ML14050A375

Summary of Analysis Supporting Pressurizer Safety Valve Technical Change

ML14050A375
Person / Time
Site:	Calvert Cliffs
Issue date:	02/13/2014
From:	Calvert Cliffs, Constellation Energy Nuclear Group, EDF Group
To:	Office of Nuclear Reactor Regulation
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ML14050A444	List: ML14050A374 ML14050A375
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ATTACHMENT (4)

SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE -

NON-PROPRIETARY

1.0 INTRODUCTION

2.0 COMPUTER CODES 3.0 ANALYSIS METHODOLOGY 4.0 EVENT DISPOSITIONS 5.0 EVENT SPECIFIC RESULTS

7.0 REFERENCES

Calvert Cliffs Nuclear Power Plant, LLC February 13,2014

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 1.0 Introduction Calvert Cliffs Nuclear Power Plant (CCNPP) includes two spring-loaded pressurizer safety valves (PSVs) for each Unit, whose main purpose is to provide Reactor Coolant System (RCS) overpressure protection.

Pressurizer safety valve RC-200 is set to open at approximately 2500 psia, and PSV RC-201 is set to open at approximately 2565 psia. With these settings, the combined capacity of the PSVs helps to ensure that the RCS pressure Safety Limit of 2750 psia is not exceeded during design basis accidents.

The primary purpose of this report is to support a PSV setpoint opening tolerance change to accommodate thermal drift associated with the PSV nominal opening settings for CCNPP Units I and 2. The change would increase the as-found tolerance on PSV RC-200 from (-1% / +2%) to (-1% / +3%) and on PSV RC-201 from +/-2% to (-2% / +3%), as well as decrease the nominal opening setpoint for PSV RC-201 from 2565 psia to 2525 psia (see table below). This table supports the proposed Technical Specification (TS) changes related to SR 3.4.10.1.

Nominal As Found As Left PSV Opening Setpoint Lift Setting and Tolerances Lift Setting and Tolerances (psia) (psia) (psia)

> 2475 and <2575 > 2475 and* 2525

-1% +3% -1% +1%

> 2475 and < 2600 > 2500 and < 2550 RC-201 2525 -2% +3% -1% +1%

2.0 Computer Codes The S-RELAP5 (Reference 1) code is an AREVA modification of the RELAP5/MOD2 code. S-RELAP5 is used for simulation of the transient system response to loss-of-coolant accident (LOCA) and non-LOCA events. Control volumes and junctions are defined which describe all major components in the primary and secondary systems that are important for the analyzed events. The S-RELAP5 hydrodynamic model is a two-dimensional, transient, two-fluid model for flow of a two-phase steam-water mixture. S-RELAP5 uses a six-equation model for the hydraulic solutions. These equations include two-phase continuity equations, two-phase momentum equations, and two-phase energy equations. The six-equation model also allows both non-homogeneous and non-equilibrium situations encountered in reactor problems to be modeled.

The RODEX2 code (References 2 and 3) is a fuel rod code which has been approved for use in the performance of LOCA and non-LOCA analyses. The RODEX2 code simulates the thermal and mechanical response of a fuel rod in a coolant channel as a function of exposure for the normal and power ramp conditions encountered in pressurized water reactors. The code incorporates models to describe the thermal-hydraulic condition of the fuel rod in a flow channel; the gas release, swelling, densification, and cracking in the pellet; the gap conductance; the radial thermal conduction; the free volume and gas pressure internal to the fuel rod; the fuel and cladding deformations; and the cladding corrosion. The calculations are performed on a time-incremental basis with conditions being updated at each calculated increment.

3.0 Analysis Methodology The AREVA methodology for evaluating non-LOCA transients is described in Reference 1. The non-LOCA analysis methodology to be applied for the PSV setpoint tolerance change has been previously reviewed and approved by the Nuclear Regulatory Commission. This report includes the required elements for approval of the application of these methodologies to CCNPP.

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY For each non-LOCA transient event analysis, the nodalization, chosen parameters, conservative input, and sensitivity studies are reviewed for applicability to the PSV setpoint tolerance change in compliance with the Safety Evaluation Report for the non-LOCA topical report (Reference 1).

" The nodalization used for the calculations supporting the PSV setpoint tolerance change is specific to CCNPP Units 1 and 2, and is in accordance with the Reference I methodology, with the only area of renodalization being between the pressurizer and the PSVs. Typically, the inlet piping to the PSV is insignificant in length; however, there is a significantly large length of PSV inlet piping at CCNPP, which was specifically modeled.

• The parameters and equipment states are chosen to provide a conservative estimate of the challenge to the acceptance criteria. The biasing and assumptions for key input parameters are analyzed as sensitivity studies.

" The S-RELAP5 code assessments in Reference 1 validated the ability of the code to predict the response of the primary and secondary systems to Chapter 14 non-LOCA transients and accidents. The only additional sensitivity studies required analyzed the process variable biasing; no additional model sensitivity studies are needed for this application.

The Reference 1 methodology requires that the events be run at a I 1; however, in an attempt to maximize the peak RCS and Main Steam System (MSS) pressures, [

] were considered in the analyses. By using [

] is expected to be bounded, which would result in a conservative conclusion. The methodology also recommends biasing the [

]; however, [ ]

were considered in the analyses.

3.1 ParameterBiasing A set of the process variables used in event analyses is biased to assure that the results are conservative.

A process variable is a measured plant condition that is input to plant observation and control systems.

Process variables are also input to the Reactor Protection System (RPS) and to TS Limiting Conditions for Operating (LCOs). These process variables undergo electronic processing to determine the measured plant state, which adds uncertainty to the measured value. In addition, some process variables are used as a target for a control system, and may have a control deadband associated with them.

The biasing of process variables is consistent with the guidelines of the Standard Review Plan (SRP).

When there is a TS limit for a parameter to be biased, the allowed operating range and uncertainty for the power level being considered will be conservatively bounded. These parameters are either process variables that are input to RPS trips, or are LCOs on initial conditions. Other process variables that do not have a TS limit, but may significantly affect the results of the transient calculations, are also biased to bound allowed operating ranges and uncertainties.

The direction in which a conservative bias will be applied is dependent on the event and criterion being analyzed. The process variable biasing for each over-pressurization case is supported by [

] The following process variables were considered for biasing: [

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 3.2 Thermal Conductivity Degradation As discussed in Section 3.2 of Reference 1, the S-RELAP5 non-LOCA model uses user-specified input for fuel thermal properties (thermal conductivity vs. temperature, volumetric heat capacity vs.

temperature) for the core heat structures. The core heat structures are used to determine fuel temperature for Doppler feedback and average core fuel surface heat flux.

For the Control Element Assembly (CEA) Ejection event, the fuel thermal conductivity input for both the heat structures will be based on the following equations (from Reference 4), which conservatively account for degradation of this parameter with exposure. Thermal conductivity degradation is an important parameter for the CEA Ejection event, since it is partially mitigated by Doppler feedback, whereas the other overpressure events do not have a significant increase in fission power and are therefore less affected.

The thermal conductivity relationship for fully dense fuel is:

Where The correction for porosity is a function of temperature as follows:

E ]1 Where E J1 The U0 2 thermal conductivity relationship is adapted to gadolinia fuels using the following modifications. The ratio of the thermal conductivity of gadolinia bearing fuel to U0 2 fuel r(z,T) =

X(UGd)U0 2 / X(U0 2) is:

E J 3

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Where E ]

L

[ ] is given by the following expression:

E ]

I 4.0 Event Dispositions In general, the change in PSV setpoint tolerance will only affect analyses which have a primary pressure excursion large enough that the PSVs open (currently 2500 psia plus 2% uncertainty, or 2550 psia).

There are three SRP /Updated Final Safety Analysis Report (UFSAR) events required to be analyzed for primary and secondary overpressure: Loss of External Electrical Load (SRP 15.2. 1I/UFSAR 14.5), Loss of Normal Feedwater Flow (SRP 15.2.7/UFSAR 14.6), and Feedwater System Pipe Break (SRP 15.2.8/UFSAR 14.26). In addition, CEA Ejection (SRP 15.4.8/IJFSAR 14.13) is analyzed only for peak primary pressure, since the event does not challenge the secondary system pressure limit.

The following sections describe the events listed in the UFSAR and provide the technical basis for whether the proposed PSV tolerance TS change has an effect on the events, and whether reanalysis is required. Table 4-1 provides a summary of the dispositions and events requiring reanalysis. The SRP/UFSAR acceptance criteria are addressed by the use of AREVA methodology for all applicable events.

