evaluation model (References 15.6.5-40, 15.6.5-41 and 15.6.5-48) consists of two principal computer codes. The appropriate conservatisms, prescribed by Appendix K of 10 CFR 50, are incorporated. The sensitivity analyses, including time step analyses required by the NRC were run. The computer codes are:
: 1) The RODEX2 -2A code was utilized to determine the initial fuel stored energy and gap conditions for the initialization of the system blowdown and hot rod response calculations.
: 2) S-RELAP5 code was used to predict the thermal-hydraulic response of the primary and secondary sides of the reactor system and the hot rod response.
The Harris Nuclear Plant is a Westinghouse designed 3-loop PWR, having three hot leg pipes, three inverted U-tube generators, and three cold leg pipes with one reactor coolant pump in each cold leg.
The reactor coolant system of the plant is nodalized in the S-RELAP5 model into control volumes representing reasonably homogeneous regions, interconnected by flow paths or "junctions." The model includes three identical accumulators, a pressurizer, and three steam Amendment 63                                                                      Page 130 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 generators with both primary and secondary sides modeled. All three loops of the plant were simulated separately in order to provide an accurate representation of the plant. A steam generator tube plugging level of 3% was assumed. The HHSI pumps and lines were explicitly modeled to simulate minimum injection flow rates prescribed by the Technical Specifications.
Since the system pressure does decay to the shutoff head of the Low Head Safety Injection (LHSI) pumps in the SBLOCA analysis, LHSI pumps and lines were included in the S-RELAP5 model. The primary coolant pump performance curves were characteristic of Westinghouse pumps.
The heat generation rate in the S-RELAP5 reactor core model was determined from reactor kinetics equations with actinide and decay heating as prescribed by Appendix K.
Single failure criteria were satisfied by the assumed loss of a diesel generator, which results in the disabling of one Charging/Safety Injection Pump (CSIP) and one of the two motor-driven auxiliary feedwater pumps. Although three CSIPs may be physically installed, only two of the three pumps are electrically connected and considered operational at any given time.
Therefore, in the analysis, only one CSIP was available to mitigate the event and the analytically assumed pump performance bounds any combination of two of the possible three CSIPs that may have been in service at the initiation of the event.
For the Harris Nuclear Plant, the Reactor Coolant Pumps (RCPs) are manually tripped. The RCP trip criteria given in the Emergency Operating Procedure (Reference 15.6.5-43) are listed in Table 15.6.5-10. In the analysis, the RCPs were tripped at reactor scram. Tripping the RCPs at scram is conservative as this assumption impedes the loop seal clearing event and hence allows additional mass to escape from the system prior to the clearing of the loop seals. The reduced system water inventory, subsequently, contributes to the severity of the transient.
A conservative top skewed axial power shape was used in this analysis. The shape was adjusted to be consistent with the Technical Specification FH and F limits.
Important system parameters used in this analysis are given in Table 15.6.5-10.
15.6.5.3.3 Small break LOCA results SBLOCA break spectrum calculations were performed for break diameters from 0.75 inch to 9.0 inches in one of the cold legs of the reactor coolant system. The break spectrum calculations were based on the nominal auxiliary feedwater flow from one motor-driven pump.
The break spectrum calculations were performed at a nominal primary Tavg of 588.8&deg;F.
Predicted event times from the break spectrum calculations are summarized in Table 15.6.5-11a. Results from S-RELAP5 hot rod response calculations for the limiting case are presented in Table 15.6.5-11b.
Review of the predicted event times in Table 15.6.5-11a shows that for the limiting break size (2.6-inch diameter) the RCP trip criterion is satisified and break uncovery occurs at approximately 15 minutes into the event. Thus, well over 5 minutes is available for operator action to manually trip the RCPs prior to break uncovery.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 The 0.75-inch break represents the "small" SBLOCA for Harris. The coolant inventory loss rate for the 0.75-inch break case was slower than the makeup capacity of the SI system such that core uncovery was limited. The core level was eventually recovered and hot rod heatup was limited. Therefore, this transient was less severe.
The 9.0-inch break represents the "large" SBLOCA for Harris. The 9.0 inch break case experienced a rapid depressurization to the accumulator pressure which limited the length of time the core was uncovered and the depth of the core uncovery.
The 2.6-inch break represents the "medium" SBLOCA for Harris. The results show the 2.6-inch break to be the limiting break because it resulted in a slow rate of depressurization to the accumulator pressure, exposing the core for a long period of time, and causing the most severe fuel heatup.
System responses from the ANF-RELAP calculations for the limiting break are shown in Figures 15.6.5-30 through 15.6.5-35. The primary and secondary pressure responses are shown in Figure 15.6.5-30. The primary pressure decreased immediately after break initiation. Reactor scram occurred when the primary pressure reached 1934.7 psia. The secondary pressure increased rapidly after break initiation as the reactor scrammed and steam generator isolation took place. The secondary pressure continued to increase until the steam generator safety valves opened, causing the secondary pressure to stabilize. At approximately 878 seconds, liquid was expelled from the loop seal piping, allowing steam to flow directly to the break, which caused the primary pressure to decrease more rapidly.
The break flow rate is shown in Figure 15.6.5-31. The two-phase flow regime from 400 seconds to 878 seconds (loop seal clears) is characterized by an essentially fixed flow with a large oscillation band. The final single-phase flow regime is steam only. While a relatively large reduction in break mass flow occurs at the onset of this regime, the break volumetric flow actually increases, thereby causing an increase in the primary system depressurization rate.
Oscillations in the break flow rate after 2020 seconds were caused by the initiation of accumulator flow, shown in Figure 15.6.5-34.
The downcomer and hot assembly collapsed liquid levels are shown in Figure 15.6.5-32. The mixture level remains above the core over the first 1300 seconds until the combination of decreasing pressure and inventory causes the level to begin to decrease. Sustained core uncovery begins at approximately 1300 seconds. The level continues to fall until accumulator flow is finally activated at 2020 seconds and the core begins to recover.
The total HHSI flow is shown in Figure 15.6.5-33. Flow begins at approximately 60 seconds and increases as primary system pressure decreases.
Flow from the accumulators, shown in Figure 15.6.5-34, begins at 2020 seconds and terminates the event.
The reactor vessel fluid mass, shown in Figure 15.6.5-35, declines rapidly after event initiation.
After loop seal clearing at approximately 878 seconds, the amount of mass in the primary system continues to decline but at a reduced rate. The minimum primary system mass occurs at approximately 2038 seconds. By 2038 seconds, accumulator flow is active and the primary system mass begins to increase.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 S-RELAP5 calculated cladding temperature for the limiting break is shown in Figure 15.6.5-36.
The PCT for the limiting break SBLOCA analysis (2.6-inch break) was calculated to be 1664&deg;F with a maximum local cladding oxidation of 2.2196%. Results from S-RELAP5 hot rod response calculations for the limiting break are presented in Table 15.6.5-11b.
15.6.5.3.4 Small break LOCA summary of results and conclusions The SBLOCA analysis for the Harris Nuclear Plant identified the 2.6-inch diameter break to be the limiting break size. The PCT was calculated to be 1664&deg;F with a maximum local cladding oxidation of 2.2196% for the 2.6-inch break size. The calculation was performed at Tavg of 588.8&deg;F. Additional errors were reported by AREVA, with a revised PCT of 1695&deg;F.
In Reference 15.6.5-53, HNP requested a change to Technical Specifications to increase the as-found lift setting tolerance for the mains steam line code safety valves from +/- 1% to +/- 3%. A consequence of the increased MSSV setpoint tolerance is a reduction of the credited AFW flow in the safety analyses at the lowest lifting MSSV setpoint plus tolerance. The change to AFW flow is a reduction from 390 gpm to 374 gpm. In the SBLOCA transients, secondary pressure rises to the MSSV setpoint upon reactor/turbine trip and remains there until primary phase change at the break occurs with a commensurate increase in energy release from the primary system. Early in a SBLOCA event, an increase in the MSSV setpoint tolerance can affect the energy balance during the transient because it results in a secondary heat sink temperature change. The higher setpoints of the MSSV's cause less heat transfer from the primary system and higher primary pressure. This results in less HPSI flow into the system, an earlier core uncovery, and more extensive cladding heatup. The estimated impact of this change on the SBLOCA analysis calculated peak cladding temperature is +32&deg;F, with a revised PCT of 1727&deg;F.
The analysis supports full power operation of the Harris Nuclear Plant at 2958 MWt (2948 MWt plus 10 MWt uncertainty) with a steam generator tube plugging level of up to 3.0% at full-power nominal Tavg of 588.8&deg;F. The analysis supports an FH of 1.73, and an        of 2.52 with an axially independent power peaking limit curve.
Operation of the Harris Nuclear Plant with AREVA fuel within the above stated criteria assures that the NRC acceptance criteria for SBLOCA (10 CFR 50.46(b)) will be met with the ECCS for the Harris Nuclear Plant.
15.6.5.4    Radiological Consequences Analysis of a Postulated Large Break Loss of Coolant Accident An abrupt failure of the main reactor coolant pipe is assumed to occur and it is assumed that the emergency core cooling features fail to prevent the core from experiencing significant degradation (i.e., melting). This sequence cannot occur unless there are multiple failures, and thus goes beyond the typical design basis accident that considers a single active failure.
Activity from the core is released to the containment and from there released to the environment by means of containment leakage and leakage from the emergency core cooling system.
The input parameters and assumptions are listed in Table 15.6.5-12. Activity is released from the fuel into the containment using the timing and release fractions from RG 1.183. The analysis considers the release of activity from the containment via containment leakage. In addition, once the recirculation mode of the emergency core cooling system (ECCS) is Amendment 63                                                                      Page 133 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 established, activity in the sump solution may be released to the environment by means of leakage from ECCS equipment into the auxiliary building. Activity of the sump solution may also be released to the environment by means of leakage into the refueling water storage tank (RWST). The total offsite and control room doses are the sum of the doses resulting from each of the postulated release paths.
The release of activity from the core occurs over a 1.8 hour interval. Table 15.6.5-13 gives the fission product release timing assumed. A wide spectrum of nuclides is taken into consideration. Table 15.0.9-1 lists the nuclides being considered for the LOCA with core melt (eight groups of nuclides). Table 15.6.5-14 gives the core fission release fractions. The iodine is mainly in the form of cesium iodide, which exists as particulate. The iodine characterization from RG 1.183 is 4.85% elemental, 0.15% organic and 95% particulate.
For the containment leakage analysis, all activity released from the fuel is assumed to be in the containment atmosphere until removed by sprays, sedimentation, radioactive decay or leakage from the containment. For the ECCS leakage analyses, all iodine activity released from the fuel is assumed to be in the sump solution until removed by radioactive decay or leakage from the ECCS.
15.6.5.4.1 Containment Modeling The containment building is modeled as two discrete volumes: sprayed and unsprayed. The volumes are conservatively assumed to be mixed only by the containment fan coolers. The containment volume is 2.344E6 ft3 with a sprayed fraction of 85.9% of the total (2.014E6 ft3).
The containment is assumed to leak at the design leak rate of 0.1% per day for the first 24 hours of the accident and then to leak at half that rate (0.05% per day) for the remainder of the 30 day period following the accident considered in the analysis.
One train of the containment spray system is assumed to operate following the LOCA. Injection spray is credited starting at 120 seconds in the event. This is conservative since it results in earlier spray termination and there is little activity in the containment at the time the sprays start.
When the RWST drains to a predetermined setpoint level, the system automatically switches to recirculation of sump liquid to provide a source for the sprays. The analysis assumed that the sprays are terminated 4.0 hours from the start of the event. The elemental iodine spray removal coefficient is 20 hr-1. Removal of elemental iodine from the containment atmosphere is assumed to be terminated when the airborne inventory drops to 0.5 percent of the total elemental iodine released to the containment (this is a DF of 200). With the RG 1.183 source term methodology this is interpreted as being 0.5 percent of the total inventory of elemental iodine that is released to the containment atmosphere over the duration of gap and in-vessel release phases. In the analysis, this occurs just before 2.0 hours. The particulate iodine spray removal coefficient is 3.94 hr-1. When the airborne inventory drops to 2 percent of the total particulate iodine released to the containment (this is a DF of 50) this removal coefficient is reduced by a factor of 10. In the analysis this occurs at 2.5 hours.
During spray operation, credit is taken for sedimentation removal of particulates in the unsprayed region. After sprays are terminated, credit for sedimentation is taken in both the sprayed and unsprayed regions. For the analysis the sedimentation removal coefficient is conservatively assumed to be 0.1 hr-1.
Amendment 63                                                                          Page 134 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 15.6.5.4.2 ECCS Leakage When ECCS recirculation is established following the LOCA, leakage is assumed to occur from ECCS equipment outside containment. There are two pathways considered for the ECCS recirculation leakage. One is the leakage directly into the Auxiliary Building and the other is back-leakage into the refueling water storage tank (RWST). Recirculation is initiated when the RWST has drained to the pre-determined setpoint level (at about 20 minutes).
It is assumed that the iodine is instantaneously mixed in the primary containment sump water at the time of release from the core.
The total ECCS recirculation leakage into the Auxiliary Building modeled in the analysis is 2 gpm (i.e., the assumed value of 1 gpm total ECCS leakage outside the containment, consistent with the amount of RCS unidentified leakage allowed by the HNP Technical Specifications, is doubled consistent with Regulatory Guide 1.183 guidance) and begins at 20 minutes. There is 2% partitioning of iodine in the leakage. Of this leakage, 1.934 gpm is inside the area served by the Reactor Auxiliary Building Emergency Exhaust System (RABEES) which filters out much of the iodine released to the atmosphere. The remaining 0.066 gpm is released outside of RABEES without filtration.
The 2 percent partitioning value for iodine releases from the fluid leaking into the ECCS area is based upon a conservative set of assumptions. In the analysis, the temperature of the sump water has been conservatively assumed to be constant at 230&deg;F although the temperature is predicted to be reduced to 212&deg;F after approximately 4 hours following the start of recirculation.
It is also assumed that this flashing fraction is equivalent to the total partitioning fraction between iodine which stays in the ECCS leakage fluid, and that which evolves into airborne iodine for release. This 2% flashing fraction/partition fraction is retained following Steam Generator Replacement/Power Uprate, even though the maximum sump fluid temperature for SGR/PUR will exceed this temperature for a short period of time.
The fraction of recirculation sump water that would flash into steam after leaking into the ECCS area has been found to be 2 percent based on the Regulatory Guide 1.183 (Reference 15.0.9-8)
Section 5.4 constant enthalpy, or "constant h" process. This reference specifies the following formula for determining the Flashing Fraction:
FF =
where:
          = enthalpy of liquid at sump fluid temperature and pressure
          = enthalpy of liquid at saturation conditions (14.7 psia, 212&deg;F)
            = heat of vaporization at 212&deg;F Treating this 2% flashing fraction as the total iodine Partition Factor, based on the 230&deg;F sump fluid temperature, is acceptable based on the following evaluation. The fraction of iodine, (the Partition Factor, or PF) that would become airborne may be calculated using the following model (Reference 15.6.5-29):
Amendment 63                                                                          Page 135 of 151
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 PF =      1700
: where, PF = partition factor S = mass fraction of steam W = mass fraction of water PC = partition coefficient, (Ci/cc liquid)/(Ci/cc gas); equivalent to the Flashing Fraction 1700 = the ratio of vapor to liquid specific volumes at 212&deg;F Standard Review Plan (SRP) 6.5.2, Rev. 1, indicates that long term iodine retention with no significant re-evolution may be assumed when the equilibrium sump pH, after mixing and dilution with the primary coolant and ECCS injection, is above 8.5. This view is supported by L.
F. Parsly (Reference 15.6.5-30) by indicating high values of PC at pH of 9 and above, when iodate formation is significant. A value of 1.765E+09 has been indicated at 212 F, pH equal to 9 and concentration of aqueous iodine of 3E-03 moles/liter. The PC indicated in SRP Section 6.5.2, Rev. 1; Figure 5.2-1 is 5E+03. Conservatively, selecting 5E+03, PF is calculated as follows:
PF =        1700      = 6.9  03 This suggests that only 0.69 percent of the iodine leaking into the ECCS area would become airborne and be removed with the exhaust. Therefore, the 2 percent value, in effect, does not account for partition and is a conservative estimate in the dose evaluation.
Allowable back-leakage from the ECCS or Containment Spray to the RWST is modeled for a variety of possible flowpaths. For the flowpaths that lead to sump fluid back-leakage entry into the RWST below the remaining liquid level, the dose analysis assumed that the total flowrate is 25 gpm. For the flowpaths that lead to the airspace above the remaining liquid level in the RWST, the dose analysis assumed that the total flowrate is between 2.2 gpm and 9.7 gpm.
These dose analysis input assumption flowrates are bounding (high) analytical values, not acceptance criteria tor measurement procedures. When comparing measured flowrates to these dose analysis input assumption flowrates, the measured flowrates should first be doubled, in accordance with the dose analysis modeling guidance of Regulatory Guide 1.183 (Reference 15.0.9-8). In addition to doubling the measured flows, other adjustments are required for applying measurement test procedure uncertainties and for ensuring that the combined measured flows do not boil upon entry to the airspace of the tank. The applicable flow paths include:
a) RWST to the RHR pumps - During recirculation isolation is provided by 1SI-320 and 1SI-322 (Train A RHR pump) and 1SI-321 and 1SI-323 (Train B RHR pump) b) High Head Safety Injection Mini-flow to the RWST - During recirculation isolation is provided by 1CS-745 and 1CS-746 (Train A CSIP) and 1CS-752 and 1CS-753 (Train B CSIP)
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Shearon Harris Nuclear Power Plant                                                  UFSAR Chapter: 15 c) RHR Test recirculation Line - During recirculation isolation is provided by 1-SI-331 (manual in-series isolation valves) d) RWST to CSIP suction header - During recirculation isolation is provided by 1CS-291 (Train A) and 1CS-292 (Train B) which are in parallel and an in-series check valve (1CS-294) e) Containment Spray Test line - During recirculation isolation is provided by 1CT 47 (Train A) and 1CT-95 (Train B) f)  RWST to the Containment Spray pumps - During recirculation isolation is provided by 1CT-26 and 1CT-27 (Train A Containment Spray Pump) and 1CT-71 and 1CT-72 (Train B Containment Spray Pump).
The iodine in the sump solution is assumed to all be in nonvolatile iodide or iodate form.
However, when the solution mixes with RWST liquid inventory, some fraction of the iodine may convert from these iodide and iodate forms into elemental (dissolved gaseous) form. The amount of iodine that converts to the elemental form is dependent both on the concentration of the total iodine (stable and radioactive isotopes) in solution and on the mixed pH of the RWST solution. The initial boron concentration of the RWST liquid inventory is ~2500 ppm. The evaluated initial pH of the RWST remaining liquid inventory is ~4.5. With this low pH, and low initial total iodine concentration, the initial conversion fraction of iodide/iodate forms to elemental iodine form is less than 0.1 %. Peak conversion rates occur at various times for the different back-leakage flowrates. For all flowpaths allowed to enter the airspace and directly to the liquid inventory, the peak conversion rate remains less than 0.1% for the duration of the event. By 24 hours into the event, the combination of increased pH and increased concentration of total iodine in the RWST liquid yields conversion rates that are less than 0.001%. In the to-air dose analysis, the 24 hour conversion rate is assumed constant for the remainder of the 30 day dose event. For the to-water dose analysis, the 84 hour conversion rate is assumed constant for the remainder of the 30 day dose event.
Elemental iodine is volatile, and will partition between the liquid and the air in the RWST gas space. The partition coefficient for elemental iodine is a function of the liquid inventory temperature. Initially, the partition coefficient for the release from the to-air flow is approximately 3 (due to the slightly cooled sump temperatures assumed for this flowpath), and for the direct to-water flow the partition coefficient is approximately 42.1. The minimum partition coefficient for the to-air flow occurs at the peak sump temperature time, with a value of about 3.
After that the coefficient rapidly rises to a value of about 10 at 24 hours. For the to-water flowpaths, the relatively colder mixture of back-leakage and RWST water provides the initially larger partition coefficient of about 42.1, which gradually drops to about 40.9 by 24 hours into the event.
The combined effects of volatile iodine gas generation and subsequent iodine gas partitioning between the liquid and gas space in the RWST are tracked by IODEX-NAI on a time-step by time-step basis, and the results are converted into an iodine gas inventory in the air space of the tank. The modeling of the air flow out of the RWST is based on backleakage displacement added to the diurnal heating and cooling cycles averaged over 24 hours. This model ignores the effect of the large heat sink provided by the mass of RWST water in the tank that would tend to moderate the effects of heating and cooling from the sunlight and atmospheric temperature variations. The IODEX-NAI process converts the airspace inventory and diurnal heating driven release into an average fraction of the total sump inventory that is released during the time-step.
Amendment 63                                                                            Page 137 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 This average release fraction is treated in LOCADOSE as a release fraction, similar to the way the containment release fraction is modeled for the containment release portion of the total LOCA dose analysis.
15.6.5.4.3 Offsite Doses The offsite doses are calculated using the assumptions and equations in Section 15.0A.1.
15.6.5.4.4      Doses to Control Room Personnel Control room inhalation doses are calculated using the following equation:
              =                        (    )
where:
DCEDE = CEDE dose via inhalation (rem)
DCFi = CEDE dose conversion factor via inhalation for isotope i (rem/Ci)(Table 15.0.9-3)
Concij = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleakage, filtered recirculation and filtered inflow (Ci-sec/m3)
(BR)j  = breathing rate during time interval j (m3/sec)(Table 15.6.5-15)
Control room external exposure doses are calculated using the following equation:
1
              =
where:
DEDE = external exposure dose via cloud immersion in rem.
GF = geometry factor, calculated based on Reference 15.6.5-23, using the equation GF= 1173V0.338 where V is the control room volume ft3 DCFi    = EDE dose conversion factor via external exposure for isotope i (rem*m3/Ci*sec)(Table 15.0.9-4)
Concij = concentration in the control room of isotope i, during time interval j, calculated dependent upon inleakage, filtered recirculation and filtered inflow (Ci-sec/m3)
Parameters used in the control room personnel dose calculations are provided in Table 15.6.5-
: 15. These parameters include the normal operation flowrates, the emergency operation flowrates, control room volume, filter efficiencies and control room operator breathing rates.
Amendment 63                                                                      Page 138 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 The inflow (filtered and unfiltered) to the control room and the control room recirculation flow are used to calculate the activity introduced to the control room and cleanup of activity from that flow.
In the event of a large break LOCA, the SI setpoint will be reached shortly after event initiation.
The SI signal causes the control room HVAC to switch from the normal operation mode to the post-accident recirculation mode of operation. It is assumed that the SI setpoint is reached immediately at the start of the event; only the 15 second delay time for switching from normal to emergency operating mode is modeled. An operator action switches the control room from the post-accident recirculation mode to the pressurization mode at 2 hours after event initiation.
CP&L Shine Dose Contribution:
As required by Reg Guide 1.183 (Reference 15.6.3-4), Section 4.2.1 on Control Room Dose Calculation Methodology, the control room dose contributions of the release plume, the containment building post-accident radionuclide inventory, and the control room HVAC filter shine doses were conservatively evaluated. The small incremental dose contributions from these sources are included in the total Control Room TEDE dose reported in Table 15.6.5-16.
15.6.5.4.5 Results The potential radiological consequences resulting from a large break LOCA have been conservatively analyzed, using assumptions and models described in previous sections. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The results are listed in Table 15.6.5-16. The resultant doses are within the guideline values. The 10CFR50.67 guideline values are 25 rem TEDE at the EAB and LPZ and 5 rem TEDE in the control room.
15.6.5.5  Radiological Consequences Analysis of a Postulated Small Break Loss-of-Coolant Accident The small break loss-of-coolant accident (SBLOCA) is not explicitly described in Regulatory Guide 1.183 (RG 1.183) as a separate design basis accident to be analyzed. Rather RG 1.183, Appendix A, states that the large break loss-of-coolant accident (LBLOCA) is assumed to be the design basis case for analyzing radiological consequences from a spectrum of loss-of-coolant accident break sizes. Therefore, the LBLOCA dose values shown in Table 15.6.5-16 bound those of the SBLOCA.
 
==REFERENCES:==
SECTION 15.6 15.6.1-1        EPRI NP-7450(A), Revision 10, RETRAN-3D - A Program for Transient Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, September 2014.
15.6.1-2        EPRI NP-2511-CCM-A, Revision 4, VIPRE A Thermal-Hydraulic Code for Reactor Cores, June 2007.
15.6.1-3        DPC-NE-3008, Revision 0, Thermal-Hydraulic Models for Transient Analysis, April 2018.
Amendment 63                                                                      Page 139 of 151
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 15.6.1-4      DPC-NE-3009, Revision 0, FSAR / UFSAR Chapter 15 Transient Analysis Methodology, April 2018.
15.6.1-5      DPC-NE-2005-PA, Revision 5, Thermal-Hydraulic Statistical Core Design Methodology, March 2016.
15.6.1-6      EMF-92-153-PA, Revision 1, HTTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel, January 2005.
15.6.3-1      Deleted by Amendment No. 43.
15.6.3-2      Letter from A. B. Cutter of Carolina Power & Light Company to Document Control Desk, USNRC, dated December 15, 1989, transmitting "LOFTTR2 Analysis for a Steam Generator Tube Rupture with Revised Operator Action Times for Shearon Harris Nuclear Power Plant," WCAP 12403 (Proprietary) WCAP 12404 (Non-Proprietary), November 1989.
15.6.3-3      Deleted by Amendment No. 51.
15.6.3-4      NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.
15.6.3-5      International Commission on Radiological Protection, "Limits for Intake of Radionuclides by Workers", ICRP Publication 30, volume 3 No. 1-4, 1979.
15.6.3-6      ENDF-223, "ENDF/B-IV Fission-Product Files: Summary of Major Nuclide Data,"
T.R. England & R.E. Schenter, October 1975.
15.6.3-7      WCAP-12403, Supplement 1 - "Steam Generator Tube Rupture Analysis for Shearon Harris Nuclear Power Plant."
15.6.3-8      Nuclear Fuel Section Design Activity 94-0030, Impact of the HNP CSIP Rotor Replacement on the SGTR Analysis.
15.6.3-9      WCAP-14778 - "Shearon Harris Nuclear Power Plant, Steam Generator Replacement/Uprate Analysis and Licensing Project, NSSS Engineering Report,"
including CP&L maintained additions to address non-Westinghouse scope items.
15.6.3-10    WOG-92-25, "Westinghouse Owners Group Steam Generator Tube Uncovery Issue", March 31, 1992.
15.6.3-11    Reference B: Attachment to WOG-93-066, Letter to Lawrence A. Walsh (Chairman WOG) from Robert C. Jones (NRC), "Westinghouse Owners Group-Steam Generator Tube Uncover Issue", March 10, 1993.
15.6.3-12    NSAL-07-11, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology."
Amendment 63                                                                    Page 140 of 151
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 15.6.3-13    License Amendment Request for Implementation of Alternative Source Term (AST). Letters dated July 17, 2001 (Serial HNP-01-107) and August 17, 2001 (Serial HNP-01-120).
15.6.5-1      "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR
: 50. Federal Register, Volume 39, Number 3, January 4, 1974.
15.6.5-2      Deleted by Amendment No. 45.
15.6.5-3      Deleted by Amendment No. 45.
15.6.5-4      Deleted by Amendment No. 45.
15.6.5-5      Deleted by Amendment No. 45.
15.6.5-6      Deleted by Amendment No. 45.
15.6.5-7      Deleted by Amendment No. 45.
15.6.5-8      Deleted by Amendment No. 46.
15.6.5-9      Deleted by Amendment No. 46.
15.6.5-10    Deleted by Amendment No. 46.
15.6.5-11    Deleted by Amendment No. 46.
15.6.5-12    Deleted by Amendment No. 46.
15.6.5-13    Deleted by Amendment No. 46.
15.6.5-14    Deleted by Amendment No. 46.
15.6.5-15    Deleted by Amendment No. 46.
15.6.5-16    Deleted by Amendment No. 46.
15.6.5-17    Deleted by Amendment No. 46.
15.6.5-18    Deleted by Amendment No. 45.
15.6.5-19    Deleted by Amendment No. 45.
15.6.5-20    Deleted by Amendment No. 45.
15.6.5-21    Deleted by Amendment No. 45.
15.6.5-22    Deleted by Amendment No. 45.
Amendment 63                                                                Page 141 of 151
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 15.6.5-23    Murphy, K. G., Campe, K. M., Nuclear Power Plant Control Room Ventilation System Design For Meeting General Criterion, 13TH AEC Air Cleaning Conference (1973).
15.6.5-24    Deleted by Amendment No. 51.
15.6.5-25    Deleted by Amendment No. 46.
15.6.5-26    Deleted by Amendment No. 45.
15.6.5-27    Deleted by Amendment No. 45.
15.6.5-28    Deleted by Amendment No. 28.
15.6.5-29    WASH 1258 "Numerical Guides for Design Objective and Limiting Conditions for Operation to Meet the Criterion As Low As Practicable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents." Volume 2, July 1973, U.S. Atomic Energy Commission.
15.6.5-30    ORNL-TM-212, Part IV, "Design Considerations of Reactor Containment Spray Systems. Calculation of Iodine-Water Partition Coefficients." L. F. Parsly, January 1970, U.S. Atomic Energy Commission.
15.6.5-31    Deleted by Amendment No. 45.
15.6.5-32    Deleted by Amendment No. 45.
15.6.5-33    Deleted by Amendment No. 46.
15.6.5-34    "RWST Boron Concentration Increase Licensing Documentation," Westinghouse Letter 92CP*-G-0018, February 17, 1992.
15.6.5-35    Deleted by Amendment No. 46.
15.6.5-36    "EMF-2103(P)(A) Revision 0, Realistic Large Break LOCA Methodology,"
Framatome ANP, Inc., April 2003.
15.6.5-37    Deleted by Amendment No. 51.
15.6.5-38    Deleted by Amendment No. 51.
15.6.5-39    Deleted by Amendment No. 46.
15.6.5-40    XN-NF-82-49(P)(A), Revision 1, "Exxon Nuclear Company Evaluation Model-EXEM PWR Small Break Model," April 1989.
15.6.5-41    XN-NF-82-49(P)(A), Revision 1, Supplement 1 , "Exxon Nuclear Company Evaluation Model- Revised EXEM PWR Small Break Model," December 1994.
Amendment 63                                                                  Page 142 of 151
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 15.6.5-42      ANF-88-054(P)(A), "PDC-3: Advanced Nuclear Fuels Corporation Power Distribution Control for Pressurized Water Reactors and Application of PDC-3 to H. B. Robinson Unit 2," Advanced Nuclear Fuels Corporation, October, 1990.
15.6.5-43      EOP-E-0, Revision 0, Reactor Trip or Safety Injection.
15.6.5-44      Deleted by Amendment No. 54.
15.6.5-45      Deleted by Amendment No. 50.
15.6.5-46      Deleted by Amendment No. 50.
15.6.5-47      Deleted by Amendment No. 50.
15.6.5-48      "AREVA NP Document EMF-2328(P)(A) Revision 0, PWR Small Break LOCA Evaluation Model, S-RELAP5 Based, March 2001," including ERRATA (January 2008).
15.6.5-49      Letter from D. Brown (AREVA) to B. Morgen (Progress Energy),"Evaluation of Condition Reports 2009-2360 and 2009-2309 for Potential Reporting Under 10 CFR 50.46; PCT Evaluations (Harris NP)," FAB09-532, dated July 10, 2009.
15.6.5-50      ANP-3011, "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis,"
August 2011.
15.6.5-51      Letter from C. Burton (Progress Energy) to NRC (HNP-12-023 dated February 23, 2012, "License Amendment Request for Revision to Technical Specification Core Operating Limits Report References for Realistic Large Break LOCA Analysis Response to Request for Additional Information."
15.6.5-52      WCAP-15398 / WCAP-15399, "Carolina Power and Light Harris Nuclear Plant Steam Generator Replacement/Uprating Analysis and Licensing Project NSSS Licensing Report," 9/00.
15.6.5-53      Letter from B.C. Waldrep (Duke Energy) to NRC (Serial HNP-15-038) dated December 17, 2015, "License Amendment Request for Main Steam Safety Valve Lift Setting Tolerance Change." (Safety Evaluation Report received by {{letter dated|date=July 25, 2016|text=letter dated July 25, 2016}}).
15.7    RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1    RADIOACTIVE WASTE GAS SYSTEM LEAK OR FAILURE 15.7.1.1    Identification of Causes The most limiting waste gas accident is defined as an unexpected and uncontrolled release to the atmosphere of the radioactive xenon and krypton fission gases that are stored in one operating waste gas decay tank with maximum curie content as discussed below. Two waste gas decay tanks may be cross-connected for a short period of time to transfer tank contents only if both tanks are isolated from Waste Gas System influents and the total curie content is Amendment 63                                                                    Page 143 of 151
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 less than the maximum curie content specified in this section. Rupture of these two tanks while cross-connected is limited by the results of this analysis. The Gaseous Waste Processing System (GWPS) is described in Section 11.3.
Since the components of the GWPS are subjected to pressures no greater than 100 psig, a failure is considered to be unlikely (design pressure 150 psig). A failure probability is further reduced since the GWPS is Safety Class 3, Seismic Category I, except for compressors which are classified as Rad-Q.
However, the probability of an accidental release resulting from such events as (1) failure of gas decay tank or associated piping, or (2) lifting and subsequent failure of a relief valve to close, is defined as a limiting fault and is analyzed to define the upper limit of a gaseous release that could result from any malfunction in the GWPS.
15.7.1.2    Analysis of Events and Consequences Assumptions and methods used in this analysis are consistent with those of Regulatory Guide 1.24 (3/23/72) as discussed in Section 1.8. Table 15.7.1-1 lists the conservative assumptions used in the analysis.
It is assumed that the plant has been operating at the power level of 2958 MWt (rated power of 2948 MWt with 0.34% uncertainty) with one percent failed fuel for an extended period sufficient to achieve equilibrium radioactive concentrations in the Reactor Coolant System. As soon as possible after shutdown, all noble gases have been removed from the Reactor Coolant System and transferred to the gas decay tank which is assumed to release its contents in an uncontrolled manner. Radiological decay is assumed only for the minimum time period required to transfer the gases from the Reactor Coolant System to the waste gas decay tank. The release is assumed to occur immediately upon completion of the waste gas transfer, releasing the entire contents of the tank to the Waste Processing Building. All of the noble gases are assumed to leak out of the building at ground level over a two hour period. The activity released to the Waste Processing Building and subsequently to the environment is given in Table 15.7.1-2.
15.7.1.3    Radiological Consequences Analysis 15.7.1.3.1 Offsite Doses The offsite doses are calculated using the assumptions and equations in Section 15.0A.1.
15.7.1.3.2 Control Room Doses The control room assumptions are provided in Section 15.6.5.4.3 and Table 15.6.5-15.
It is assumed that the control room HVAC system begins in the normal operational mode. The activity level causes a high radiation signal almost immediately. It is conservatively assumed that the post-accident recirculation control room HVAC mode is entered 15 seconds after event initiation. The control room is assumed to be placed in the pressurization mode by operator action at 2 hours after isolation signal.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.7.1.3.3 Results The radiological analysis results for the gas decay tank rupture are listed in Table 15.7.1-2. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The resultant doses are within the guideline values.
The offsite dose limit for a gas decay tank rupture is given in HNP Technical Specifications 6.8.4j as 0.5 rem whole body. This translates to a dose limit of 0.5 rem TEDE. The 10CFR50.67 limit in the control room is 5 rem TEDE.
15.7.1.4    Conclusions The dose analysis results in Table 15.7.1-2 are within the acceptance limits established for this event.
15.7.2    LIQUID WASTE SYSTEM LEAK OR FAILURE HNP Liquid Waste System leaks or failures are bounded by the Analyzed Liquid Tank Failure in Section 15.7.3.
15.7.3    POSTULATED RADIOACTIVE RELEASES DUE TO LIQUID TANK FAILURE The consequences of the postulated failure of a tank containing potentially contaminated liquid on the nearest potable water supply and the nearest surface water in an unrestricted area are discussed in Sections 2.4.12 and 2.4.13.
15.7.3.1    Results and Conclusions The analysis of this event is based on operation at 2958 MWt (rated power of 2948 MWt with 0.34% uncertainty) with 1% failed fuel. The results shown in Sections 2.4.12 and 2.4.13 meet their acceptance criteria.
15.7.4    DESIGN BASIS FUEL HANDLING ACCIDENTS 15.7.4.1    Identification of Causes and Accident Description The possibility of a fuel handling accident is remote because of the many interlocks, administrative controls, and physical limitations imposed on the fuel handling operations. All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a senior reactor operator (SRO). The analyzed Fuel Handling Accident inside containment involves dropping a spent fuel assembly resulting in the rupture of the cladding of all the fuel rods (264) in the assembly.
The projected worst case Fuel Handling Accident (FHA) in the Fuel Handling Building (FHB) involves dropping a recently discharged (100 hr decayed) PWR assembly (including the handling tool) on top of another recently discharged PWR assembly in a fuel storage rack. The dropped assembly subsequently falls over landing on BWR fuel assemblies in an adjacent storage rack. Fifty fuel rods are projected to fail in the impacted PWR assembly in storage and all of the rods (264) in the dropped assembly fail when the assembly falls over (Reference 15.7.4.5). Due to the upper bail handle of the BWR fuel assemblies extending above the top of Amendment 63                                                                      Page 145 of 151
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 the BWR storage racks, up to 52 BWR assemblies could be impacted when the dropped PWR assembly falls over. All of the rods in the impacted BWR assemblies are assumed to fail.
15.7.4.2    Radiological Consequences Analysis 15.7.4.2.1 Input Assumptions Common to both FHA in the FHB and in Containment Consistent with Regulatory Guide 1.183 (Position 1.2 of Appendix B), the radionuclides considered are xenons, kryptons, halogens, cesiums and rubidiums. The list of xenons, kryptons, and halogens considered is given in Tables 15.7.4-1 and 15.7.4-3. The cesium and rubidium are not included because they are not assumed to be released from the pool as discussed later.
The calculation of the radiological consequences following a FHA uses gap fractions of 8% for I-131, 30% for Kr-85, and 5% for all other pertinent nuclides (Reference 15.7.4-11).
Iodine species in the pool is 99.85% elemental and 0.15% organic iodine. This is based on the split leaving the fuel of 95% cesium iodide (CsI), 4.85% elemental iodine and 0.15% organic iodine. It is assumed that all CsI is dissociated in the water and re-evolves as elemental. This is assumed to occur instantaneously. Thus, 99.85% of the iodine released is elemental.
The water above the damaged fuel rods retains a large fraction of the gap activity of iodines. An overall effective decontamination factor (DF) of 200 is used.
The cesium and rubidium released from the damaged fuel rods is assumed to remain in a nonvolatile form and would not be released from the pool.
15.7.4.2.2 Postulated Fuel Handling Accident in the FHB The major assumptions and parameters used in the analysis are itemized in Table 15.7.4-1.
This analysis involves dropping a recently discharged (100 hour decay) PWR fuel assembly onto 52 Brunswick BWR fuel assemblies. This analysis also includes 50 PWR rods additionally damaged in the accident. the assembly inventory is based on the assumption that the PWR fuel assembly has been operated at 1.73 times the core average power and the BWR fuel assemblies have been operated at 1.5 times the core average power. All activity released from the fuel pool is assumed to be released to the atmosphere in two hours.
The BWR fuel inventory was conservatively evaluated at the IF-300 spent fuel shipping cask limits for GE- 7, 8, 9, 10, and 13 fuel assemblies with a maximum average lattice enrichment of 4.25 wt. % U-235 and a maximum assembly average burnup of 45 GWD/MTU. The decay time used in the analysis is 100 hours for the PWR fuel and 4 years for the BWR fuel. Thus, the analysis supports the design basis limit of 100 hours decay time prior to fuel movement.
It was determined that for the HNP specific water height above the failed fuel in the fuel handling building of 21 feet, the elemental DF would be at least 291, compared to the Reg.
Guide 1.183 allowable elemental DF of 500. Using the elemental DF 291, it was determined that overall effective DF for 21 feet of coverage would be 203. Since this continues to exceed the Reg. Guide 1.183 cited overall effective DF of 200, it remains conservative to use the overall DF of 200 in the HNP dose calculations.
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 No credit is taken for removal of iodine by filters by the spent fuel pool ventilation system operation. Credit is not taken for isolation of release paths.
The activity released from the damaged assemblies is assumed to be released to the fuel building and subsequently to the atmosphere over a 2 hour period.
15.7.4.2.3 Postulated Fuel Handling Accident in Containment A fuel assembly is assumed to be dropped in containment and damaged during refueling.
Activity released from the damaged assembly is released to the outside atmosphere through the containment openings (such as the personnel air lock door or the equipment hatch).
The major assumptions are parameters used in the analysis are itemized in Table 15.7.4-3.
This analysis involves dropping a recently discharged (100 hour decay) PWR fuel assembly. All activity released from the fuel pool is assumed to be released to the atmosphere in two hours.
The pool referred to in RG 1.183 is interpreted as the flooded reactor cavity for the purposes of evaluating the fuel handling accident in containment. No credit is taken for isolation of containment for the FHA containment.
The calculation of the radiological consequences following a FHA uses gap fractions of 8% for I-131, 30% for Kr-85, and 5% for all other pertinent nuclides (Reference 15.7.4-11).
It is assumed that all of the fuel rods in the equivalent of one fuel assembly (264 rods) are damaged to the extent that all their gap activity is released. The assembly inventory is based on the assumptions that the subject fuel assembly has been operated at 1.73 times the core average power.
The decay time used in the analysis is 100 hours.
It was determined that for HNP specific water height above the failed fuel in the containment of 22 feet, the elemental DF would be at least 382, compared to the Reg. Guide 1.183 allowable elemental DF of 500. Using the elemental DF of 382, it was determined that the overall effective DF for 22 feet of coverage would be 243. Since this continues to exceed the Reg.
Guide 1.183 cited overall effective DF of 200, it remains conservative to use the overall DF of 200 in the HNP dose calculations.
No credit is taken for removal of iodine by filters nor is credit taken for isolation of release paths.
Although the containment purge will be automatically isolated on a purge line high radiation alarm, isolation is not modeled in the analysis. The activity released from the damaged assembly is assumed to be released to the outside atmosphere over a 2 hour period. Since no filters or containment isolation is modeled, this analysis supports refueling operation with the equipment hatch or personnel air lock remaining open.
15.7.4.2.4 Offsite Doses The offsite doses are calculated using the assumptions and equations in Section 15.0A.1.
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Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 15.7.4.2.5 Control Room Doses The control room assumptions are provided in Section 15.6.5.4.3 and Table 15.6.5-15. The FHA control room doses modeled 300 cfm unfiltered inleakage.
It is assumed that the control room HVAC system begins in normal mode. The activity level in the intake duct causes a high radiation signal almost immediately. It is conservatively assumed that the post-accident recirculation control room HVAC mode is entered 15 seconds after event initiation. The control room HVAC is placed into pressurization mode at 2 hours after isolation signal.
15.7.4.2.6 Results The radiological analysis results for the FHA in FHB doses are listed in Table 15.7.4-2. The FHA in Containment doses are listed in Table 15.7.4-4. The TEDE doses have been analyzed for the worst two hours at the EAB and for the duration of the event at the LPZ and in the control room. The resultant doses are within the applicable limits. The offsite doses are less than
~25% of the 10CFR50.67 limits (i.e., 6.3 rem TEDE) and the control room dose is less than the 10CFR50.67 limit of 5 rem TEDE.
15.7.4.3    DELETED 15.7.4.3.1 DELETED 15.7.4.3.2 DELETED 15.7.4.4    Deleted 15.7.4.4.1 Deleted 15.7.4.4.2 Deleted 15.7.4.4.3 Deleted 15.7.4.4.4 Deleted 15.7.4.5    Other Fuel Handling Accidents Fuel handling drop accidents involving the other fuel handling tools (BPRA, RCCA change tool, spent fuel handling tool), and items carried by the tools have also been evaluated (Reference 15.7.4-7) and are addressed in Section 9.1. The tool drop scenarios involve dropping the tools, and items carried by the tools, onto PWR spent fuel racks, BWR spent fuel racks, and combinations of both. For all cases evaluated, the off-site dose consequences were determined to be bounded by the Fuel Handling Accident described in FSAR Section 15.7.4.2.2 which addresses a fuel handling drop accident which results in damage to 314 PWR spent fuel rods and 52 BWR spent fuel assemblies (Reference 15.7.4-7, pages 3.2.2-3.23).
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Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 15.7.5    SPENT FUEL CASK DROP ACCIDENTS 15.7.5.1    Cask Drop Into the New or Spent Fuel Pool As discussed in Section 9.1, the cask handling crane is prohibited from traveling over the new and spent fuel pools or any unprotected safety related equipment. Thus, an accident resulting from dropping a cask or other major load into the new or spent fuel pools is not credible.
15.7.5.2    Cask Drop to Flat Surface The cask has full integrity when the head is fully tensioned and the valve box covers are installed.
15.7.5.2.1 Cask with full integrity As discussed in Section 9.1, the potential drop of a spent fuel cask is limited to less than an equivalent 30 ft. drop onto a flat, essentially unyielding, horizontal surface. Since the spent fuel cask, with the valve box covers installed and the head fully tensioned, is designed to withstand such loadings, the radiological consequences of dropping the cask in this condition are not evaluated.
15.7.5.2.2 Cask with less than full integrity The loaded IF-300 series cask may be moved with the valve covers removed and, from the decon pit to the unloading pool, with only four cask head bolts installed. An evaluation of a 30-ft. drop during the movement from the decon pit to the unloading pool was performed and indicated that, while fuel components would be retained in the cask, the cask is not expected to be gas tight. Noble gas and iodine gap activity could be released to the Fuel Handling Building and subsequently to the environment. Damage to the valves caused by dropping the cask could cause the same type of release. The radiological consequences from this release were analyzed for an IF-300 series cask. This analysis utilized the worst case fuel types anticipated to be shipped in the cask, as shown in Table 15.7.5-1. The results of this analysis show that these consequences would be a small fraction of the 10CFR100 exposure guidelines. The analysis is documented in Harris Plant calculation HNP-M/FHB-1001.
 
==REFERENCES:==
SECTION 15.7 15.7.4-1        Industrial Ventilation, 8th Edition, American Conference of Governmental Industrial Hygienists.
15.7.4-2        Deleted by Amendment No. 49 15.7.4-3        Deleted by Amendment No. 51 15.7.4-4        Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000.
15.7.4-5        Westinghouse Letter, 97CP-G-0006; Christine M. Vertes to Leo Martin, dated April 9, 1997, "Limiting Fuel Handling Accident Assumptions."
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Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 15.7.4-6        Deleted by Amendment No. 49 15.7.4-7        ESR 98-00181 "Fuel Handling Tool Drop onto Spent Fuel Rack Evaluation" 15.7.4-8        Deleted by Amendment No. 51 15.7.4-9        Deleted by Amendment No. 51 15.7.4-10        CP&L Calculation HNP-M/FHB-1001 "Off-site Doses from FHB Cask Drop."
15.7.4-11        Letter from NRC (ADAMS Accession No. ML18045A060) to T.M. Hamilton (Duke Energy), Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Regarding a New Set of Fission Gas Gap Release Fractions for High Burnup Fuel Rods that Exceed the Linear Heat Generation Rate Limit Detailed in Regulatory Guide 1.183, Table 3, Footnote 11 (CAC No. MF9740; EPID L-2017-LLA-0233), dated March 26, 2018 15.8    ANTICIPATED TRANSIENTS WITHOUT SCRAM An anticipated transient without scram (ATWS) is an anticipated operational occurrence (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power) that is accompanied by a failure of the reactor trip system to shut down the reactor. A series of generic studies (References 15.8.0-1, 15.8.0-2) on ATWS showed that acceptable consequences would result provided that the turbine trips and auxiliary feedwater flow is initiated in a timely manner.
The final USNRC ATWS rule (Reference 15.8.0-3) requires that all US Westinghouse-designed plants install ATWS mitigation system actuation circuitry (AMSAC) to initiate a turbine trip and actuate auxiliary feedwater independent of the reactor trip system. SHNPP is in compliance with the final ATWS rule by virtue of having installed a NRC-approved AMSAC. For the revised steam generator replacement/uprating conditions, the generic studies in References 15.8.0-1 through 15.8.0-3 have been shown to remain applicable to HNP (Reference 15.8.0-4).
For the Measurement Uncertainty Recapture (MUR) Power Uprate, the generic studies were updated to reflect the MUR plant conditions, and to include plant-specific AMSAC setpoints and delay times. The analyses demonstrated that the effects of the MUR would not result in unacceptable consequences (Reference 15.8.0-5).
 
==REFERENCES:==
SECTION 15.8 15.8.0-1        Burnett, T.W.T., et al., "Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August 1974.
15.8.0-2        Letter from T. M. Anderson (Westinghouse) to S. H. Hanauer (USNRC),
                "Anticipated Transients Without Scram for Westinghouse Plants," NS-TMA-2182, December 1979.
15.8.0-3        ATWS Final Rule, Code of Federal Regulations 10 CFR 50.62, "Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants."
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Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 15.8.0-4      "Steam Generator Replacement/Uprating Analysis and Licensing Project - NSSS Licensing Report," WCAP-15398 (Proprietary) and WCAP-15399 (Non-proprietary), June 2000.
15.8.0-5      "Harris Nuclear Plant (CQL) Engineering Report for the Anticipated Transient Without Scram (ATWS) Update Program, LTR-TA-18-1, Revision 1, January 2018.
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Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 TABLE                                        TITLE 15.0.1-1  ACCIDENT CATEGORY USED FOR EACH CHAPTER 15 EVENT 15.0.3-1  DELETED BY AMENDMENT NO. 48 15.0.3-2  DELETED BY AMENDMENT NO. 48 15.0.3-3  DELETED BY AMENDMENT NO. 48 15.0.3-4  DELETED BY AMENDMENT NO. 48 15.0.3-5  COMPONENT RESPONSE TIME, SETPOINT, AND CAPACITY UTILIZED IN ACCIDENT ANALYSES 15.0.3-6  DELETED BY AMENDMENT NO. 63 15.0.6-1  DELETED BY AMENDMENT NO. 48 15.0.6-2  TRIP SETPOINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES 15.0.7-1  DELETED BY AMENDMENT NO. 48 15.0.8-1  PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS 15.0.9-1  CORE TOTAL FISSION PRODUCT ACTIVITIES 15.0.9-2  RCS COOLANT FISSION PRODUCT CONCENTRATIONS 15.0.9-3  COMMITTED EFFECTIVE DOSE EQUIVALENT DOSE CONVERSION FACTORS 15.0.9-4  EFFECTIVE DOSE EQUIVALENT DOSE CONVERSION FACTORS 15.0.9-5  NUCLIDE DECAY CONSTANTS 15.0.9-6  IODINE SPIKE APPEARANCE RATES (CURIES/MINUTE) 15.0.9-7  IODINE SPECIFIC ACTIVITIES ( Ci/gm) 15.0.13-1 SINGLE FAILURES ASSUMED FOR ACCIDENTS OF MODERATE 15.0.13-2 SINGLE FAILURES FOR NON-CONDITION II EVENTS 15.1.2-1  INPUT PARAMETERS AND BIASING FOR INCREASE IN FEEDWATER FLOW 15.1.2-2  KEY OPERATING PARAMETERS FOR INCREASE IN FEEDWATER FLOW 15.1.2-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR INCREASE IN FEEDWATER FLOW 15.1.2-4  EVENT
 
==SUMMARY==
FOR INCREASE IN FEEDWATER FLOW - LIMITING CASE (HFP, EOC, MANUAL ROD CONTROL) 15.1.2-5  DELETED BY AMENDMENT NO. 51 15.1.3-1  INPUT PARAMETERS AND BIASING FOR INCREASE IN STEAM FLOW Amendment 63                                                                  Page 1 of 7
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE                                      TITLE 15.1.3-2  KEY OPERATING PARAMETERS FOR INCREASE IN STEAM FLOW 15.1.3-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR INCREASE IN STEAM FLOW 15.1.3-4  EVENT
 
==SUMMARY==
FOR INCREASE IN STEAM FLOW - (EXCESS LOAD) - MINIMUM (BOC)
FEEDBACK CASE 15.1.4-1  DELETED BY AMENDMENT NO. 48 15.1.5-1  EQUIPMENT REQUIRED FOLLOWING A RUPTURE OF A MAIN STEAM LINE 15.1.5-2  DELETED BY AMENDMENT NO. 48 15.1.5-3  SEQUENCE OF EVENTS FOR LIMITING MAIN STEAM LINE BREAK CASE - HZP WITH OFFSITE POWER WITH THE STUCK ROD 15.1.5-4  DELETED BY AMENDMENT NO. 51 15.1.5-5  PARAMETERS USED IN STEAM LINE BREAK RADIOLOGICAL ANALYSIS 15.1.5-6  RADIOLOGICAL CONSEQUENCES OF A POSTULATED MAIN STEAM LINE BREAK 15.1.5-7  STEAM LINE BREAK 15.2.3-1  INPUT PARAMETERS AND BIASING FOR TURBINE TRIP 15.2.3-2  KEY OPERATING PARAMETERS FOR TURBINE TRIP 15.2.3-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR TURBINE TRIP 15.2.3-4  EVENT
 
==SUMMARY==
FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE 15.2.3-5  EVENT
 
==SUMMARY==
FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE 15.2.3-6  EVENT
 
==SUMMARY==
FOR TURBINE TRIP MDNBR CASE 15.2.6-1  INPUT PARAMETERS AND BIASING LOSS OF NON- EMERGENCY AC POWER TO THE STATION AUXILIARIES 15.2.6-2  KEY OPERATING PARAMETERS FOR LOSS OF NON- EMERGENCY AC POWER TO THE STATION AUXILIARIES 15.2.6-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR LOSS OF NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES 15.2.6-4  EVENT
 
==SUMMARY==
FOR LOSS OF NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES 15.2.6-5  PARAMETERS USED IN LOSS OF AC POWER RADIOLOGICAL ANALYSIS 15.2.6-6  RADIOLOGICAL CONSEQUENCES OF A LOSS OF NON-EMERGENCY AC POWER TO THE PLANT AUXILIARIES Amendment 63                                                                  Page 2 of 7
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE                                      TITLE 15.2.7-1  INPUT PARAMETERS AND BIASING LOSS OF NORMAL FEEDWATER FLOW 15.2.7-2  KEY OPERATING PARAMETERS FOR LOSS OF NORMAL FEEDWATER 15.2.7-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR LOSS OF NORMAL FEEDWATER 15.2.7-4  EVENT
 
==SUMMARY==
FOR LOSS OF NORMAL FEEDWATER 15.2.8-1  INPUT PARAMETERS AND BIASING FOR FEEDWATER LINE BREAK 15.2.8-2  KEY OPERATING PARAMETERS FOR FEEDWATER LINE BREAK EVENT 15.2.8-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR FEEDWATER LINE BREAK EVENT 15.2.8-4  EVENT
 
==SUMMARY==
FOR FEEDLINE BREAK WITH OFFSITE POWER AVAILABLE 15.2.8-5  EVENT
 
==SUMMARY==
FOR FEEDLINE BREAK WITH LOSS OF OFFSITE POWER 15.2.8-6  FEEDLINE BREAK 15.2.8-7  EVENT
 
==SUMMARY==
FOR FEEDLINE BREAK LONG-TERM CORE COOLING WITH OFFSITE POWER LOST 15.2.8-8  EVENT
 
==SUMMARY==
FOR FEEDLINE BREAK PEAK PRIMARY PRESSURE 15.3.1-1  DELETED BY AMENDMENT NO. 48 15.3.2-1  INPUT PARAMETERS AND BIASING FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-2  KEY OPERATING PARAMETERS FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-4  DELETED BY AMENDMENT NO. 63 15.3.3-1  INPUT PARAMETERS AND BIASING FOR LOCKED ROTOR 15.3.3-2  KEY OPERATING PARAMETERS FOR LOCKED ROTOR 15.3.3-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR LOCKED ROTOR 15.3.3-4  EVENT
 
==SUMMARY==
FOR LOCKED ROTOR OVERPRESSURIZATION CASE 15.3.3-5  EVENT
 
==SUMMARY==
FOR LOCKED ROTOR MDNBR CASE 15.3.3-6  PARAMETERS USED IN THE LOCKED ROTOR RADIOLOGICAL ANALYSIS 15.3.3-7  RADIOLOGICAL CONSEQUENCES OF A LOCKED ROTOR Amendment 63                                                              Page 3 of 7
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE                                        TITLE 15.4.1-1  INPUT PARAMETERS AND BIASING FOR UNCONTROLLED BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITION 15.4.1-2  KEY OPERATING PARAMETERS FOR UNCONTROLLED BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITION 15.4.1-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR UNCONTROLLED BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITIONS 15.4.1-4  EVENT
 
==SUMMARY==
FOR UNCONTROLLED RCCA BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER STARTUP CONDITIONS 15.4.1-5  DELETED BY AMENDMENT NO. 51 15.4.2-1  DNB ANALYSIS INPUT PARAMETERS AND BIASING FOR UNCONTROLLED BANK WITHDRAWAL AT POWER 15.4.2-1a PRIMARY SIDE OVERPRESSURIZATION ANALYSIS INPUT PARAMETERS AND BIASING FOR UNCONTROLLED BANK WITHDRAWAL AT POWER 15.4.2-2  DNB ANALYSIS KEY OPERATING PARAMETERS FOR UNCONTROLLED BANK WITHDRAWAL AT POWER 15.4.2-2a PRIMARY SIDE OVERPRESSURIZATION ANALYSIS KEY OPERATING PARAMETERS FOR UNCONTROLLED BANK WITHDRAWAL AT POWER 15.4.2-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR UNCONTROLLED BANK WITHDRAWAL AT POWER 15.4.2-4  EVENT
 
==SUMMARY==
FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT POWER LIMITING MDNBR CASE (BOC, 10.8 PCM/SEC INSERTION RATE) 15.4.2-5  EVENT
 
==SUMMARY==
FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT POWER LIMITING PRIMARY SIDE OVERPRESSURIZATION CASE WITH BOC KINETICS AND THE MOST LIMITING REACTIVITY INSERTION RATE 15.4.3-1  INPUT PARAMETERS AND BIASING FOR ROD DROP 15.4.3-2  KEY OPERATING PARAMETERS FOR ROD DROP 15.4.3-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR ROD DROP 15.4.3-4a EVENT
 
==SUMMARY==
FOR A LIMITING DROPPED ROD CASE AT EOC 15.4.3-5  PARAMETERS USED IN THE SINGLE RCCA WITHDRAWAL ACCIDENT 15.4.3-6  INPUT PARAMETERS AND BIASING FOR SINGLE RCCA WITHDRAWAL 15.4.3-6a RADIOLOGICAL CONSEQUENCES OF A SINGLE RCCA WITHDRAWAL 15.4.3-7  KEY OPERATING PARAMETERS FOR SINGLE RCCA WITHDRAWAL 15.4.3-8  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR SINGLE RCCA WITHDRAWAL Amendment 63                                                                Page 4 of 7
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE                                      TITLE 15.4.3-9  EVENT
 
==SUMMARY==
FOR 100% RTP SINGLE RCCA WITHDRAWAL REPRESENTATIVE CASE 15.4.4-1  DELETED BY AMENDMENT NO. 51 15.4.6-1  DELETED BY AMENDMENT NO. 45 15.4.6-2  ADMINISTRATIVE CONTROLS TO PREVENT DILUTION 15.4.6-3  DELETED BY AMENDMENT NO. 48 15.4.7-1  KEY OPERATING PARAMETERS FOR INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITION 15.4.8-1  INPUT PARAMETERS AND BIASING FOR ROD EJECTION ACCIDENTS 15.4.8-2  KEY OPERATING PARAMETERS FOR ROD EJECTION ACCIDENTS (MDNBR CASES) 15.4.8-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR ROD EJECTION ACCIDENTS 15.4.8-4a DELETED BY AMENDMENT NO. 56 15.4.8-4b EVENT
 
==SUMMARY==
FOR CONTROL ROD EJECTION LIMITING CASE 15.4.8-5  PARAMETERS USED FOR THE ROD CLUSTER ASSEMBLY EJECTION RADIOLOGICAL ANALYSIS 15.4.8-6  RADIOLOGICAL CONSEQUENCES OF A ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT 15.5.1-1  ASSUMED STATE OF PLANT SYSTEMS 15.5.1-2  INPUT BIASING 15.5.1-3  SEQUENCE OF EVENTS (BOC MDNBR CASE) 15.5.1-4  SEQUENCE OF EVENTS (EOC MDNBR CASE) 15.5.1-5  SEQUENCE OF EVENTS (PRESSURIZER OVERFILL CASE) 15.6.1-1  INPUT PARAMETERS AND BIASING FOR INADVERTENT OPENING OPENING OF A PRESSURIZER PRESSURE SAFETY OR PORV 15.6.1-2  KEY OPERATING PARAMETERS FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR PORV 15.6.1-3  RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR INADVERTENT OPENING OF A PRESSURIZER SAFETY OR PORV 15.6.1-4  EVENT
 
==SUMMARY==
FOR INADVERTENT OPENING OF A PRESSURIZER SAFETY OR PORV 15.6.2-1  PARAMETERS USED FOR LETDOWN LINE BREAK OUTSIDE CONTAINMENT RADIOLOGICAL ANALYSIS Amendment 63                                                                Page 5 of 7
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE                                        TITLE 15.6.2-2  RADIOLOGICAL CONSEQUENCES OF A LETDOWN LINE BREAK OUTSIDE CONTAINMENT 15.6.3-1  SHNPP SGTR ANALYSIS, OPERATOR ACTION TIMES FOR MARGIN TO OVERFILL ANALYSIS 15.6.3-2  SHNPP SGTR ANALYSIS, SEQUENCE OF EVENTS MARGIN TO OVERFILL ANALYSIS 15.6.3-3  OPERATOR ACTION TIMES FOR SGTR OFFSITE DOSE ANALYSIS 15.6.3-4  SEQUENCE OF EVENTS OFFSITE DOSE ANALYSIS 15.6.3-5  SGTR MASS RELEASE RESULTS 15.6.3-6  PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE 15.6.3-7  DELETED BY AMENDMENT NO. 51 15.6.3-8  DELETED BY AMENDMENT NO. 51 15.6.3-9  DELETED BY AMENDMENT NO. 51 15.6.3-10  OFFSITE ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES 15.6.3-11  DELETED BY AMENDMENT NO. 51 15.6.3-12  DELETED BY AMENDMENT NO. 51 15.6.3-13  RADIOLOGICAL CONSEQUENCES OF A SGTR 15.6.5-1  KEY PARAMETERS FOR LBLOCA 15.6.5-2  EVENT TIMES FOR LBLOCA 15.6.5-3  SBLOCA SYSTEM ANALYSIS PARAMETERS 15.6.5-4  DELETED BY AMENDMENT NO. 48 15.6.5-5  DELETED BY AMENDMENT NO. 48 15.6.5-6  DELETED BY AMENDMENT NO. 48 15.6.5-7  DELETED BY AMENDMENT NO. 48 15.6.5-8  DELETED BY AMENDMENT NO. 48 15.6.5-9  DELETED BY AMENDMENT NO. 48 15.6.5-10  DELETED BY AMENDMENT NO. 58 15.6.5-11a SEQUENCES OF EVENTS DURING SBLOCA (LIMITING CASE) 15.6.5-11b SBLOCA ANALYSIS RESULTS (LIMITING CASE)
Amendment 63                                                                Page 6 of 7
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE                                        TITLE 15.6.5-12 PARAMETERS USED FOR LARGE BREAK LOCA RADIOLOGICAL ANALYSIS 15.6.5-13 LBLOCA CORE FISSION PRODUCT RELEASE TIMING 15.6.5-14 LBLOCA CORE FISSION PRODUCT RELEASE FRACTIONS 15.6.5-15 CONTROL ROOM PARAMETERS USED FOR RADIOLOGICAL ANALYSIS 15.6.5-16 RADIOLOGICAL CONSEQUENCES OF A POSTULATED LARGE BREAK LOCA 15.6.5-17 DELETED BY AMENDMENT NO. 63 15.6.5-18 DELETED BY AMENDMENT NO. 63 15.7.1-1  ASSUMPTIONS FOR WASTE GAS DECAY TANK RELEASE ACCIDENT ANALYSIS 15.7.1-2  RADIOLOGICAL CONSEQUENCES OF A WASTE GAS DECAY TANK RELEASE 15.7.2-2  DELETED BY AMENDMENT NO. 51 15.7.2-3  DELETED BY AMENDMENT NO. 51 15.7.4-1  PARAMETERS USED IN FUEL HANDLING ACCIDENT INSIDE THE FUEL HANDLING BUILDING RADIOLOGICAL ANALYSIS 15.7.4-2  RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL HANDLING ACCIDENT IN THE FUEL HANDLING BUILDING 15.7.4-3  PARAMETERS USED IN A FUEL HANDLING ACCIDENT INSIDE CONTAINMENT RADIOLOGICAL ANALYSIS 15.7.4-4  RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL HANDLING ACCIDENT INSIDE CONTAINMENT 15.7.4-5  DELETED BY AMENDMENT NO. 49 15.7.4-6  DELETED BY AMENDMENT NO. 49 15.7.4-7  DELETED BY AMENDMENT NO. 49 15.7.4-8  DELETED BY AMENDMENT NO. 51 15.7.4.9  DELETED BY AMENDMENT NO. 51 15.7.5-1  FUEL CONDITIONS ANALYZED FOR IF-300 SERIES CASK DROP Amendment 63                                                                Page 7 of 7
 
Shearon Harris Nuclear Power Plant                                                        UFSAR Chapter: 15 TABLE 15.0.1-1 ACCIDENT CATEGORY USED FOR EACH CHAPTER 15 EVENT FSAR Event (a)
Designation                                      Event Name                                      Condition 15.1        INCREASE IN HEAT REMOVAL BY SECONDARY SYSTEM 15.1.1      Decrease in Feedwater Temperature                                                    II (AOO) 15.1.2      Increase in Feedwater Flow                                                            II (AOO) 15.1.3      Increase in Steam Flow                                                                II (AOO) 15.1.4      Inadvertent Opening of a Steam Generator Relief or Safety Valve                      II (AOO) 15.1.5      Steam Line Break                                                                      IV (PA) 15.2        DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM 15.2.1      Steam Pressure Regulator Malfunction                                                      N/A 15.2.2      Loss of External Load                                                                II (AOO) 15.2.3      Turbine Trip                                                                          II (AOO) 15.2.4      Inadvertent Closure of MSIV's                                                        II (AOO) 15.2.5      Loss of Condenser Vacuum                                                              II (AOO) 15.2.6      Loss of Nonemergency AC Power                                                        II (AOO) 15.2.7      Loss of Normal Feedwater                                                              II (AOO) 15.2.8      Feedline Break                                                                        IV (PA) 15.3        DECREASE IN REACTOR COOLANT SYSTEM FLOW 15.3.1      Partial Loss of Forced Reactor Coolant Flow                                          II (AOO) 15.3.2      Complete Loss of Forced Reactor Coolant Flow                                          III (PA) 15.3.3      RCP Shaft Seizure (Locked Rotor)                                                      IV (PA) 15.3.4      RCP Shaft Break                                                                        IV (PA) 15.4        REACTIVITY AND POWER DISTRIBUTION ANOMALIES 15.4.1      Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup            II (AOO)
Condition 15.4.2      Uncontrolled RCCA Bank Withdrawal at Power                                            II (AOO) 15.4.3      RCCA Misoperation
: 1) Dropped Rod/Bank                                                                II (AOO)
: 2) Single Rod Withdrawal                                                            III (PA)
: 3) Statically Misaligned RCCA                                                      II (AOO) 15.4.4      Startup of an Inactive RCP at an Incorrect Temperature                                II (AOO) 15.4.5      A Malfunction or Failure of the Flow Controller in a BWR Loop that Results in an          N/A Increased Reactor Coolant Flow Rate 15.4.6      CVCS Malfunction that Results in a Decrease in the Boron Concentration in the        II (AOO)
Reactor Coolant 15.4.7      Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position          III (PA) 15.4.8      Spectrum of Rod Cluster Control Assembly Ejection Accidents                            IV (PA) 15.5        INCREASE IN REACTOR COOLANT INVENTORY 15.5.1      Inadvertent Operation of the ECCS During Power Operation                              II (AOO) 15.5.2      CVCS Malfunction that Increases RCS Inventory                                        II (AOO) 15.6        DECREASE IN REACTOR COOLANT INVENTORY 15.6.1      Inadvertent Opening of a Pressurizer Safety or PORV                                  II (AOO)
Amendment 61                                                                                    Page 1 of 2
 