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Table 4-1: Summary of UFSAR Chapter 14 Event Dispositions SRP Event Event Event Description Condition Disposition Number 15.1.1 14.7 Decrease in Feedwater Temperature 11(AOO) Bounded by AOR 15.1.2 14.7 Increase in Feedwater Flow 11(AOO) Bounded by AOR 14.12 Increase in Feedwater Flow to one Steam Generator (SG) 11(AOO) Bounded by AOR 15.1.3 14.4 Increase in Steam Flow II (AOO) Bounded by AOR 15.1.4 14.12 Inadvertent Opening of a SG Relief or Safety Valve 11(AOO) Bounded by AOR 15.1.5 14.14 Main Steam Line Break IV (PA) Bounded by AOR 15.2.1 14.5 LOEL II (AOO) Requires re-analysis 15.2.2 14.5 Turbine Trip 11 (AOO) Bounded by LOEL (UFSAR 14.5) 15.2.3 14.5 Loss of Condenser Vacuum 11(AOO) Bounded by LOEL (UFSAR 14.5) 15.2.4 14.5 Inadvertent Closure of Main Steam Isolation Valves (MSIVs) II (AOO) Bounded by LOEL (UFSAR 14.5) 14.12 Inadvertent Closure of Main Steam Isolation Valve to one SG II (AOO) Bounded by AOR (Asymmetric Loss of Load) 15.2.5 14.5 Steam Pressure Regulator Malfunction or Failure that Results II (AOO) Bounded by LOEL in Decreasing Steam Flow (UFSAR 14.5) 15.2.6 14.10 Loss of Non-Emergency AC Power to the Station Auxiliaries II (AOO) Bounded by LOEL (UFSAR 14.5) and LOFW (UFSAR 14.6) 15.2.7 14.6 LOFW Flow II (AOO) Requires re-analysis 14.12 LOFW to one SG (Asymmetric Loss of Feedwater) II (AOO) Bounded by LOEL (UFSAR 14.5) Relative to RCS Overpressure 15.2.8 14.26 Feedwater System Pipe Break IV (PA) Requires re-analysis 15.3.1 14.9 Loss of Forced Reactor Coolant Flow 1I(AOO) Bounded by AOR 15.3.3 14.16 Reactor Coolant Pump (RCP) Shaft Seizure IV (PA) Bounded by AOR 15.3.4 N/A RCP Shaft Break IV (PA) Not in Licensing Basis 15.4.1 14.2 Uncontrolled RCCA Bank Withdrawal from Subcritical or II (AOO) Bounded by LOEL Low Power Startup (UFSAR 14.5) Relative to RCS Overpressure 15.4.2 14.2 Uncontrolled Control Rod Assembly Bank Withdrawal at II (AOO) Bounded by LOEL Power (UFSAR 14.5) Relative to RCS Overpressure 15.4.3.1 14.2 RCCA Misoperation - Withdrawal of a Single Full-Length III (PA) Not a credible event RCCA 15.4.3.2 14.2 RCCA Misoperation - Static Misalignment of a Single RCCA II (AOO) Not a credible event 15.4.3.3 14.11 RCCA Misoperation - Dropped RCCA and RCCA Bank II (AOO) Bounded by AOR 15.4.4 14.1 Startup of an Inactive Reactor Coolant Loop at an Incorrect II (AOO) Not allowed by TS Temperature 15.4.5 N/A Recirculation Loop at Incorrect Temperature or Flow Not in Licensing Basis Controller Malfunction 15.4.6 14.3 CVCS Malfunction that Results in a Decrease in Boron II (AOO) Bounded by AOR Concentration in the Reactor Coolant 5

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Table 4-1: Summary of UFSAR Chapter 14 Event Dispositions SRP SP UFSAR Event Event Event Description Condition Disposition Number 15.4.7 N/A Inadvertent Loading of a Fuel Assembly into an Improper III (PA) Not in Licensing Basis Location 15.4.8 14.13 Spectrum of RCCA Ejection Accidents IV (PA) Requires re-analysis 15.4.9 N/A Spectrum of Rod Drop Accidents Not applicable to CCNPP 15.5.1 N/A Inadvertent Operation of ECCS II (AOO) Not in Licensing Basis 15.5.2 14.25 CVCS Malfunction that Increases Reactor Coolant Inventory II (AOO) Bounded by AOR 15.6.1 14.8 Inadvertent Opening of Pressurizer Safety or Power Operated 11 (AOO) Bounded by AOR Relief Valve (PORV) 15.6.2 N/A Radiological Consequences of the Failure of Small Lines III (PA) Not in Licensing Basis Carrying Primary Coolant Outside of Containment 15.6.3 14.15 Steam Generator Tube Rupture (SGTR) IV (PA) Bounded by AOR 15.6.4 N/A Spectrum of Boiling Water Reactor Steam Piping Failures Not applicable to CCNPP Outside Containment 15.6.5 14.17.3 Small Break Loss-of-Coolant Accidents IV (PA) Bounded by AOR 14.17.2 Loss-of-Coolant Accidents IV (PA) Bounded by AOR 15.7.1 14.22 Radioactive Waste Gas System Leak or Failure Bounded by AOR 15.7.2 14.23 Liquid Waste System Leak or Failure Bounded by AOR 15.7.3 14.23 Postulated Radioactivity Releases Due to Liquid Tank Failure Ill (PA) Bounded by AOR 15.7.4 14.18 Design Basis Fuel Handling Accidents IV (PA) Bounded by AOR 15.7.5 5.6.1.5 Spent Fuel Cask Drop Accidents III (PA) Bounded by AOR 15.7.6 N/A Spent Fuel Pit Water Loss Not in Licensing Basis 14.19 Turbine Generator Overspeed Incident N/A Bounded by AOR 14.20 Containment Response N/A Bounded by AOR 14.24 Maximum Hypothetical Accident N/A Bounded by AOR AOO Anticipated Operational Occurrence LOFW Loss of Feedwater AOR Analysis of Record PA Postulated Accident LOEL Loss of External Load 4.1 Increase in Heat Removal by the Secondary System (SRP 15.1) 4.1.1 Excess Feedwater Heat Removal (SRP 15.1.1, 15.1.2 / UFSAR 14.7, 14.12.2.1)

The Excess Feedwater Heat Removal event is defined as a reduction in SG feedwater temperature without a corresponding reduction in steam flow from the SGs. This could be caused by the loss of one or more of the feedwater heaters, or due to a feedwater controller malfunction at steady-state that causes an increase in feedwater flow.

The most limiting Excess Feedwater Heat Removal event is postulated to occur at Hot Full Power (HFP) and is caused by the assumed loss of both high pressure feedwater heaters, resulting in a decrease in feedwater temperature to the SGs. The cooler water entering the SGs causes the SG temperature and pressure to slowly decrease, and more heat is extracted from the RCS. This results in a reduction in RCS temperature and pressure and a reduction in pressurizer level. With a negative Moderator Temperature Coefficient (MTC), the core power increases and reaches a new steady-state with no reactor trip. The RCS pressure reaches a new steady value, which is below the initial RCS pressure and well below the 6

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY PSV setpoints. Thus, there will be no challenge to opening the PSVs. The plant response for the Excess Feedwater Heat Removal event initiated by an increase in feedwater flow to 155% of rated full power flow is similar to the event initiated by the loss of both high pressure feedwater heaters, however, the UFSAR indicates that the resultant increase in power is less for the event initiated by an increase in feedwater flow. The resultant RCS pressure will be similar to the event initiated by the loss of both high pressure feedwater heaters and will be well below the PSV setpoints. Thus, there will be no challenge to opening the PSVs.

The asymmetric Excess Feedwater Flow event is defined in UFSAR Section 14.12.2.1. The asymmetric Excess Feedwater Flow event is initiated at HFP by a malfunction in one of the feedwater controllers, which instantaneously fully opens the feedwater regulator valve to one SG. The full opening of the feedwater regulator valve causes additional subcooled feedwater to enter the SG which lowers the SG temperature and pressure and cold leg temperature. This results in a reduction in steam flow from the affected SG. The unaffected SG picks up part of the load which decreases the cold leg temperature.

There is a small increase in core power and the event is terminated by the Asymmetric Steam Generator Pressure Trip.

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.1.2 Increase in Steam Flow (SRP 15.1.3 / UFSAR 14.4)

An Excess Load event is defined as any rapid uncontrolled increase in SG steam flow other than a steam line break. The most limiting Excess Load event at HFP is an inadvertent opening of the atmospheric dump and turbine bypass valves. The limiting Excess Load event at zero power is postulated to be initiated by the full opening of the turbine control valves. The immediate response to the increased steam flow demand is a rapid decrease in SG pressure and temperature. The RCS temperatures and pressure consequently decrease. With a negative MTC for the HFP case, there will be an increase in power which will slow down the decrease in RCS temperature and pressure. However, the RCS pressure during the event will be significantly below the initial RCS pressure. Therefore, there is no challenge to opening the PSVs.

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.1.3 Inadvertent Opening of a SG Relief or Safety Valve (SRP 15.1.4 / UFSAR 14.12.2.3)

This event is defined in UFSAR Section 14.12.2.3 as an asymmetric Excess Load event. The asymmetric Excess Load event is initiated at HFP by the inadvertent opening of a single secondary safety valve on one SG. The excess load on a single SG causes its pressure and temperature to decrease, which results in a decrease in core inlet temperature. With a negative MTC, an increase in core power occurs and a new steady state condition will be reached to match the excess load demand. There will be no significant increase in RCS pressure and no challenge to opening the PSVs.

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 4.1.4 Main Steam Line Break (SRP 15.1.5 / UFSAR 14.14)

The Main Steam Line Break event is characterized by a rupture in the steam system which increases the rate of heat extraction by the SGs and causes a decrease in RCS temperature and pressure. Since the RCS pressure decreases for this event, there is no challenge to opening the PSVs.

Since there is no increase in RCS pressure, there is no challenge to the reactor coolant pressure boundary.

The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.2 Decreasein Heat Removal by the Secondary System (SRP 15.2) 4.2.1 Loss of External Electrical Load (SRP 15.2.1 / UFSAR 14.5 and 14.12.2.4)

An LOEL event is defined as any event that results in a reduction in the SGs heat removal capacity through the loss of secondary steam flow. Closure of all MSIVs, turbine stop valves, or turbine control valves will cause a Loss of Load event. An Asymmetric Loss of Load, caused by the closure of a single MSIV, presents a less severe challenge to the RCS pressure boundary since the loss of heat removal from one SG is not as significant as the loss of heat removal from both SGs. Of the three types of valves, the turbine stop valves have the quickest closure time. The UFSAR identifies the most limiting Loss of Load event as a turbine trip without a concurrent reactor trip or an inadvertent closure of the turbine stop valves at HFP.

An immediate termination of steam flow to the turbine results in a rapid increase in SG pressure and temperature with RCS adding heat and without any steam being extracted from the SG. With no credit for the steam dump and bypass system the SG pressure rapidly increases to the Main Steam Safety Valve (MSSV) setpoints. This causes a rapid increase in the RCS temperature. The increase in RCS temperature causes the pressurizer level and pressure to increase. With no credit for PORVs and sprays, the pressurizer pressure will increase to the PSV setpoints and the PSVs will open. An increase in the PSV tolerance by 1% will adversely affect the peak RCS pressure reached for this event.

Since there is an increase in RCS pressure, there may be a challenge to the reactor coolant pressure boundary. The increase in PSV tolerance may impact the results of this event; therefore, this event will require re-analysis for the RCS and MSS overpressure cases. The LOEL event is a Condition II (AOO) event, and as such it is verified that an incident of moderate frequency does not generate a more serious plant condition without other faults occurring independently.

4.2.2 Turbine Trip (SRP 15.2.2 / UFSAR 14.5)

Based on the discussion in UFSAR Section 14.5.1, this event is bounded by the LOEL event analyzed as a Turbine Trip event.

4.2.3 Loss of Condenser Vacuum (SRP 15.2.3 / UFSAR 14.5)

Based on the discussion in UFSAR Section 14.5.1, this event is bounded by the LOEL event analyzed as a Turbine Trip event.

4.2.4 Inadvertent Closure of Main Steam Isolation Valves (SRP 15.2.4 / UFSAR 14.5)

Based on the discussion in UFSAR Section 14.5.1, this event is expected to be bounded by the LOEL event analyzed as a Turbine Trip event. However, the closure of all MSIVs is considered in the reanalysis of the LOEL event.

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 4.2.5 Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow (SRP 15.2.5 / UFSAR 14.5)

Rapid closure of the turbine stop valves assumed in the LOEL event (UFSAR 14.5) produces a more severe transient response than the slower closing turbine control valves. Consequently, the LOEL event, analyzed as a Turbine Trip event (UFSAR 14.5) bounds this event.

4.2.6 Loss of Non-Emergency AC Power to the Station Auxiliaries (SRP 15.2.6 / UFSAR 14.10)

The UFSAR defines the most limiting Loss of AC Power event as a loss of turbine load at HFP with offsite AC power unavailable. The immediate system response is similar to a simultaneous Loss of Load event, LOFW event, and Loss of Coolant Flow event. In the short term, the SG pressure will rapidly increase to the MSSV setpoints and the RCS temperature and pressure will increase. However, a reactor trip will occur within about 1 sec on low RCS flow. Once the control rods are inserted and the core power is at decay heat levels, the increase in RCS pressure is mitigated. Because the LOEL event creates a significant SG heat removal/power generation mismatch, the LOEL event bounds the Loss of AC Power event for peak RCS and MSS pressures.

Due to natural circulation and the associated higher long-term average RCS temperature for the Loss of AC Power event relative to the LOFW event, the Loss of AC Power event will result in a slightly higher pressurizer level than for the LOFW event. However, the LOFW event does not significantly challenge pressurizer overfill, and the reactor trips much later for a LOFW event than for a Loss of AC Power event; therefore, the Loss of AC Power event will not challenge pressurizer overfill. The Loss of AC Power event is bounded by the LOFW event regarding long-term decay heat removal (minimum SG inventory) because the LOFW event has continued RCP operation, which adds heat to the RCS.

The Loss of AC Power event is bounded by the LOEL event (UFSAR 14.5), and does not require re-analysis for RCS or MSS overpressure.

4.2.7 Loss of Feedwater Flow (SRP 15.2.7 / UFSAR 14.6, 14.12)

The LOFW event is defined as a reduction in feedwater flow to the SGs without a corresponding reduction in steam flow from the SGs. The most limiting LOFW Flow event at HFP is an inadvertent and instantaneous closure of both feedwater regulating valves. This causes the largest steam flow and feedwater flow mismatch and results in the most rapid reduction in SG inventory. This event is characterized by an increase in SG pressure and temperature until the MSSV setpoints are reached.

Correspondingly, the RCS temperature and pressure increases until a reactor trip on either the High Pressurizer Pressure (HPP) trip or SG low level trip occurs. The reactor trip rapidly decreases the core heat flux to decay heat levels. This terminates the initial rise in RCS temperature and pressure and SG pressure. Shortly after reactor trip, the MSSVs open followed by opening of the PSVs. However, the RCS pressure overshoot beyond the PSV setpoints is limited since reactor trip has already occurred.

The RCS pressure overshoot beyond the PSV setpoints in the LOFW event is bounded by the RCS pressure overshoot in the LOEL event (UFSAR 14.5). For the LOEL event there is an immediate isolation of the secondary system causing a significant power generation/SG heat removal mismatch prior to reactor trip. However, for the LOFW event there is no significant power generation/SG heat removal mismatch prior to reactor trip. Thus, the LOEL event produces a higher rate of increase in RCS temperature and pressure and a larger RCS pressure overshoot beyond the PSV setpoints than does the LOFW event.

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY The increase in PSV tolerance will not affect the secondary overpressure case since parameters are biased such that the PSVs do not open for this case. Likewise, the PSVs do not open for the minimum SG inventory case.

An asymmetric LOFW Flow event is defined in UFSAR Section 14.12.2.2. The asymmetric LOFW Flow event is initiated at HFP by a malfunction in one of the feedwater controllers which instantaneously shuts the feedwater regulator valve to one SG. The loss of feedwater will cause the temperature and pressure in the affected SG to increase in response to the decreasing SG level. The temperature and pressure in the unaffected SG also increases in response to the increase in turbine header pressure. There will be a slight asymmetry in core inlet temperature. There will be no significant change in core power, no significant increase in RCS pressure, and no challenge to opening of the PSVs.

Since there is an increase in RCS pressure, there may be a challenge to the reactor coolant pressure boundary. The increase in PSV tolerance may impact the results of this event; therefore, this event is reanalyzed, although it is not expected to be limiting. The LOFW Flow event is analyzed to ensure the RCS and MSS pressures remain below the acceptable limit. An analysis is also performed to verify the action of the RPS, auxiliary feedwater actuation signal (AFAS), and auxiliary feedwater (AFW) systems are adequate to ensure that the SGs provide for long term heat removal and to ensure an incident of moderate frequency does not generate a more serious plant condition without other faults occurring independently.

4.2.8 Feedwater System Pipe Break (SRP 15.2.8 / UFSAR 14.26)

The Feedwater Line Break (FLB) event is defined as a break in the MFW system piping. Depending on the break size, the event may be a cooldown or a heatup event. The FLB event is analyzed to demonstrate that the RCS pressure limit is not exceeded. The limiting break size for RCS pressurization causes a cessation of feedwater flow to both SGs and rapid inventory loss in the affected SG, resulting in lower SG liquid inventory in the SG boiler region which rapidly reduces the heat removal capacity. The reduced heat transfer capability results in a rapid increase in RCS temperature and pressure. When analyzed crediting only the HPP trip, the event will challenge the RCS pressure limit. The RCS heatup will continue after a HPP reactor trip due to a total loss of heat transfer in the affected SG or low SG pressure due to the depressurization from the break. The rapid rise in RCS pressure causes the PSVs to open and significant pressure overshoot beyond the PSV setpoints will occur before the core heat flux is significantly reduced by the scram, which mitigates the increase in pressure. The increase in the PSV tolerance will adversely affect the peak RCS pressure for this event.

Since there is an increase in RCS pressure, there may be a challenge to the reactor coolant pressure boundary. The increase in PSV tolerance may impact the results of this event; therefore, this event will require re-analysis for RCS overpressure. The FLB is analyzed to verify that the automatic initiation of AFW in combination with operator action at 10 minutes to increase AFW flow is sufficient to provide a continued heat sink for the removal of decay heat.

4.3 Decreasein RCS Flow Rate (SRP15.3) 4.3.1 Loss of Forced Reactor Coolant Flow (SRP 15.3.1 / UFSAR 14.9)

The most limiting Loss-of-Coolant Flow event is a concurrent loss of power to all four RCPs. The immediate system response to the coastdown of the pumps is a rapid decrease in coolant mass flow rate through the core followed by a reactor trip on low flow. The reactor trips on low RCS flow within 1 sec of event initiation. There is no significant increase in RCS temperature and pressure for this event.

Therefore, there is no challenge to opening of the PSVs.

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.3.2 Reactor Coolant Pump Shaft Seizure (SRP 15.3.3 / UFSAR 14.16)

The most limiting seized rotor event is an instantaneous RCP shaft seizure at HFP. The immediate system response is a rapid reduction to three pump flow. Due to the reduction in core flow, the core coolant temperatures up the core increase. Assuming a positive MTC, there is an increase in core power.