Shearon Harris Nuclear Power Plant                                                    UFSAR Chapter: 15 TABLE 15.0.1-1 ACCIDENT CATEGORY USED FOR EACH CHAPTER 15 EVENT FSAR Event (a)
Designation                                    Event Name                                    Condition 15.6.2      Radiological Consequences of the Failure of Small Lines Carrying Primary Coolant  II (AOO)
Outside Containment 15.6.3      Radiological Consequences of Steam Generator Tube Rupture                          IV (PA) 15.6.4      Radiological Consequences of a Main Steam Line Failure Outside Containment            N/A 15.6.5      Loss of Coolant Accidents
: 1) Large Break LOCA                                                                IV (PA)
: 2) Small Break LOCA                                                                III (PA) 15.7        RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1      Radioactive Waste Gas System Leak or Failure                                        III (PA) 15.7.2      Liquid Waste System Leak or Failure                                                III (PA) 15.7.3      Postulated Radioactive Releases Due to Liquid Tank Failure                          III (PA) 15.7.4      Fuel Handling Accidents                                                            IV (PA) 15.7.5      Spent Fuel Cask Drop Accidents                                                        Not Credible/III Amendment 61                                                                                Page 2 of 2
 
Shearon Harris Nuclear Power Plant                                                                                                UFSAR Chapter: 15 TABLE 15.0.3-5 COMPONENT RESPONSE TIME, SETPOINT, AND CAPACITY UTILIZED IN ACCIDENT ANALYSES Response      Nominal    Setpoint Item                    Time      Setpoint  Uncertainty                                Total Capacity Pressurizer Safety Valves (3)            1.13 sec    2485 psig      +/-3%(a)  minimum required: 380,000 lbm/hr each valve of saturated steam at valve loop seal clearing time                                                    Setpoint Pressurizer Power Operated Relief Valves Compensated PORV (1)                      2 sec      100 psid              236,000 lbm/hr at 2155 psia and saturation temperature Non-Compensated PORVs (2)                  2 sec      2335 psig Steam Line Relief Valves (15)                ---
Group 1                                          1170.0 psig    +/-3%      881,980 lbm/hr(c)
Group 2                                          1185.0 psig    +/-3%      893,160 lbm/hr(c)
Group 3                                          1200.0 psig    +/-3%      904,330 lbm/hr(c)
Group 4                                          1215.0 psig    +/-3%      915,500 lbm/hr(c)
Group 5                                          1230.0 psig    +/-3%      926,670 lbm/hr(c)
Pressurizer Sprays                            ---
Spray Initiates                                    2260 psig              350 gpm/valve at 2485 psig and 650&deg;F Full On                                            2310 psig Main Steam Isolation Valves Signal Delay                            2 sec Closing Time                            5 sec Main Feedwater Isolation Valves Signal Delay                            2 sec Close Time                              8 sec Auxiliary Feedwater                      61.5 sec(d)                          390 gpm(e)
Auxiliary Feedwater Isolation Valves(f)    41 sec Amendment 63                                                                                                                                Page 1 of 2
 
Shearon Harris Nuclear Power Plant                                                                                                    UFSAR Chapter: 15 Table 15.0.3-5 (Continued) a The +/-3% uncertainty associated with the pressurizer safety valve setpoint includes an allowance of +1% for set pressure shift. The remainder of the uncertainty is an allowance for drift (+2% / -3%) which exceeds the allowance provided in Tech. Spec. 3.4.2.2 (+/-1%).
b Deleted c
Capacity at 3% accumulation d
Maximum response time.
e The safety analyses support an AFW flow rate of 374 gpm from 1 MDAFW pump at the lowest lifting MSSV setpoint plus 3% tolerance. See Section 15.0 for more details.
f Values used in Westinghouse analysis of Steam Generator Tube Rupture are slightly different. (See Sections 15.0 and 15.6.3).
Amendment 63                                                                                                                                  Page 2 of 2
 
Shearon Harris Nuclear Power Plant                                                                                                UFSAR Chapter: 15 TABLE 15.0.6-2 TRIP SETPOINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Tech. Spec.              Tech. Spec.            Analysis                      Response Item                        Trip Setpoint          Total Allowance              Value                        Timea Power Range, Neutron Flux High Setting                            108% of RTP        + (0.0458) (120% of RTP) 113.5% of RTP          0.5 sec Low Setting                              25% of RTP          + (0.0783) (120% of RTP) 34.4% of RTP High Positive Flux Rate                    5% of RTP with a    + (0.0233) (120% of RTP) +4.5% of RTP with a    0.5 sec time constant  2 sec                          time constant  2 sec High Pressurizer Water Level              87% of instrument  + (0.08)(100%)          95% of instrument span  2.0 sec span High Pressurizer Pressure                  2385 psig          + (0.04625) (800 psi)    2422 psig              2.0 sec Low Pressurizer Pressure                  1960 psig          - (0.04625) (800 psi)    1923 psig              2.0 sec Lead Time Constant, 4                  2.0 sec                                        1.8 secd Lag Time Constant, 5                  1.0 Sec                                        1.1 sec(d)
Low Primary Coolant Flow                  91.7% of full flow  - (0.0308) (120%)        88% of full flow        1.0 sec Low-Low Steam Generator Level              25.0% of Narrow    - (0.089) (100%)        16.1%  spane            3.5 sec Range Span            - (0.25%) (100%)        0.0 spanf Undervoltage - Reactor Coolant Pumpsg      5148 volts                                                          1.5 sec Underfrequency - Reactor Coolant Pumps    57.5 Hz            - (0.05) (10Hz)          57.0 Hz                0.6 sec Turbine Trip and Main Feedwater Isolation  78.0% of Narrow    + (0.22) (100%)          100% span              2.5 sec (Turbine Trip) on SG Water Level - High-High            Range Span                                                            10 sec (Feedwater Isolation)
Reactor Trip on SI                        SI Actuation                                  SI Actuation          0.0h Over Temperature Ti T0                                                                                    T/To = 1.0          4.75 sec RTD lag time and 1.25 sec delay K1                                      1.175                + (0.09) (150%)          1.32 K2                                      0.0224/&deg;F                                      0.0224/&deg;F K3                                      0.0010/psig                                    0.0010/psig 1                                      0.0 sec                                        0.0 sec 2                                      0.0 sec                                        0.0 sec 3                                      4.0 sec                                        4.4 sec 4                                      22.0 sec              -10%                    19.8 sec 5                                      4.0 sec              +10%                    4.4 sec Amendment 63                                                                                                                                Page 1 of 2
 
Shearon Harris Nuclear Power Plant                                                                                                        UFSAR Chapter: 15 TABLE 15.0.6-2 TRIP SETPOINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Tech. Spec.                            Tech. Spec.                                      Analysis Item                        Trip Setpoint                          Total Allowance                                    Value 6                  0.0 sec                                                                          0.0 sec f1 (I)            0 when +9%  I  -21%                                                            0 when +9%  I  -21%
3.18% per neg. % I                                                              3.18% per neg. % I 1.712% per pos. % I                                                              1.712% per pos. % I Over Power T                                                                                            T/ToAll Chapters = 1.0 T0 K4                  1.10                                      + (0.033) (150%)                      1.15 K5                  0.02/&deg;F for increasing average temperature                                        0.02/&deg;F for increasing average temperature 0.0 for decreasing average temperature                                            0.0 for decreasing average temperature K6                  0.002/&deg;F for T > T                                                              0.002/&deg;F for T > T 0.0 for T  T                                                                    0.0 for T  T 7                  13.0 sec                                                                          11.7 sec f2 (I)            0.0 when +21%  I  -21%                                                        0.0 when +21%  I  -21%
3.5% per negative %I                                                            3.5% per negative %I 3.5% per positive %I                                                            3.5% per positive %I a
The Reactor Trip System Response Time is defined as the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.
b-c Deleted d
A specific undervoltage setpoint was not assumed in the analysis.
e For loss of normal feedwater.
f For feedwater line break.
g A specific undervoltage setpoint was not assumed in the analysis h
The feedwater line break analysis assumes a 0.0 sec reactor trip delay following a 2.0 sec SI signal generation delay.
i The Tech Spec values for the K constants and time constants are provided in the COLR. The Tech Spec values listed here are representative values.
Values used in Westinghouse analysis of Steam Generator Tube Rupture are slightly different. (See Sections 15.0 and 15.6.3).
Amendment 63                                                                                                                                        Page 2 of 2
 
Shearon Harris Nuclear Power Plant                                                                                                    UFSAR Chapter: 15 TABLE 15.0.8-1 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS Incident                  Reactor Trip Functions      ESF Actuation Functions        Other Equipment                  ESF Equipment 15.1 Increase in Heat Removed by the Secondary System Feedwater system malfunctions    Power range high flux, high-high High-high steam generator  Feedwater isolation valves                        -
that result in an increase in    steam generator level, manual    level-produced feed-water feedwater flow                                                    isolation and turbine trip Excessive increase in            Power range high flux, over-                    -          Pressurizer self-actuated safety                  -
secondary steam flow              temperature T, overpower T,                              valves, steam generator safety manual                                                      valves Inadvertent opening of a steam    Low pressurizer pressure, manual Low pressurizer pressure,  Feedwater isolation valves,      Auxiliary Feedwater System, generator relief or safety valve  SIS, overpower T, power range  low compensated steam line steam line isolation valves      Safety Injection System high flux                        pressure, Hi-1 containment pressure, manual Steam system piping failure      SIS, low pressurizer pressure,  Low pressurizer pressure,  Feedwater isolation valves,      Auxiliary Feedwater System, manual, overpower T, power      low compensated steam line steam line pressure, isolation  Safety Injection System, range high flux                  pressure, Hi-1 and Hi-3    valves                          Containment Heat Removal containment pressure,                                      System manual 15.2 Decrease in Heat Removal by the Secondary System Loss of external electrical load/ High pressurizer pressure over-                -          Pressurizer safety valves, steam                  -
turbine trip                      temperature T, manual, steam                              generator safety valves generator low-low level, high pressurizer water level Loss of non-emergency AC          Steam generator low-low level,  Steam generator low-low    Steam generator safety valves    Auxiliary Feedwater System power to the station auxiliaries  manual, OTT                    level Loss of normal feedwater flow    Steam generator low-low level,  Steam generator low-low    Steam generator safety valves    Auxiliary Feedwater System manual, OTT                    level Amendment 63                                                                                                                                      Page 1 of 3
 
Shearon Harris Nuclear Power Plant                                                                                                          UFSAR Chapter: 15 TABLE 15.0.8-1 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS Incident                  Reactor Trip Functions          ESF Actuation Functions          Other Equipment                  ESF Equipment Feedwater system pipe break      Steam generator low-low level,      High Containment pressure,  Steam line isolation valves,    Auxiliary Feedwater System high pressurizer pressure, SIS,      steam generator low-low    feedline isolation, pressurizer  Safety Injection System over-temperature T, manual          water level, low            self-actuated safety valves compensated steam line      steam generator safety valves pressure High steamline differential pressure 15.3 Decrease in Reactor Coolant System Flow Rate Partial & complete loss of forced Low flow, undervoltage                              -            Steam generator safety valves                    -
reactor coolant flow              underfrequency, manual Reactor coolant pump shaft        Low flow, manual                                    -            Pressurizer safety valves, steam                  -
seizure (locked rotor)                                                                            generator safety valves 15.4 Reactivity & Power Distribution Anomalies Uncontrolled rod cluster control  Power range high flux, manual,                      -                            -                                -
assembly bank withdrawal from    high positive flux rate, overpower a subcritical or low power        T, overtemperature T startup condition Uncontrolled rod cluster control  Power range high Overtemperature                    -            Pressurizer safety valves, steam                  -
assembly bank withdrawal at      T, high pressurizer pressure,                                  generator safety valves power                            manual, high pressurizer level, overpower T Rod cluster control assembly      Overtemperature T, manual, high                    -                Main steam safety valves,                    -
misoperation                      pressurizer level                                                    pressurizer safety valves Start up of an inactive reactor  Low flow interlocked with P-8,                      -                            -                                -
coolant loop at an incorrect      manual temperature Chemical & Volume Control        Source range high flux, power                      -            Low insertion limit annunciators                  -
System malfunction that results  range high flux, overtemperature                                for boration, Source Range in a decrease in boron            T, manual                                                      count rate (while shut down) concentration in the reactor coolant Spectrum of rod cluster control  Power range high flux, high positive                -                            -                                -
Amendment 63                                                                                                                                          Page 2 of 3
 
Shearon Harris Nuclear Power Plant                                                                                                        UFSAR Chapter: 15 TABLE 15.0.8-1 PLANT SYSTEMS AND EQUIPMENT AVAILABLE FOR TRANSIENT AND ACCIDENT CONDITIONS Incident                    Reactor Trip Functions        ESF Actuation Functions          Other Equipment                  ESF Equipment assembly ejection accidents        flux rate, manual, overtemperature T, overpower T, low pressurizer pressure 15.5 Increase in Reactor Coolant Inventory Inadvertent operation of the      Low pressurizer pressure, manual                -                              -              Safety Injection System ECCS during power operation        safety injection trip 15.6 Decrease in Reactor Coolant Inventory Inadvertent opening of a          Pressurizer low pressure,                        -                              -                                -
pressurizer safety or relief valve overtemperature T, manual Steam generator tube failure      Low pressurizer pressure,          Engineered Safety Features Service Water System,            Emergency Core Cooling overtemperature T (See Section    Actuation System          Component Cooling Water          System, Auxiliary Feedwater 15.6.3)                                                      System, steam generator shell    System, Emergency Power side fluid operating system,      System steam generator safety and/or relief valves, steam line stop valves, pressurizer relief valves (PORV's).
Loss of coolant accidents          Reactor Trip System                Engineered Safety Features Service Water System,            Emergency Core Cooling resulting from the spectrum of                                        Actuation System          Component Cooling Water          System, Auxiliary Feedwater postulated piping breaks within                                                                  System, steam generator safety    System, Containment Heat the reactor coolant pressure                                                                    and/or relief valves              Removal System, Emergency boundary                                                                                                                          Power System Amendment 63                                                                                                                                          Page 3 of 3
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.0.9-1 CORE TOTAL FISSION PRODUCT ACTIVITIES Based on 2958 MWt Isotope  Activity (Ci)          Isotope Activity (Ci)
I-131    8.02E+07              Cs-134  1.53E+07 I-132    1.16E+08              Cs-136  4.27E+06 I-133    1.64E+08              Cs-137  9.17E+06 I-134    1.80E+08              Rb-86  1.78E+05 I-135    1.53E+08 Ru-103  1.22E+08 Kr-85    8.62E+05              Ru-105  8.39E+07 Kr-85m    2.19E+07              Ru-106  4.15E+07 Kr-87    4.22E+07              Rh-105  7.66E+07 Kr-88    5.95E+07              Mo-99  1.47E+08 Xe-131m    8.96E+05              Tc-99m  1.29E+08 Xe-133    1.60E+08 Xe-133m    5.12E+06                Y-90  7.14E+06 Xe-135    3.83E+07                Y-91  1.04E+08 Xe-135m    3.21E+07                Y-92  1.08E+08 Xe-138    1.37E+08                Y-93  1.24E+08 Nb-95  1.38E+08 Te-127    8.45E+06              Zr-95  1.37E+08 Te-127m    1.10E+06              Zr-97  1.36E+08 Te-129    2.53E+07              La-140  1.50E+08 Te-129m    3.76E+06              La-141  1.34E+08 Te-131m    1.16E+07              La-142  1.30E+08 Te-132    1.14E+08              Nd-147  5.38E+07 Sb-127    8.55E+06              Pr-143  1.22E+08 Sb-129    2.57E+07              Am-241  1.06E+04 Cm-242  3.44E+06 Ce-141    1.35E+08              Cm-244  3.21E+05 Ce-143    1.25E+08 Ce-144    1.01E+08              Sr-89  8.10E+07 Pu-238    2.58E+05              Sr-90  6.82E+06 Pu-239    2.38E+04              Sr-91  9.97E+07 Pu-240    3.26E+04              Sr-92  1.07E+08 Pu-241    1.02E+07              Ba-139  1.47E+08 Np-239    1.57E+09              Ba-140  1.42E+08 Amendment 61                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                              UFSAR Chapter: 15 TABLE 15.0.9-2 RCS Coolant Fission Product Concentrations Based on 1% Fuel Defects Isotope        Activity (Ci/gm)
I-131                1.71 I-132                2.47 I-133              7.234 I-134            5.67E-01 I-135                1.84 Kr-85m                1.73 Kr-85              1.06E+01 Kr-87                1.10 Kr-88                3.21 Xe-131m                3.41 Xe-133m                4.86 Xe-133              2.76E+02 Xe-135m              4.36E-01 Xe-135                8.52 Xe-138              6.30E-01 Cs-134                1.55 Cs-136                3.21 Cs-137                1.61 Rb-86              1.97E-02 These fission product concentrations are based on projected RCS concentrations based on 1% fuel failure. They may be converted to Dose Equivalent I-131 using the dose conversion factors in ICRP-30 (Reference 15.0.9-12). For those events whose initial conditions are based on RCS Iodine concentrations at the Technical Specification Dose Equivalent I-131 limits (instead of the 1% fuel failure initial condition), see Table 15.0.9-7.
Amendment 61                                                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.0.9-3 Committed Effective Dose Equivalent Dose Conversion Factors Isotope  DCF (rem/curie) Isotope  DCF (rem/curie)
I-131      3.29E4      Cs-134      4.62E4 I-132      3.81E2      Cs-136      7.33E3 I-133      5.85E3      Cs-137      3.19E4 I-134      1.31E2      Rb-86      6.63E3 I-135      1.23E3 Ru-103      8.95E3 Kr-83m        N/A      Ru-105      4.55E2 Kr-85m        N/A      Ru-106      4.77E5 Kr-85        N/A      Rh-105      9.56E2 Kr-87        N/A        Mo-99      3.96E3 Kr-88        N/A      Tc-99m      3.3E1 Xe-131m        N/A Xe-133m        N/A        Y-90      8.44E3 Xe-133        N/A        Y-91      4.89E4 Xe-135m        N/A        Y-92      7.80E2 Xe-135        N/A        Y-93      2.15E3 Xe-138        N/A        Nb-95      5.81E3 Zr-95      2.37E4 Te-127      3.18E2      Zr-97      4.33E3 Te-127m      2.15E4      La-140      4.85E3 Te-129m      2.39E4      La-141      5.81E2 Te-129        9.0E1      La-142      2.53E2 Te-131m        6.4E3      Nd-147      6.85E3 Te-132      9.44E3      Pr-143    1.09E4 Sb-127      6.04E3      Am-241      4.44E8 Sb-129      6.44E2      Cm-242      1.73E7 Cm-244      2.48E8 Ce-141      8.96E3 Ce-143      3.39E3      Sr-89      4.14E4 Ce-144      3.74E5      Sr-90      1.3E6 Pu-238      3.92E8      Sr-91      1.66E3 Pu-239        4.3E8      Sr-92      8.1E2 Pu-240        4.3E8      Ba-139      1.7E2 Pu-241      8.26E6      Ba-140      3.74E3 Np-239      2.51E3 Amendment 61                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.0.9-4 Effective Dose Equivalent Dose Conversation Factors DCF                        DCF 3                          3 Isotope      (rem-m /Ci-sec)  Isotope  (rem-m /Ci-sec)
I-131          6.734E-2      Cs-134        0.2801 I-132          0.4144      Cs-136        0.3922 I-133          0.1088      Cs-137        0.1066 I-134          0.4810        Rb-86      1.780E-2 I-135          0.2953 Kr-83m          5.550E-6      Ru-103      8.325E-2 Kr-85m          2.768E-2      Ru-105        0.1410 Kr-85          4.403E-4      Ru-106          0.0 Kr-87          0.1524      Rh-105      1.376E-2 Kr-88          0.3774        Mo-99      2.694E-2 Xe-131m          1.439E-3      Tc-99m      2.179E-2 Xe-133m          5.069E-3 Xe-133          5.772E-3        Y-90      7.030E-4 Xe-135m          7.548E-2        Y-91      9.620E-4 Xe-135          4.403E-2        Y-92      4.810E-2 Xe-138          0.2135        Y-93      1.776E-2 Nb-95        0.1384 Te-127          8.954E-4      Zr-95        0.1332 Te-127m          5.439E-4      Zr-97      3.337E-2 Te-129m          5.735E-3      La-140        0.4329 Te-129          1.018E-2      La-141      8.843E-3 Te-131m          0.2594      La-142        0.5328 Te-132          3.811E-2      Nd-147      2.290E-2 Sb-127          0.1232        Pr-143      7.770E-5 Sb-129          0.2642      Am-241      3.027E-3 Cm-242      2.105E-5 Ce-141          1.269E-2      Cm-244      1.817E-5 Ce-143          4.773E-2 Ce-144          3.156E-3      Sr-89      2.860E-4 Pu-238          1.806E-5      Sr-90      2.786E-5 Pu-239          1.569E-5      Sr-91        0.1277 Pu-240          1.758E-5      Sr-92        0.2512 Pu-241          2.683E-7      Ba-139      8.029E-3 Np-239          2.845E-2      Ba-140      3.175E-2 Amendment 61                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 TABLE 15.0.9-5 Nuclide Decay Constants Decay Constant            Decay Constant
                                      -1                      -1 Isotope        (hr )      Isotope      (hr )
I-131      0.00359        Cs-134    3.84E-5 I-132        0.303      Cs-136      2.2E-3 I-133        0.0333      Cs-137    2.64E-6 I-134        0.791        Rb-86    1.55E-3 I-135        0.105 Kr-83m        0.379        Ru-103    7.35E-4 Kr-85m        0.155        Ru-105      0.156 Kr-85      7.37E-6      Ru-106    7.84E-5 Kr-87        0.547      Rh-105    1.96E-2 Kr-88        0.248        Mo-99    1.05E-2 Xe-131m      0.00241        Tc-99m      0.115 Xe-133m        0.0130 Xe-133      0.00546          Y-90    1.08E-2 Xe-135m          2.72        Y-91    4.94E-4 Xe-135        0.0756        Y-92      0.196 Xe-138          2.93        Y-93    0.0686 Nb-95    8.22E-4 Te-127      7.41E-2        Zr-95    4.51E-4 Te-127m      2.65E-4        Zr-97      4.1E-2 Te-129m        8.6E-4      La-140    1.72E-2 Te-129        0.598      La-141      0.176 Te-131m      2.31E-2      La-142      0.45 Te-132      8.86E-3      Nd-147    2.63E-3 Sb-127        7.5E-3        Pr-143    2.13E-3 Sb-129          0.16      Am-241      1.83E-7 Cm-242      1.77E-4 Ce-141      8.89E-4      Cm-244      4.37E-6 Ce-143        0.021 Ce-144      1.02E-4        Sr-89    5.72E-4 Pu-238      9.02E-7        Sr-90    2.72E-6 Pu-239      3.29E-9        Sr-91      0.073 Pu-240      1.21E-8        Sr-92      0.256 Pu-241        5.5E-6      Ba-139      0.502 Np-239        0.0123      Ba-140    2.27E-3 Amendment 61                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 Table 15.0.9-6 Iodine Spike Appearance Rates (Curies/Minute)
Based on 1 Ci/gm of D.E. I-131 Primary Coolant Activity I-131    I-132 I-133 I-134 I-135 335 times the equilibrium rate  127.6  422.4  608.4 186.3 197.3 500 times the equilibrium rate  190.5  657.0  915.0 294.0 301.0 Amendment 61                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 Table 15.0.9-7 Iodine Specific Activities (Ci/gm)
Primary Coolant Based on 1.0 and 60.0 Ci/gm of D.E. I-131 Secondary Coolant Based on 0.1 Ci/gm of D.E. I-131 Primary Coolant                  Secondary Coolant Nuclide                1 Ci/gm                  60 Ci/gm          0.1 Ci/gm I-131                  0.570                    34.20              0.0570 I-132                  0.823                    49.38              0.0823 I-133                  2.408                    144.48              0.2408 I-134                  0.189                    11.34              0.0189 I-135                  0.613                    36.78              0.0613 Amendment 61                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                UFSAR Chapter: 15 TABLE 15.0.13-1 SINGLE FAILURES ASSUMED FOR ACCIDENTS OF MODERATE FREQUENCY Event Description                  Section          Worst Failure Assumed                  Effect Feedwater temperature reduction                    15.1.1                      (1)                          none Excessive feedwater flow                            15.1.2      One protection train                        none Excessive steam flow                                15.1.3                      (1)                          none Inadvertent secondary depressurization              15.1.4      One Safety injection train          delays boron to core Loss of external load                              15.2.2      One protection train                        none Turbine trip                                        15.2.3      One protection train                        none Inadvertent closure of MSIV                        15.2.4      One protection train                        none Loss of condenser vacuum                            15.2.5      One Protection train                        none Loss of ac power                                    15.2.6      Loss of two auxiliary feedwater      increases primary pumps (3) with and without                heatup isolation of one SG (4)
Loss of normal feedwater                            15.2.7      Loss of two auxiliary feedwater      increases primary pumps (3)                                  heatup Loss of forced reactor coolant flow              15.3.1 & 2    One protection train                        none RCCA bank withdrawal from subcritical              15.4.1      One protection train                        none RCCA bank withdrawal at power                      15.4.2      One protection train                        none Dropped RCCA, dropped RCCA bank                    15.4.3      One NIS channel                      core power is not reduced on power mismatch Statically misaligned RCCA                          15.4.3                      (2)                          none Uncontrolled boron dilution                        15.4.6      Malfunction of RMWS flow              Reduces time to indication so that actual flow is        criticality higher than indicated Inadvertent ECCS operation at power                15.5.1      One protection train                        none Increase in RCS inventory                          15.5.2      One protection train                        none Inadvertent RCS depressurization                    15.6.1      One protection train                        none Failure of small lines carrying primary coolant    15.6.2                      (2)                          none outside containment
 
(1) No protective action required.
(2) No transient analysis involved.
(3) The Framatome analysis assumes only one AFW pump is available to mitigate the Loss of Feedwater and Loss of AC Power Events even though no single failure has been identified that would result in loss of two AFW pumps (see discussion in FSAR Section 15.0.13).
(4) No single failure has been identified that would lead to the combination of AFW pump failures and inadvertent steam generator isolation that is assumed in this analysis. The combination of failures in the analysis was simply a matter of analytical expediency and does not constitute a legitimate licensing basis system configuration (see discussion in FSAR Section 15.0.13.a.9).
Amendment 63                                                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                    UFSAR Chapter: 15 TABLE 15.0.13-2 SINGLE FAILURES FOR NON CONDITION II EVENTS Event Description            Section          Worst Failure Assumed                Effect Single Rod Withdrawal                15.4.3    NIS channel nearest to withdrawn          None RCCA Inadvertent Fuel Loading              15.4.7    (1) (2)                                    None LOCA (Small Break)                    15.6.5    Loss of One SI Train                    Higher PCT Gaseous Rad Waste Failure            15.7.1    (1)                                        None Liquid Rad Waste Failure              15.7.2    (1)                                        None Liquid Tank Failure                  15.7.3    (1)                                        None Fuel Cask Drop                        15.7.5    (1)                                        None Steamline Rupture                    15.1.5    Loss of One SI Train                Delay Boron to Core Feedline Rupture                      15.2.8    (3)                                    Delay Cooling Locked Rotor                          15.3.3    Loss of One Protection Train              None Shaft Break                          15.3.4    Loss of One Protection Train              None Rod Ejection                          15.4.8    One NIS channel                      Delay reactor trip Steam Generator Tube Rupture          15.6.3    Failed-open PORV on ruptured    Increased Break Flow and SG                              Increased Steam Release Fuel Handling Accident                15.7.4    (1)                                        None
 
(1) No Protective Action Required.
(2) No Transient Analysis Involved.
(3) See discussion in Section 15.2.8 for event-specific failure scenarios.
Amendment 63                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                        UFSAR Chapter: 15 TABLE 15.1.2-1 INPUT PARAMETERS AND BIASING FOR INCREASE IN FEEDWATER FLOW Parameter                                        HFP Cases                                                  HZP Case Core Power                                              2958 MW(rated + 0.34%)                                            10-9 rated Reactor Coolant System Pressure                                  Nominal                                                    Nominal Pressurizer Level                                                Nominal                                                    Nominal Core Average Temperature                          Nominal at 2958 MW (rated + 0.34%)                                Nominal at HZP Power Reactor Coolant Flow                                      Tech. Spec. Minimum                                        Tech. Spec. Minimum Steam Generator Pressure                          Nominal at 2958 MW (rated + 0.34%)                                Nominal at HZP Power Initial Feedwater Flow Rate                        Nominal at 2958 MW (rated + 0.34%)                                          ~0 Feedwater Temperature                                            Nominal                                                    Nominal Cycle Exposure                                                BOC & EOC                                                      EOC Moderator Temperature Coefficient                            Tech. Spec. Limits                                        Tech. Spec. Limit Doppler Coefficient                                            0.8
* BOC                                                  0.8
* EOC 0.8
* EOC Delayed Neutron Fraction                                    Nominal for Exposure                                            Nominal
/                                                        Nominal for Exposure                                            Nominal Reactor Trip Reactivity Insertion        Minimum (bounds the most reactive rod stuck out of the core Minimum (bounds the most reactive rod stuck out of the core)
Pellet to Clad Heat Transfer Coefficient                          Mean                                                      Mean Rod Position Controller                            Auto for BOC Manual & Auto for EOC                                        Manual Pressurizer Heaters                                              Disable                                                    Disable Pressurizer Spray                                                Available                                                  Available Pressurizer PORVs                                                Available                                                  Available Main Feedwater                                                    Auto                                                        Auto Auxiliary Feedwater                                              Available                                                  Disable Amendment 61                                                                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                          UFSAR Chapter: 15 TABLE 15.1.2-2 KEY OPERATING PARAMETERS FOR INCREASE IN FEEDWATER FLOW Parameter                                      HFP Value    HZP Value Initial Reactor Power (MW)                      2958.0        2.9E-6 Initial Pressurizer Pressure (psia)              2250          2250 Initial Pressurizer Level (% of level span)      60.0          25.0 Initial Tavg (&deg;F)                                588.8        557.7 Initial Total RCS Flow Rate (lbm/s)              30390        30351 Initial Steam Generator Pressure (psia)          977          1099 Initial Feedwater Flow Rate per SG (lbm/s)      1210            ~0 (2)
Feedwater Temperature (&deg;F)                      440.0            40
 
(2)
Safety analyses support operating with a reduced main feedwater temperature of 375&deg;F at full power.
Amendment 61                                                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 TABLE 15.1.2-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR INCREASE IN FEEDWATER FLOW Parameter                  BOC          EOC Moderator Temperature Coefficient (pcm/&deg;F)    +5      -50(HFP)
Doppler Coefficient (pcm/&deg;F)                -1.072  -1.256(HFP) 0.006146    0.005142
                      -1
              / (sec )                                  398.5      298.0 Scram Worth (pcm)                            1770      1770 Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.1.2-4 EVENT
 
==SUMMARY==
FOR INCREASE IN FEEDWATER FLOW -
LIMITING CASE (HZP, EOC, MANUAL ROD CONTROL)
Event                                                    Time (sec)
Initiate Transient (Step Increase in Feedwater Flow)      0.0 High Flux Reactor Trip                                    18.9 Turbine Trip (on reactor trip)                            19.4 MDNBR                                                    21.8 High-High Steam Generator Level Signal (MFW terminated)  47.7 Amendment 61                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 TABLE 15.1.3-1 INPUT PARAMETERS AND BIASING FOR INCREASE IN STEAM FLOW Core Cooling Parameter            Minimum Feedback Case      Maximum Feedback Case Core Power                      SCD                          SCD Pressurizer Pressure                SCD                          SCD Pressurizer Level                  High                        High RCS Average Temperature                SCD                          SCD Reactor Coolant System Flow Rate            SCD                          SCD Fuel Temperature                  High                        High Steam Generator Narrow Range Level          Nominal                      Nominal Steam Generator Tube Plugging            Low                          Low Feedwater Temperature                Nominal                      Nominal Cycle Exposure                    BOC                          EOC Moderator Temperature Coefficient    BOC Least Negative            EOC Most Negative Doppler Coefficient          BOC Least Negative          EOC Least Negative Delayed Neutron Fraction            Minimum                      Minimum
                      /l                      Minimum                      Minimum Pressurizer Heaters              Disabled                    Disabled Pressurizer Spray              Insensitive                  Insensitive Pressurizer Power-Operated Relief        Insensitive                  Insensitive Valves Pressurizer Safety Valves            Nominal                      Nominal Main Steam Safety Valves              Nominal                      Nominal Main Feedwater                  Automatic                    Automatic Auxiliary Feedwater              Insensitive                  Insensitive Rod Position Controller            Automatic                    Automatic Amendment 63                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                      UFSAR Chapter: 15 TABLE 15.1.3-2 KEY OPERATING PARAMETERS FOR INCREASE IN STEAM FLOW Parameter                                Core Cooling Minimum and Maximum Feedback Cases Initial Core Power (MW)                            2948 Initial Pressurizer Pressure (psia)                      2250 Initial Pressurizer Level (% of level span)                    66.75 Initial RCS Average Temperature (&deg;F)                        588.8 Initial Reactor Coolant System Flow Rate (gpm)                    296,380 Feedwater Temperature (&deg;F)                            440(a)
Initial Main Steam System Flow Rate per SG                      1211.2 (lbm/s)
(a)
Safety analyses support operating with a reduced main feedwater temperature of 375&deg;F at full power.
Amendment 63                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                  UFSAR Chapter: 15 TABLE 15.1.3-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR INCREASE IN STEAM FLOW Core Cooling Parameter                Minimum Feedback Case        Maximum Feedback Case BOC                          EOC Moderator Temperature                      0                          -57.5 Coefficient (pcm/&deg;F)
Doppler Coefficient (pcm/&deg;F)                -1.09                        -1.25 Delayed Neutron Fraction                  0.0055                      0.0045
                        -1
                / (sec )                        366.67                        250.00 Scram Worth (pcm)                        N/A(a)                        N/A(a)
(a) There is no reactor trip experienced in the increase in steam flow transient.
Amendment 63                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                            UFSAR Chapter: 15 TABLE 15.1.3-4 EVENT
 