An almost immediate reactor trip on low flow occurs, which mitigates the increase in power. There is a mild increase in RCS pressure, but not significant enough to challenge opening of the PSVs.

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.3.3 Reactor Coolant Pump Shaft Break (SRP 15.3.4 / UFSAR N/A)

This event is not part of the licensing basis of the plant.

4.4 Reactivity and PowerDistributionAnomalies (SRP 15.4) 4.4.1 Uncontrolled RCCA Bank Withdrawal from Subcritical or Low Power Startup (SRP 15.4.1 /

UFSAR 14.2)

The Uncontrolled CEA Bank Withdrawal event from subcritical or low power is initiated at Hot Zero Power (HZP) critical conditions. The CEA bank withdrawal causes the power to increase at an exponential rate. The event initiated from subcritical conditions is less severe due to the high rate-of-change of power (startup rate) trip. The event initiated from HZP critical conditions experiences a Variable High Power (VHP) reactor trip. The increase in core power causes an increase in RCS temperature and pressure. This event has the potential for opening the PSVs for an RCS overpressure case. The limited time at power prior to reactor trip, the reduced decay heat load after trip, and the lower initial RCS temperature combine to result in a smaller heat removal requirement. A reactor trip on VHP trip or HPP, along with PSV capacity, is sufficient to maintain peak RCS pressure well below the overpressure limit.

4.4.2 Uncontrolled Control Rod Assembly Bank Withdrawal at Power (SRP 15.4.2 / UFSAR 14.2)

The Uncontrolled CEA Bank Withdrawal at Power event is initiated at HFP conditions. The CEA bank withdrawal causes the core power to increase at an increasing rate depending on the worth of the CEAs.

The event will trip on the VHP trip, the HPP trip, or the Thermal Margin/Low Pressure trip. The increase in core power causes an increase in RCS temperature and pressure. This event has the potential for opening the PSVs for an RCS overpressure case. The CEA Bank Withdrawal at Power event is not limiting with respect to peak RCS pressure. With no loss of secondary load or feedwater and no loss of offsite power, a reactor trip on VHP trip or HPP, along with PSV capacity, is sufficient to maintain peak RCS pressure well below the overpressure limit.

The RCS pressure overshoot beyond the PSV setpoints in the CEA Bank Withdrawal at Power event is bounded by the RCS pressure overshoot in the LOEL event (UFSAR 14.5). For the LOEL event there is an immediate isolation of the secondary system causing a significant power generation/SG heat removal mismatch prior to reactor trip. However, for the CEA Bank Withdrawal at Power event there is no significant power generation/SG heat removal mismatch prior to reactor trip. Thus, the LOEL event 11

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY produces a higher rate of increase in RCS temperature and pressure and a larger RCS pressure overshoot beyond the PSV setpoints than does the CEA Bank Withdrawal at Power event.

Since the CEA Bank Withdrawal at Power event is bounded by the Loss of Load event (UFSAR 14.5) for the RCS overpressure case, the CEA Bank Withdrawal at Power event does not require re-analysis.

4.4.3 Rod Cluster Control Assembly (RCCA) Misoperation (SRP 15.4.3 / UFSAR 14.2, 14.11) 4.4.3.1 Withdrawal of a Single Full-Length RCCA (UFSAR 14.2)

The CEA motion inhibit prevents single CEAs from being misaligned, as stated in UFSAR Section 14.2.

Therefore, analysis of this event is unnecessary.

4.4.3.2 Static Misalignment of a Single RCCA (UFSAR 14.2)

The CEA motion inhibit prevents single CEAs from being misaligned, as stated in UFSAR Section 14.2.

Therefore, analysis of this event is unnecessary 4.4.3.3 Dropped RCCA and RCCA Bank (UFSAR 14.11)

The CEA Drop event is defined as an uncontrolled insertion of a CEA. The most limiting CEA Drop event is an uncontrolled CEA insertion at HFP. This event is characterized by a decrease in core power and a corresponding decrease in RCS temperatures and pressure. Therefore, this event does not challenge opening of the PSVs.

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.4.4 Startup of an Inactive Reactor Coolant Loop at an Incorrect Temperature (SRP 15.4.4 /

UFSAR 14.1)

The TSs do not permit part loop operation during power operation. Therefore, this event does not challenge the reactor coolant pressure boundary and is not affected by the increased PSV tolerance.

4.4.5 Recirculation Loop at Incorrect Temperature or Flow Controller Malfunction (SRP 15.4.5 /

UFSAR N/A)

This event is not part of the licensing basis of the plant.

4.4.6 CVCS Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant (SRP 15.4.6 / UFSAR 14.3)

System calculations are not performed for this event. There is no significant change in RCS pressure for this event. Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The increase in PSV tolerance does not affect the results of this event.

4.4.7 Inadvertent Loading of a Fuel Assembly into an Improper Location (SRP 15.4.7 / UFSAR N/A)

This event is not part of the licensing basis of the plant.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 4.4.8 Spectrum of RCCA Ejection Accidents (SRP 15.4.8 /UFSAR 14.13)

The CEA Ejection event is defined as a rapid, uncontrolled, total withdrawal of a single or dual CEA.

The most limiting CEA ejection is a rapid total withdrawal of the highest worth CEA. The system response includes an increase in core power and an increase in RCS temperature and pressure. Since there is an increase in RCS pressure, there may be a challenge to the reactor coolant pressure boundary.

The increase in PSV tolerance may impact the results of this event; therefore, this event will require reanalysis for the RCS overpressure case.

4.4.9 Spectrum of Rod Drop Accidents (SRP 15.4.9 / UFSAR N/A)

This event is not applicable to pressurized water reactors.

4.5 Increases in RCS Inventory (SRP 15.5) 4.5.1 Inadvertent Operation of ECCS (SRP 15.5.1 / UFSAR N/A)

This event is not a part of the licensing basis.

4.5.2 CVCS Malfunction that Increases Reactor Coolant Inventory (SRP 15.5.2 / UFSAR 14.25)

The Excessive Charging event is assumed to occur by inadvertent initiation of charging flow. The event is analyzed to assure that the operator has at least 15 minutes from initiation of a high pressurizer level alarm to take corrective action and terminate the event prior to filling the pressurizer solid. The AOR described in UFSAR Section 14.25.2 is strictly an inventory hand calculation that does not present a challenge to the reactor coolant pressure boundary. The increase in PSV tolerance does not affect the results of this event.

4.6 Decreasesin RCS Inventory (SRP 15.6) 4.6.1 Inadvertent Opening of Pressurizer Safety or Power Operated Relief Valve (SRP 15.6.1 /

UFSAR 14.8)

An RCS Depressurization event is defined as a rapid, uncontrolled decrease in RCS pressure other than a loss of coolant. The most limiting RCS depressurization at HFP is an inadvertent opening of both PORVs. This event is characterized by a depressurization of the RCS. Therefore, there is no challenge to opening the PSVs.

Since there is no increase in RCS pressure, there is no challenge to the PSVs. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.6.2 Small Break Loss-of-Coolant-Accident (SRP 15.6.5 / UFSAR 14.17.3)

A Small Break LOCA event is initiated by a break in the RCS pressure boundary. This event is characterized by a rapid decrease in RCS pressure. Therefore, there is no challenge to opening the PSVs.

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 4.6.3 Steam Generator Tube Rupture (SRP 15.6.3 / UFSAR 14.15)

The SGTR event is initiated by a penetration of the barrier between the RCS and MSS. This event is characterized by a decrease in RCS pressure. Therefore, there is no challenge to opening the PSVs.

Since there is no increase in RCS pressure, there is no challenge to the PSVs. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.6.4 Spectrum of Boiling Water Reactor Steam Piping Failures Outside Containment (SRP 15.6.4 /

UFSAR N/A)

This event is not applicable to pressurized water reactors.

4.6.5 Loss-of-Coolant-Accident (SRP 15.6.5 / UFSAR 14.17.2)

A large break LOCA event is initiated by a break in the RCS pressure boundary. This event is characterized by a rapid decrease in RCS pressure. Therefore, there is no challenge to opening the PSVs.

Since there is no increase in RCS pressure, there is no challenge to the PSVs. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.7 RadioactiveReleasefrom a Subsystem or Component (SRP 15.7) 4.7.1 Radioactive Waste Gas System Leak or Failure (SRP 15.7.1 / UFSAR 14.22)

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.7.2 Postulated Radioactivity Releases Due to Liquid Tank Failure (SRP 15.7.3 / UFSAR 14.23)

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.7.3 Design Basis Fuel Handling Accidents (SRP 15.7.4 / UFSAR 14.18)

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.7.4 Spent Fuel Cask Drop Accidents (SRP 15.7.5 / UFSAR 5.6.1.5)

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 4.8 Other UFSAR Chapter 14 Analyses 4.8.1 Turbine-Generator Overspeed Incident (UFSAR 14.19 / UFSAR 5.3.1.2)

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.8.2 Containment Response (UFSAR 14.20)

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

4.8.3 Maximum Hypothetical Accident (UFSAR 14.24)

Since there is no significant increase in RCS pressure (i.e., no degradation in margin to PSV lift setpoints), there is no challenge to the reactor coolant pressure boundary. The PSV setpoint change and increase in PSV tolerance does not affect the results of this event.

5.0 Event-Specific Results 5.1 Loss of ExternalElectricalLoad (SRP 15.2.1 / UFSAR 14.5) 5.1.1 Event Description An LOEL event is defined as any event that results in a reduction in the SGs heat removal capacity through the loss of secondary steam flow. Closure of all MSIVs, turbine stop valves, or turbine control valves will cause a LOEL event. Of the three types of valves in the steam lines between the SG and the high pressure turbine, closure of the MSIVs leads to the greatest secondary over-pressurization event.

The most limiting LOEL event with respect to peak RCS pressure is a turbine trip without a concurrent reactor trip or an inadvertent closure of the turbine stop valves at HFP. A turbine trip would result in the closure of the turbine stop valves.

With the inability of the SG to remove the heat from the RCS, the RCS temperature will rapidly begin to increase. The pressurizer pressure and level will increase with the increasing RCS temperature. To maximize RCS pressure, no credit is taken for the pressurizer pressure and level control system.

Consequently, the pressure will rapidly approach the PORV setpoint and the HPP trip setpoint.

To maximize secondary peak pressure, credit is taken for the pressurizer pressure control system. This will delay the HPP trip and thus add more energy to the secondary system.

The overpressure portion of the analysis of the event is over once the reactor has been shut down by the HPP reactor trip and RCS and MSS pressures have peaked.

5.1.2 Input Parameters The key input parameters used in the analysis of the LOEL are consistent with or conservative relative to the approved methodology (Reference 1). The differences between the new analysis and the AOR are presented in Table 5-1 and Table 5-2 for the RCS and MSS overpressure cases, respectively. Note that only the limiting analysis values are presented in these tables.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY

" Initial Conditions - The LOEL to both SGs event was analyzed at HFP, since the resultant loss of steam load was the greatest and presented the most significant challenge to the safety valve performance for primary and secondary system pressure.

For the RCS overpressure case, the biasing is as follows: [

For the MSS overpressure case, the biasing is as follows: [

• Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. Both RCS and MSS overpressure events used kinetics parameters that are limiting for the time-in-life as chosen, that were biased to maximize the increase in reactor power during the transient.

" RPS Trips and Delays - The LOEL event was analyzed without a direct reactor trip following turbine trip. This assumption conservatively delayed reactor trip until conditions in the RCS resulted in a HPP trip. The RPS trip setpoint and response time were conservatively biased to delay the actuation of the trip function. In addition, control rod insertion was delayed to account for the CEA holding coil delay time.

• Pressurizer Pressure Control - The availability of the pressurizer pressure control system was treated differently for primary and secondary overpressure. To maximize primary pressure, the pressurizer pressure control system was not available; to maximize secondary pressure, the control system was modeled to delay the reactor trip.

• Primary and Secondary Safety Valves - The PSVs and MSSVs were conservatively modeled.

The PSV setpoints were set to the maximum value allowed by the proposed TS change, [

] Each bank of MSSVs was set to I

• Gap Conductance - Although this event is not sensitive to fuel rod gap conductance and fuel thermal properties, an average core gap conductance was used with BOC fuel thermal properties.

" Steam Generator Tube Plugging - Both minimum and maximum SG tube plugging were considered for the LOEL event. To maximize primary pressure, a high level of SG tube plugging was used, whereas no SG tube plugging was used to maximize secondary pressure.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Table 5-1: Initial Conditions/Input Parameters for the LOEL Event - Maximum RCS Pressure Parameter Units AOR New Analysis Initial Core Power MWt 2754 2754 Initial Core Inlet Temperature OF 546 546 Vessel Flow Rate gpm 422,250 412,000 Initial Pressurizer Pressure psia 2164 2164 Initial Pressurizer Level'  % span 68.8 67.2 MTC pcm/*F +1.5 +1.5 Number of Plugged SG Tubes  % per SG 0 10 CEA Worth at Trip pcm -5000 -5000 Time to 90% Insertion of SCRAM Rods sec 3.1 3.1 Reactor Regulating System Operating Mode Manual Manual Steam Dump and Bypass System Operating Mode Inoperative Inoperative MSSV Opening Pressure psia 1020 1029.25 Pressurizer Pressure Control System Operating Mode Manual Manual Pressurizer Level Control System Operating Mode Manual Manual Turbine Stop Valve Stroke Time sec 0.15-2 0 Reactor Trip and Setpoint Various HPP @ 2420 psia HPP @ 2420 psia Table 5-2: Initial Conditions/Input Parameters for the LOEL Event - Maximum MSS Pressure Parameter Units AOR New Analysis Initial Core Power MWt 2754 2754 Initial Core Inlet Temperature OF 550 550 Vessel Flow Rate gpm 370,000 370,000 Initial Pressurizer Pressure psia 2164 2164 Initial Pressurizer Level'  % span 33.0 32.2 MTC Pcm/°F +1.5 +1.5 Number of Plugged SG Tubes  % per SG 0 0 CEA Worth at Trip pcm -5000 -5000 Time to 90% Insertion of SCRAM Rods sec 3.1 3.1 Reactor Regulating System Operating Mode Manual Manual Steam Dump and Bypass System Operating Mode Inoperative Inoperative MSSV Opening Pressure psia 1020 1029.25 Pressurizer Pressure Control System Operating Mode Automatic Automatic Pressurizer Level Control System Operating Mode Automatic Manual Turbine Stop Valve Stroke Time sec 0.15-2 0.05-2 Reactor Trip and Setpoint Various HPP @ 2420 psia HPP @ 2420 psia Note that the initial pressurizer level for the AOR is calculated by interpolating the span of 116 to 242 in (429 to 952 ft3) using the initial pressurizer volume from the AOR and noting that the pressurizer level span is 360 inches total.

The following list provides justification for the differences in the current AOR and the new analysis.

RCS Over-Pressure

1. RCS Flow Rate - The current AOR uses the core lift RCS flow rate; however, the highest flow rate at steady-state operation would be 412,000 gpm. This value accounts for cycle-to-cycle variation.

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2. SG Tube Plugging - The new analysis assumes a high degree of SG tube plugging, to further limit heat removal from the primary system. This is a conservative assumption.

3. MSSV Opening Pressure - The new analysis delays the opening of the MSSV to account for at least a 3% tolerance above the nominal lift setpoint. The new analysis' MSSV model and setpoints are as follows:

Bank 1:1029.25 psia Bank 2:1049.7 psia Bank 3:1064.7 psia

- Bank I

- Banks 2 and 3 100 90 80 70 60 50 40 ~--p

-4#-- -- 4-4 --

o-r4 4I- 4--I -- +/- I--4+-

30 20 10 0

92 94 96 98 100 102 104 106 108

% of Maximum TS Pressure

4. Turbine Stop Valve Stroke Time - A faster turbine stop valve stroke time is expected to produce slightly higher peak RCS pressures.

5. MSSV Inlet Piping - The new analysis models the MSSV inlet piping explicitly, whereas the AOR accounted for the pressure drop caused by the inlet piping by adding an additional bias to the accumulation on each set of valves.

MSS Over-Pressure

1. MSSV Opening Pressure - The new analysis delays the opening of the MSSV to account for at least a 3% tolerance above the nominal lift setpoint. The new analysis' MSSV setpoints are as follows:

Bank 1: 1029.25 psia Bank 2: 1049.7 psia Bank 3: 1064.7 psia 18

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY

2. Turbine Stop Valve Stroke Time - The new analysis considers a wider range of stroke times than the AOR, which ensures conservatism.

3. MSSV Inlet Piping - The new analysis models the MSSV inlet piping explicitly, whereas the AOR accounted for the pressure drop caused by the inlet piping by adding an additional bias to the accumulation on each set of valves.

4. Pressurizer Level Control System - No credit is taken for the makeup or letdown system. If this system were in automatic mode, it would act to lower the peak pressurizer level, which would lessen the RCS peak pressure.

5.1.3 Acceptance Criteria This event is classified as an AOO. The principally challenged criteria are listed below.

1. The pressures in the reactor coolant and MSSs should be less than 110% of design values.

2. Fuel cladding integrity should be maintained by ensuring Specified Acceptable Fuel Design Limits (SAFDLs) are not exceeded. This is demonstrated by assuring that the minimum calculated Departure from Nucleate Boiling (DNB) ratio is not less than the 95/95 DNB correlation limit.

3. An incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently.

4. Radiological consequences do not exceed 10 Code of Federal Regulations (CFR) 50.67 guidelines.

The proposed change in PSV setpoint tolerance does not adversely affect the margin to the SAFDLs; therefore, the LOEL event was not analyzed for DNB or fuel centerline melt (FCM).

The radiological consequences of opening the MSSVs to mitigate this event would produce negligible site boundary doses compared to the 10 CFR 50.67 guidelines; as such, no mass and energy data for use in radiological consequences is required.

Therefore, the acceptance criteria are met if the peak reactor coolant and MSS pressures remain below 110% of their design values and the pressurizer does not become water-solid.

5.1.4 Description of Analyses and Evaluations The LOEL event was performed with the approved non-LOCA methodology (Reference 1), as discussed in Section 3.0. The S-RELAP5 code was used to model the key system components and calculate core response and fluid conditions (flow rates, temperatures, and pressures). The peak system pressures include appropriate elevation corrections.

There are two main cases run for this event: one biased for peak primary pressure, and one biased for peak secondary pressure. [

J Only the results of the limiting cases are presented.

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SUMMARY

OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY In addition to the cases analyzed at HFP, cases were analyzed to verify the applicability of TS Table 3.7.1-1 power levels for MSSVs inoperable.