==SUMMARY==
FOR INCREASE IN STEAM FLOW (EXCESS LOAD) - MINIMUM (BOC) FEEDBACK CASE Event              Time (sec) 10% Step Increase in Steam Flow  0.001 Peak Power              164.3 MDNBR                404.0 Amendment 63                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                      UFSAR Chapter: 15 TABLE 15.1.5-1 EQUIPMENT REQUIRED FOLLOWING A RUPTURE OF A MAIN STEAM LINE Short Term (Required for Mitigation of Accident)                Hot Standby                        Required for Cooldown Reactor trip and safeguards actuation channels      Auxiliary Feedwater System including      Steam generator power operated including sensors, circuitry, and processing        pumps, water supply, and system valves    relief valves (can be manually equipment (the protection circuits used to trip the and piping (this system must be placed    operated locally).
reactor on undervoltage, underfrequency, and        in service to supply water to operable turbine trip may be excluded).                      steam generators no later than 10 minutes after the incident).
Safety Injection System including the pumps, the    Containment ventilation cooling units. Control for defeating automatic refueling water storage tank, the boron injection                                              safety injection actuation during a tank, and the systems valves and piping.                                                      cooldown and depressurization.
Diesels generators and emergency power              Capability for obtaining a reactor coolant Residual Heat Removal System distribution equipment.                            system sample.                            including pumps, heat exchanger, and system valves and piping necessary to cool and maintain the Reactor Coolant System in a cold shutdown condition.
Emergency Service Water System.
Containment safeguards cooling equipment.
Auxiliary Feedwater System including pumps, water supplies, piping and valves.
AFW Isolation System Amendment 61                                                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 TABLE 15.1.5-3 SEQUENCE OF EVENTS FOR LIMITING MAIN STEAM LINE BREAK MDNBR CASE -
HZP WITH OFFSITE POWER WITH THE STUCK ROD Time(s)        Event 0.0            Reactor at EOC HZP conditions 0.0            Double-ended guillotine break in main steam line occurs 0.0            SIS low steam line pressure setpoint reached 2.0            Full AFW flow to affected steam generator 7.0            MSIVs closed 10.0          MFW flow stops following SI signal generation 14.0          Criticality occurs (total reactivity > 0 $)
37.0          HHSI pumps at rated speed (37 s delay) 244.0          Peak neutron power reached (739 MW) 251.0          Borated water has filled SI lines and begins to enter cold legs 600.0          Simulation terminated Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                        UFSAR Chapter: 15 TABLE 15.1.5-5 Parameters Used in Steam Line Break Radiological Analysis Reactor coolant noble gas activity prior to accident (% fuel defect level)                  1.0 Reactor coolant iodine activity prior to accident (Ci/gm of DE I-131)
Pre-accident iodine spike                                                            60 Accident-initiated iodine spike (Ci/gm of DE I-131)                                  1.0 Reactor coolant iodine appearance rate increase due to the accident-initiated spike (times  500(Table 15.0.9-6) equilibrium rate)
Duration of accident-initiated iodine spike (hr)                                            5.0 SG tube leak rate (gpm total)                                                                1 SG tube leak rate to affected (faulted) SG (gpm)                                      0.35 SG tube leak rate to unaffected (intact) SGs (gpm)                                    0.65 Steam release from faulted SG to environment during first two minutes (lbm)                                                              162,000 Time to release initial mass in faulted SG (min)                                            2 Steam releases from intact SGs (lbm) 0 - 2 hours                                                                          401,000 2 - 8 hours                                                                          917,000
      > 8 hours                                                                            0 Time to cool RCS below 212&deg;F and stop releases from faulted SG (hr)                          40 Time to cool RCS below 350&deg;F and stop releases from intact SGs (hr)                          8 SG iodine water/steam partition coefficient Faulted SG                                                                            1 Intact SGs                                                                            0.01 Iodine chemical form after release to atmosphere (%)
Elemental                                                                            97 Organic                                                                              3 Particulate (cesium iodide)                                                          0 RCS mass (lbm)                                                                              4.11E5 Intact Secondary SG Side mass (lbm/per SG)                                                  115,585 Faulted SG mass (lbm)                                                                        162,000 Amendment 63                                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 TABLE 15.1.5-6 Radiological Consequences of a Postulated Main Steam Line Break For the pre-accident iodine spike:
Exclusion Area Boundary*                      0.14 rem TEDE Low Population Zone                            0.15 rem TEDE Control Room                                  0.28 rem TEDE For the accident-initiated iodine spike:
Exclusion Area Boundary*                      0.69 rem TEDE Low Population Zone                            1.04 rem TEDE Control Room                                  1.75 rem TEDE
*The exclusion area boundary doses reported are for the worst two hour period, determined to be from 0.0 to 2.0 hours for the pre-accident iodine spike, and from 4.9 to 6.9 hours for the accident-initiated iodine spike.
Amendment 63                                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                                UFSAR Chapter: 15 TABLE 15.1.5-7 STEAMLINE BREAK Impact of    Impact of Operator's Failure to Instructions Given to the        Components and        Single Active    Take Action or the Operator Required Operator      Alarms to Alert the Operator to    Delay Time    Operator for Performing    Instrumentation Necessary to  Component      Taking a Closely Related but Action            Initiate a Particular Action      Assumed        the Required Action        Complete Indicated Action    Failure              Erroneous Action A. Identify the faulted A. No specific alarms provided    A. Within 10    A. Identify the faulted  A.1 Steamline pressure        None          A. The operator does not isolate steam generator        for this function. Primary          minutes        steam generator    indicators                                  AFW to any steam generator or and isolate            indication to the operator is                      by comparing                                                    isolates AFW to wrong steam auxiliary feedwater    steamline pressure indication.                      steamline          A.2 Steam Generator AFW      None          generator; the faulted steam to that steam          A possible alarm is the steam                      pressures.          isolation valves                            generator will continue to generator.              flow feed flow mismatch.                            Terminate                                                      blowdown.
auxiliary feedwater A.3 Steam generator level    None to that steam      indicators generator by shutting the AFW isolation valves.
B. The Operator      B. No specific alarms for this    B. N/A        B. The conditions for    B.1 Pressurizer level        None          B.1 The operator fails to must reset the        purpose. Primary indications                        resetting Safety    indicators                                  modulate SI pumps after the Safety Injection      to the operator are:                                Injection are given                                            pressurizer level returns to the and manually          Pressurizer level, Pressurizer                      to the operator. B.2 Pressurizer pressure      None          indicating range: Water relief thru control the          Pressure and RCS                                    The operator is    indicators                                  pressurizer relief valves may repressurization      temperature. Possible alarms                        instructed to                                                  occur.
of RCS and            include:                                            manually control    B.3 RCS Temperature          None maintain normal        - High Pressurizer Level                          the high head SI    Indicators                                  B.2 The operator stops SI before pressure control.      - High Pressurizer Pressure.                      pumps and re-                                                  peak reactivity is reached: If establish normal    B.4 High Head Safety                        criticality is attained the core pressurizer level  Injection Pumps              None          power will increase until it control.                                                        reaches equilibrium with the steam demand.
Amendment 61                                                                                                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                    UFSAR Chapter: 15 TABLE 15.2.3-1 INPUT PARAMETERS AND BIASING FOR TURBINE TRIP Short-Term Core            Peak Primary            Peak Secondary Parameter                Cooling                  Pressure                  Pressure Core Power                        SCD                      High                      High Pressurizer Pressure              SCD                        Low                      High Pressurizer Level                High                      High                      High RCS Average                      SCD                        Low                      High Temperature Reactor Coolant Flow              SCD                      Low                      High Fuel Temperature              Maximum                  Maximum                  Maximum Steam Generator                Nominal                  Nominal                  Nominal Narrow Range Level Steam Generator Tube          Maximum                  Maximum                    Minimum Plugging Feedwater Temperature          Nominal                  Nominal                  Nominal Cycle Exposure                    BOC                      BOC                      BOC Moderator Temperature    BOC Least Negative        BOC Least Negative        BOC Least Negative Coefficient Doppler Temperature      BOC Least Negative        BOC Least Negative        BOC Least Negative Coefficient Delayed Neutron                Maximum                  Maximum                  Maximum Fraction
    /                            Maximum                  Maximum                  Maximum Reactor Trip Reactivity  Minimum with most        Minimum with most        Minimum with most Insertion              reactive rod stuck out of reactive rod stuck out of reactive rod stuck out of the core                  the core                  the core Pressurizer Heaters            Disabled                  Available                Disabled Pressurizer Spray              Available                Disabled                  Available Pressurizer Power-              Available                Disabled                  Available Operated Relief Valves Pressurizer Safety                Low                      High                      Low Valves Main Steam Safety                High                      High                      High Valves Main Feedwater          Isolated on Turbine Trip  Isolated on Turbine Trip  Isolated on Turbine Trip Auxiliary Feedwater            Disabled                  Disabled                  Disabled Amendment 63                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 TABLE 15.2.3-2 KEY OPERATING PARAMETERS FOR TURBINE TRIP Parameter          Short-Term Core Cooling  Peak Primary Pressure  Peak Secondary Pressure Initial Core Power (MW)              2948                    2958                    2958 Initial Pressurizer 2250                    2200                    2300 Pressure (psia)
Initial Pressurizer Level 67.5                  81.75                  66.75
(% of level span)
Initial RCS Average 588.8                  583.7                  539.9 Temperature (&deg;F)
Initial Reactor Coolant 296,380                290,000                  321,000 Flow (gpm)
Feedwater Temperature 440a                    440a                    440a
(&deg;F) a Safety analyses support operating with a reduced main feedwater temperature of 375&deg;F at full power.
Amendment 63                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 TABLE 15.2.3-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR TURBINE TRIP Peak Primary and Secondary Short-Term Core Cooling Parameter                                              Pressure Cases BOC BOC Moderator Temperature Coefficient 0.0                      0.0 (pcm/&deg;F)
Doppler Coefficient (pcm/&deg;F)              -1.09                    -1.09 Delayed Neutron Fraction                0.007                    0.007
            / (sec-1)                      466.67                    466.67 Scram Worth (pcm)                      5000                    5000 Amendment 63                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 TABLE 15.2.3-4 EVENT
 
==SUMMARY==
FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE Event                                      Time (seconds)
Turbine Trip Initiates                                                      0.01 High Pressurizer Pressure Trip Setpoint Reached                              5.0 Pressurizer Safety Valve Start to Open                                      6.8 High Pressurizer Pressure - Reactor Trip (control rods begin to insert)      7.0 Peak Primary System Pressure - Bottom of Rx Vessel                          8.1 Pressurizer Safety Valve Start to Close                                      9.6 Steam Generator MSSV Bank 1 (Loops 1, 2, 3) Open                            10.7 Steam Generator MSSV Bank 2 (Loops 1, 2, 3) Open                            11.7 Steam Generator MSSV Bank 3 (Loops 1, 2, 3) Open                            13.9 Steam Generator MSSV Bank 3 (Loop 2) Close                                  45.0 Steam Generator MSSV Bank 3 (Loops 1, 3) Close                              45.0 End of Simulation                                                          50.0 Amendment 63                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                              UFSAR Chapter: 15 TABLE 15.2.3-5 EVENT
 
==SUMMARY==
FOR TURBINE TRIP SECONDARY OVERPRESSURIZATION CASE Event                                                      Time (sec)
Turbine Trip Initiates                                    0.01 Pressurizer Spray On                                      1.6 Pressurizer PORV Begin Cycling                            2.2 - 15.7 Steam Generator MSSV Bank 1 (Loop 1, 3) Open              4.7 Steam Generator MSSV Bank 1 (Loop 2) Open                  4.7 Steam Generator MSSV Bank 2 (Loop 2) Open                  5.4 Steam Generator MSSV Bank 2 (Loop 1, 3) Open              5.4 Steam Generator MSSV Bank 3 (Loop 1, 3) Open              6.4 Steam Generator MSSV Bank 3 (Loop 2) Open                  6.5 Steam Generator MSSV Bank 4 (Loop 2) Open                  9.0 Steam Generator MSSV Bank 4 (Loop 1, 3) Open              9.0 2/3 OTDT Trip Setpoint Reached                            11.4 OTDT - Reactor Trip (control rods begin to insert)        12.7 Steam Generator MSSV Bank 5 (Loop 2) Open                  14.0 Steam Generator MSSV Bank 5 (Loop 1, 3) Open              14.0 Pressurizer Spray Off                                      15.7 Peak Secondary System Pressure - Bottom of SG-1 Downcomer  18.1 Steam Generator MSSV Bank 5 (Loop 1, 2) Close              32.0 Steam Generator MSSV Bank 5 (Loop 3) Close                32.0 Steam Generator MSSV Bank 4 (Loop 2) Close                33.7 Steam Generator MSSV Bank 4 (Loop 1, 3) Close              33.7 Steam Generator MSSV Bank 3 (Loop 2) Close                36.5 Steam Generator MSSV Bank 3 (Loop 1, 3) Close              36.6 Steam Generator MSSV Bank 2 (Loop 2) Close                43.6 Steam Generator MSSV Bank 2 (Loop 1, 3) Close              43.6 End of Simulation                                          50.0 Amendment 63                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.2.3-6 EVENT
 
==SUMMARY==
FOR TURBINE TRIP DNB CASE Event                        Time (seconds)
Turbine Trip Initiates                                0.01 Pressurizer Spray On                                  1.6 Pressurizer PORVs begin cycling                    3.5 - 15.8 Steam Generator MSSV Bank 1 (Loop 1, 2, 3) Open        8.1 Steam Generator MSSV Bank 2 (Loop 2) Open              9.0 Steam Generator MSSV Bank 2 (Loop 1, 3) Open          9.0 Steam Generator MSSV Bank 3 (Loop 2) Open            10.5 Steam Generator MSSV Bank 3 (Loop 1, 3) Open          10.5 2/3 OTDT Trip Setpoint Reached                        12.0 OTDT - Reactor Trip (control rods begin to insert)    13.3 Steam Generator MSSV Bank 4 (Loop 1, 2, 3) Open      13.9 Time of MDNBR                                        14.0 Pressurizer Spray Off                                16.7 Steam Generator MSSV Bank 4 (Loop 2) Close            35.9 Steam Generator MSSV Bank 4 (Loop 1, 3) Close        35.9 Steam Generator MSSV Bank 3 (Loop 2) Close            38.4 Steam Generator MSSV Bank 3 (Loop 1, 3) Close        38.4 Steam Generator MSSV Bank 2 (Loop 2) Close            44.1 Steam Generator MSSV Bank 2 (Loop 1, 3) Close        44.2 End of Simulation                                    50.0 Amendment 63                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                  UFSAR Chapter: 15 TABLE 15.2.6-1 INPUT PARAMETERS AND BIASING LOSS OF NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES Parameter Core Power                                            2958 MW (rated + 0.34%)
Reactor Coolant System Pressure                      Nominal Pressurizer Level                                    Nominal+Uncertainty Core Average Temperature                              Nominal at 2958 MW (rated + 0.34%)
Reactor Coolant Flow                                  Tech. Spec. Minimum Steam Generator Pressure                              Nominal at 2958 MW (rated + 0.34%)
Feedwater Flow Rate                                  Nominal at 2958 MW (rated + 0.34%)
Feedwater Temperature                                Nominal Steam Generator Level                                Nominal Cycle Exposure                                        BOC Moderator Temperature Coefficient                    Tech. Spec. Limit Doppler Coefficient                                  0.8
* BOC Delayed Neutron Fraction                              Nominal BOC
/                                                  Nominal BOC Reactor Trip Reactivity Insertion                    Minimum (bounds the most reactive rod stuck out of the core)
Pellet to Clad Heat Transfer Coefficient              Mean Rod Position Controller                              Manual Pressurizer Heaters                                  Available Pressurizer Spray                                    Available Pressurizer PORVs                                    Available Main Feedwater                                        Auto until failure to deliver FW flow initiates transient Auxiliary Feedwater                                  1 of 2 MDAFW pumps Available TDAFW pump Disable Amendment 61                                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.2.6-2 KEY OPERATING PARAMETERS FOR LOSS OF NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES Parameter                                                        Value Initial Reactor Power (MW)                                      2958 Initial Pressurizer Pressure (psia)                              2250 Initial Pressurizer Level (% of level span)                      66.8 Initial Total Tavg (&deg;F)                                          588.8 Initial Total RCS Flow Rate (lbm/s)                              30390 Initial Steam Generator Pressure (psia)                          981 Initial Feedwater Flow Rate per SG (lbm/s)                      1216 Feedwater Temperature (&deg;F)                                      440.0(b)
 
(b)
Safety analyses support operating with a reduced main feedwater temperature of 375&deg;F at full power.
Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 TABLE 15.2.6-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR LOSS OF NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES Parameter                                      BOC Value Moderator Temperature Coefficient (pcm/&deg;F)      +5 Doppler Coefficient (pcm/&deg;F)                    -0.80 0.006146
/ (sec-1)                                    398.5 Scram Worth (pcm)                                1770 Amendment 61                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 TABLE 15.2.6-4 EVENT
 
==SUMMARY==
FOR LOSS OF NON-EMERGENCY AC POWER TO THE STATION AUXILIARIES Event                Time (sec)
Total loss of MFW occurred                              0.0 Pressurizer PORV opened                                11.5 Reactor tripped                                    41.3(OTT) 50.9 (LSGL)
Main Turbine tripped                                  41.35 RCPs tripped on loss of offsite power                  41.37 MSSVs opened                                          41.5 Maximum pressurizer level                              46.0 AFW actuation signal on Low-Low SG Level occured      47.4 Pressurizer PORV closed                                47.5 Motor-driven AFW pump started, blowndown isolated      108.9 Maximum post-trip RCS average temperature              1,245 Minimum SG liquid inventory occurred                  1,320 End of calculation                                    10,000 Amendment 62                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                          UFSAR Chapter: 15 TABLE 15.2.6-5 Parameters Used in Loss of AC Power Radiological Analysis Reactor coolant noble gas and alkali metal activity prior to accident (% fuel defect level)          1.0 Reactor coolant iodine activity prior to accident (Ci/gm of DE I-131)
Pre-accident iodine spike                                                                        60 Accident-initiated iodine spike (Ci/gm of DE I-131)                                              1.0 Reactor coolant iodine appearance rate increase due to the accident-initiated spike          500 (Table 15.0.9-6)
(times equilibrium rate)
Duration of accident-initiated iodine spike (hr)                                                    5.0 Secondary coolant iodine activity prior to accident (Ci/gm of DE I-131)                            0.1 Secondary coolant alkali metal activity prior to accident (% of primary concentration)              10 Release Modeling SG tube leak rate (gpm total)                                                                        1 Steam release to environment (lbm) 0 - 2 hours                                                                                    378,000 2 - 8 hours                                                                                    965,000
  > 8 hours                                                                                          0 SG iodine and alkali metal water/steam partition coefficient                                        0.01 Iodine chemical form after release to atmosphere (%)
Elemental                                                                                        97 Organic                                                                                            3 Particulate (cesium iodide)                                                                        0 RCS mass (lbm)                                                                                    4.11E5 Secondary Side mass (lbm/per SG)                                                                  115,585 Amendment 61                                                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.2.6-6 RADIOLOGICAL CONSEQUENCES OF A LOSS OF NON-EMERGENCY AC POWER TO THE PLANT AUXILIARIES For the pre-accident iodine spike:
Exclusion Area Boundary*                    0.013 rem TEDE Low Population Zone                        0.01 rem TEDE Control Room                                0.02 rem TEDE For the accident-initiated iodine spike:
Exclusion Area Boundary*                    0.05 rem TEDE Low Population Zone                        0.02 rem TEDE Control Room                                0.05 rem TEDE
*The exclusion area boundary doses reported are for the worst two hour period, determined to be from 6.0 to 8.0 hours.
Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                    UFSAR Chapter: 15 TABLE 15.2.7-1 INPUT PARAMETERS AND BIASING LOSS OF NORMAL FEEDWATER FLOW Parameter Core Power                                2958 MW (rated + 0.34%)
Reactor Coolant System Pressure            Nominal Pressurizer Level                          Nominal+Uncertainty Core Average Temperature                  Nominal at 2958 MW (rated + 0.34%)
Reactor Coolant Flow                      Tech. Spec. Minimum Steam Generator Pressure                  Nominal at 2958 MW (rated + 0.34%)
Feedwater Flow Rate                        Nominal at 2958 MW (rated + 0.34%)
Feedwater Temperature                      Nominal Steam Generator Level                      Nominal Cycle Exposure                            BOC Moderator Temperature Coefficient          Tech. Spec. Limit Doppler Coefficient                        0.8
* BOC Delayed Neutron Fraction                  Nominal BOC
/                                        Nominal BOC Reactor Trip Reactivity Insertion          Minimum (bounds the most reactive rod stuck out of the core)
Pellet to Clad Heat Transfer Coefficient  Mean Rod Position Controller                    Manual Pressurizer Heaters                        Available Pressurizer Spray                          Available Pressurizer PORVs                          Available Main Feedwater                            Auto until failure to deliver FW flow initiates transient Auxiliary Feedwater                        1 of 2 MDAFW pumps Available TDAFW pump Disable Safety Injection                          HHSI Available Amendment 61                                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.2.7-2 KEY OPERATING PARAMETERS FOR LOSS OF NORMAL FEEDWATER Parameter                                                        Value Initial Reactor Power (MW)                                      2958 Initial Pressurizer Pressure (psia)                              2250 Initial Pressurizer Level (% of level span)                      66.8 Initial Tavg (&deg;F)                                                588.8 Initial Total RCS Flow Rate (lbm/sec)                            30390 Initial Steam Generator Pressure (psia)                          982 Initial Feedwater Flow Rate per SG (lbm/s)                      1216 Feedwater Temperature (&deg;F)                                      440.0(b)
 
(b)
Safety analyses support operating with a reduced main feedwater temperature of 375&deg;F at full power.
Amendment 61                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                            UFSAR Chapter: 15 TABLE 15.2.7-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR LOSS OF NORMAL FEEDWATER Parameter                                            BOC Value Moderator Temperature Coefficient (pcm/&deg;F)            +5 Doppler Coefficient (pcm/&deg;F)                          -0.80 0.006146
/ (sec-1)                                          398.5 Scram Worth (pcm)                                    1770 Amendment 61                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.2.7-4 EVENT
 
==SUMMARY==
FOR LOSS OF NORMAL FEEDWATER Event                    Time (s)
Total loss of MFW occured                            0.0 Pressurizer PORV opened                              11.5 41.3(OTT)
Reactor tripped 52.7 (LSGL)
Main Turbine tripped                                41.35 MSSVs opened                                        41.5 Maximum pressurizer level                            45.5 Maximum post-trip RCS average temperature            46.0 Pressurizer PORV closed                              47.5 AFW actuation signal on Low-Low SG level occured    49.2 Motor Driven AFW pump started, blowdown isolated    110 Minimum SG liquid inventory occured                2,965 Amendment 62                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.2.8-1 INPUT PARAMETERS AND BIASING FOR FEEDWATER LINE BREAK Short-Term Core      Long-Term Core      Peak Primary Parameter Cooling              Cooling            Pressure Core Power                  SCD                  High                High Pressurizer Pressure            SCD                  Low                High Pressurizer Level              Low                  Low                High RCS Average Temperature            SCD                  High                High Reactor Coolant Flow            SCD                  Low                Low Fuel Temperature              High                  High                High Steam Generator Narrow            Low                  Low                Low Range Level Steam Generator Tube              Low                  Low                High Plugging Feedwater Temperature          Nominal              Nominal            Nominal Cycle Exposure                BOC                  EOC                BOC Moderator Temperature    BOC Least Negative    EOC Most Negative  BOC Least Negative Coefficient Doppler Temperature      BOC Least Negative  EOC Least Negative  BOC Least Negative Coefficient Delayed Neutron Fraction        Maximum              Maximum            Maximum
              /l                  Maximum              Maximum            Maximum Reactor Trip Reactivity    Minimum with most    Minimum with most  Minimum with most Insertion          reactive rod stuck    reactive rod stuck  reactive rod stuck out of the core      out of the core    out of the core Pressurizer Heaters          Disabled            Disabled            Available Pressurizer Spray            Available            Available            Disabled Pressurizer Power-Operated        Available            Disabled            Disabled Relief Valves Pressurizer Safety Valves          Low                  Low                High Main Steam Safety Valves          High                  High                High Main Feedwater          Lost with FLB        Lost with FLB      Lost with FLB Auxiliary Feedwater          Insensitive      Minimum Available    Minimum Available Safety Injection          Insensitive        No HHSI Flow      Maximum HHSI Delivered          Available Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.2.8-2 KEY OPERATING PARAMETERS FOR FEEDWATER LINE BREAK Short-Term Core      Long-Term Core        Peak Primary Parameter Cooling              Cooling              Pressure Initial Core Power (MW)          2948                2958                  2958 Initial Pressurizer 2250                2200                  2300 Pressure (psia)
Initial Pressurizer Level 53.25                53.25                81.75
(% of level span)
Initial RCS Average 588.8                593.9                593.9 Temperature Initial Reactor Coolant 296,380              290,000                290,000 Flow (gpm)
Feedwater Temperature 440(a)                440(a)                440(a)
(&deg;F)
 
(a) Safety analyses support operating with a reduced main feedwater temperature of 375&deg;F at full power.
Amendment 63                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE 15.2.8-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR FEEDWATER LINE BREAK Short-Term Core        Long-Term Core Peak Primary Pressure Parameter              Cooling                Cooling BOC BOC                    EOC Moderator                                  -57.5 HFP, ARO Temperature                0.0                                      0.0 Coefficient (pcm/&deg;F)                            -50.0 HFP, ARI Doppler Coefficient
                              -1.09                  -1.25            -1.09 (pcm/&deg;F)
Delayed Neutron 0.007                0.006              0.007 Fraction
      / l (sec-1)          466.67                  333.33              466.67 Scram Worth (pcm)            5000                  5000              5000 Amendment 63                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.2.8-4 EVENT
 
==SUMMARY==
FOR FEEDLINE BREAK SHORT-TERM CORE COOLING Event                    Time (sec)
MFW line break initiated at SG-1                                  0.01 Minimum DNBR observed                                              7.6 SIS low steam line pressure setpoint reached                      8.1 Reactor trip on SIS                                              10.1 MSIS on low steam line pressure                                  10.1 Turbine trip on reactor trip; loss of offsite power              10.1 RCPs trip on loss of offsite power                                13.1 Intact SGs isolated from blowdown through ruptured SG            15.1 Faulted SG NR level off scale low                                25.9 Transient simulation terminated                                  30.0 Amendment 63                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.2.8-5 EVENT
 
==SUMMARY==
FOR FEEDLINE BREAK LONG-TERM CORE COOLING WITH OFFSITE POWER AVAILABLE Event                                                          Time (sec)
MFW line break initiated at SG-1                                  0.01 SIS low steam line pressure setpoint reached                        8.2 Reactor trip on SIS                                                10.2 MSIS on low steam line pressure                                    10.2 Turbine trip on reactor trip                                      10.2 Intact SGs isolated from blowdown through ruptured SG              15.2 Faulted SG NR level off scale low                                  26 First intact SG MSSV opened                                        29 AFW isolation on high steam pressure differential plus delay        57 AFW flow began to two intact SGs                                    70 Operator control of AFW assumed (30 minutes)                      1800 Transient simulation terminated                                  14400 Gradual primary system cooldown maintained RCPs on HHSI off (operator action)
SG heat removal capacity exceeds decay heat load Amendment 63                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                UFSAR Chapter: 15 TABLE 15.2.8-6 FEEDLINE BREAK Impact of Operator's Failure Components and                      to Take Action or Alarms to Alert                      Instructions Given    Instrumentation  Impact of Single      the Operator the operator to                        to the Operator      Necessary to        Active        Taking a Closely Required            initiate a        Delay Time    for Performing the      Complete        Component            Related but Operator Action  Particular Action      Assumed        Required Action    Indicated Action      Failure        Erroneous Action A. The operator    A. The operator      Within 30 minutes B. Maintain        B.1 Steam        None              B.1 The operator controls AFW to    will use individual                    proper S/G level    Generator AFW                        fails to modulate the intact steam  S/G level                              in intact S/Gs. If  modulation valves                    AFW flows to generators and    indicators to                          possible,          and controls.                        intact steam controls          control AFW flow                      maximize AFW                                            generators:
cooldown.          to each of the                        flow to intact S/Gs                                      Overfilling of a steam generators.                      to help lower                                            steam generator Hight level and                        primary                                                  may occur.
low level alarms                      temperature are provided.
B. The operator    The operator will    Within 30 minutes Limit safety        HHSI isolation    None              Liquid will be controls safety    use pressurizer                        injection and use  valves                              discharged from injection          level indication to                    letdown flow if                                          the pressurizer prevent/limit liquid                  necessary to                                            with loss of RCS discharge through                      prevent liquid-                                          pressure control.
the safety valves.                    solid pressurizer conditions Amendment 61                                                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE 15.2.8-7 EVENT
 
==SUMMARY==
FOR FEEDLINE BREAK LONG-TERM CORE COOLING WITH OFFSITE POWER LOST Event                                                              Time (sec)
MFW line break initiated at SG-1                                      0.01 SIS low steam line pressure setpoint reached                          8.2 Reactor trip on SIS                                                  10.2 MSIS on low steam line pressure                                      10.2 Turbine trip on reactor trip; loss of offsite power                  10.2 RCPs trip on loss of offsite power                                    13.2 Intact SGs isolated from blowdown through ruptured SG                15.2 Faulted SG NR level off scale low                                      26 First intact SG MSSV opened                                            27 AFW isolation on high steam pressure differential plus delay          57 AFW flow began to two intact SGs                                      70 Operator control of AFW assumed (30 minutes)                          1800 Transient simulation terminated                                      3600 Gradual primary system cooldown maintained RCPs coasting in natural circulation flow HHSI off (operator action)
SG heat removal capacity exceeds decay heat load Amendment 63                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.2.8-8 EVENT
 
==SUMMARY==
FOR FEEDLINE BREAK PEAK PRIMARY PRESSURE Event                                                                Time (sec)
MFW line break initiated at SG-1                                        0.01 SIS low steam line pressure setpoint reached                              7.9 Reactor trip on SIS                                                      9.9 MSIS on low steam line pressure                                          9.9 Turbine trip on reactor trip                                              9.9 Intact SGs isolated from blowdown through ruptured SG                    15 HHSI flow initiated based on low steam line pressure and 10 sec delay      20 Minimum pressurizer pressure (2148 psia)                                  22 First intact SG MSSV opened                                              24 Faulted SG NR level off scale low                                        26 AFW isolation on high steam pressure differential plus delay              57 AFW flow began to two intact SGs                                          69 Pressurizer SRVs first cycle                                              292 Pressurizer liquid full                                                  726 Maximum RV pressure (2643 psia) observed                                1286 Operator control of HHSI and AFW assumed (30 minutes)                    1800 HHSI terminated by operator due to high pressurizer liquid level        1800 Transient simulation terminated                                          3600 Decay heat removal by AFW and SG MSSVs HHSI off (operator action)
RCPs on Pressurizer water solid Amendment 63                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.3.2-1 INPUT PARAMETERS AND BIASING FOR LOSS OF FORCED REACTOR COOLANT FLOW Parameter                              Short-Term Core Cooling Core Power                                        SCD Pressurizer Pressure                                  SCD Pressurizer Level                                    Low RCS Average Temperature                                  SCD Reactor Coolant Flow                                  SCD Fuel Temperature                                    High Steam Generator Narrow Range Level                          Nominal Steam Generator Tube Plugging                              High Feedwater Temperature                                Nominal Cycle Exposure                                      BOC Moderator Temperature Coefficient                    BOC Least Negative Doppler Temperature Coefficient                    BOC Least Negative Delayed Neutron Fraction                              Maximum
                      /l                                        Maximum Minimum with most reactive rod stuck out of Reactor Trip Reactivity Insertion the core Pressurizer Heaters                                Disabled Pressurizer Spray                                  Available Pressurizer Power-Operated Relief Valves                      Available Pressurizer Safety Valves                              Nominal Main Steam Safety Valves                              Nominal Auxiliary Feedwater                                Insensitive Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 TABLE 15.3.2-2 KEY OPERATING PARAMETERS FOR LOSS OF FORCED REACTOR COOLANT FLOW Parameter                                Short-Term Core Cooling Initial Core Power (MW)                                    2948 Initial Pressurizer Pressure (psia)                              2250 Initial Pressurizer Level (% of level span)                          53.25 Initial RCS Average Temperature (&deg;F)                                588.8 Initial Reactor Coolant System Flow Rate (gpm)                            296,380 Feedwater Temperature (&deg;F)                                    440 (a)
(a) Safety analyses support operating with a reduced main feedwater temperature of 375 &deg;F at full power.
Amendment 63                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE 15.3.2-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR LOSS OF FORCED REACTOR COOLANT FLOW Short-Term Core Cooling Parameter BOC Moderator Temperature Coefficient (pcm/&deg;F)                    0.0 Doppler Coefficient (pcm/&deg;F)                          -1.09 Delayed Neutron Fraction                            0.007
                  / l (sec-1)                              466.67 Scram Worth (pcm)                                  5000 Amendment 63                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                            UFSAR Chapter: 15 TABLE 15.3.2-4 EVENT
 
==SUMMARY==
FOR LOSS OF FORCED REACTOR COOLANT - UNDERFREQUENCY CASE Event                            Time (sec)
Initiate underfrequency event                            0.0 Reactor Scram (Underfrequency)                            1.5 Turbine Trip                                              2.0 Pressurizer Spray Actuates                                2.5 MDNBR                                                    3.3 Pressurizer Spray Closes                                  6.5 End of Simulation                                        10.0 Amendment 63                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 TABLE 15.3.3-1 INPUT PARAMETERS AND BIASING FOR LOCKED ROTOR Parameter              Short-Term Core Cooling    Peak Primary Pressure Core Power                        SCD                        High Pressurizer Pressure                  SCD                        High Pressurizer Level                  Low                        High RCS Average Temperature                  SCD                        High Reactor Coolant Flow                  SCD                        Low Fuel Temperature                    High                      High Steam Generator Narrow Range              Nominal                    Nominal Level Steam Generator Tube Plugging                High                      High Feedwater Temperature                  Nominal                    Nominal Cycle Exposure                    BOC                        BOC Moderator Temperature            BOC Least Negative        BOC Least Negative Coefficient Doppler Temperature              BOC Least Negative        BOC Least Negative Coefficient Delayed Neutron Fraction              Maximum                    Maximum
                /l                      Maximum                    Maximum Reactor Trip Reactivity Insertion Minimum with most reactive Minimum with most reactive rod stuck out of the core  rod stuck out of the core Pressurizer Heaters                Disabled                  Available Pressurizer Spray                  Available                  Disabled Pressurizer Power-Operated              Available                  Disabled Relief Valves Pressurizer Safety Valves              Nominal                    Nominal Main Steam Safety Valves                Nominal                    Nominal Auxiliary Feedwater                Insensitive                Insensitive Amendment 63                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 TABLE 15.3.3-2 KEY OPERATING PARAMETERS FOR LOCKED ROTOR Parameter                  Short-Term Core Cooling            Peak Primary Pressure Initial Core Power (MW)                        2948                              2958 Initial Pressurizer Pressure 2250                              2300 (psia)
Initial Pressurizer Level (% of 53.25                            81.75 level span)
Initial RCS Average 588.8                            593.8 Temperature Initial Reactor Coolant Flow 296,380                          290,000 (gpm)
Feedwater Temperature (&deg;F)                        440(a)                            440(a)
(a) Safety analyses support operating with a reduced main feedwater temperature of 375 &deg;F at full power.
Amendment 63                                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                            UFSAR Chapter: 15 TABLE 15.3.3-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR LOCKED ROTOR Short-Term Core Cooling Peak Primary Pressure Parameter BOC                  BOC Moderator Temperature 0.0                  0.0 Coefficient (pcm/&deg;F)
Doppler Coefficient (pcm/&deg;F)            -1.09                -1.09 Delayed Neutron Fraction                0.007                0.007
        /l (sec-1)                      466.67                466.67 Scram Worth (pcm)                      5000                5000 Amendment 63                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                              UFSAR Chapter: 15 TABLE 15.3.3-4 EVENT
 