A pressurizer overfill analysis is also performed, biased appropriately, to determine the peak pressurizer level following a LOEL.

5.1.5 Results Primary Side Pressure The sequence of events for the limiting primary side overpressure LOEL event is given in Table 5-3. The peak RCS pressure in this case is 2706.6 psia, which is below the overpressure limit of 110% of design pressure (i.e., 2750 psia).

The transient response is shown in Figures 5-1 through 5-6. These Figures show reactor power, core average heat flux, pressurizer level, RCS temperatures, maximum SG pressures, and maximum RCS pressure, respectively.

Secondary Side Pressure The sequence of events for the limiting secondary side overpressure LOEL event is given in Table 5-4.

The peak MSS pressure in this case is 1101.8 psia, which is below the overpressure limit of 110% of design pressure (i.e., 1116.5 psia).

The transient response is shown in Figures 5-7 through 5-12. These Figures show reactor power, core average heat flux, pressurizer pressure, pressurizer level, RCS temperatures, and maximum SG pressures, respectively.

5.1.6 Conclusions The LOEL analysis has been analyzed with conservatively biased process variables, conservative equipment assumptions, and limiting kinetics parameters. The conclusions from these analyses are that the RPS and other safety-related systems are adequate to ensure the reactor coolant and MSS pressure boundary limits will not be exceeded as a result of this event.

Additionally, the conclusions from the calculations for the MSSV inoperable verify the power levels assumed for 1, 2, or 3 MSSVs inoperable per SG shown in TS Table 3.7.1-1.

The pressurizer overfill analysis concluded that the pressurizer will not become water-solid following an LOEL.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Table 5-3: Primary Side Overpressure LOEL Sequence of Events Event Time (sec) Value Event Initiation 0.0 MSSVs Open 5.2 5.5 High Pressurizer Pressure Trip Setpoint 6.17 2420.0 psia Peak Reactor Power 7.55 102.5% RTP Scram Rods Fall 7.58 --

PSV RC-200 Opens 8.10 PSV RC-201 Opens 8.87 Peak RCS Pressure 8.75 2706.6 psia Peak Secondary Pressure 9.2 1094.8 psia PSV RC-201 Closes 10.8 --

Peak Pressurizer Level 11.25 75.32% span PSV RC-200 Closes 11.6 --

Peak Reactor Vessel Inlet Temperature 14.55 565.9 0F Table 5-4: Secondary Side Overpressure LOEL Sequence of Events Event Time (sec) Value Event Initiation 0.0 --

MSSVs Open 8.2 --

8.3 High Pressurizer Pressure Trip Setpoint 15.4 2420 psia Peak Reactor Power 16.2 102.8% RTP Scram Rods Fall 16.8 --

Peak RCS Pressure 18.6 2517.2 psia Peak Pressurizer Level 19.8 44.1% span Peak Secondary Pressure 22.6 1101.8 psia Peak Reactor Vessel Inlet Temperature 24.6 568.8°F 21

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 120 110 100 -

90 Cr 80 70 0

60 C 50

' 40 n-30 20 10 0

0 20 40 Time (s)

Figure 5-1: Primary Side Overpressure LOEL - Reactor Power 140.0

_ 120.0 I-t- 100.0 X

LL

, 80.0 a) 60.0 0) 240.0 20.0 0.0 0 2 40 0 20 4 Time (s)

Figure 5-2: Primary Side Overpressure LOEL - Core Average Heat Flux 22

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 100 80 k 0

60

-J

-v 0~ 40 F

-J 20 k 0

0 20 40 Time (s)

Figure 5-3: Primary Side Overpressure LOEL - Pressurizer Level 610 600 590 580 S570 I-E 560 550 540 530 Time (s)

Figure 5-4: Primary Side Overpressure LOEL - RCS Temperatures 23

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 1200 110% Design Pressure 1100

'A) 0.

1000 k a) a)

0~ -- o Maximum SG-1 Pressure

.. Maximum SG-2 Pressure 900 4

800 0 20 40 Time (s)

Figure 5-5: Primary Side Overpressure LOEL - Maximum SG Pressures 2800 2600 2400 U) 22-2000 1800 0 20 40 Time (s)

Figure 5-6: Primary Side Overpressure LOEL - Maximum Primary Pressure 24

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0.

90 80

. 70 (D

30 60 5-05 400 30 20 10 0

0 10 20 30 40 50 Time (s)

Figure 5-7: Secondary Side Overpressure LOEL - Reactor Power 140.0 E'

I-- 120.0 I-100.0 X

LL i 80.0 (D

60.0 a)

• 40.0 0

20.0 0.01 0 10 20 30 40 50 Time (s)

Figure 5-8: Secondary Side Overpressure LOEL - Core Average Heat Flux 25

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 2800 2600

• 2400 2200 2000 1800 0 10 20 30 40 50 Time (s)

Figure 5-9: Secondary Side Overpressure LOEL - Pressurizer Pressure 01 1 80 I 60 I 03

-J

-u 0~ 40 -~

-J 20 k n

0 0 10 20 30 40 50 Time (s)

Figure 5-10: Secondary Side Overpressure LOEL - Pressurizer Level 26

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 640 620 600 LL 02 580 E

'- 560 540 520 0 10 20 30 40 50 Time (s)

Figure 5-11: Secondary Side Overpressure LOEL- RCS Temperatures 1150 1050 950 850 0 10 20 30 40 50 Time (s)

Figure 5-12: Secondary Side Overpressure LOEL - Maximum SG Pressures 5.2 Loss of FeedwaterFlow (SRP 15.2.7/ UFSAR 14.6) 5.2.1 Event Description A LOFW Flow event is defined as a reduction in feedwater flow to a SG without a corresponding reduction in steam flow from the SG. The closure of the feedwater regulating valves, the loss of 27

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY condensate or feedwater pumps, or a pipe break in the condensate or feedwater systems during steady-state operation would result in a LOFW Flow event.

The most limiting LOFW Flow event at HFP is an inadvertent closure of both feedwater regulating valves. An instantaneous closure of the regulating valves would cause the largest steam and feedwater flow mismatch and result in the most rapid reduction in the SG inventory.

The immediate system response is a steady decrease in SG liquid inventory. The temperature in the SG will increase after the loss of subcooled feedwater flow, which causes the SG pressure to increase correspondingly until the MSSV opening setpoint is reached. The RCS temperature and pressure will increase due to a mismatch in primary-to-secondary heat transfer. The reactor coolant expands, surging into the pressurizer. With the SG liquid inventory depleting and RCS pressure rising, a reactor trip will occur on either HIPP or low SG water level.

Pressures will continue to increase in the RCS and MSS until the point at which the PSV and MSSV setpoints are reached and the valves lift to relieve pressure. Neither the turbine bypass valves nor atmospheric dump valves (ADVs) are credited, as this would relieve pressure. The pressurizer PORVs are assumed to be unavailable for this event, since crediting them would allow the RCS to relieve pressure. The reactor trip rapidly decreases the core heat flux to decay heat levels. This will terminate the initial rise in RCS temperature, RCS pressure, and SG pressure, marking the end of the RCS and MSS over-pressurization events.

5.2.2 Input Parameters The key input parameters used in the analysis of the LOFW Flow are consistent with or conservative relative to the approved methodology (Reference 1). The differences between the new analysis and the AOR are presented in Table 5-5 and Table 5-6 for the RCS and MSS overpressure cases, respectively.

Note that only the limiting analysis values are presented in these tables.

• Initial Conditions - The LOFW Flow event was analyzed at HFP, since the resultant loss of feedwater was the greatest and presented the most significant challenge to the safety valve performance for primary and secondary system pressure.

For the RCS overpressure case, the biasing is as follows: [

For the MSS overpressure case, the biasing is as follows: [

" Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. All cases used kinetics parameters that are limiting for the time-in-life as chosen, that were biased to maximize the increase in reactor power during the transient.

• RPS Trips and Delays - The event is protected by the HPP and low SG level trip. The setpoints and response times for these trips were conservatively biased to delay the actuation of the trip function. In addition, control rod insertion was delayed to account for the CEA holding coil delay time. Note that the low SG level trip was conservatively disabled for the RCS overpressure case.

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• Pressurizer Control - The pressurizer sprays, heaters, and PORVs are disabled for the RCS overpressure case. Only the pressurizer sprays are enabled for the MSS overpressure case.

• Primary and Secondary Safety Valves - The PSVs and MSSVs were conservatively modeled.

The PSV setpoints were set to the maximum value allowed by the proposed TS change, [

I Each bank of MSSVs was set to the maximum of the TS range for each bank of valves. The valves were modeled to [

l

" Secondary Pressure Control - The ADVs were disabled for this event. The MSSVs are set to open at the TS maximum allowed lift setpoint.

• Steam Generator Blowdown - SG blowdown is credited for the RCS overpressure case, and disabled for the MSS overpressure case.

• Offsite Power - A loss of offsite power (LOOP) is assumed for the RCS overpressure case, and not assumed for the MSS overpressure case.

• Auxiliary Feedwater - The AFW response time precludes this system from actuating prior to a turnaround in either primary or secondary pressure.

" Steam Generator Tube Plugging - The RCS overpressure case uses a high level of SG tube plugging to limit primary-to-secondary heat transfer, while the MSS overpressure case uses a low level of SG tube plugging.