==SUMMARY==
FOR LOCKED ROTOR OVERPRESSURIZATION CASE Event                                                  Time (s)
Initiate transient (seizure of RCP-1 motor)            0.00 Low RCS flow signal                                    0.04 Reactor scram                                          1.04 Main turbine trip                                      1.05 Reverse flow in affected loop                          1.45 Pressurizer safety valves open                          3.18 Maximum primary side pressure                          3.90 Remaining RCS pumps tripped                            4.05 Maximum pressurizer level                              5.30 Amendment 63                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.3.3-5 EVENT
 
==SUMMARY==
FOR LOCKED ROTOR MDNBR CASE Event                                                  Time (sec)
Initiate transient (seizure of RCP-2 motor)            0.00 Low RCS flow signal                                    0.04 Pressurizer spray valve opens                          1.0 Reactor scram                                          1.04 Reverse flow in affected loop                          1.4 Main turbine trip                                      1.54 Pressurizer PORV open                                  2.0 Pressurizer PORV closes                                2.3 Maximum primary side pressure                          2.5 Minimum DNBR                                            3.0 Maximum pressurizer level                              4.1 Remaining RCS pumps tripped                            4.54 Pressurizer spray valve closes                          5.0 Amendment 63                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                          UFSAR Chapter: 15 TABLE 15.3.3-6 PARAMETERS USED IN THE LOCKED ROTOR RADIOLOGICAL ANALYSIS Source Term Core Activity                                                                          See Table 15.0.9-1 Fraction of fuel rods in core assumed to fail for dose considerations                  8 Centerline melted fuel (%)                                                            0 Radial peaking factor                                                                  1.73 Gap Fractions (% of core activity)
I-131                                                                              8 Kr-85                                                                              10 Other Iodine and Noble Gas nuclides                                                5 Alkali Metals                                                                      12 Iodine Chemical form after released to atmosphere (%)
Elemental                                                                          97 Organic                                                                            3 Particulate (cesium iodide)                                                        0 Reactor coolant noble gas activity prior to accident (%fuel defect level)              1.0 Secondary coolant iodine activity prior to accident (Ci/gm of DE I-131)              0.1 Secondary coolant alkali metal activity prior to accident (% of primary concentration) 10 Release Modeling SG tube leak rate (gpm total)                                                          1 Steam release to environment (lbm) 0 - 2 hours                                                                        378,000 2 - 8 hours                                                                        965,000
  >8 hours                                                                            0 SG iodine and alkali metal water/steam partition coefficient                          0.01 RCS mass (lbm)                                                                        4.11E5 Secondary Side mass (lbm/per SG)                                                      115,585 Amendment 63                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.3.3-7 RADIOLOGICAL CONSEQUENCES OF A LOCKED ROTOR Exclusion Area Boundary*                                  1.08 rem TEDE Low Population Zone                                      0.70 rem TEDE Control Room                                              1.29 rem TEDE
* The exclusion area boundary doses reported are for the worst two hour period, determined to be from 6.0 to 8.0 hours.
Amendment 63                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.4.1-1 INPUT PARAMETERS AND BIASING FOR UNCONTROLLED BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITION Parameter                            Short-Term Core Cooling Core Power                                        Zero Pressurizer Pressure                                  SCD Pressurizer Level                                    Low RCS Average Temperature                                  SCD Reactor Coolant Flow                                  SCD Fuel Temperature                                    N/A Steam Generator Narrow Range Level                            High Steam Generator Tube Plugging                                Low Feedwater Temperature                                Nominal Cycle Exposure                                    BOC Moderator Temperature Coefficient                    BOC Least Negative Doppler Temperature Coefficient                    BOC Least Negative Delayed Neutron Fraction                              Maximum Prompt Neutron Lifetime                              Maximum Minimum with most reactive rod stuck out of Reactor Trip Reactivity Insertion the core Differential Worth                                Maximum Pressurizer Heaters                                Disabled Pressurizer Spray                                Available Pressurizer Power-Operated Relief Valves                      Available Pressurizer Safety Valves                                Low Main Steam Safety Valves                                  Low Auxiliary Feedwater                              Insensitive Reactor Coolant Pumps Operational                              2 Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.4.1-2 KEY OPERATING PARAMETERS FOR UNCONTROLLED BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITION Parameter                        Short-Term Core Cooling Initial Core Power (MW)                          2.9E-6 Initial Pressurizer Pressure (psia)                      2250 Initial Pressurizer Level (% of level span)                  18.25 Initial RCS Average Temperature (&deg;F)                      557.0 Initial Reactor Coolant System Flow Rate (gpm)
Two combined loops with pumps operating            212,344 Loop without pump operating        -25,354 Feedwater Temperature (&deg;F)                            100 Amendment 63                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.4.1-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR UNCONTROLLED BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER CONDITIONS Short-Term Core Cooling Parameter (BOC)
Moderator Temperature Coefficient (pcm/&deg;F)                      +5.0 Doppler Coefficient (pcm/&deg;F)              Least Negative Doppler Defect Table Delayed Neutron Fraction                                0.007
                / l (sec-1)                                    350.0 Scram Worth (pcm)                                    1770 Maximum Reactivity Insertion Rate (pcm/s)                        55 Amendment 63                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.4.1-4 EVENT
 
==SUMMARY==
FOR UNCONTROLLED RCCA BANK WITHDRAWAL FROM A SUBCRITICAL OR LOW POWER STARTUP CONDITIONS Event                                                            Time (sec)
RCCA Withdrawal Initiated                                        30.0 Power Range High Neutron Flux Trip (low setting) Setpoint Reached 436.0 Scram Initiated                                                  436.5 Peak Nuclear Power                                                436.5 MDNBR                                                            437.0 Simulation End                                                    600.0 Amendment 63                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.4.2-1 INPUT PARAMETERS AND BIASING FOR UNCONTROLLED BANK WITHDRAWAL AT POWER Parameter              Short-Term Core Cooling    Peak Primary Pressure Core Power                        SCD                        Low Pressurizer Pressure                    SCD                        Low Pressurizer Level                    Low                        High RCS Average Temperature                    SCD                        High Reactor Coolant Flow                    SCD                      Nominal Fuel Temperature                      Low                        Low Steam Generator Narrow Range                  High                    Nominal Level Steam Generator Tube Plugging                High                      High Feedwater Temperature                  Nominal                    Nominal Cycle Exposure                    BOC/EOC                      BOC Moderator Temperature          BOC/EOC Least Negative        BOC Least Negative Coefficient Doppler Temperature            BOC/EOC Least Negative        BOC Least Negative Coefficient Delayed Neutron Fraction                Maximum                    Maximum
                /l                        Maximum                    Maximum Reactor Trip Reactivity Insertion  Minimum with most reactive Minimum with most reactive rod stuck out of the core  rod stuck out of the core Bank Reactivity Insertion Rate            Spectrum                  Maximum Pressurizer Heaters                  Disabled                  Available Pressurizer Spray                  Available                  Disabled Pressurizer Power-Operated                Available                  Disabled Relief Valves Pressurizer Safety Valves                Low                        High Main Steam Safety Valves                  High                      High Main Feedwater                      Manual                    Manual Auxiliary Feedwater                  Disabled                  Disabled Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.4.2-2 KEY OPERATING PARAMETERS FOR UNCONTROLLED BANK WITHDRAWAL AT POWER DNB ANALYSIS Parameter            100% RTP                50% RTP                10% RTP Short-Term Core        Short-Term Core        Short-Term Core Cooling                Cooling                Cooling Initial Core Power (MW)            2948                  1474                    294.8 Initial Pressurizer            2250                  2250                    2250 Pressure (psia)
Initial Pressurizer Level          53.25                  35.75                  21.75
(% of level span)
Initial RCS Average            588.8                  572.9                  560.18 Temperature Initial Reactor Coolant        296,380                296,380                296,380 Flow (gpm)
Feedwater Temperature              440(a)                  370                    223
(&deg;F)
(a) Safety analyses support operating with a reduced main feedwater temperature of 375 &deg;F at full power.
Amendment 63                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.4.2-2a KEY OPERATING PARAMETERS FOR UNCONTROLLED BANK WITHDRAWAL AT POWER PEAK PRIMARY PRESSURE ANALYSIS 8% RTP                12% RTP Parameter Peak Primary Pressure  Peak Primary Pressure Initial Core Power (MW)                235.84                353.76 Initial Pressurizer Pressure 2199.7                  2199.7 (psia)
Initial Pressurizer Level (% of 81.75                  81.75 level span)
Initial RCS Average Temperature              564.54                565.82 Initial Reactor Coolant Flow 296,380                296,380 (gpm)
Feedwater Temperature (&deg;F)                  198.4                  245.0 Amendment 63                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE 15.4.2-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR UNCONTROLLED BANK WITHDRAWAL AT POWER BOC / EOC              BOC / EOC        BOC / EOC Parameter 100% RTP                50% RTP        8% - 10% RTP Moderator Temperature                0 / -25                0 / -16        3.2 / -11 Coefficient (pcm/&deg;F)
Doppler Coefficient                                                Least Negative
                            -1.09 / -1.25          -1.16 / -1.31 (pcm/&deg;F)                                                    Doppler Defect Table Delayed Neutron 0.007 / 0.006          0.007 / 0.006    0.007 / 0.006 Fraction
      / l (sec-1)        466.67 / 333.33        466.67 / 333.33    466.67 / 333.33 Scram Worth (pcm)              5000                    3250            3250 Maximum Bank Reactivity Insertion        16 / 22.5            18.2 / 25.7      22 / 34.8 Rate (pcm/s)
Amendment 63                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.4.2-4 EVENT
 
==SUMMARY==
FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT POWER LIMITING MDNBR CASE WITH BOC KINETICS AND THE MOST LIMITING REACTIVITY INSERTION RATE AMONG A SPECTRUM OF ANALYZED VALUES.
Event                      Time(s)
Bank withdrawal began                                0.0 Pressurizer spray valve open                        10.9 Pressurizer PORVs start the cycle of opening and 19.3 - 46.4 closing MSSVs Banks 1 to 5 start opening                36.5 - 46.4 OTDT reactor trip setpoint reached                  42.2 Scram occurred (control rods start to insert)      43.4 MDNBR occurs                                        44.2 Simulation ends                                      60 Amendment 63                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                              UFSAR Chapter: 15 TABLE 15.4.2-5 EVENT
 
==SUMMARY==
FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT POWER LIMITING PRIMARY SIDE OVERPRESSURIZATION CASE WITH BOC KINETICS AND THE MOST LIMITING REACTIVITY INSERTION RATE Event                  Time(s)
Bank withdrawal initiates                        0.0 Pressurizer backup / control heaters on High pressurizer pressure trip setpoint reached  18.5 Scram occurs (control rods start to insert)      20.5 Turbine trips on reactor trip                    20.5 PSVs open                                        22.6 Maximum primary side pressure occurs            22.7 PSVs close                                      24.0 Simulation ends                                  50.0 Amendment 63                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.4.3-1 INPUT PARAMETERS AND BIASING FOR ROD DROP Parameter                              Short-Term Core Cooling Core Power                                        SCD Pressurizer Pressure                                    SCD Pressurizer Level                                    Low RCS Average Temperature                                    SCD Reactor Coolant Flow                                    SCD Fuel Temperature                                      Low Steam Generator Narrow Range Level                            Nominal Steam Generator Tube Plugging                                High Feedwater Temperature                                  Nominal Cycle Exposure                                BOC/MOC/EOC Moderator Temperature Coefficient                  BOC/MOC/EOC Least Negative Doppler Temperature Coefficient                  BOC/MOC/EOC Least Negative Delayed Neutron Fraction                              Maximum
                      /l                                        Maximum Minimum with most reactive rod stuck out of Reactor Trip Reactivity Insertion the core Control Bank Worth                                  Maximum Dropped Rod Worth                                  Spectrum Pressurizer Heaters                                  Disabled Pressurizer Spray                                  Available Pressurizer Power-Operated Relief Valves                      Available Pressurizer Safety Valves                              Nominal Main Steam Safety Valves                                Nominal Auxiliary Feedwater                                Available Rod Position Controller                              Automatic Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 TABLE 15.4.3-2 KEY OPERATING PARAMETERS FOR ROD DROP Parameter                                  Short-Term Core Cooling Initial Core Power (MW)                                      2948 Initial Pressurizer Pressure (psia)                                2250 Initial Pressurizer Level (% of level span)                            53.25 Initial RCS Average Temperature (&deg;F)                                  588.8 Initial Reactor Coolant System Flow Rate                              296,380 (gpm)
Feedwater Temperature (&deg;F)                                      440(a)
(a) Safety analyses support operating with a reduced main feedwater temperature of 375 &deg;F at full power.
Amendment 63                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 TABLE 15.4.3-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR ROD DROP Parameter                BOC                    MOC        EOC Moderator Temperature                0                    -12        -25 Coefficient (pcm/&deg;F)
Doppler Coefficient
                              -1.09                -1.15      -1.25 (pcm/&deg;F)
Delayed Neutron 0.007                0.0065      0.006 Fraction
      / l (sec-1)            466.67                393.94      333.33 Scram Worth (pcm)            5000                  5000      5000 Dropped Rod Worth 120 - 190            160 - 280    275 - 450 (pcm)
Amendment 63                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 TABLE 15.4.3-4a EVENT
 
==SUMMARY==
FOR A LIMITING DROPPED ROD CASE AT EOC Event                                        Time (sec)
Start Transient Simulation                  0 Initiate Rod Drop                            1.00E-06 Initiate Control Bank D Withdrawal          2.7 Initiate Pressurizer Spray                  28.2 Reach Minimum DNB Ratio                      33 Initiate Control Bank D Insertion            37.4 Reactor Trip (OPDT 2/3)                      41.9 Initiate Turbine Trip                        42.9 Terminate Pressurizer Spray                  45.1 End Transient Simulation                    50.0 Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                          UFSAR Chapter: 15 TABLE 15.4.3-5 PARAMETERS USED IN THE SINGLE RCCA WITHDRAWAL RADIOLOGICAL ANALYSIS Source Term Core Activity                                                                          See Table 15.0.9-1 Fraction of fuel rods in core assumed to fail for dose considerations                  9 Centerline melted fuel (%)                                                            0 Radial peaking factor                                                                  1.73 Gap Fractions (% of core activity)
Iodine and Noble Gas nuclides                                                      10 Alkali Metals                                                                      12 Iodine Chemical form after released to atmosphere (%)
Elemental                                                                          97 Organic                                                                            3 Particulate (cesium iodide)                                                        0 Reactor coolant noble gas activity prior to accident (% fuel defect level)            1.0 Secondary coolant iodine activity prior to accident (Ci/gm of DE I-131)              0.1 Secondary coolant alkali metal activity prior to accident (% of primary concentration) 10 Release Modeling SG tube leak rate (gpm total)                                                          1 Steam release to environment (lbm) 0 - 2 hours                                                                        378,000 2 - 8 hours                                                                        965,000
    >8 hours                                                                          0 SG iodine and alkali metal water/steam partition coefficient                          0.01 RCS mass (lbm)                                                                        4.11E5 Secondary Side mass (lbm/per SG)                                                      115,585 Amendment 63                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.4.3-6 INPUT PARAMETERS AND BIASING FOR SINGLE RCCA WITHDRAWAL Parameter                          Short-Term Core Cooling Core Power                                    SCD Pressurizer Pressure                                SCD Pressurizer Level                                Low RCS Average Temperature                                SCD Reactor Coolant Flow                                SCD Fuel Temperature                                  Low Steam Generator Narrow Range Level                        Nominal Steam Generator Tube Plugging                            High Feedwater Temperature                              Nominal Cycle Exposure                                  BOC Moderator Temperature Coefficient                BOC Least Negative Doppler Temperature Coefficient                  BOC Least Negative Delayed Neutron Fraction                          Maximum
                        /l                                  Maximum Reactor Trip Reactivity Insertion      Minimum with most reactive rod stuck out of the core Reactivity Insertion Rate                        Maximum Withdrawable Rod Worth                            Maximum Pressurizer Heaters                            Available Pressurizer Spray                              Disabled Pressurizer Power-Operated Relief Valves                    Available Pressurizer Safety Valves                          Nominal Main Steam Safety Valves                              High Auxiliary Feedwater                            Disabled Amendment 63                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.4.3-6a RADIOLOGICAL CONSEQUENCES OF SINGLE RCCA WITHDRAWAL Exclusion Area Boundary*    1.64 rem TEDE Low Population Zone        1.17 rem TEDE Control Room                1.97 rem TEDE
*The exclusion area boundary doses reported are for the worst two hour period, determined to be from 6.0 to 8.0 hours.
Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 TABLE 15.4.3-7 KEY OPERATING PARAMETERS FOR SINGLE RCCA WITHDRAWAL Parameter                                  Short-Term Core Cooling Initial Core Power (MW)                                      2948 Initial Pressurizer Pressure (psia)                                2250 Initial Pressurizer Level (% of level span)                            53.25 Initial RCS Average Temperature (&deg;F)                                  588.8 Initial Reactor Coolant System Flow Rate                              296,380 (gpm)
Feedwater Temperature (&deg;F)                                      440(a)
(a) Safety analyses support operating with a reduced main feedwater temperature of 375 &deg;F at full power.
Amendment 63                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.4.3-8 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR SINGLE RCCA WITHDRAWAL Short-Term Core Cooling Parameter BOC Moderator Temperature Coefficient (pcm/&deg;F)                  0 Doppler Coefficient (pcm/&deg;F)                        -1.09 Delayed Neutron Fraction                          0.007
                    / l (sec-1)                            466.67 Scram Worth (pcm)                              5000 Maximum RCCA Reactivity Insertion Rate (pcm/s)                2.25 Maximum RCCA Withdrawable Rod Worth (pcm)                      70 Amendment 63                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 TABLE 15.4.3-9 EVENT
 
==SUMMARY==
FOR 100% RTP SINGLE RCCA WITHDRAWAL REPRESENTATIVE MDNBR CASE Event                      Time (s)
Bank withdrawal began                                        0.0 Pressurizer PORVs start the cycle of opening and closing  26.1 - 133.6 OTDT reactor trip setpoint reached                          134.8 MDNBR occurs                                                136.0 Control rods start to insert                                136.0 Simulation ends                                              166.0 Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                          UFSAR Chapter: 15 TABLE 15.4.6 2 ADMINISTRATIVE CONTROLS TO PREVENT DILUTION VALVE NO. DESCRIPTION                                    REQUIRED POSITION DURING REFUELING 1-8455        Reactor Makeup Water to CVCS                    Lock closed; may be opened to permit Makeup Control System makeup to Refueling Water Storage Tank provided valves 1-FCV-113B and 1-FCV-114A are maintained closed with their main control board control switches in "shut" position, and manual valves 1-8441, 1-8454 and 1-8439 are locked closed.
1-8308        Boric Acid Batch Tank Outlet                    Locked closed; may be opened provided the boron concentration of the boric acid batch tank  *ppmB and valve 1-8302 is closed.
1-8302        Reactor Makeup Water to Boric Acid Batch        Lock closed, may be opened provided valve 1-8308 is Tank                                            closed.
1-8513        Resin Sluice to CVCS Demineralizers            Lock closed.
1-8629A      Boron Recycle Evaporator Feed Pump to          Lock closed.
Charging/SI Pumps 1-7054        CVCS Letdown to Boron Thermal                  Closed with main control board control switch in "shut" Regeneration System                            position, and BTRS function selector switch maintained in "off" position; no lock required.
1-7004        Reactor Makeup Water to Boron Thermal          Lock closed.
Regeneration System 1-7052        Resin Sluice to BTRS Demineralizers            Lock closed.
1-8545        Boron Thermal Regeneration System              Opened with main control board control switch Bypass                                          maintained in "open" position; no lock required.
* The boron concentration determined by the plant's Technical Specifications for Refueling.
Amendment 63                                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.4.7-1 KEY OPERATING PARAMETERS FOR INADVERTENT LOADING AND OPERATION OF A FUEL ASSEMBLY IN AN IMPROPER POSITION Parameter                          Short-Term Core Cooling Initial Core Power (MW)                              2948 Initial Pressurizer Pressure (psia)                        2250 Initial RCS Average Temperature (&deg;F)                          588.8 Initial Reactor Coolant System Flow Rate                      296,380 (gpm)
Amendment 63                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 TABLE 15.4.8-1 INPUT PARAMETERS AND BIASING FOR ROD EJECTION ACCIDENTS Parameter                        Non-SCD Cases*                  SCD Cases**
Spectrum Core Power                                                            SCD (1e-7 %RTP - 100.34 %RTP)
Pressurizer Pressure                        Low                            SCD Pressurizer Level                          N/A                          Low RCS Average Temperature                        High                          SCD Reactor Coolant System Flow Rate                    Low                            SCD Fuel Temperature                        Spectrum                          Low Steam Generator Narrow Range N/A                        Nominal Level Steam Generator Tube Plugging                      N/A                          High Feedwater Temperature                          N/A                        Nominal Cycle Exposure                          Spectrum                        BOC Moderator Temperature Coefficient        Spectrum Least Negative          BOC Least Negative Doppler Coefficient              Spectrum Least Negative          BOC Least Negative Delayed Neutron Fraction                      Minimum                      Maximum
                  /l                                N/A                        Maximum Minimum with most Minimum with most reactive rod      reactive rod stuck out Reactor Trip Reactivity Insertion        stuck out of the core and          of the core and ejected rod out of the core      ejected rod out of the core Ejected Rod Worth                        Maximum                        Maximum Pressurizer Heaters                          N/A                        Available Pressurizer Spray                          N/A                        Disabled Pressurizer Power-Operated Relief N/A                        Available Valves Pressurizer Safety Valves                        N/A                        Nominal Main Steam Safety Valves                        N/A                        Nominal Auxiliary Feedwater                          N/A                        Disabled
*Non-SCD biasing for VIPRE-01 DNB analysis of cases which experience a flux-based reactor trip
**Parameters applicable to the RETRAN-3D/VIPRE-01 analysis performed to evaluate a case which did not experience a flux-based trip Amendment 63                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                            UFSAR Chapter: 15 TABLE 15.4.8-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR ROD EJECTION LIMITING CASE (EOC, 50% RTP) 50% RTP Case Parameter EOC Moderator Temperature Coefficient (pcm/&deg;F)              -16 Doppler Coefficient (pcm/&deg;F)                    -1.31 Delayed Neutron Fraction                      0.0045 Ejected Rod Worth (pcm)                        375 Amendment 63                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.4.8-4B EVENT
 
==SUMMARY==
FOR CONTROL ROD EJECTION LIMITING CASE (EOC, 50% RTP)
Event                                                    Time(s)
Reactor at EOC 50% RTP Conditions, RCCA Election          0.0 Power Range High Positive Neutron Flux Rate Trip          0.06 RCCA Fully Ejected                                        0.08 Peak Power                                                0.12 Maximum Fuel Pellet Enthalpy                              1.13 Maximum Fuel Temperature                                  1.27 Transient Terminated                                      3.0 Amendment 63                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                          UFSAR Chapter: 15 TABLE 15.4.8-5 Parameters Used for the Rod Cluster Control Assembly Ejection Radiological Analysis Source Team Core Activity                                                                              See Table 15.0.9-1 Fraction of fuel rods in core that fail (% of core)                                        10 Gap Fractions (% of core activity)
Iodine                                                                            10 Noble Gas                                                                          10 Alkali Metals                                                                      12 Fraction of fuel melting (% of core)                                                        0 Radial peaking factor                                                                      1.73 Reactor coolant noble gas and alkali metal activity prior to accident (% fuel defect level) 1.0 Reactor coolant iodine activity prior to accident (Ci/gm of DE I-131)                      1.0 Secondary coolant iodine activity prior to accident (Ci/gm of DE I-131)                    0.1 Secondary coolant alkali metal activity prior to accident (% of primary concentration)      10 Containment Leakage Release Path Containment net free volume (ft3)                                                          2.344E6 Containment leak rates (weight %/day) 0 - 24 hours                                                                      0.1
        > 24 hours                                                                        0.05 Iodine chemical form in containment (%)
Elemental                                                                          4.85 Organic                                                                            0.15 Particulate (cesium iodide)                                                        95 Spray removal in containment                                                                Not Credited Sedimentation removal in containment (hr-1)
Iodines                                                                            Not Credited Alkali metals                                                                      0.1 Primary to secondary Leakage Release Path SG tube leak rate (gpm total)                                                              1.0 Steam release to environment (lbm)*
0 - 2 hours                                                                        378,000
        > 2 hours                                                                          0 SG iodine and alkali metal water/steam partition coefficient                                0.01 Iodine Chemical form after release to atmosphere (%)
Elemental                                                                          97 Organic                                                                            3 Particulate (cesium iodide)                                                        0
*A separate case was run with secondary steam release continuing for 8 hours (2 - 8 hour steam release = 965,000 lbm) per Reg. Guide 1.183, Appendix H, Section 7 with no containment release. The reported dose consequence results bound the 8 hour secondary release (only) dose consequence.
Amendment 63                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.4.8-6 Radiological Consequences of a Rod Cluster Control Assembly Ejection Accident Exclusion Area Boundary*                  2.94 rem TEDE Low Population Zone                        4.32 rem TEDE Control Room                              2.75 rem TEDE
*The exclusion area boundary doses reported are for the worst two hour period, determined to be from 0.0 to 2.0 hours.
Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.5.1-1 ASSUMED STATE OF PLANT SYSTEMS Parameter                                    PZR Overfill Automatic Rod Control                          Manual Pressurizer Heaters                            Disabled Pressurizer Spray                            Available Pressurizer PORVs                      Available at first, then assumed to lose motive power Steam Bypass Valves                        Not Modeled Steam Atmospheric Dump Valves              Not Modeled Main Feedwater                        Auto (no consequence)
Auxiliary Feedwater                          Available Charging/SIPumps                    2 Pumps (Maximum Flow)
SI Boron Concentration              Maximum per Tech. Specs.
(2600 ppm)
SI Unborated Purge Volume (total)              23.13 ft3 SI Fluid Temperature                            40&deg;F Letdown Flow                                Not Modeled Turbine Control                        Auto / Load Demand Cycle Exposure                                  EOC Initial Core Power                            2958 MWt (rated + 0.34%)
RCS Average Temperature                        588.8 &deg;F Steam Generator Tube Plugging                    3%
Level RCS Flow                                Tech. Spec. minimum (293,540 gpm)
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Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE 15.5.1-2 INPUT BIASING Parameter                                        PZR Overfill Case Scram Reactivity Worth                          Minimum allowed shutdown margin and the most reactive rod stuck out of the core High Pressurizer Trip Setpoint                  Nominal + Uncertainty Low Pressurizer Trip Setpoint                    Nominal - Uncertainty Moderator Temperature Reactivity Coefficient    EOC Bounding Minimum Doppler Reactivity Coefficient                  EOC Bounding Minimum Delayed Neutron Data                            Nominal EOC Prompt Neutron Lifetime (/)                    Nominal EOC Pressurizer Safety Valve Open Setpoint          Nominal + Tolerance Pressurizer Safety Valve Stroke Time            Nominal (Initial lift includes delay for loop seal purge)
Non-Compensated Pressurizer PORV Open Setpoint  Nominal MS Safety Valve Open Setpoint                    Nominal + Tolerance Initial Pressurizer Level                        Nominal - Uncertainty Initial Pressurizer Pressure                    Nominal Amendment 63                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 TABLE 15.5.1-5 SEQUENCE OF EVENTS (PRESSURIZER OVERFILL CASE)
Event                                                                Time (seconds)
Inadvertent actuation of HHSI system: (maximum flow from 2 SI pumps)        0.0 RPS trip signal on SIS signal                                              2.025 Main turbine trip                                                          2.05 Pressurizer spray actuates                                                  3.2 Compensated pressurizer PORV opens                                          3.5 Pressurizer spray terminated                                                6.5 Compensated pressurizer PORV closes                                        7.5 Minimum pressurizer pressure (2200.1 psia)                                  8.5 Minimum pressurizer level (52.87% of span)                                  10.0 Non-borated water cleared from SI lines                                    ~27.0 Compensated PORV opens                                                      578.5 Pressurizer level = 100% span                                              591.5 Pressurizer liquid level = 463.15 inches                                    900.0 Pressurizer PORVs assumed disabled                                          900.0 Pressurizer pressure = 2267 psia                                            900.0 Pressurizer filled with liquid (level = 463.38 inches)                      918.5 Pressurizer SRV opens                                                      945.0 Pressurizer SRV closes                                                      947.0 Pressurizer pressure ranging between 2450 and 2550 psia                    950-1200 Minimum SRV inlet temperature = ~ 564&deg;F (while valves are open)            950-1200 Transient terminated                                                        1200.0 Amendment 61                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.6.1-1 INPUT PARAMETERS AND BIASING FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR PORV Parameter                            Short-Term Core Cooling Core Power                                      SCD Pressurizer Pressure                                SCD Pressurizer Level                                  Low RCS Average Temperature                                SCD Reactor Coolant Flow                                SCD Fuel Temperature                                  High Steam Generator Narrow Range Level                        Nominal Steam Generator Tube Plugging                              Low Feedwater Temperature                              Nominal Cycle Exposure                                    BOC Moderator Temperature Coefficient                  BOC Least Negative Doppler Temperature Coefficient                    BOC Most Negative Delayed Neutron Fraction                            Maximum
                      /l                                      Maximum Minimum with most reactive rod stuck out of Reactor Trip Reactivity Insertion the core Pressurizer Heaters                              Disabled Pressurizer Spray                                Available Pressurizer Power-Operated Relief Valves                    Available Pressurizer Safety Valves                            Nominal Main Steam Safety Valves                            Nominal Auxiliary Feedwater                              Available Amendment 63                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 TABLE 15.6.1-2 KEY OPERATING PARAMETERS FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR PORV Parameter                                  Short-Term Core Cooling Initial Core Power (MW)                                      2948 Initial Pressurizer Pressure (psia)                                2250 Initial Pressurizer Level (% of level span)                            53.25 Initial RCS Average Temperature (&deg;F)                                  588.8 Initial Reactor Coolant System Flow Rate                              296,380 (gpm)
Feedwater Temperature (&deg;F)                                      440 (a)
(a) Safety analyses support operating with a reduced main feedwater temperature of 375 &deg;F at full power.
Amendment 63                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 TABLE 15.6.1-3 RANGE OF NEUTRONICS PARAMETERS SUPPORTED BY ANALYSIS FOR INADVERTENT OPENING OF A PRESSURIZER SAFETY OR PORV Short-Term Core Cooling Parameter BOC Moderator Temperature Coefficient (pcm/&deg;F)                      0.0 Doppler Coefficient (pcm/&deg;F)              Most Negative Doppler Defect Table Delayed Neutron Fraction                              0.007
                  / l (sec-1)                                466.67 Scram Worth (pcm)                                    5000 Amendment 63                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 TABLE 15.6.1-4 EVENT
 