• Operator Actions - There are no operator actions required for the overpressure events, since they are mitigated by an RPS trip and (passive) safety valves.

• Single Failure - No single failures are assumed for the overpressure events, since they are mitigated by safety-grade systems (RPS and spring-loaded safety valves) that are either redundant or act passively.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Table 5-5: Initial Conditions/Input Parameters for the LOFW Flow Event - Maximum RCS Pressure Parameter Units AOR New Analysis Initial Core Power MWt 2754 2754 Pump Heat MWt 17 Explicitly modeled Initial Core Inlet Temperature OF 546 546 Vessel Flow Rate gpm 422,250 412,000 Initial Pressurizer Pressure psia 2164 2164 Initial Pressurizer Level'  % span 68.8 67.2 Pressurizer Spray -- No No MTC pcm/*F +1.5 +1.5 Number of Plugged SG Tubes  % per SG 0 10 CEA Worth at Trip pcm -4400 -5000 Reactor Trip and Setpoint Various HPP @ 2420 psia HPP @ 2420 psia Trip Delay Time sec 0.9 0.9 LOAC at Time of Trip -- Yes Yes MSSV Setpoints psia 1020.0 1019.7 1051.4 1049.7 1071.0 1064.7 Table 5-6: Initial Conditions/Input Parameters for the LOFW Flow Event - Maximum MSS Pressure Parameter Units AOR New Analysis Initial Core Power MWt 2754 2754 Pump Heat MWt 17 Explicitly modeled Initial Core Inlet Temperature OF 550 550 Vessel Flow Rate gpm 370,000 370,000 Initial Pressurizer Pressure psia 2164 2164 Initial Pressurizer Level'  % span 32.0 32.2 Pressurizer Spray -- No Yes MTC pcm/IF +1.5 +1.5 CEA Worth at Trip pcm -4400 -5000 Reactor Trip and Setpoint HPT @ 2420 psia HPT @ 2420 psia Various Low SG level @ 55" Low SG level @

BNWL 116.4" BNWL LOAC at Time of Trip -- No No MSSV Setpoints 1020.0 1019.7 psia 1051.4 1049.7 1071.0 1064.7 BNWL - Below Normal Water Level Note that the initial pressurizer level for the AOR is calculated by interpolating the span of 116 to 242 in (429 to 952 ft3) using the initial pressurizer volume from the AOR and noting that the pressurizer level span is 360 inches total.

The following list provides justification for the differences in the current AOR and the new analysis.

RCS Overpressure

1. RCS Flow Rate - The current AOR uses the core lift RCS flow rate; however, the highest flow rate at steady-state operation would be 412,000 gpm. This value accounts for cycle-to-cycle variation.

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2. SG Tube Plugging - The new analysis assumes a high degree of SG tube plugging, to further limit heat removal from the primary system. This is a conservative assumption.

3. CEA Worth at Trip - The CEA worth used in the new analysis bounds the value for minimum HFP worth assuming the most-reactive rod is stuck out of the core.

4. MSSV Setpoints - The setpoints used in the new analysis bound the maximum as-found lift settings in TS SR 3.7.1.1.

5. MSSV Inlet Piping - The new analysis models the MSSV inlet piping explicitly, whereas the AOR accounted for the pressure drop caused by the inlet piping by adding an additional bias to the accumulation on each set of valves.

MSS Overpressure

1. Pressurizer Spray - The pressurizer spray acts to reduce the pressurizer pressure, thereby delaying a reactor trip on HPP.

2. CEA Worth at Trip - The CEA worth used in the new analysis bounds the value for minimum HFP worth assuming the most-reactive rod is stuck out of the core.

3. Low SG Level Setpoint - The new analysis uses a setpoint that is more conservative than the value in the current AOR.

4. MSSV Setpoints - The setpoints used in the new analysis bound the maximum as-found lift settings in TS SR 3.7.1.1.

5. MSSV Inlet Piping - The new analysis models the MSSV inlet piping explicitly, whereas the AOR accounted for the pressure drop caused by the inlet piping by adding an additional bias to the accumulation on each set of valves.

5.2.3 Acceptance Criteria This event is classified as an AOO. The principally challenged criteria are below.

1. The pressures in the reactor coolant and MSSs should be less than 110% of design values.

2. Fuel cladding integrity should be maintained by ensuring SAFDLs are not exceeded. This is demonstrated by assuring that the minimum calculated DNB ratio is not less than the 95/95 DNB correlation limit.

3. An incident of moderate frequency should not generate a more serious plant condition without other faults occurring independently.

4. Radiological consequences do not exceed 10 CFR 50.67 guidelines.

The proposed change in PSV setpoint tolerance does not adversely affect the margin to the SAFDLs; therefore, the LOFW Flow event is not analyzed for DNB or FCM.

The radiological consequences of opening the MSSVs to mitigate this event would produce negligible site boundary doses compared to the 10 CFR 50.67 guidelines; as such, no mass and energy data for use in radiological consequences is required.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Therefore, the acceptance criteria are met if the peak reactor coolant and MSS pressures remain below 110% of their design values, the pressurizer does not become water-solid, and the AFW flow rate is sufficient to remove decay heat.

5.2.4 Description of Analyses and Evaluations The LOFW Flow event was performed with the approved non-LOCA methodology (Reference 1), as discussed in Section 3.0. The S-RELAP5 code was used to model the key system components and calculate core response and fluid conditions (flow rates, temperatures, and pressures). The peak system pressures include appropriate elevation corrections.

There are two main cases run for this event: one biased for peak primary pressure and one biased for peak secondary pressure. [

] Only the results of the limiting cases are presented herein.

Pressurizer overfill and long-term decay heat removal analyses are also performed, biased appropriately relative to the acceptance criterion being challenged.

5.2.5 Results Primary Side Pressure The sequence of events for the limiting RCS overpressure LOFW Flow event is given in Tale 5-7. The peak RCS pressure in this case is 2658.9 psia, which is below the overpressure limit of 110% of design pressure (i.e., 2750 psia).

The transient response is shown in Figures 5-13 through 5-18. These Figures show reactor power, pressurizer level, RCS temperatures, maximum SG pressures, maximum RCS pressure, and SG levels, respectively.

Secondary Side Pressure The sequence of events for the limiting MSS overpressure LOFW Flow event is given in Table 5-8. The peak MSS pressure in this case is 1086.9 psia, which is below the overpressure limit of 110% of design pressure (i.e., 1116.5 psia).

The transient response is shown in Figures 5-19 through 5-24. These Figures show reactor power, pressurizer pressure, pressurizer level, RCS temperatures, maximum SG pressures, and SG levels, respectively.

5.2.6 Conclusions The LOFW Flow analyses have been analyzed with conservatively biased process variables, conservative equipment assumptions, and limiting kinetics parameters. The conclusions from these analyses are that the RPS and other safety-related systems are adequate to ensure the reactor coolant and MSS pressure boundary limits will not be exceeded as a result of this event.

Additionally, the pressurizer overfill analysis concluded that the pressurizer does not become water-solid, and the long term decay heat removal analysis concluded that the AFW flow rate is sufficient to remove decay heat, following a LOFW.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Table 5-7: Primary Side Overpressure LOFW Flow Sequence of Events Event Time (see) Value Event Initiation 0.0 --

High Pressurizer Pressure Trip Setpoint 40.5 2420 psia Reactor and Turbine Trip 41.4 --

MSSVs Open 41.7 --

Scram Rods Fall 41.9 --

Peak Reactor Power 41.9 101.4% RTP PSV RC-200 Opens 44.3 --

Peak RCS Pressure 45.0 2658.9 psia Peak MSS Pressure 46.8 1072.4 psia Peak Pressurizer Level 47.8 74.3% span PSV RC-200 Closes 47.8 --

Peak Reactor Vessel Inlet Temperature 50.6 563.9 0 F SG Blowdown Isolation 68.2 Table 5-8: Secondary Side Overpressure LOFW Flow Se uence of Events Event Time (sec) Value Termination of feedwater flow 0.0 --

Low SG Level Trip Setpoint 26.3 0.0% NR Level Reactor and Turbine Trip 27.2 MSSVs Open 27.5 --

Peak Reactor Power 27.7 101.2% RTP Scram Rods Fall 27.7 --

Peak RCS Pressure 30.1 2356.7 psia Peak Pressurizer Level 31.2 38.1% span Peak Secondary Pressure 33.8 1086.9 psia Peak Reactor Vessel Inlet Temperature 35.6 565.9 0F 33

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 120 110 100 90 80 70 60 0

50 40 cc 30 20 10 0

0 20 40 60 80 100 Time (s)

Figure 5-13: Primary Side Overpressure LOFW Flow - Reactor Power 100 80

- 60

-J 40 F

20 F 0

0 20 40 60 80 100 Time (s)

Figure 5-14: Primary Side Overpressure LOFW Flow - Pressurizer Level 34

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 610 600 - - e Average T,,

A...... "

- .- -* Average T.*

590 580 570 L a.

...... 0 ............

..... a .......

a 560 .....

1-550 540 530 0 20 40 60 80 100 Time (s)

Figure 5-15: Primary Side Overpressure LOFW Flow - RCS Temperatures 1200 1100 Ca 1000 ca a.