==SUMMARY==
FOR INADVERTENT OPENING OF A PRESSURIZER SAFETY OR PORV Event                                                  Time (sec)
Pressurizer safety relief valve opened                  0.01 Low pressurizer pressure trip setpoint reached          20.1 Reactor trip                                            22.1 Minimum departure from nucleate boiling ratio          22.6 Turbine Trip on SCRAM                                  23.1 Steam dump on turbine trip                              23.1 SI Signal on low pressurizer pressure                  32.8 Main feedwater isolation signal                        32.8 Auxiliary feedwater starts on SI Signal                32.8 Calculation terminated                                  50.0 Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 TABLE 15.6.2-1 PARAMETERS USED FOR LETDOWN LINE BREAK OUTSIDE CONTAINMENT RADIOLOGICAL ANALYSIS Reactor coolant noble gas activity prior to accident (% fuel defect level)          1.0 Reactor coolant iodine activity prior to accident (Ci/gm of DE I-131)              1.0 Reactor coolant iodine appearance rate increase due to the accident-initiated spike (times equilibrium rate)                                                500 (Table 15.0.9-6)
Letdown line break flow (gpm)                                                      200 Duration of letdown line break (minutes)                                            30 Break flow flashing fraction                                                        0.4 Amendment 61                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.6.2-2 RADIOLOGICAL CONSEQUENCES OF A LETDOWN LINE BREAK OUTSIDE CONTAINMENT Exclusion Area Boundary*          2.31 rem TEDE Low Population Zone                0.52 rem TEDE Control Room                      1.04 rem TEDE
* The exclusion area boundary doses reported are for the worst two hour period, determined to be from 0.0 to 2.0 hours.
Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 TABLE 15.6.3-1 SHNPP SGTR ANALYSIS OPERATOR ACTION TIMES FOR MARGIN TO OVERFILL ANALYSIS Action                                                    Time Intervals Isolate auxiliary feedwater flow to ruptured SG          Maximum of 8.8 minutes or LOFTTR2 calculated time to reach 30% narrow range level in the ruptured SG Isolate steam flow from ruptured SG                      Maximum of 12 minutes or LOFTTR2 calculated time to reach 30% narrow range level in the ruptured SG Operator action time to initiate cooldown                5 min Cooldown                                                  Calculated by LOFTTR2 Operator action time to initiate depressurization        4 min Depressurization                                          Calculated by LOFTTR2 Operator action time to initiate SI termination          3 min SI termination and pressure equalization                  Calculated by LOFTTR2 Amendment 61                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                            UFSAR Chapter: 15 TABLE 15.6.3-2 SHNPP SGTR ANALYSIS SEQUENCE OF EVENTS MARGIN TO OVERFILL ANALYSIS Event                                              Time (sec)
SG Tube Rupture                                    0 Reactor Trip                                      112 Safety Injection                                  145 AFW Flow to Ruptured SG Isolated                  528 Ruptured SG Isolated                              720 RCS Cooldown Initiated                            1020 RCS Cooldown Terminated                            1356 RCS Depressurization Initiated                    1598 RCS Depressurization Terminated                    1680 SI Terminated                                      1860 Break Flow Terminated                              2700 Amendment 61                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 TABLE 15.6.3-3 OPERATOR ACTION TIMES FOR SGTR OFFSITE DOSE ANALYSIS Action                              Time (min)
Isolate auxiliary feedwater flow to Maximum of 10 minutes or LOFTTR2 calculated time to ruptured SG                        reach 30% narrow range level in the ruptured SG Isolate steam flow from ruptured SG Maximum of 12 minutes or LOFTTR2 calculated time to reach 30% narrow range level in the ruptured SG Close block valve to isolate failed 20 min after valve fails to close at the time of isolation of open PORV on ruptured S/G          ruptured S/G Operator action time to initiate    5 min cooldown Cooldown                            Calculated by LOFTTR2 Operator action time to initiate    4 min depressurization Depressurization                    Calculated by LOFTTR2 Operator action time to initiate SI 3 min termination SI termination and pressure        Calculated by LOFTTR2 equalization Amendment 61                                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 TABLE 15.6.3-4 SEQUENCE OF EVENTS OFFSITE DOSE ANALYSIS Event                                                      Time (sec)
SG Tube Rupture                                            0 Reactor Trip                                                114 Safety Injection                                            178 AFW Flow to Ruptured SG Isolated                            600 Ruptured SG Isolated                                        720 Ruptured SG PORV Fails Open                                722*
Ruptured SG PORV Block Valve Closed                        1922 RCS Cooldown Initiated                                      2224*
RCS Cooldown Terminated                                    2996 RCS Depressurization Initiated                              3236 RCS Depressurization Terminated                            3312 SI Terminated                                              3492 Break Flow Terminated                                      4652
*The actual times listed above and simulated in the analysis were slightly longer than those obtained using the operator action times of Table 15.6.3 3, due to computer program limitations for modelling operator actions.
Amendment 61                                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 TABLE 15.6.3-5 SGTR MASS RELEASE RESULT TOTAL MASS FLOW (POUNDS)
Time Period Time of Reactor                        2 Hours to Time Trip to Time at      Time at Which    at Which RCS Time Zero to      which Break        Break Flow is    Reaches RHR Time of Reactor        Flow is        Terminated to 2      In-Service Trip*          Terminated*            Hours          Conditions*
Ruptured SG
-Condenser              128,300              0                  0                  0
-Atmosphere                0              138,300                0              35,100,0
-Feedwater              123,400            33,000                0 Intact SGs
-Condenser              254,100              0                  0                  0
-Atmosphere                0              176,900              183,300          862,800
-Feedwater              254,100          292,400              201,800          894,900 Break Flow                4900            163,000                0                  0 Flashed Break              830              10,013                0                  0 Flow
* For dose consequence analysis, reactor trip occurs at 114 seconds; break flow is terminated at 4652 seconds; RHR conditions are reached at 8 hours.
Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                          UFSAR Chapter: 15 TABLE 15.6.3-6 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE I. Source Data A. Total Unit Thermal Output, MWt                            2970.4*
B. Total steam generator tube leakage prior to accident, gpm  1.0 C. Reactor coolant iodine activity:
: 1. Accident-Initiated Spike                                The initial RC iodine activities are presented in Table 15.0.9-7. The iodine appearance rates assumed for the accident-initiated spike are presented in Table 15.0.9-6.
: 2. Pre-Accident Spike                                      Primary coolant iodine activities based on 60 Ci/gm of D.E. I-131 are presented on Table 15.0.9-7.
D. Noble Gas Activity                                        Primary coolant noble gas activities based on 1 percent fuel defects are presented in Table 15.0.9-2. No noble gases are contained in the secondary system.
E. Secondary system initial activity                          Dose equivalent of 0.1 Ci/gm of I-131, presented in Table 15.0.9-7.
F. Reactor coolant initial mass, grams                        1.73 x 108 G. Steam generator initial mass (each), grams                3.93 x 107 (faulted), 5.19 x 107 (intact)
H. Offsite power                                              Lost at time of reactor trip I. Primary-to-secondary leakage duration for intact SG, hours 8 J. Species of iodine                                          97 percent elemental, 3 percent organic II. Activity Release Data A. Ruptured steam generator
: 1. Initial Primary-to-secondary leakage, gpm              0.3
: 2. Ruptured flow flashing fraction                        See Figure 15.6.3-19
: 3. Rupture flow                                            See Table 15.6.3-5 & Figure 15.6.3-13
: 4. Flashed rupture flow                                    See Table 15.6.3-5 & Figure 15.6.3-22
: 5. Steam releases                                          See Table 15.6.3-5 & Figure 15.6.3-23 An additional 41,310 lbm/hr to TDAFW pump is modeled until ruptured SG isolation.
: 6. Iodine partition factor for rupture flow Non-flashed                                      100 Flashed                                          1.0 B. Intact steam generators
: 1. Primary-to-secondary leakage, gpm                      0.7
: 2. Steam releases                                          See Table 15.6.3-5 & Figure 15.6.3-23
: 3. Iodine partition factor                                100 C. Condenser
: 1. Iodine partition factor                                100 D. Atmospheric Dispersion Factors                            See Table 15.6.3-10 Amendment 63                                                                                          Page 1 of 2
 
Shearon Harris Nuclear Power Plant                                                              UFSAR Chapter: 15 Table 15.6.3-6 (Continued)
Notes:
* Includes 12.4 MWt of NSSS heat generation.
These analyses were done by Westinghouse for the steam generator replacement and power uprate program.
Details of the inputs used in the calculations are discussed in the NSSS Licensing Report (Reference 15.6.5-52). Both the analyses consider core powers up to 2900 MWt with an additional 2% increase applied. Therefore, the increase in the core power to 2958 MWt due to MUR has adequately been accounted for in these analyses. Note that the margin-to-overfill analysis as defined in the FSAR limits the full power Tavg value to a minimum of 588.8&deg;F. This limitation continues to apply with the MUR power uprating.
Amendment 63                                                                                              Page 2 of 2
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.6.3-10 OFFSITE ATMOSPHERIC DISPERSION FACTORS AND BREATHING RATES Exclusion Area                              Breathing Rate Boundary*            Low Population Zone      (Sec/m3)
Time (hrs.)              x/Q (Sec/m3)              x/Q (Sec/m3)      [Ref. 15.6.3-4]
0-2                  6.17 x 10-4                1.4 x 10-4        3.5 x 10-4 2-8                        -                    1.0 x 10-4        3.5 x 10-4 8-24                      -                    1.0 x 10-4        1.8 x 10-4 24-96                      -                    5.9 x 10-5        2.3 x 10-4
          >96                        -                    2.4 x 10-5        2.3 x 10-4
* The exclusion area boundary atmospheric dispersion factor is conservatively applied during all time intervals in the determination of the limiting two hour period Amendment 61                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 TABLE 15.6.3-13 RADIOLOGICAL CONSEQUENCES OF A SGTR Accident Initiated Iodine Spike TEDE Doses Exclusion Area Boundary*              1.30 Low Population Zone                  0.33 Control Room                          0.63 Pre-Accident Iodine Spike TEDE Doses Exclusion Area Boundary*              2.22 Low Population Zone                  0.53 Control Room                          1.16
* The exclusion area boundary doses reported are for the worst two hour period, determined to be from 0.0 to 2.0 hours for both the pre-accident iodine spike and accident-initiated iodine spike cases.
Amendment 63                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                      UFSAR Chapter: 15 TABLE 15.6.5-1 KEY PARAMETERS FOR LBLOCA Event                                      Operating Range 1.0        Plant Physical Description 1.1 Fuel a) Cladding outside diameter                0.376 in.
b) Cladding inside diameter                0.328 in.
c) Cladding thickness                      0.024 in.
d) Pellet outside diameter                  0.3215 in.
e) Pellet density                          96 percent of theoretical f) Active fuel length                      144 in.
g) Gd2 O3 concentration                    2, 4, 6, 8 w/o 1.2 RCS a) Flow resistance                          Analysis Analysis assumes location giving most limiting PCT b) Pressurizer location (broken loop) c) Hot assembly location                    Anywhere in core d) Hot assembly type                        17 x 17 e) SG tube plugging                          3 percent 2.0        Plant Initial Operating Conditions 2.1 Reactor Power 1
a) Nominal reactor power                    2958 MWt 2
b) FQ                                        2.52 3
c) FH                                      1.73 2.2 Fluid Conditions a) Loop flow                                109.2 Mlbm/hr  M  115.3 Mlbm/hr b) RCS average temperature                  582.0&deg;F  T  594.8&deg;F 4
c) Upper head temperature                  ~ Tcold Temperature d) Pressurizer pressure                    2200 psia  P 2288 psia Event                                      Operating Range e) Pressurizer level                        Gaussian distribution mean of 60% and standard deviation of 7.4%. Distribution on the high side up to a
92% .
f) Accumulator pressure                    599.7 psia  P  679.7 psia 3              3 g) Accumulator liquid volume                994.6 ft  V  1029.4 ft a
The estimated impact to the LBLOCA PCT is 0&deg;F for reducing the sampled pressurizer level high-side limit to 81% to bound the Technical Specitication pressurizer water level LCO of 75% indicated span.
Amendment 61                                                                                      Page 1 of 2
 
Shearon Harris Nuclear Power Plant                                                          UFSAR Chapter: 15 TABLE 15.6.5-1 KEY PARAMETERS FOR LBLOCA Event                                          Operating Range h) Accumulator temperature                      80&deg;F  T  130&deg;F (It's coupled with containment temperature) i) Accumulator resistance fL/D                  As-built piping configuration j) Minimum ECCS boron                            2400 ppm 3.0      Accident Boundary Conditions a) Break location                              Any RCS piping location b) Break type                                  Double-ended guillotine or split c) Break size (each side, relative to cold leg  0.26  A  1.0 full pipe area (split) pipe area)                                      0.26  A  1.0 full pipe area (guillotine) d) Worst single-failure                        Loss of Diesel (one train of ECCS) e) Offsite power                                On or Off f) ECCS pumped injection temperature            125&deg;F g) HHSI pump delay                              29 s (w/ offsite power) 29 s (w/o offsite power) h) LHSI pump delay                              29 s (w/ offsite power) 29 s (w/o offsite power) i) Containment pressure                        14.7 psia, nominal value j) Containment temperature                      80&deg;F  T  130&deg;F k) Containment sprays delay                    0s l) Containment spray water temperature          40&deg;F 1
Consistent with rated core power of 2948 MWt and 0.34% uncertainty.
2 The peaking factor includes measurement and engineering uncertainty.
3 The    value is listed in COLR; while 4% measurement uncertainty is a Technical Specifications limit.
4 Upper head temperature will change based on sampling of RCS temperature.
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Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 TABLE 15.6.5-2 EVENT TIMES FOR LBLOCA Event                                      Time(s)
Break Opened                                0.0 RCP Trip                                    N/A SIAS Issued                                0.4 Start of Broken Loop Accumulator Injection  8.8 Start of Intact Loop Accumulator Injection  12.5 and 12.5 (Loops 2 and 3 respectively)
Beginning of Core Recovery                  25.4 (Beginning of Reflood)
Broken Loop HHSI Delivery Began            29.4 Intact Loop HHSI Delivery Began            29.4 and 29.4 (Loop 2 and 3 respectively)
LHSI Available                              29.4 Broken Loop LHSI Delivery Began            29.4 Intact Loop LHSI Delivery Began            29.4 and 29.4 (Loop 2 and 3 respectively)
Broken Loop Accumulator Emptied            35.0 Intact Loop Accumulators Emptied            37.0 and 35.9 (Loops 2, and 3 respectively)
PCT Occurred (1935&deg;F)                      131.3 Transient Calculation Terminated            829.7 Amendment 61                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 TABLE 15.6.5-3 SBLOCA SYSTEM ANALYSIS PARAMETERS Parameter                                      Value Reactor Power, MWt                                                        2958a Radial Peaking Factor (  ) (includes uncertainty)                        1.73 Total Power Peaking Factor (    ) (includes uncertainty)                  2.52 RCS Flow Rate (minimum) (gpm)                                            293540 Pressurizer Pressure (nominal), psia                                        2250 RCS Operating Temperature (nominal), &deg;F                                    588.8 Accumulator Pressure (minimum), psia                                      599.7 Accumulator Fluid Temperature (maximum), &deg;F                                130.0 3
Accumulator Water Volume (nominal), ft.                                    1012 SG Tube Plugging, %                                                          3 SG Secondary Pressure (nominal), psia                                        985 MFW Temperature at 100% RTP (nominal), &deg;F                                  440.0 AFW Temperature (maximum), &deg;F                                                120 AFW Pump Delay Time on SIAS (LOOP), sec                                    61.5 HHSI, and LHSI/RHR Fluid Temperature, &deg;F                                    125 Pressurizer Pressure - Low Reactor Trip (minimum), psia                  1934.7 (includes uncertainty)
Reactor Scram Delay on Low Pressurizer Pressure, (maximum)                  2.0 sec SIAS Activation Setpoint Pressure (minimum), psia                        1714.7 HHSI Pump Delay Time on SIAS (LOOP), sec                                    29 LHSI Pump Delay Time on SIAS (LOOP), sec                                    29 MSSV lift pressures (nominal; does not inclde 2%b tolerance), psia        1184.7 1199.7 1214.7 1229.7 1244.7 a
Includes 0.34% (10 MWt) uncertainty.
b Safety analyses support operating with an MSSV setpoint tolerance of +/-3%. See Section 15.0 for more details.
Amendment 61                                                                              Page 1 of 2
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 Table 15.6.5-3 (Continued)
HHSI Flow                                LHSI Flow RCS Cold Leg      Flow per Loop  RCS Cold Leg    Intact Loop Flow  Broken Loop Pressure (psia)        (gpm)      Pressure (psia)  per Loop (gpm)    Flow (gpm) 0.00            165.82            0.00            924.49          1813.98 15.00            165.82          25.37            924.49          1813.98 398.83            151.94          35.43            888.15          1741.90 646.81            141.37          50.07            833.24          1632.95 829.46            134.06          70.06            753.65          1475.03 1012.11            126.52          90.03            655.05          1278.47 1142.57            120.60          100.02            600.22          1169.09 1390.45            108.60          105.21            569.63          1107.78 1638.33              95.47        110.07            539.84          1048.26 1755.74              88.49        115.72            503.46          973.90 1886.20              80.36        131.51            356.47          657.00 2003.62              70.44        142.79            108.86          137.47 2134.08              57.37        144.79            36.94            44.67 2251.50              42.19        144.89              0.00            0.00 2378.23              20.30 2378.33              0.00 Amendment 61                                                              Page 2 of 2
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 TABLE 15.6.5-11a SEQUENCE OF EVENTS DURING SBLOCA (LIMITING CASE)
Event Description for:                                  Time (sec)
Limiting Break Diameter = 2.60 inches (Limiting Break area = 0.03687 ft2)
Break open                                                    0.0 Low PZR Pressure Trip                                        19.6 RX, LOOP, RCPs, MFWP, and Turbine Trip                      21.6 Low PZR Pressure SIAS Setpt                                  30.6 HHSI Flow Begins                                            59.6 Setpt to start Aux. FW Pmp                                  107.5 Loop seal 1 clears                                          878 Loop seal 2 clears                                          N/A Loop seal 3 clears                                          N/A Break uncovers                                              -900 Core uncovery begins                                        -1300 Hot Rod rupture occurs                                      2019 Accumulator Injection begins                                2020 Minimum RV mass occurs                                      2038 PCT occurs                                                  2060 Amendment 61                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 TABLE 15.6.5-11b SBLOCA ANALYSIS RESULTS (LIMITING CASE)
Parameter                              Value Break diameter (in)                    2.60 Peak Clag Temperature (&deg;F)            1664 Time of PCT (sec)                      2060 PCT Elevation (ft)                    11.13 Time of Rupture (sec)                  2019 Core Wide Oxidation (%)                0.0556 Local Maximum Oxidation (%)            2.2196 Amendment 61                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                        UFSAR Chapter: 15 TABLE 15.6.5-12 PARAMETERS USED FOR LARGE BREAK LOCA RADIOLOGICAL ANALYSIS Source Term Core Activity                                                        See Table 15.0.9-1 Activity release fractions and timing                                See Tables 15.6.5-13 & Table 15.6.5-14 Iodine chemical form in containment (%)
Elemental                                                  4.85 Organic                                                    0.15 Particulate (cesium iodide)                                95 Containment 3
Containment net free volume (ft )                                    2.344E6 3
Containment sprayed volume (ft )                                    2.014E6 Fan cooler units Number in operation                                                  2 Flow rate (per unit)                                                31,250 Containment leak rates (weight %/day) 0 - 24 hours                                              0.10
          > 24 hours                                                0.05 Spray Operation Time to initiate sprays                                    120.0 seconds Time to terminate spray operation                          4.0 hours Spray flow rates (gpm)                                              1730 Spray fall height (ft)                                              125
                          -1 Removal Coefficients (hr )
Spray elemental iodine removal                            20.0 Spray particulate removal                                  3.94 Sedimentation particulate removal                          0.1 (after spray termination in sprayed region and from start of event in unsprayed region)
Containment sump volume (gal)                                        3.595E5 Time to initiate ECCS recirculation (min)                            20 ECCS leak rate to Auxiliary Building (total, gpm)                    2 Inside RABEES (gpm)                                        1.934 Outside RABEES (gpm)                                      0.066 ECCS leak rate to RWST analysis values (gpm)                        1.5 Directly to water inventory                                          25.0 Directly to air space above water                                    2.2 - 9.7 (pathway dependent)
Airborne fraction for ECCS leakage to RWST (%)
Variable Time < 24 hours                                  <0.1%
Variable After 24 hours                                    <0.001%
Partition Coefficient for Elemental Iodine for ECCS leakage to RWST Variable - Release from mixed RWST liquid inventory to air 42.1 - 40.9 Variable - Release from sump liquid entering air          3 - 10 space direclty RABEES filter efficiencies (%)
Elemental                                                            95 Organic                                                              95 Particulate                                                          95 Amendment 62                                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 TABLE 15.6.5-13 LBLOCA CORE FISSION PRODUCT RELEASE TIMING Release Phase                      Duration Coolant Activity                  10 to 30 seconds Gap Activity                      0.5 hour Early In-vessel                    1.3 hour Amendment 61                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.6.5-14 LBLOCA CORE FISSION PRODUCT RELEASE FRACTIONS Gap Release              Early In-Vessel Noble gases                            0.05                        0.95 Halogens                                0.05                        0.35 Alkali Metals                          0.05                        0.25 Tellurium group                        0                            0.05 Barium, Strontium                      0                            0.02 Noble Materials (Ruthenium group)      0                            0.0025 Cerium group                            0                            0.0005 Lanthanides                            0                            0.0002 The RG 1.183 source term assumes a release of gap activity (5% of core) followed by the in-vessel release as defined.
Amendment 61                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                    UFSAR Chapter: 15 TABLE 15.6.5-15 CONTROL ROOM PARAMETERS USED FOR RADIOLOGICAL ANALYSIS Volume (ft3)                                                        71,000 Normal Ventilation Flow Rates (cfm)
Filtered Makeup Flow Rate                                          0.0 Filtered Recirculation Flow Rate                                    0.0 Unfiltered Makeup Flow Rate                                        1050.0 Unfiltered Recirculation Flow Rate (Not modeled-no impact on analyses)
Post Accident Recirculation Flow Rates (cfm)
Filtered Makeup Flow Rate                                          0.0 Filtered Recirculation Flow Rate                                    3600.0 Unfiltered Inleakage                                                300 Unfiltered Recirculation Flow Rate (Not modeled-no impact on analyses)
Pressurization Mode Flow Rates (cfm)
Filtered Makeup Air Flow Rate                                      400.0 Filtered Recirculation Flow Rate                                    3200.0 Unfiltered Inleakage                                                300 Unfiltered Recirculation Flow Rate (Not modeled-no impact on analyses)
Filter Efficiencies (%)
Elemental                                                          99 Organic                                                            99 Particulate                                                        99 CR Radiation Monitor Sensitivity (Ci/ml for Xe-133)                3.0E-6 CR Radiation Monitor Location                                        Emergency & normal air intakes Delay to Initiate Switchover of Post-Accident signal                15 seconds Recirculation HVAC mode after radiation Operator Action Time to Switch to Pressurization Mode                2 hours Breathing Rate - Duration of the Event (m3/sec)                      3.5E-4 Amendment 63                                                                                      Page 1 of 2
 
Shearon Harris Nuclear Power Plant                                                            UFSAR Chapter: 15 Table 15.6.5-15 (Continued)
Control Room Atmosphere Dispersion Factors for Large Break LOCA and RCCA Ejection Accident (Containment Pathway) (sec/m3)*
0 - 8 hours                2.04E-3 8 - 24 hours                5.80E-4 1 - 4 days                  1.63E-4 4 - 30 days                6.16E-6 Control Room Atmospheric Dispersion Factors for RWST vent release following a Large Break LOCA (sec/m3)**
0 - 8 hours                9.18E-3 8 - 24 hours                2.61E-3 1 - 4 days                  7.31E-4 4 - 30 days                2.77E-5 Control Room Atmospheric Dispersion Factors for all accidents except Large Break LOCA and RCCA Ejection Accident (Containment Pathway) (sec/m3)***
0 - 8 hours                4.08E-3 8 - 24 hours                1.16E-3 1 - 4 days                  3.25E-4 4 - 30 days                1.23E-5 Occupancy Factors****
0 - 24 hours                1.0 1 - 4 days                  0.6 4 - 30 days                0.4
*These atmospheric dispersion factors incorporated a reduction factor of 4 based on credit for dual manual selection emergency air intakes being applied in accordance with SRP 6.4 Section III.3.d.(4).(ii).
**These atmospheric dispersion factors do not incorporate a reduction factor to add conservatism for this particular release point/receptor pair.
***These atmospheric dispersion factors incorporate a reduction factor of 2 based on least credit for dual emergency air intakes being applied in accordance with SRP 6.4 Section III.3.d.(4).(ii). These could be further reduced by a factor of 2, similar to those used for the large break LOCA doses. The calculated analysis results using these factors therefore are conservative by a factor of 2.
****These occupancy factors have been conservatively incorporated in the atmospheric dispersion factors. This is conservative since it does not allow the benefit of reduced occupancy for activity already present in the control room from earlier periods.
Amendment 63                                                                                              Page 2 of 2
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.6.5-16 RADIOLOGICAL CONSEQUENCES OF A POSTULATED LARGE BREAK LOCA Exclusion Area Boundary*                  7.69 rem TEDE Low Population Zone                      4.92 rem TEDE Control Room**                            3.33 rem TEDE
*The exclusion area boundary dose reported is for the worst two hour period, determined to be from 0.4 hours to 2.4 hours.
**The control room dose includes the immersion dose, the ECCS leakage doses, and the indirect shine doses to operators.
Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.7.1-1 ASSUMPTIONS FOR WASTE GAS DECAY TANK RELEASE ACCIDENT ANALYSIS Design Basis Assumption Source Data:
: 1. Power level prior to accident is 2958 MWt (rated power of 2948 MWt with 0.34% uncertainty)
: 2. RCS radioactive concentrations are maximum values based on 1 percent failed fuel
: 3. All gases stripped from processing one entire RCS volume are passed to the gas decay tank which fails. The flash tank is assumed to remove 100 percent of the noble gases and 0.1 percent of the iodines that enter it.
: 4. A decontamination factor of 10 is assumed for the CVCS purification ion exchanger for iodine
: 5. Accident occurs immediately following a cold shutdown Activity Release:
: 1. All gases released from tank leak from the RAB at ground level within 2 hour period Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.7.1-2 RADIOLOGICAL CONSEQUENCES OF A WASTE GAS DECAY TANK RELEASE Gas Decay Tank Isotopes                            Tank Inventory (Ci)
Kr-83m                                              19.1 Kr-85m                                              138.0 Kr-85                                              4100.0 Kr-87                                              46.0 Kr-88                                              172.0 Xe-131m                                            775.0 Xe-133m                                            903.0 Xe-133                                              58500.0 Xe-135m                                            56.6 Xe-135                                              900.0 Xe-138                                              5.16 Exclusion Area Boundary*                            0.30 rem TEDE Low Population Zone                                  0.07 rem TEDE Control Room                                        0.04 rem TEDE
* The exclusion area boundary dose reported is for the worst two hour period, determined to be from 0.0 to 2.0 hours.
Amendment 63                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 TABLE 15.7.4-1 PARAMETERS USED IN FUEL HANDLING ACCIDENT INSIDE THE FUEL HANDLING BUILDING RADIOLOGICAL ANALYSIS Radial peaking factor (PWR fuel)                      1.73 (BWR fuel)                                            1.5 Fuel damaged (number of assemblies)                    1.2 PWR (314 rods) + 52 BWR
[NOTE: All damaged PWR rods assumed to exceed 6.3 kW/ft above 54 GWD/MTU burnup]
Time from shutdown before fuel movement (PWR)(hr)      100 (BWR fuel) (yr)                                        4 Activity in the damaged fuel assemblies (Ci)
I-131                                          7.21E5 I-133                                          7.59E4 I-135                                          5.57E1 Kr-85                                          1.41E5 Xe-131m                                        9.06E3 Xe-133m                                        1.77E4 Xe-133                                        1.19E6 Xe-135                                        2.41E2 Gap Fractions (% of core activity)
I-131                                          8 Kr-85                                          30 Other Iodine and Noble Gas nuclides            5 Water depth                                            21 feet Overall pool iodine scrubbing factor                  200 Iodine chemical form in release to atmosphere (%)
Elemental                                      70 Organic                                        30 Particulate                                    0 Spent Fuel Pool Ventilation System                    No filtration assumed Isolation of release No isolation assumed Time to release all activity (hours) 2 Activity Released, (Ci)
I-131                                          2.884E2 I-133                                          1.898E1 I-135                                          1.393E-2 Kr-85                                          4.230E4 Xe-131m                                        4.530E2 Xe-133m                                        8.850E2 Xe-133                                        5.950E4 Xe-135                                        1.205E1 Amendment 62                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.7.4-2 RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL HANDLING ACCIDENT IN THE FUEL HANDLING BUILDING Exclusion Area Boundary*                          2.42 rem TEDE Low Population Zone                                0.55 rem TEDE Control Room                                      1.15 rem TEDE
*The exclusion area boundary dose reported is for the worst two hour period, determined to be from 0 to 2 hours.
Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 TABLE 15.7.4-3 Parameters Used in a Fuel Handling Accident Inside Containment Radiological Analysis Radial peaking factor                                              1.73 Fuel damaged (number of assemblies)                                1
[NOTE: All damaged fuel rods assumed to exceed 6.3 kW/ft above 54 GWD/MTU burnup]
Time from shutdown before fuel movement (hr)                      100 Activity in the damaged fuel assembly (Ci)
I-131                                                    6.06E5 I-133                                                    6.38E4 I-135                                                    4.68E1 Kr-85                                                    8.82E3 Xe-131m                                                  7.61E3 Xe-133m                                                  1.49E4 Xe-133                                                    9.97E5 Xe-135                                                    2.03E2 Gap Fractions (% of core activity)
I-131                                                    8 Kr-85                                                    30 Other Iodine and Noble Gas nuclides                      5 Water depth                                                        22 feet Overall pool iodine scrubbing factor                              200 Iodine chemical form in release to atmosphere (%)
Elemental                                                70 Organic                                                  30 Particulate                                              0 Filter efficiency                                                  No filtration assumed Isolation of release                                              No isolation assumed Time to release all activity (hours)                              2 Activity Released (Ci)
I-131                                                    2.424E2 I-133                                                    1.595E1 I-135                                                    1.170E-2 Kr-85                                                    2.646E3 Xe-131m                                                  3.805E2 Xe-133m                                                  7.450E2 Xe-133                                                    4.985E4 Xe-135                                                    1.015E1 Amendment 62                                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 TABLE 15.7.4-4 RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL HANDLING ACCIDENT INSIDE CONTAINMENT Exclusion Area Boundary*                  2.02 rem TEDE Low Population Zone                      0.46 rem TEDE Control Room                              0.97 rem TEDE
*The exclusion area boundary dose reported is for the worst two hour period, determined to be from 0 to 2 hours.
Amendment 63                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                              UFSAR Chapter: 15 TABLE 15.7.5-1 FUEL CONDITIONS ANALYZED FOR IF-300 SERIES CASK DROP Assembly Average Fuel Type          Burn-up (GWD/MTU)      Enrichment (w/o) Cooling (Decay) (yrs)
PWR Fuel Types:
PWR 15X15                        35                  2.33                2.5 PWR 15X15                        35                  5.00                2.5 PWR 15X15                        45                  2.33                5 PWR 15X15                        45                  5.00                5 BWR Fuel Types:
BWR 7x7                          35                  2.33                3 BWR 7x7                          35                  5.00                3 BWR 8x8                          35                  2.33                3 BWR 8x8                          35                  5.00                3 BWR 8x8R                        35                  2.33                3 BWR 8x8R                        35                  5.00                3 GE- 7,8,9,10 & 13                45                  3.19                4 GE- 7,8,9,10 & 13                45                  4.25                4 Amendment 61                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 FIGURE                                          TITLE 15.0.3-1  DELETED BY AMENDMENT NO. 48 15.0.3-2  DELETED BY AMENDMENT NO. 48 15.0.4-1  DELETED BY AMENDMENT NO. 48 15.0.5-1  DELETED BY AMENDMENT NO. 48 15.0.5-2  DELETED BY AMENDMENT NO. 48 15.0.5-3  DELETED BY AMENDMENT NO. 48 15.0.5-4  ROD POSITION VERSUS TIME AFTER ROD DROP BEGINS UTILIZED IN AREVA ANALYSES 15.0.5-5  NORMALIZED RCCA REACTIVITY WORTH VERSUS ROD INSERTION UTILIZED IN AREVA ANALYSES 15.0.5-6  NORMALIZED RCCA REACTIVITY WORTH VERSUS TIME AFTER ROD DROP BEGINS UTILIZED IN AREVA ANALYSES 15.0.10-1 DELETED BY AMENDMENT NO. 48 15.0A.1-1 CONTAINMENT LEAKAGE DOSE MODEL 15.1.2-1  REACTOR POWER FOR INCREASE IN FEEDWATER FLOW - LIMITING CASE 15.1.2-2  PRIMARY SYSTEM TEMPERATURES FOR INCREASE IN FEEDWATER FLOW - LIMITING CASE 15.1.2-3  REACTIVITIES FOR INCREASE IN FEEDWATER FLOW - LIMITING CASE 15.1.2-4  PRESSURIZER PRESSURE FOR INCREASE IN FEEDWATER FLOW - LIMITING CASE 15.1.2-5  STEAM GENERATOR COLLAPSED LEVEL FOR INCREASE IN FEEDWATER FLOW - LIMITING CASE 15.1.2-6  DELETED BY AMENDMENT NO. 51 15.1.2-7  DELETED BY AMENDMENT NO. 51 15.1.2-8  DELETED BY AMENDMENT NO. 51 15.1.2-9  DELETED BY AMENDMENT NO. 51 15.1.2-10 DELETED BY AMENDMENT NO. 51 15.1.3-1  REACTOR POWER FOR INCREASE IN STEAM FLOW - BOC CASE 15.1.3-2  PRESSURIZER PRESSURE FOR INCREASE IN STEAM FLOW - BOC CASE 15.1.3-3  REACTIVITIES FOR INCREASE IN STEAM FLOW - BOC CASE 15.1.3-4  VESSEL AVERAGE TEMPERATURE FOR INCREASE IN STEAM FLOW - BOC CASE 15.1.4-1  DELETED BY AMENDMENT NO. 48 Amendment 63                                                                Page 1 of 12
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 FIGURE                                          TITLE 15.1.4-2  DELETED BY AMENDMENT NO. 48 15.1.4-3  DELETED BY AMENDMENT NO. 48 15.1.4-4  DELETED BY AMENDMENT NO. 48 15.1.4-5  DELETED BY AMENDMENT NO. 48 15.1.5-1  STEAM GENERATOR PRESSURES FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD 15.1.5-2  BREAK FLOW RATES FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD 15.1.5-3  STEAM GENERATOR INVENTORIES FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD 15.1.5-4  PRESSURIZER PRESSURE FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD 15.1.5-5  REACTOR POWER FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD 15.1.5-6  REACTIVITY COMPONENTS FOR THE HZP MSLB WITH OFFSITE POWER AND WITH A STUCK ROD 15.2.3-1  REACTOR POWER FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE 15.2.3-2  AVERAGE TEMPERATURES FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE 15.2.3-3  PRESSURIZER PRESSURE AND LOWER HEAD PRESSURE FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE 15.2.3-4  PRESSURIZER LEVEL FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE 15.2.3-5  REACTOR POWER FOR TURBINE TRIP MDNBR CASE 15.2.3-6  AVERAGE TEMPERATURES FOR TURBINE TRIP MDNBR CASE 15.2.3-7  PRESSURIZATION PRESSURE AND LOWER HEAD PRESSURE FOR TURBINE TRIP MDNBR CASE 15.2.3-8  DELETED BY AMENDMENT NO. 51 15.2.3-9  REACTOR POWER FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE 15.2.3-10 AVERAGE TEMPERATURES FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE 15.2.3-11 PRESSURIZER LEVEL FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE 15.2.3-12 STEAM GENERATOR LOWER SHELL PRESSURE FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE 15.2.6-1  DELETED BY AMENDMENT NO. 63 Amendment 63                                                              Page 2 of 12
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 FIGURE                                        TITLE 15.2.6-2  DELETED BY AMENDMENT NO. 63 15.2.6-3  DELETED BY AMENDMENT NO. 63 15.2.6-4  DELETED BY AMENDMENT NO. 63 15.2.6-5  DELETED BY AMENDMENT NO. 63 15.2.6-6  DELETED BY AMENDMENT NO. 51 15.2.7-1  REACTOR POWER FOR LOSS OF NORMAL FEEDWATER FLOW 15.2.7-2  PRESSURIZER LEVEL FOR LOSS OF NORMAL FEEDWATER FLOW 15.2.7-3  PRESSURIZER PRESSURE FOR LOSS OF NORMAL FEEDWATER FLOW 15.2.7-4  RCS TEMPERATURES FOR LOSS OF NORMAL FEEDWATER FLOW 15.2.7-5  STEAM GENERATOR MASS INVENTORY FOR LOSS OF NORMAL FEEDWATER FLOW 15.2.7-6  DELETED BY AMENDMENT NO. 51 15.2.8-1  REACTOR POWER FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-2  PRESSURIZER LEVEL FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-3  PRESSURIZER PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-4  LOOP 1 PRIMARY SYSTEM TEMPERATURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-5  LOOP 2 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-6  LOOP 3 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-7  STEAM GENERATOR PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-8  STEAM GENERATOR NARROW RANGE LEVEL FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-9  REACTOR COOLANT SYSTEM FLOW FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER 15.2.8-10 DELETED BY AMENDMENT NO. 63 15.2.8-11 REACTOR POWER FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                            Page 3 of 12
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 FIGURE                                        TITLE 15.2.8-12 PRESSURIZER LEVEL FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-13 PRESSURIZER PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-14 LOOP 1 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-15 LOOP 2 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-16 LOOP 3 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-17 STEAM GENERATOR PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-18 STEAM GENERATOR NARROW RANGE LEVEL FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-19 REACTOR COOLANT SYSTEM FLOW FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-20 TOTAL PRESSURIZER SAFETY VALVE FLOW FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER 15.2.8-21 REACTOR POWER FOR FEEDWATER SYSTEM PIPE BREAK SHORT-TERM CORE COOLING CASE 15.2.8-22 PRESSURIZER PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK SHORT-TERM CORE COOLING CASE 15.2.8-23 REACTOR COOLANT SYSTEM FLOW FOR FEEDWATER SYSTEM PIPE BREAK SHORT-TERM CORE COOLING CASE 15.2.8-24 PRIMARY SYSTEM AVERAGE TEMPERATURE FOR FEEDWATER SYSTEM PIPE BREAK SHORT-TERM CORE COOLING CASE 15.2.8-25 REACTOR POWER FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE 15.2.8-26 PRESSURIZER LEVEL FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE 15.2.8-27 PRESSURIZER PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE 15.2.8-28 PRIMARY SYSTEM AVERAGE TEMPERATURE FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE 15.2.8-29 STEAM GENERATOR PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE 15.2.8-30 STEAM GENERATOR NARROW RANGE LEVEL FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                            Page 4 of 12
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 FIGURE                                        TITLE 15.2.8-31 REACTOR COOLANT SYSTEM FLOW FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE 15.2.8-32 TOTAL PRESSURIZER RELIEF FLOW FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE 15.3.1-1  DELETED BY AMENDMENT NO. 51 15.3.1-2  DELETED BY AMENDMENT NO. 51 15.3.1-3  DELETED BY AMENDMENT NO. 51 15.3.1-4  DELETED BY AMENDMENT NO. 42 15.3.1-5  DELETED BY AMENDMENT NO. 27 15.3.1-6  DELETED BY AMENDMENT NO. 27 15.3.1-7  DELETED BY AMENDMENT NO. 27 15.3.1-8  DELETED BY AMENDMENT NO. 27 15.3.2-1  REACTOR POWER FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-2  CORE AVERAGE HEAT FLUX FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-3  PRESSURIZER PRESSURE FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-4  PRESSURIZER LEVEL FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-5  REACTOR COOLANT SYSTEM MASS FLOW RATE FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-6  CORE INLET AND OUTLET TEMPERATURES FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-7  TOTAL CORE REACTIVITY FOR LOSS OF FORCED REACTOR COOLANT FLOW 15.3.2-8  DELETED BY AMENDMENT NO. 27 15.3.3-1  REACTOR POWER FOR LOCKED ROTOR OVERPRESSURIZATION CASE 15.3.3-2  PRESSURIZER LEVEL FOR LOCKED ROTOR OVERPRESSURIZATION CASE 15.3.3-3  REACTOR COOLANT SYSTEM MASS FLOW RATE FOR LOCKED ROTOR OVERPRESSURIZATION CASE 15.3.3-4  AVERAGE FLUID TEMPERATURES FOR LOCKED ROTOR OVERPRESSURIZATION CASE 15.3.3-5  TOTAL CORE REACTIVITY FOR LOCKED ROTOR OVERPRESSURIZATION CASE 15.3.3-6  MAXIMUM PRIMARY SYSTEM PRESSURE FOR LOCKED ROTOR OVERPRESSURIZATION CASE 15.3.3-7  REACTOR POWER LEVEL FOR LOCKED ROTOR MDNBR CASE Amendment 63                                                              Page 5 of 12
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 FIGURE                                          TITLE 15.3.3-8  CORE AVERAGE HEAT FLUX FOR LOCKED ROTOR MDNBR CASE 15.3.3-9  PRESSURIZER PRESSURE FOR LOCKED ROTOR MDNBR CASE 15.3.3-10 PRESSURIZER LEVEL FOR LOCKED ROTOR MDNBR CASE 15.3.3-11 REACTOR COOLANT SYSTEM MASS FLOW RATE FOR LOCKED ROTOR MDNBR CASE 15.3.3-12 AVERAGE FLUID TEMPERATURES FOR LOCKED ROTOR MDNBR CASE 15.3.3-13 TOTAL CORE REACTIVITY FOR LOCKED ROTOR MDNBR CASE 15.3.3-14 DELETED BY AMENDMENT NO. 42 15.3.3-15 DELETED BY AMENDMENT NO. 42 15.4.1-1  REACTOR POWER FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT HZP 15.4.1-2  REACTIVITY COMPONENTS FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT HZP 15.4.1-3  CORE AVERAGE HEAT FLUX FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT HZP 15.4.1-4  LOOP AVERAGE FLUID AND FUEL TEMPERATURES FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT HZP 15.4.1-5  DELETED BY AMENDMENT NO. 51 15.4.1-6  DELETED BY AMENDMENT NO. 51 15.4.2-1  MINIMUM DNBR VS REACTIVITY INSERTION RATE FOR RCCA BANK WITHDRAWAL AT BOC 10%
POWER LEVEL CASE 15.4.2-2  REACTOR POWER FOR LIMITING RCCA BANK WITHDRAWAL CASE 15.4.2-3  REACTIVITIES FOR LIMITING RCCA BANK WITHDRAWAL CASE 15.4.2-4  DELETED BY AMENDMENT NO. 58 15.4.2-5  CORE POWER BASED ON ROD SURFACE HEAT FLUX FOR LIMITING RCCA BANK WITHDRAWAL CASE 15.4.2-6  PRIMARY TEMPERATURES FOR LIMITING RCCA BANK WITHDRAWAL CASE 15.4.2-7  PRESSURIZER LEVEL FOR LIMITING RCCA BANK WITHDRAWAL CASE 15.4.2-8  RCS MASS FLOW RATE FOR LIMITING RCCA BANK WITHDRAWAL CASE 15.4.2-9  PRESSURIZER PRESSURE FOR LIMITING RCCA BANK WITHDRAWAL CASE 15.4.2-10 PRESSURE AT BOTTOM OF REACTOR VESSEL FOR LIMITING RCCA BANK WITHDRAWAL PRIMARY SIDE OVERPRESSURIZATION CASE 15.4.3-1  REACTOR POWER FOR LIMITING DROPPED ROD CASE Amendment 63                                                              Page 6 of 12
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 FIGURE                                        TITLE 15.4.3-2  AVERAGE CORE HEAT FLUX FOR LIMITING DROPPED ROD CASE 15.4.3-3  PRESSURIZER PRESSURE FOR LIMITING DROPPED ROD CASE 15.4.3-4  PRIMARY TEMPERATURES FOR LIMITING DROPPED ROD CASE 15.4.3-5  DELETED BY AMENDMENT NO. 50 15.4.3-6  CORE INLET FLOW FOR LIMITING DROPPED ROD CASE 15.4.3-7  PRESSURIZER LEVEL FOR LIMITING DROPPED ROD CASE 15.4.3-8  REACTIVITY FOR LIMITING DROPPED ROD CASE 15.4.3-9  DELETED BY AMENDMENT NO. 50 15.4.3-10 ROD POSITION FOR LIMITING DROPPED ROD CASE 15.4.3-11 ROD SPEED FOR LIMITING DROPPED ROD CASE 15.4.3-12 REACTOR POWER FOR SINGLE UNCONTROLLED RCCA WITHDRAWAL AT HFP 15.4.3-13 CORE AVERAGE HEAT FLUX FOR SINGLE UNCONTROLLED RCCA WITHDRAWAL AT HFP 15.4.3-14 AVERAGE FLUID TEMPERATURES FOR SINGLE UNCONTROLLED RCCA WITHDRAWAL AT HFP 15.4.3-15 PRESSURIZER PRESSURE FOR SINGLE UNCONTROLLED RCCA WITHDRAWAL AT HFP 15.4.4-1  DELETED BY AMENDMENT NO. 51 15.4.4-2  DELETED BY AMENDMENT NO. 51 15.4.4-3  DELETED BY AMENDMENT NO. 51 15.4.4-4  DELETED BY AMENDMENT NO. 51 15.4.4-5  DELETED BY AMENDMENT NO. 42 15.4.8-1  DELETED BY AMENDMENT NO. 56 15.4.8-2  DELETED BY AMENDMENT NO. 56 15.4.8-3  DELETED BY AMENDMENT NO. 56 15.4.8-4  DELETED BY AMENDMENT NO. 56 15.4.8-5  DELETED BY AMENDMENT NO. 56 15.4.8-6  DELETED BY AMENDMENT NO. 56 15.4.8-7  DELETED BY AMENDMENT NO. 56 15.4.8-8  DELETED BY AMENDMENT NO. 56 Amendment 63                                                            Page 7 of 12
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 FIGURE                                          TITLE 15.4.8-9  REACTOR POWER FOR RCCA EJECTION - EOC 50% POWER CASE 15.4.8-10 DELETED BY AMENDMENT NO. 63 15.4.8-11 TOTAL CORE REACTIVITY FOR RCCA EJECTION - EOC 50% POWER CASE 15.4.8-12 DELETED BY AMENDMENT NO. 63 15.4.8-13 DELETED BY AMENDMENT NO. 63 15.4.8-14 DELETED BY AMENDMENT NO. 63 15.4.8-15 DELETED BY AMENDMENT NO. 63 15.4.8-16 DELETED BY AMENDMENT NO. 63 15.5.1-1  DELETED BY AMENDMENT NO. 63 15.5.1-2  DELETED BY AMENDMENT NO. 63 15.5.1-3  DELETED BY AMENDMENT NO. 63 15.5.1-4  DELETED BY AMENDMENT NO. 63 15.5.1-5  DELETED BY AMENDMENT NO. 63 15.5.1-6  DELETED BY AMENDMENT NO. 63 15.5.1-7  DELETED BY AMENDMENT NO. 63 15.5.1-8  DELETED BY AMENDMENT NO. 63 15.5.1-9  DELETED BY AMENDMENT NO. 63 15.5.1-10 DELETED BY AMENDMENT NO. 63 15.5.1-11 DELETED BY AMENDMENT NO. 63 15.5.1-12 DELETED BY AMENDMENT NO. 63 15.5.1-13 DELETED BY AMENDMENT NO. 63 15.5.1-14 DELETED BY AMENDMENT NO. 63 15.5.1-15 PRESSURIZER LIQUID LEVEL (PRESSURIZER OVERFILL CASE) 15.5.1-16 REACTOR COOLANT SYSTEM TEMPERATURES (PRESSURIZER OVERFILL CASE) 15.5.1-17 PRESSURIZER PRESSURE (PRESSURIZER OVERFILL CASE) 15.5.1-18 SAFETY INJECTION FLOW RATE (PRESSURIZER OVERFILL CASE) 15.5.1-19 REACTOR POWER (PRESSURIZER OVERFILL CASE)
Amendment 63                                                              Page 8 of 12
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 FIGURE                                          TITLE 15.5.1-20 PRESSURIZER SAFETY RELIEF VALVE INLET TEMPERATURE (PRESSURIZER OVERFILL CASE) 15.5.1-21 PRESSURIZER SAFETY RELIEF VALVE INLET PRESSURE (PRESSURIZER OVERFILL CASE) 15.6.1-1  REACTOR POWER FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE 15.6.1-2  CORE AVERAGE HEAT FLUX FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE 15.6.1-3  PRESSURIZER PRESSURE FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE 15.6.1-4  PRESSURIZER LEVEL FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE 15.6.1-5  REACTOR COOLANT SYSTEM MASS FLOW RATE FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE 15.6.1-6  REACTOR COOLANT SYSTEM TEMPERATURE FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE 15.6.1-7  TOTAL CORE REACTIVITY FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE 15.6.3-1  PRESSURIZER LEVEL - MARGIN TO OVERFILL ANALYSIS 15.6.3-2  RCS PRESSURE - MARGIN TO OVERFILL ANALYSIS 15.6.3-3  SECONDARY PRESSURE - MARGIN TO OVERFILL ANALYSIS 15.6.3-4  INTACT LOOP HOT AND COLD LEG TEMPERATURES - MARGIN TO OVERFILL ANALYSIS 15.6.3-5  PRIMARY TO SECONDARY BREAK FLOW MARGIN TO OVERFILL ANALYSIS 15.6.3-6  RUPTURED SG WATER VOLUME MARGIN TO OVERFILL ANALYSIS 15.6.3-7  PRESSURIZER LEVEL-OFFSITE RADIATION DOSE ANALYSIS 15.6.3-8  PRESSURIZER PRESSURE - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-9  SECONDARY PRESSURE - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-10 INTACT LOOP HOT & COLD LEG TEMPERATURES - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-11 RUPTURED LOOP HOT AND COLD LEG TEMPERATURES - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-12 DIFFERENTIAL PRESSURE BETWEEN RCS AND RUPTURED SG - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-13 PRIMARY TO SECONDARY BREAK FLOW - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-14 RUPTURED SG WATER VOLUME - OFFSITE RADIATION DOSE ANALYSIS Amendment 63                                                                Page 9 of 12
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 FIGURE                                          TITLE 15.6.3-15 RUPTURED SG WATER MASS - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-16 RUPTURED SG MASS RELEASE RATE TO THE ATMOSPHERE - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-17 INTACT SGS MASS RELEASE RATE TO THE ATMOSPHERE - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-18 IODINE TRANSPORT MODEL - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-19 BREAK FLOW FLASHING FRACTION - OFFSITE RADIATION DOSE ANALYSIS 15.6.3-20 DELETED BY AMENDMENT NO. 51 15.6.3-21 DELETED BY AMENDMENT NO. 51 15.6.3-22 TOTAL FLASHED BREAK FLOW - SGTR OFFSITE RADIATION DOSE ANALYSIS 15.6.3-23 RUPTURED SG MASS RELEASE RATE TO THE ATMOSPHERE - SGTR OFFSITE RADIATION DOSE ANALYSIS 15.6.3-24 INTACT SG MASS RELEASE RATE TO THE ATMOSPHERE - SGTR OFFSITE RADIATION DOSE ANALYSIS 15.6.5-1  NORMALIZED AXIAL DEPENDENCE OF F VERSUS ELEVATION 15.6.5-2  NORMALIZED REACTOR THERMAL POWER FOR LBLOCA (LIMITING CASE) 15.6.5-3  ECCS FLOWS (ACCUMULATOR, CHARGING, SI AND RHR) FOR LBLOCA 15.6.5-3a DELETED BY AMENDMENT NO. 50 15.6.5-4  INTACT LOOP AND BROKEN LOOP HHSI FLOW FOR LBLOCA (LIMITING CASE 15.6.5-5  INTACT LOOP AND BROKEN LOOP LHSI FLOW FOR LBLOCA (LIMITING CASE) 15.6.5-6  UPPER PLENUM PRESSURE FOR LBLOCA (LIMITING CASE) 15.6.5-7  TOTAL BREAK FLOW RATE FOR LBLOCA (LIMITING CASE) 15.6.5-8  AVERAGE CORE INLET MASS FLUX FOR LBLOCA (LIMITING CASE) 15.6.5-9  HOT ASSEMBLY INLET MASS FLUX FOR LBLOCA (LIMITING CASE) 15.6.5-10 PCT NODE VOID FRACTION FOR LBLOCA (LIMITING CASE) 15.6.5-11 PCT NODE AVERAGE FUEL, CLADDING SURFACE AND FUEL CENTERLINE TEMPERATURES FOR LBLOCA (LIMITING CASE) 15.6.5-12 PCT NODE HEAT TRANSFER COEFFICIENT FOR LBLOCA (LIMITING CASE) 15.6.5-13 PCT NODE HEAT FLUX FOR LBLOCA (LIMITING CASE) 15.6.5-14 CONTAINMENT AND LOOP PRESSURES FOR LBLOCA (LIMITING CASE)
Amendment 63                                                              Page 10 of 12
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 FIGURE                                            TITLE 15.6.5-15 UPPER PLENUM PRESSURE FOR LBLOCA (LIMITING CASE) 15.6.5-16 DOWNCOMER COLLAPSED LIQUID LEVEL FOR LBLOCA (LIMITING CASE) 15.6.5-17 DELETED BY AMENDMENT NO. 58 15.6.5-18 CORE COLLAPSED LIQUID LEVEL FOR LBLOCA (LIMITING CASE) 15.6.5-19 DELETED BY AMENDMENT NO. 58 15.6.5-20 PEAK AND RUPTURE LOCATION CLADDING TEMPERATURES FOR LBLOCA (LIMITING CASE) 15.6.5-21 DELETED BY AMENDMENT NO. 50 15.6.5-22 DELETED BY AMENDMENT NO. 50 15.6.5-23 DELETED BY AMENDMENT NO. 50 15.6.5-24 DELETED BY AMENDMENT NO. 50 15.6.5-25 DELETED BY AMENDMENT NO. 50 15.6.5-26 DELETED BY AMENDMENT NO. 50 15.6.5-27 DELETED BY AMENDMENT NO. 50 15.6.5-28 DELETED BY AMENDMENT NO. 48 15.6.5-29 DELETED BY AMENDMENT NO. 48 15.6.5-30 SYSTEM PRESSURES FOR LIMITING BREAK (2.6 INCH SBLOCA) 15.6.5-31 BREAK FLOW RATE FOR LIMITING BREAK (2.6 INCH SBLOCA) 15.6.5-32 DOWNCOMER AND HOT ASSEMBLY COLLAPSED LIQUID LEVELS FOR LIMITING BREAK (2.6 INCH SBLOCA) 15.6.5-33 HHSI FLOWS FOR LIMITING BREAK (2.6 INCH SBLOCA) 15.6.5-34 COMBINED ACCUMULATOR FLOW FOR LIMITING BREAK (2.6 INCH SBLOCA) 15.6.5-35 TOTAL RCS MASS AND REACTOR VESSEL MASS FOR LIMITING BREAK (2.6 INCH SBLOCA) 15.6.5-36 HOT ROD CLADDING TEMPERATURE FOR LIMITING BREAK (2.6 INCH SBLOCA) 15.6.5-37 DELETED BY AMENDMENT NO. 46 15.6.5-38 DELETED BY AMENDMENT NO. 46 15.6.5-39 DELETED BY AMENDMENT NO. 46 15.6.5-40 DELETED BY AMENDMENT NO. 46 Amendment 63                                                              Page 11 of 12
 
Shearon Harris Nuclear Power Plant        UFSAR Chapter: 15 FIGURE                              TITLE 15.6.5-41 DELETED BY AMENDMENT NO. 46 15.6.5-42 DELETED BY AMENDMENT NO. 46 15.6.5-43 DELETED BY AMENDMENT NO. 46 15.6.5-44 DELETED BY AMENDMENT NO. 46 15.6.5-45 DELETED BY AMENDMENT NO. 45 15.6.5-46 DELETED BY AMENDMENT NO. 45 15.6.5-47 DELETED BY AMENDMENT NO. 45 15.6.5-48 DELETED BY AMENDMENT NO. 45 15.6.5-49 DELETED BY AMENDMENT NO. 45 15.6.5-50 DELETED BY AMENDMENT NO. 45 15.6.5-51 DELETED BY AMENDMENT NO. 45 15.6.5-52 DELETED BY AMENDMENT NO. 45 15.6.5-53 DELETED BY AMENDMENT NO. 45 15.6.5-54 DELETED BY AMENDMENT NO. 45 15.6.5-55 DELETED BY AMENDMENT NO. 45 15.6.5-56 DELETED BY AMENDMENT NO. 45 15.6.5-57 DELETED BY AMENDMENT NO. 45 15.6.5-58 DELETED BY AMENDMENT NO. 51 15.7.4-1  DELETED BY AMENDMENT NO. 49 Amendment 63                                  Page 12 of 12
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.0.5-6 NORMALIZED RCCA REACTIVITY WORTH VERSUS TIME AFTER ROD DROP BEGINS Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 FIGURE 15.0A.1-1 CONTAINMENT LEAKAGE DOSE MODEL
                                                                    ----+-----2 J
4 NOTES,
: 1. CONTAINMENT SPRAY WlTH SODIUM HYORO:XIOE.
: 2. CONTAINMENT LEAKAGE TO ATMOSPHERE.
J. UNSPRAYED REGION OF CONTAINMENT.
: 4. AIR EXCHAfllGE vrA CONTAINl.4ENT FAN COOlER SYST;EM.
Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.1.2-1 REACTOR POWER FOR INCREASE IN FEEDWATER FLOW-LIMITING CASE 4000 3500 I s:-. 3000 .
        .~
ai 2500 '
::0 Cl.
o 2000 ti .
tO 11) a: 1500 1000 .
500 20      30    40    50    60 70 90  100 Time (s)
Amendment 61                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                                                          UFSAR Chapter: 15 FIGURE 15.1.2-2 PRIMARY SYSTEM TEMPERATURES FOR INCREASE IN FEEDWATER FLOW -
LIMITING CASE 590
                                                                                                                                            --            Tavg Overall 580                                                                                                                          ~
                                                                                                                                            -
* Tavg Loop 1
                                                                                                                                            ---- **      Tavg Loop 2 570                                                                                                                          --*- -      Tavg Loop 3
                                                                                                                                            - -- -      T Loop 1 CL LL'
        ;sso
:::i iQi-540 Q,
          ; 530 1-520 480 .............._._._.........................................................................""""--aJL.a..La..l~
0            l0              20                30                40                50            60              70                BO                90              100 Time (s)
Amendment 61                                                                                                                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                                                                        UFSAR Chapter: 15 FIGURE 15.1.2-3 REACTIVITIES FOR INCREASE IN FEEDWATER FLOW - LIMITING CASE
                                                                                                                                                                          --- <Total>
3                                                                                                    __,.... -~- -*......                                    - * - <Doppler>
                                                              ,r,                      // __ .* ,*                                                        ~-.. , .....:..~.... -- <Moderator>
                                                          ,/        \,              ;'                                                                                          -*-~~------~ .. ~- -"*--~-
J*
            -3
            -4 ........ ..............................................................................__.__.L,....L._._..__...............................
o                                                      so                                      so                  eo                                              ao                                      mo
                      ~                                                                                                                                    ..:.i....i............................__._..........................,
10                    20                                          40                                                            10                                          90 Time {s)
Amendment 61                                                                                                                                                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                                          UFSAR Chapter: 15 FIGURE 15.1.2-4 PRESSURIZER PRESSURE FOR INCREASE IN FEEDWATER FLOW - LIMITING CASE 250Q,..............,..-r-r--.-.-,.......,.......,........-.....,.........-,-....,,-.-.--..-.-.,,...,-..,.,...,l'"T""l'.......,.."T""l""..-r-.....---,m.....-T"T"...,
I 24()0 2300  I
        ....... 2200
        ~
        "iii 0.
        !::i 2100
        ,1t,1 I,?
CL    2000 1900 :
1800 1700 O 10          20                  30                40              50                00            70          80            90        100 Time (s)
Amendment 61                                                                                                                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 FIGURE 15.1.2-5 STEAM GENERATOR COLLAPSED LEVEL FOR INCREASE IN FEEDWATER FLOW -
LIMITING CASE 110 100 90 80 70
        ~
0 lQ) 60
        ..J
        *5 50
        ,::i tr J
40 30
                                                          -          SG-1 (Affected)
                                                          -
* SG-2 (Una1fected) 20                                          ...... .. SG- 3 (Unaffected) 10 0                                                                        -
0  10    20      30    40    50    60      70      80      90  100 Time (s)
Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.1.3-1 REACTOR POWER FOR INCREASE IN STEAM FLOW - BOC CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.1.3-2 PRESSURIZER PRESSURE FOR INCREASE IN STEAM FLOW - BOC CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.1.3-3 REACTIVITIES FOR INCREASE IN STEAM FLOW - BOC CASE Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.1.3-4 VESSEL AVERAGE TEMPERATURE FOR INCREASE IN STEAM FLOW - BOC CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.1.5-1 STEAM GENERATOR PRESSURES FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.1.5-2 BREAK FLOW RATES FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.1.5-3 STEAM GENERATOR INVENTORIES FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE WITH A STUCK ROD Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.1.5-4 PRESSURIZER PRESSURE FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.1.5-5 REACTOR POWER FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.1.5-6 REACTIVITY COMPONENTS FOR THE HZP MSLB WITH OFFSITE POWER AVAILABLE AND WITH A STUCK ROD Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.3-1 REACTOR POWER FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.3-2 AVERAGE TEMPERATURES FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.3-3 PRESSURIZER PRESSURE AND LOWER HEAD PRESSURE FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.2.3-4 PRESSURIZER LEVEL FOR TURBINE TRIP PRIMARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.3-5 REACTOR POWER FOR TURBINE TRIP MDNBR CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.3-6 AVERAGE TEMPERATURES FOR TURBINE TRIP MDNBR CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.3-7 PRESSURIZATION PRESSURE AND LOWER HEAD PRESSURE FOR TURBINE TRIP MDNBR CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.3-9 REACTOR POWER FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.3-10 AVERAGE TEMPERATURES FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.3-11 PRESSURIZER LEVEL FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.3-12 STEAM GENERATOR LOWER SHELL PRESSURE FOR TURBINE TRIP SECONDARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                              UFSAR Chapter: 15 FIGURE 15.2.6-1 REACTOR POWER FOR LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 4000 3000
  ~
  ~
Cl>
  ~
0  2000 a.
  ~
Cl>
C:
i:
1000 0
0  1000    2000  3000  4000    5000 6000 7000 8000  9000 10000 Time (sec)
Amendment 62                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 FIGURE 15.2.6-2 PRESSURIZER LEVEL FOR LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 1000  2000  3000    ooo  0000    eooo 1000 aooo e    10000 Tims (see)
Amendment 62                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 FIGURE 15.2.6-3 PRESSURIZER PRESSURE FOR LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 2400 2300
  *a 220*
  .,.._~
    ;;;ii (1.. 2100
    ,IIJ!
    ;;;ii tLli
    ,i,,,
I      2000 n.
118100 1() 10001 .:moo 3  *o  40 o  5 o eo  o 1000 aooo sooo  1o oo Time, (see)
Amendment 62                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                              UFSAR Chapter: 15 FIGURE 15.2.6-4 RCS TEMPERATURES FOR LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES
                                                              - Th
                                                        -Tooldl
                                                        = = ~ Tavg ti:'
60ID 1 i
    ',w E
  ~
U)    SB~Ji 0
0::
540 0  1000 2000  30001 40ll0    5000  1uoo1  BOCm  9000  1cmoo Tl Amendment 62                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                  UFSAR Chapter: 15 FIGURE 15.2.6-5 STEAM GENERATOR MASS INVENTORY FOR LOSS OF NONEMERGENCY AC POWER TO THE STATION AUXILIARIES 120000                                                ~      86.-1
                                                              -    ,SG-2
                                                              - - ---SG-3 11000001 60000 40000
        .2000 0
0  1OIJO 2.000    3000  40    5000 60 . 7000    8000    ,000  0I 0 Time (see)
Amendment 62                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 FIGURE 15.2.7-1 REACTOR POWER FOR LOSS OF NORMAL FEEDWATER FLOW 3000 '
  'P!
:52*
1000 I
lo.
0 mono
                ~
          -0      1000  2000  3000  4000    5000    6000 7000 SO 0  9000 Ti  (sec),
Amendment 62                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                        UFSAR Chapter: 15 FIGURE 15.2.7-2 PRESSURIZER LEVEL FOR LOSS OF NORMAL FEEDWATER FLOW
                                                                                                                                    -  cvAR TO
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                                                                                                                      -L.. L. .L.----...__-------
6000      1000    ao o  '91JO!D  1001m n*          (SCC)
Amendment 62                                                                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                            UFSAR Chapter: 15 FIGURE 15.2.7-3 PRESSURIZER PRESSURE FOR LOSS OF NORMAL FEEDWATER FLOW 2300 2200 ffl,
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Amendment 62                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                                        UFSAR Chapter: 15 FIGURE 15.2.7-4 RCS TEMPERATURES FOR LOSS OF NORMAL FEEDWATER FLOW 6- 0  r--r---.-.- r -..........--.--.-r--r-r--r---.-r-,.......,..---.----,r-r-,.......---,-,...--.--r-r---,-,...--.-.........---,-,.-r--.-.----_,H,_~-,--,-Let-
                                                                                                                                                                  * -(C=v..,..,:A--=R,...,t...,.5=l-
                                                                                                                                                  - ..... Cold Lev tCV                    H 13)
                                                                                                                                                    - - - A i;,ra      fCVAR U 71
                                                                                                                                                                                              --i 560 1 0 .________..__ _ _ _ _......,_ _......,_ _.._.._ _...__ _......,_________
0          1000                21!l00          30(10          * -400                  5000              61JOO                7[100        B    !()      !JOOO                    1OIJIJD e (sec)
Amendment 62                                                                                                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                          UFSAR Chapter: 15 FIGURE 15.2.7-5 STEAM GENERATOR MASS INVENTORY FOR LOSS OF NORMAL FEEDWATER FLOW 12001lO ~~~~.........- - - - - - - - - - - - - - - - - - -                                                                ...........          , -~""'20"'"'0-1,-
                                                                                                                                      --5--G_1_{C:...'V,-AR SG2 ~CY:AR 6200]
                                                                                                                                    - -        56]:JC\IAR 1200 100000
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c                    1000                                                        z1000    soo      soolll    aoo          aooo      ,90 o              i 0000 rne~see)1 Amendment 62                                                                                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.8-1 REACTOR POWER FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.8-2 PRESSURIZER LEVEL FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.8-3 PRESSURIZER PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-4 LOOP 1 PRIMARY SYSTEM TEMPERATURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-5 LOOP 2 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-6 LOOP 3 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-7 STEAM GENERATOR PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-8 STEAM GENERATOR NARROW RANGE LEVEL FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-9 REACTOR COOLANT SYSTEM FLOW FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH OFFSITE POWER Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.2.8-11 REACTOR POWER FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.2.8-12 PRESSURIZER LEVEL FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.2.8-13 PRESSURIZER PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.8-14 LOOP 1 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.8-15 LOOP 2 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.8-16 LOOP 3 PRIMARY SYSTEM TEMPERATURES FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-17 STEAM GENERATOR PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.8-18 STEAM GENERATOR NARROW RANGE LEVEL FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-19 REACTOR COOLANT SYSTEM FLOW FOR FEEDWATER SYSTEM PIPE BREAK LONG-TERM CORE COOLING CASE WITH LOSS OF OFFSITE POWER Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                  UFSAR Chapter: 15 FIGURE 15.2.8-21 REACTOR POWER FOR FEEDWATER SYSTEM PIPE BREAK SHORT-TERM CORE COOLING CASE Amendment 63                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-22 PRESSURIZER PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK SHORT-TERM CORE COOLING CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-23 REACTOR COOLANT SYSTEM FLOW FOR FEEDWATER SYSTEM PIPE BREAK SHORT-TERM CORE COOLING CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-24 PRIMARY SYSTEM AVERAGE TEMPERATURE FOR FEEDWATER SYSTEM PIPE BREAK SHORT-TERM CORE COOLING CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.2.8-25 REACTOR POWER FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.2.8-26 PRESSURIZER LEVEL FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-27 PRESSURIZER PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-28 PRIMARY SYSTEM AVERAGE TEMPERATURE FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-29 STEAM GENERATOR PRESSURE FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-30 STEAM GENERATOR NARROW RANGE LEVEL FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-31 REACTOR COOLANT SYSTEM FLOW FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.2.8-32 TOTAL PRESSURIZER RELIEF FLOW FOR FEEDWATER SYSTEM PIPE BREAK PEAK PRIMARY PRESSURE CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.2-1 REACTOR POWER FOR LOSS OF FORCED REACTOR COOLANT FLOW Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.2-2 CORE AVERAGE HEAT FLUX FOR LOSS OF FORCED REACTOR COOLANT FLOW Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.2-3 PRESSURIZER PRESSURE FOR LOSS OF FORCED REACTOR COOLANT FLOW Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.3.2-4 PRESSURIZER LEVEL FOR LOSS OF FORCED REACTOR COOLANT Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.2-5 REACTOR COOLANT SYSTEM MASS FLOW RATE FOR LOSS OF FORCED REACTOR COOLANT FLOW Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.2-6 CORE INLET AND OUTLET TEMPERATURES FOR LOSS OF FORCED REACTOR COOLANT FLOW Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.3.2-7 TOTAL CORE REACTIVITY FOR LOSS OF FORCED REACTOR COOLANT FLOW Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.3-1 REACTOR POWER FOR LOCKED ROTOR OVERPRESSURIZATION CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.3-2 PRESSURIZER LEVEL FOR LOCKED ROTOR OVERPRESSURIZATION CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.3-3 REACTOR COOLANT SYSTEM MASS FLOW RATE FOR LOCKED ROTOR OVERPRESSURIZATION CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.3.3-4 AVERAGE FLUID TEMPERATURES FOR LOCKED ROTOR OVERPRESSURIZATION CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.3.3-5 TOTAL CORE REACTIVITY FOR LOCKED ROTOR OVERPRESSURIZATION CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.3.3-6 MAXIMUM PRIMARY SYSTEM PRESSURE FOR LOCKED ROTOR OVERPRESSURIZATION CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.3.3-7 REACTOR POWER LEVEL FOR LOCKED ROTOR MDNBR CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.3.3-8 CORE AVERAGE HEAT FLUX FOR LOCKED ROTOR MDNBR CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.3-9 PRESSURIZER PRESSURE FOR LOCKED ROTOR MDNBR CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.3.3-10 PRESSURIZER LEVEL FOR LOCKED ROTOR MDNBR CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.3.3-11 REACTOR COOLANT SYSTEM MASS FLOW RATE FOR LOCKED ROTOR MDNBR CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.3.3-12 AVERAGE FLUID TEMPERATURES FOR LOCKED ROTOR MDNBR CASE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.3.3-13 TOTAL CORE REACTIVITY FOR LOCKED ROTOR MDNBR CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                  UFSAR Chapter: 15 FIGURE 15.4.1-1 REACTOR POWER FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT HZP Amendment 63                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.4.1-2 REACTIVITY COMPONENTS FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT HZP Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.4.1-3 CORE AVERAGE HEAT FLUX FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT HZP Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.4.1-4 LOOP AVERAGE FLUID AND FUEL TEMPERATURES FOR UNCONTROLLED RCCA BANK WITHDRAWAL AT HZP Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.4.2-1 MINIMUM DNBR VS REACTIVITY INSERTION RATE FOR RCCA BANK WITHDRAWAL AT BOC 10% POWER LEVEL CASES Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.2-2 REACTOR POWER FOR LIMITING RCCA BANK WITHDRAWAL CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.4.2-3 REACTIVITIES FOR LIMITING RCCA BANK WITHDRAWAL CASE Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.2-5 CORE POWER BASED ON ROD SURFACE HEAT FLUX FOR LIMITING RCCA BANK WITHDRAWAL CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.2-6 PRIMARY TEMPERATURES FOR LIMITING RCCA BANK WITHDRAWAL CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.2-7 PRESSURIZER LEVEL FOR LIMITING RCCA BANK WITHDRAWAL CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.2-8 RCS MASS FLOW RATE FOR LIMITING RCCA BANK WITHDRAWAL CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.2-9 PRESSURIZER PRESSURE FOR LIMITING RCCA BANK WITHDRAWAL CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.4.2-10 PRESSURE AT BOTTOM OF REACTOR VESSEL FOR LIMITING RCCA BANK WITHDRAWAL PRIMARY SIDE OVERPRESSURIZATION CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.4.3-1 REACTOR POWER FOR LIMITING DROPPED ROD CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.4.3-2 AVERAGE CORE LINEAR HEAT GENERATION RATE FOR LIMITING DROPPED ROD CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.3-3 PRESSURIZER PRESSURE FOR LIMITING DROPPED ROD CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.3-4 PRIMARY TEMPERATURES FOR LIMITING DROPPED ROD CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.4.3-6 CORE INLET FLOW FOR LIMITING DROPPED ROD CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.4.3-7 PRESSURIZER LEVEL FOR LIMITING DROPPED ROD CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.4.3-8 REACTIVITY FOR LIMITING DROPPED ROD CASE Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.4.3-10 ROD POSITION FOR LIMITING DROPPED ROD CASE Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.4.3-11 ROD SPEED FOR LIMITING DROPPED ROD CASE Amendment 63                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.3-12 REACTOR POWER FOR SINGLE UNCONTROLLED RCCA WITHDRAWAL AT HFP Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.3-13 CORE AVERAGE HEAT FLUX FOR SINGLE UNCONTROLLED ROD WITHDRAWAL AT HFP Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.4.3-14 AVERAGE FLUID TEMPERATURES FOR SINGLE UNCONTROLLED RCCA WITHDRAWAL AT HFP Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.3-15 PRESSURIZER PRESSURE FOR SINGLE UNCONTROLLED ROD WITHDRAWAL AT HFP Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                      UFSAR Chapter: 15 FIGURE 15.4.8-9 REACTOR POWER FOR RCCA EJECTION - EOC 50% POWER CASE Amendment 63                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.4.8-11 TOTAL CORE REACTIVITY FOR RCCA EJECTION - EOC 50% POWER CASE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.5.1-15 PRESSURIZER LIQUID LEVEL (PRESSURIZER OVERFILL CASE) i500
                                                ~----_
0        200        400      tl00      aoo        1200 Time (sec)
Amendment 61                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                              UFSAR Chapter: 15 FIGURE 15.5.1-16 REACTOR COOLANT SYSTEM TEMPERATURES (PRESSURIZER OVERFILL CASE) air:,-~Q'      RCS Hot Leg oo-----ci::i  RCS Cold Leg
                                                                                .ta.ta-....r.a BCS Averaga fiQO .__..___.__-'L-_.___.__...L.-__.____,_---'-__.___.___,_____._____,_____.__,__,.___.___.__........................__ _
400              6DO              aaD              tOOD Time ($ec}
Amendment 61                                                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.5.1-17 PRESSURIZER PRESSURE (PRESSURIZER OVERFILL CASE) 0                    400      600*    BDD  1000    12DO Time (sec)
Amendment 61                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 FIGURE 15.5.1-18 SAFETY INJECTION FLOW RATE (PRESSURIZER OVERFILL CASE)
                                                      ,0..CI- ~ C l  RCS Laop 1 l"'lioo-~o    RCS Loop 2 a-p.-.....ii.a Ra.3 Loop 3
                    .21)()      .4{IO      &JO      800              1000      1200 Time (sec)
Amendment 61                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.5.1-19 REACTOR POWER (PRESSURIZER OVERFILL CASE)
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                                        *Time (sec)
Amendment 61                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                      UFSAR Chapter: 15 FIGURE 15.5.1-20 PRESSURIZER SAFETY RELIEF VALVE INLET TEMPERATURE (PRESSURIZER OVERFILL CASE) 700 ____________________________...........,...__.....,._ __,___,_ __
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                                                  -  ...a...~=.1..~=--1._.,__,,_,.__..._....,__,__...._.....____,,.__....J.......i.aa=.L......,...._.__.____.,
0                                                              &00                eoo                          10 00                1200 Time (sec)
Amendment 61                                                                                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                              UFSAR Chapter: 15 FIGURE 15.5.1-21 PRESSURIZER SAFETY RELIEF VALVE INLET PRESSURE (PRESSURIZER OVERFILL CASE)
:2700 m
I 2500 2250 f,l;i
    ~
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    ~
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1'15-0 Q                400      1600      8.01)              1~00 T'ime (sec)
Amendment 61                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.6.1-1 REACTOR POWER FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.6.1-2 CORE AVERAGE HEAT FLUX FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.6.1-3 PRESSURIZER PRESSURE FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.6.1-4 PRESSURIZER LEVEL FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.6.1-5 REACTOR COOLANT SYSTEM FLOW RATE FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.6.1-6 REACTOR COOLANT SYSTEM TEMPERATURE FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE Amendment 63                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                        UFSAR Chapter: 15 FIGURE 15.6.1-7 TOTAL CORE REACTIVITY FOR INADVERTENT OPENING OF A PRESSURIZER PRESSURE SAFETY OR POWER OPERATED RELIEF VALVE Amendment 63                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                    UFSAR Chapter: 15 FIGURE 15.6.3-1 PRESSURIZER LEVEL - MARGIN TO OVERFILL ANALYSIS SHNPl' $1eam Generator Tube Rnptllre
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500      1000          1500          2000  2500    3000 Time (s)
The HNP MTO analysis has been amended to account for a TDAFW pump speed controller failure that results in increased AFW delivery over the first 528 seconds. The final MTO is unchanged (66 cu ft) but the level history in this figure applies to the original analysis.
Amendment 61                                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                              UFSAR Chapter: 15 FIGURE 15.6.3-2 PRESSURIZER PRESSURE - MARGIN TO OVERFILL ANALYSIS SHNl'l' Ste~rn Generator Tube Rupture 2soo---r----------------- ----~~--.. .
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                        .._,_, 1500 * ' . * *'.' '' ' ' * ' '' ' ' ' *. '' ' ' '
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                          <V D....
500 .. .. " . . .
10:)0          1500              2000                2500              3000 Time (s)
The HNP MTO analysis has been amended to account for a TDAFW pump speed controller failure that results in increased AFW delivery over the first 528 seconds. The final MTO is unchanged (66 cu ft) but the pressure history in this figure applies to the original analysis.
Amendment 61                                                                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                              UFSAR Chapter: 15 FIGURE 15.6.3-3 SECONDARY PRESSURE - MARGIN TO OVERFILL ANALYSIS SHNPP Steam Ge11erato.r Tube R\1ptnrt Ruptur~~ Sleom Gen~rotqr l~loct St~om CenerQlot' 1200-------------- ----~---------,
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0      500      1000                  1500        2000 2500                3000 Time (s}
The HNP MTO analysis has been amended to account for a TDAFW pump speed controller failure that results in increased AFW delivery over the first 528 seconds. The final MTO is unchanged (66 cu ft) but the pressure history in this figure applies to the original analysis.
Amendment 61                                                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                                                                              UFSAR Chapter: 15 FIGURE 15.6.3-4 INTACT LOOP HOT AND COLD LEG TEMPERATURES - MARGIN TO OVERFILL ANALYSIS Ho I            l c 13 Cold              Leg 550 ......- - -- - - -~ -- - - - - - -- -- - - - - - -----,
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The HNP MTO analysis has been amended to account for a TDAFW pump speed controller failure that results in increased AFW delivery over the first 528 seconds. The final MTO is unchanged (66 cu ft) but the temperature histories in this figure applies to the original analysis.
Amendment 61                                                                                                                                                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                      UFSAR Chapter: 15 FIGURE 15.6.3-5 PRIMARY TO SECONDARY BREAK FLOW MARGIN TO OVERFILL ANALYSIS SHNPP Sttam Ge~to* lUbe ,Rupture 50-r---------------- ------------,
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0                500            1000              1500                      2000              2500        3000 Time (s)
The HNP MTO analysis has been amended to account for a TDAFW pump speed controller failure that results in increased AFW delivery over the first 528 seconds. The final MTO is unchanged (66 cu ft) but the break flow history in this figure applies to the original analysis.
Amendment 61                                                                                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                                    UFSAR Chapter: 15 FIGURE 15.6.3-6 RUPTURED SG WATER VOLUME MARGIN TO OVERFILL ANALYSIS oooc,~-------------------------------
                          ---- ~----------- ~---~-- ----~
                                                                                                                                                      --~~
                                                                                                                                                *-----~
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* I I *  *~ * ** : I ~ * *
* 1000 0---~~~---~~~~-.~~-~~-r-~*~*~-~-r--~~~~-.-~*~*-~~1-1 0              500                      1000                    150D                  2000                        2.500              3000 Time (s)
The HNP MTO analysis has been amended to account for a TDAFW pump speed controller failure that results in increased AFW delivery over the first 528 seconds. The final MTO remains as shown (66 cu ft) but the volume history in this figure applies to the original analysis.
Amendment 61                                                                                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                              UFSAR Chapter: 15 FIGURE 15.6.3-7 PRESSURIZER LEVEL - OFFSITE RADIATION DOSE ANALYSIS SrlNPP Steam Generator Tube Rupture 10Q .......- -- - - - - - - -- -- - - - - - - - - ,
8,0
            -en i:::.
60
            ~
              ~
II)
            -~
              ";;J m        40 en cu 0...
20 -
0  -l,1................-+-...............+--'-_._."-+-.......- . . -........"-+...,__._._........,........_.&.-;-......_........__.
0                600          12 001 1300        ,2 400        3000                                              41:!0D Time (s)
Amendment 61                                                                                                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                            UFSAR Chapter: 15 FIGURE 15.6.3-8 PRESSURIZER PRESSURE - OFFSITE RADIATION DOSE ANALYSIS SHNPP Steom Generator Tube Rupture 2400 ..,.....-. . . - - - - - - - - - --    - -  -  -  - ---,
2300 2.200 2100 2000
* _g t 9,00 a
                *~ 11,00 l
                =,
(I}
1100 Q,)
                -~ 1600
                =s qi
                ~ 1S00 a.
130 I 1
uoo 1100 0      600  12 00 1800    240-0 3000  "600  42 00 uoo lime (s)
Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                            UFSAR Chapter: 15 FIGURE 15.6.3-9 SECONDARY PRESSURE - OFFSITE RADIATION DOSE ANALYSIS SHNPP Steomi Generator Tube Rupture
                        - - - Fhiptuttd Sl!.orn Ctrlterotor
                        ... - - -    lnlcct stem, Gencrotors 1200 ...----    - - -~ - - - - - - - - - - - - - - - - - - ,
11;0 0
                                                  ~~--- .... ,... ..... ,
1ooo, 900 aoo
                -*~
                'o (l)    700
:::I                                                  1
                                                                        *I (I)
(/)
600
                  ~
i::-
                  '='    500    ._
                                                                          ''t
                  ~
                                                                            \
C 0
                  ~
(/)      (1)0 -                                              \
                                                                                  \
                                                                                    \
JOO                                                        \
4' 200 10 D 0
1 BOO      2400            3000 lime (s)
Amendment 61                                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                    UFSAR Chapter: 15 FIGURE 15.6.3-10 INTACT LOOP HOT & COLD LEG TEMPERATURES - OFFSITE RADIATION DOSE ANALYSIS SHNPP Steam Generator Tube Rupture
                    - - -* Hot:t.ag
                    - - - - *Cotd leg 650 ...,
8DC l
L
                                                                      \
                                                                      '\
                                                                          '\
                                                                            \
                                                                              \
                                                                                \                  , _ . .... lliiiiliiii,;a,..__  I
                                                                                    ' ... ..... ,~
            ......,                                                                  '- .f'-
g      400
            .5 3SO 0      6,00          12 00  1800  240 0                300 01  3600      4200                    "4800 lime (s)
Amendment 61                                                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                        UFSAR Chapter: 15 FIGURE 15.6.3-11 RUPTURED LOOP HOT AND COLD LEG TEMPERATURES - OFFSITE RADIATION DOSE ANALYSIS SHNPP Steam Generator Tube Rupture
                    - - - H'atl.J!g
                    - -*- -  Cold  l.e9 es,o - - - - - - -- - - - - - - - - - - - - -- -
600 11.i...
g'    550    .*-- --,
Q
                                              \
                                                \
                                                  \                                  ,,,.
                                                  ' \                            / ~
                                                      \
                                                        ;,,,                  I
                                                              '                I
                                                                "'- ,        I JSO JOO -r-.........i..."'--'--1....L...11,...1..-'-l....l,._,_~ .....,h~..,.J...~ - '-'-'-1-i-.i..-i...1-1!-,..J....&...l.-1 0          600                              1800 2400 JOOO 3600 200 4800 Time (sJ Amendment 61                                                                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 FIGURE 15.6.3-12 DIFFERENTIAL PRESSURE BETWEEN RCS AND RUPTURED SG - OFFSITE RADIATION DOSE ANALYSIS SHNPP Sleom Gen erot.or Tube Rupture 1100  1 1 &90 1*soo 1 HIO uo a 1'200 mo "i'.i) 1 c..
                  ,000
            ~
            -::,    900 en u,
a.:.* eoo Cl.I
            !I...
          ~          70 0
          -c Cl.!
I....
            ,cp      tiOO
                    )00 400 lQiO 200    -
100 0
                                          ,,                                    I'
                    -100
                          - 0  6,00  I :iHHJ  UOO    2400  lOCQ  J-600  -UQO  *UD*O Time (s)
Amendment 61                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 FIGURE 15.6.3-13 PRIMARY TO SECONDARY BREAK FLOW - OFFSITE RADIATION DOSE ANALYSIS SHNPP Steam Gene ratOf' lobe Rupture 60 - - - -- - - -~ - -- - - - - - - - - - .
SO*
                <.Ji
                ~
            ......... JO E
              ..0
                ~
o    ,
u:
              ~          20 a
m
              *~
10 IJ 18~0  2400    3000  3B0O  4200  ~800 Time (s)
Amendment 61                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 FIGURE 15.6.3-14 RUPTURED SG WATER VOLUME - OFFSITE RADIATION DOSE ANALYSIS SHNPP *Steam Generator lube Rupture 6000  --------~--------------:1
            -g 5UO (b
E
                ;:::I 4000 1
              ~
JB C
3:
i-0 C      3000 (I,>
c:;:
d)
C)
E 2
              <n      2000
              ~.....
15'..
:)
et::
HJDO
* 0  6C O  120,0  1 BOO    2
* 80 3000 .3600 U00  4800 lime (s)
Amendment 61                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 FIGURE 15.6.3-15 RUPTURED SG WATER MASS - OFFSITE RADIATION DOSE ANALYSIS SHNPP Steam Generator Tube Rupture 1800il0- - - - ~ ~ - - ~ - ~ - - - - - - - - - - - ,
150000 *
* ts
              ~      120000
              ~
              ~
                ....0
              ]        90000 ai *
(.:)
JOOOD 0    600        1800    2~00  3000 lime (s}
Amendment 61                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 FIGURE 15.6.3-16 RUPTURED SG MASS RELEASE RATE TO THE ATMOSPHERE - OFFSITE RADIATION DOSE ANALYSIS
                              .SHNPP Steam Generator Tube Rupture 600 ~- - - - -- .....................- - - - - - - - - - - - . . . . . . ,
            ~
0 (I)
            -    400
            ~
i=i.        '
1:i.. 100.
* I e:::*        ''
            ~
0  6 00  1200        1800        24 00  ,3000  3600  4z&deg;00    4BIJ ,0 lime (s)
Amendment 61                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                    UFSAR Chapter: 15 FIGURE 15.6.3-17 INTACT SGS MASS RELEASE RATE TO THE ATMOSPHERE - OFFSITE RADIATION DOSE ANALYSIS SHNP? Steam Generator Tube Rupture 1200 - -- - - - -- ~- --- - - -- - - - - - - - ~
            ~ 1000 ' I G>
f.n E
            -=-*
            ..C 200 -      I
                      ,(I .JJJ....?!:::,,,,~,,,i,,,,,,ib.f,...1....1....,1..~...!...L-J.....I....U...t.,..L.LJ-11-LJ.1-J~..L.J...,J:--1 0          EillO        1200          tebo 2400 lGGO 3600 noo                                        ,4800
                                                                        - rme (s)
Amendment 61                                                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                        UFSAR Chapter: 15 FIGURE 15.6.3-18 IODINE TRANSPORT MODEL - OFFSITE RADIATION DOSE ANALYSIS
                                -11 l'flASH tffT01'                                    -  *~-  ,~
II VAPOR&_ I                                  r-,
PRIMARY DROPL:ETS                                            s a        ..                                        T                  I E
* T WATER I
NOT                                    . A,            ,A, RASH:ED r-1 ,~ y
                                                            .... ...          --      r it        ~,  1,_
1 WATER
                                                  ~
S1'&#xa3;AM      I M
s        I  M' 0
I I
p s
A            p C            H i~              E IA e
                                                      *PAATITION I
Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                              UFSAR Chapter: 15 FIGURE 15.6.3-19 BREAK FLOW FLASHING FRACTION - OFFSITE RADIATION DOSE ANALYSIS SHNPP Sleom Generator Tube Rupture l
                            . rn ........- - - - - - - - - - - - - - - - - -*- - -- -......
                            - 1&
                            . 14
                .g C:
12 y
0 I.A...
Q'I          .I
                .ii
                  'C" 0
i;.:
                  ~  . 3[ *- 01 0
Ci::
                  ,CJ ClJ d5
* H-0 I 0 -a,...i...;i:.-'-f.-'-..&...&...l--.li,,,,,.L-L-4-L-L..i.....l-Ll--'--l'-f-.&..i..L..,l...!,-'--'-4..i..~-11
                                    *(I            tUJlj      1200            UDO        200          JODO        HOO          UCO      4800 lime (s)
Amendment 61                                                                                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                UFSAR Chapter: 15 FIGURE 15.6.3-22 TOTAL FLASHED BREAK FLOW - SGTR OFFSITE RADIATION DOSE ANALYSIS 1200,Q 11 OIHI 10 00-0 9Q0(1 E    8 0 !0,(1 '
              ..c
            .e-,
                ~    700~
Li:
              ~
                ''E'  6000
              ~
              ~
[.A C
5006 C:
0 4, QO{I 0
:,ooa 20:CG 10*0Q
                              -0 0 l i 00 1  1:200 1B 11-0  2 H)~  J ODO JSDO  420~  ii 6,Ga Time: (s)
Amendment 61                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                    UFSAR Chapter: 15 FIGURE 15.6.3-23 RUPTURED SG MASS RELEASE RATE TO THE ATMOSPHERE - SGTR OFFSITE RADIATION DOSE ANALYSIS 600 ....
Amendment 61                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 FIGURE 15.6.3-24 INTACT SG MASS RELEASE RATE TO THE ATMOSPHERE - SGTR OFFSITE RADIATION DOSE ANALYSIS 1200 - - -- - - - -- -- - - - - - - - -....
          -*Iv!'
            *c i
          ..S:    ""0 I:
                      '~"'
1200  18 H  :H ,~O ~ 000 315,Q:O -4 200- ~e-00 Time (s)*
Amendment 61                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                UFSAR Chapter: 15 FIGURE 15.6.5-1 NORMALIZED AXIAL DEPENDENCE OF            VERSUS ELEVATION l~ - - --        - - -- - --          - - -~ ~ ~ -- - - -- - - - .
0 +--------,r---~-----'t'......;___ _ _ _ _ _ _ _ _ _ __      , - -_ ____.
0            2                                      a      10          12 Bevallon in Core (Jtl Amendment 61                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 FIGURE 15.6.5-2 NORMALIZED THERMAL REACTOR POWER FOR LBLOCA (LIMITING CASE)
Normalized Power 0.0      [ ~ FractiollQf ~nitial Core Power I
              -0.5
              -1.0
              -1 .5 100  200  300      400      500    600 700 800  000 Time (a)
Amendment 61                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                                    UFSAR Chapter: 15 FIGURE 15.6.5-3 ECCS FLOWS (ACCUMULATOR, CHARGING, SI AND RHR) FOR LBLOCA (LIMITING CASE)
ECCS Flows 4000  r------r----.--                            '"T-----.-                            ----r--......--......-----.--- - , - ~
                                                                                                            - - Loop 1 (broken)
                                                                                                              ---- Locp2
                                                                                                              - - ~ Loop3 3000  I If I
D::
2000 ii u..
1000
                          **"I. ****--' - * -,,.,*- -*~ _,. .,,. _ * .., , . , n . .._.,. _ _. ., u ,_ * - - * - - - -*-**- - - -- --~ - .
* e *r, .... ._
_.-..,..........,........,,...,....____                                      ----~                        -          -
200                                400                              600                      800          1000 Time (s}
Amendment 61                                                                                                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                      UFSAR Chapter: 15 FIGURE 15.6.5-4 INTACT LOOP AND BROKEN LOOP HHSI FLOW FOR LBLOCA (LIMITING CASE)
HHSI Flows 30 ,---,---,---.--r-
                                      , - ......._ ...........,_..,...__,....-r---r-  , ----r--r--...--T"""""""O---,.--, - -_..........,
25
                                                                            -              Locp -1 toroken)
                                                                            .... ...... ~ Loop2
                                                                            - ~ - Loop3 20"'
                                                ~---*----*-*-*~- ------~-ot 10 20                  40                      60                  80          100            120 Tme(s)
Amendment 61                                                                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                            UFSAR Chapter: 15 FIGURE 15.6.5-5 INTACT LOOP AND BROKEN LOOP LHSI FLOW FOR LBLOCA (LIMITING CASE)
LHSI Flows 1-*                      (I=
Loop    1 (bro~ ]
                                      *--*- Leep        2+ 3 250                                        *---.,~-...,r-***..._-,. ..., - - -*.,..  .. '  -  . .      __ ..._,
200
                  -I 100 ,.
50 .I 0 ..._.__.....__......_........,_____________________...._.__................- - ~ ~
o              ~                  ~          ~                                ~      100                    1W TJme (s)
Amendment 61                                                                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 FIGURE 15.6.5-6 UPPER PLENUM PRESSURE FOR LBLOCA (LIMITING CASE)
Upper Plenum Pressure 2500 .--------.---,------.-------,----,,--__________..-__                ~
2000                                  I-      Upper Plenum  I
            -.e
            .m ii) 1500 I
              =-
1000 500 0 ~--------'-------'---._____----1-_......._----1,_.........__"----__.____,
0          5            10          15        20          25        30 Time (s)
Amendment 61                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                      UFSAR Chapter: 15 FIGURE 15.6.5-7 TOTAL BREAK FLOW RATE FOR LBLOCA (LIMITING CASE)
Break Flow 80 ..------------.----.---......-----,.---.------,------,,---,
_ - -** Vessel Side
                                                                              - - - - Pump Side
                                                                              - - - Total 60 n
Q 40 i
B JI Cl a:::
j ll.. 20
                      .;2Q ~' _......______.__.........__..L.-..-__.__  __.___~..._______.__ __._______,
0            200            400            600            800          1000 nnie (s)
Amendment 61                                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                    UFSAR Chapter: 15 FIGURE 15.6.5-8 AVERAGE CORE INLET MAS FLUX FOR LBLOCA (LIMITING CASE)
Average Core Inlet Mass Flux 750
                        !-          Average Corel 500 i-250 B
0
    ~
LL, i    -2.50
          -500
          -750 r-1000  ......__,._____._____.i...__--'-----''-'----'--_,______,___,__.____,___ ___.___ _.__  ____.__ _~
0          100        200              300          400        5DO    600    700    800      900 Time (s)
Amendment 61                                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                  UFSAR Chapter: 15 FIGURE 15.6.5-9 HOT ASSEMBLY INLET MASS FLUX FOR LBLOCA (LIMITING CASE)
Hot Assembly Inlet Mass Flux 1000 ......-....--,----.,......--..---,----,----,----,----..,.........-,---,--.,--..---r--..--..---,
750 I-          Hot As.semblyl 500 250
      'ip Ne E
0
        ~
Li:
a
:i
            *250
            -500 w1000 ..___..___,__~.,____~..._~_.___~.,__~...__,.____...__...........
0          100        200        300        400        500        600    700    800      900 Time ,(s)
Amendment 61                                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                                            UFSAR Chapter: 15 FIGURE 15.6.5-10 PCT NODE VOID FRACTION FOR LBLOCA (LIMITING CASE)
Void Fraction I ,,
0.90 0.80 i:::
:a
      ,0
      &#xa3;
:2
      ~
0.70 0.60
                              - - Fraction of R 0 .50 .__...__...IL..-__.__-----,L........._  _L._....,___.___._...........~ - L . . J . . ..............................___..,........UW-.L--___,
0          100        200            300          400              500          600                  700                800          900 Time (s)
Amendment 61                                                                                                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                        UFSAR Chapter: 15 FIGURE 15.6.5-11 PCT NODE AVERAGE FUEL, CLADDING SURFACE AND FUEL CENTERLINE TEMPERATURES FOR LBLOCA (LIMITING CASE)
Comparison of PCT Node Temperatures
                .2500 ,-.-----.--~-.-.........--------.-..---..,...-----,--..--,.-----,,--..-.
                                    - - Fual Avg. of Fuel Centerline to Clad Surface ,
                                    -*---- Clad SUrface 2000              - - - Fuel Centerline Ci:'
1500 1) 500
                                                                              "\. .-- ***~
I      -
                            ~      ~        ~      ~      ~      ~      ~          ~    ~
Time (s)
Amendment 61                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                              UFSAR Chapter: 15 FIGURE 15.6.5-12 PCT NODE HEAT TRANSFER COEFFICIENT FOR LBLOCA (LIMITING CASE)
Heat Transfer Coefficient 1.20
        .....u::-
1 Cl!I
          ~ 0.90 e.
1 I
f I,.,.
0.60 C:
          ~
i
:c
* 0.30 L-- liTC' o.oo 0 100  200  300  400    500      700 800  QOO Time(s)
Amendment 61                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                  UFSAR Chapter: 15 FIGURE 15.6.5-13 PCT NODE HEAT FLUX FOR LBLOCA (LIMITING CASE)
Heat Flux 40 1-ToFluidl
* 10 0 ._____-L...-___...L--_-.1-.______.L-_._----L..___.___.______..___J~.__..__..__i 0    100    200      300    400        500      600    700    800      900 Iiroe ~)
Amendment 61                                                                                Page 1 of 1
 
Shearon Harris Nuclear Power Plant                            UFSAR Chapter: 15 FIGURE 15.6.5-14 CONTAINMENT AND LOOP PRESSURES FOR LBLOCA (LIMITING CASE)
Containment    ..an
                                          . d Loop Pressures.
100 90 200        400          600  800    1000 Time (s)
Amendment 61                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 FIGURE 15.6.5-15 UPPER PLENUM PRESSURE FOR LBLOCA (LIMITING CASE)
Upper Plenum Pressure 3000 ,---.---,- -- -,. - --            ......, - - - - -,- -----.
2000 1000 0  I.A 0          200      400        600              800    1000 Tine(s)
Amendment 61                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                        UFSAR Chapter: 15 FIGURE 15.6.5-16 DOWNCOMER COLLAPSED LIQUID LEVEL FOR LBLOCA (LIMITING CASE)
Downcomer Liquid Level 25
                                                  -          sector 1 (broken)
                                                  .._..,._.. Sector 2
                                                  - --- Sedar3
                                                  - - - Averaae 20 10 5
0 1.-....__,___- - 1_ _..._.....__--'_~___.____.__------1._~....1,..__._ _,____.__,
0      100    200      300    400        500      600    700      800    900 Time(,)
Amendment 61                                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                UFSAR Chapter: 15 FIGURE 15.6.5-18 CORE COLLAPSED LIQUID LEVEL FOR LBLOCA (LIMITING CASE)
Cote Liquid Level 8                                        -
* Hot Assembly
                                                        - - - - Center Core
                                                        - - - Average Core
                                                        - - - Outer Core 6
e 4
2 0 ~,..._,..._~.._......_...__...__......._......._.......___........_........__.___~ ~ - -
0      100      200      300        400        500    600            700    800  900 Time(s)
Amendment 61                                                                                              Page 1 of 1
 
Shearon Harris Nuclear Power Plant                          UFSAR Chapter: 15 FIGURE 15.6.5-20 PEAK CLADDING TEMPERATURES FOR LBLOCA (LIMITING CASE)
PCT 2000~~-------- - ~ ~ - - - - - - - ~ - ~
1500 .
          -f
                                                    \
          .5
          .c I
:E                                        j 500 I
I L
200      400        600    800  1000 lime (s)
Amendment 61                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                    UFSAR Chapter: 15 FIGURE 15.6.5-30 SYSTEM PRESSURES FOR LIMITING BREAK (2.6 INCH SBLOCA)
System Pressures 2500.0
                                                            - . RV *UpperHad
                                                            -----+SG-1
                                                              ~    SG-2
                                                            -....SG-3
:i J
; 1500.0 J
1000.0  I soo.o .
1000        1500      .2000          2500 Time(*)
Amendment 61                                                                  Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                      UFSAR Chapter: 15 FIGURE 15.6.5-31 BREAK FLOW RATE FOR LIMITING BREAK (2.6 INCH SBLOCA)
Liquld&Vapor Break Flow eoo.o
                                                              ---- Uquld Break Flow
                                                              ---+ V~por Break Flow
}
II 4100.0
:I l
200.0 500          1000        1500      2000          2500          3000 Time(&)
Amendment 61                                                                      Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                          UFSAR Chapter: 15 FIGURE 15.6.5-32 DOWNCOMER AND HOT ASSEMBLY COLLAPSED LIQUID LEVELS FOR LIMITING BREAK (2.6 INCH SBLOCA)
Level
____. DC Ltve1
* A.,_.
20.0                                                      _,.Hcit~Lwel 10.0 500  1000  1500  2000  2.500  3000  3!500 .ooo  4500    eooo  5500  eoco Tnne(s)
Amendment 61                                                                        Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                            UFSAR Chapter: 15 FIGURE 15.6.5-33 HHSI FLOWS FOR LIMITING BREAK (2.6 INCH SBLOCA)
HHSI Mass Flow Rates
                                                                            ---Loop1
                                                                            -1.oop2
                                                                            - -*l..oop3 20.0
_.1.r_..,-.
              .-.....J'-.....~....... .---***-
r l
i 10.0 ,
I I
0.0 0                          1000              1500            2000                              3000 Tme (a),
Amendment 61                                                                                            Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                                                                            UFSAR Chapter: 15 FIGURE 15.6.5-34 COMBINED ACCUMULATOR FLOW FOR LIMITING BREAK (2.6 INCH SBLOCA)
Total .Accumulator Flow
:zoo,.o
.~
I I  100.0 o~-----------------------------------____.
    -100.0 '---'.........__.__...__.__........__.______........,___,_.........__.__--.......
0                                              1000 1500
                                                                                                --- - .......--1..------.. . . . . . . . .__.___.__.,___.
2000          UGO Tlme(1)
Amendment 61                                                                                                                                    Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 FIGURE 15.6.5-35 TOTAL RCS MASS AND REACTOR VESSEL MASS FOR LIMITING BREAK (2.6 INCH SBLOCA)
Total Primary Mass 500000.0  .---------.------.----=~~:__-r---------
400000.0
                                                                    -Res
                                                                    ---+RQl;IQf1/4taol 300000,Q, 200000.0 100000.0 2'000*.0                    4CIOD.O                  eotXI.O Amendment 61                                                                          Page 1 of 1
 
Shearon Harris Nuclear Power Plant                                              UFSAR Chapter: 15 FIGURE 15.6.5-36 HOT ROD CLADDING TEMPERATURE FOR LIMITING BREAK (2.6 INCH SBLOCA)
Peak Cladding Temperature 1800.0  r-- --.--,----.----r---r----.----.--- ..---.......- --.---.--------,
__. Hot Red Heat Sit. node 29 0 10.83 ft
                                                *--+ Hot Rod Heat Str, node 30010.88 ft
            'I                                  --... Hot Rod Heat St,. node 31 011.13 ft 1eoc:to  1                                -4'HotRodHNtStt.node32011.38 ft
                                                ~    fwtRodHe.tStr. node 33011.83 ft 1AOO.O  1
  ,t 1200.0 I  ,m>>.O
      !00.0 600.0 G-.00 1000      2000        3000                                            6000 Time C*>
Amendment 61                                                                              Page 1 of 1}}

Latest revision as of 13:54, 11 December 2024

Amendment 63 to Update Final Safety Analysis Report, Chapter 15, Accident Analysis
ML20147A032
Person / Time
Site: Harris Duke Energy icon.png
Issue date: 05/15/2020
From:
Duke Energy Progress
To:
Office of Nuclear Reactor Regulation
Shared Package
ML20147A016 List:
References
RA-20-0134
Download: ML20147A032 (485)
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