900 800 0 20 40 60 80 100 Time(s)

Figure 5-16: Primary Side Overpressure LOFW Flow - Maximum SG Pressures 35

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 2800 2600 2400

$D c/)

o 2200 a:

2000 1800 0

Time (s)

Figure 5-17: Primary Side Overpressure LOFW Flow - Maximum RCS Pressure 100 80

- 60 40 20 0

100 Time (s)

Figure 5-18: Primary Side Overpressure LOFW Flow - SG Levels 36

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 120 110 100 90 80 70 60 0) 50 40 cc 30 20 10 0

0 20 40 60 80 100 Time (s)

Figure 5-19: Secondary Side Overpressure LOFW Flow - Reactor Power 2800 2600 I C, 2400.H.g..Prssurizer 2 40 0 ý--- ---------------- ---------------------........................ P.ress...re..S.e.tpoirn*t ............

42)

(L. 2200 2000 F 1800 0 20 40 60 80 100 Time (s)

Figure 5-20: Secondary Side Overpressure LOFW Flow - Pressurizer Pressure 37

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 100 80 I 60 k a)

-J 0* 40

,v k

-J 20 0

0 20 40 60 80 100 Time (s)

Figure 5-21: Secondary Side Overpressure LOFW Flow - Pressurizer Level 620 610 -- Average T.,

.......- Average T.

600 . Average T.,

- 590 I580o... .......... ........ ........ *.......

CL 570 560 - ....

55 0 . .- --. --

540 0 20 40 60 80 100 Time (s)

Figure 5-22: Secondary Side Overpressure LOFW Flow - RCS Temperatures 38

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 1200 110% Design Pressure 1100 1000 0-

-e Maximum SG-1 Pressure Maximum SG-2 Pressure -.

900 800 0 20 40 60 80 100 Time (s)

Figure 5-23: Secondary Side Overpressure LOFW Flow - Maximum SG Pressures 100 80 g 60

_J 40 20 0

0 100 Time (s)

Figure 5-24: Secondary Side Overpressure LOFW Flow - SG Levels 39

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 5.3 FeedwaterSystem Pipe Break (SRP 15.2.8 / UFSAR 14.26) 5.3.1 Event Description The FLB event is defined as a major break in a main feedwater line that is sufficiently large to prevent maintaining SG secondary side water inventory in the affected SG. A FLB between the SG and the upstream feedwater line check valve is usually the worst case, as blowdown of the SG secondary side water cannot be isolated. A FLB upstream of the feedwater check valve transitions to the LOFW event as soon as the check valve closes.

The FLB event can be considered as a heat-up event, a cool-down event, or a combination of both. There can be an initial, short, heat-up transient when the feedwater flow stops. This phase is terminated by a reactor trip. This heat-up portion of the transient produces the so-called "first peak." This first peak is expected to produce the maximum RCS pressure. Following the reactor trip, the RCS begins to cool down as a result of the heat removal from the affected SG. Once the feed ring uncovers, steam may flow to the break from both SGs prior to MSIV closure. The RCS pressure may decrease enough to cause the Safety Injection system to activate. The cooldown portion of the transient is terminated by dryout of the affected SG, which dramatically reduces the heat removal from the RCS.

5.3.2 Input Parameters The key input parameters used in the analysis of the FLB event are consistent with or conservative relative to the approved methodology (Reference 1). The differences between the new analysis and the AOR are presented in Table 5-9. Note that only the limiting analysis values are presented in this table.

• Initial Conditions - The FLB event was analyzed at HFP, since the resultant loss of feedwater was the greatest and presented the most significant challenge to the safety valve performance for primary system pressure. The biasing is as follows: [

• Reactivity Feedback - The reactivity feedback coefficients were biased according to the approved methodology. The analysis used kinetics parameters that are limiting for the time-in-life as chosen, that were biased to maximize the increase in reactor power during the transient.

• RPS Trips and Delays - The event is protected by the HPP trip. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, control rod insertion was delayed to account for the CEA holding coil delay time. Note that the low SG pressure trip was credited for larger break sizes, and that the low SG level trip was conservatively disabled for this analysis.

• Pressurizer Control - The pressurizer sprays, heaters, and PORVs are disabled.

• Secondary Pressure Control - The ADVs were disabled for this event. The MSSVs are set to open at the TS maximum allowed lift setpoint.

• Steam Generator Blowdown -SG blowdown is credited.

• Offsite Power - A LOOP is assumed.

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" Auxiliary Feedwater - The AFW response time precludes this system from actuating prior to a turnaround in primary pressure.

• Steam Generator Tube Plugging - A high level of SG tube plugging to limit primary-to-secondary heat transfer is assumed.

• Operator Actions - There are no operator actions required for the short-term over-pressurization events, since they are mitigated by an RPS trip and (passive) safety valves.

• Single Failure -No single failures are assumed for this analysis, since it is mitigated by safety-grade systems (RPS and spring-loaded safety valves) that are either redundant or act passively.

Table 5-9: Initial Conditions/Input Parameters for the FLB Event Parameter Units AOR New Analysis Initial Core Power MWt 2771 (Includes 17 MWt 2754 (Pump heat pump heat) explicitly modeled)

Initial Core Inlet Temperature OF 550 535 Vessel Flow Rate gpm 370,000 412,000 Initial Pressurizer Pressure psia 2164 2164 Initial Pressurizer Level'  % span 68.8 32.2 MTC pcm/°F +1.5 +1.5 Reactor Trip and Setpoint Various HPP @ 2470 psia HPP @ 2470 psia AFAS inches BNWL 265.2 n/a CEA Worth at Trip pcm -5000 -5000 Reactor Regulating System Operating Mode Manual Manual Steam Dump and Bypass System Operating Mode Manual Manual Pressurizer Pressure Control System Operating Mode Manual Manual Pressurizer Level Control System Operating Mode Manual Manual Note that the initial pressurizer level for the AOR is calculated by interpolating the span of 116 to 242 in (429 to 952 Wt)using the initial pressurizer volume from the AOR and noting that the pressurizer level span is 360 inches total.

AOR/New Analysis Input Reconciliation

1. RCS Inlet Temperature - The current AOR uses the maximum inlet temperature; however, the new analysis determined that a lower inlet temperature was more limiting.

2. RCS Flow Rate - The current AOR uses the TS minimum RCS flow rate; however, the new analysis determined that a higher flow rate was more limiting.

3. Initial Pressurizer Level - A minimum initial pressurizer level was determined to be limiting (by delaying the HPP trip) for the new analysis.

4. MSSV Inlet Piping - The new analysis models the MSSV inlet piping explicitly, whereas the AOR accounted for the pressure drop caused by the inlet piping by adding an additional bias to the accumulation on each set of valves.

5. Safety Valve Model - The PSV model is consistent with that presented in LOEL RCS Overpressure (Section 5.1.2), and the MSSV model is consistent with that described for LOFW Flow in Section 5.2.2.

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6. Low SG Pressure RPS Trip - The new analysis credits this TS trip to drop rods for larger break sizes where the SG pressure drops more rapidly. The limiting case credited the HPP trip only.

5.3.3 Acceptance Criteria This event is classified as a PA. These faults, at worst, may result in the failure of a small fraction of the fuel rods.

1. Pressures in the RCS and MSS are maintained below 110% of the design pressures (UFSAR 14.1.1.2).

2. Any fuel damage calculated to occur must be sufficiently limited such that the core will remain in place and intact with no loss of core cooling capability. Preclusion of fuel failure is demonstrated by delivering sufficient AFW to remove core decay heat such that there is no significant heatup of the RCS following reactor trip.

3. Any activity release must be such that the calculated doses at the site boundary are a small fraction of the 10 CFR 50.67 guidelines.

The proposed change in PSV setpoint tolerance does not adversely affect the margin to the SAFDLs; therefore, the FLB event was not analyzed for DNB or FCM.

Therefore, the acceptance criteria are met if the peak reactor coolant and MSS pressures remain below 110% of their design values and the AFW flow rate is sufficient to remove decay heat.

5.3.4 Description of Analyses and Evaluations The FLB event was performed with the approved non-LOCA methodology (Reference 1), as discussed in Section 3.0. The S-RELAP5 code was used to model the key system components and calculate core response and fluid conditions (flow rates, temperatures, and pressures). The peak system pressures include appropriate elevation corrections.

[

] Only the results of the RCS overpressure limiting case are presented.

An analysis was performed to verify the MSS overpressure acceptance criterion is met. A long-term decay heat removal analysis was also performed to analyze the ability of the AFW system to adequately remove RCP and decay heat following a FLB event.

5.3.5 Results The sequence of events for the limiting primary side FLB event is given in Table 5-10. The peak RCS pressure in this case is 2730.7 psia, which is below the overpressure limit of 110% of design pressure (i.e., 2750 psia).

The transient response is shown in Figures 5-25 through 5-31. These Figures show reactor power, core average heat flux, pressurizer level, RCS temperatures, SG pressures, maximum RCS pressure, and SG levels, respectively.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 5.3.6 Conclusions The FLB analyses have been analyzed with conservatively biased process variables, conservative equipment assumptions, and limiting kinetics parameters. The conclusions from these analyses are that the RPS and other safety-related systems are adequate to ensure the reactor coolant and MSS pressure boundary limits will not be exceeded as a result of this event.

Additionally, the long term decay heat removal analysis concluded that the AFW flow rate is sufficient to remove RCP and decay heat following a FLB event.

Table 5-10: Primary Side Overpressure FLB Sequence of Events (0.02 ft2 Break Size*)

Event Time (see) Value Event Initiation 0.0 --

Low-Low SG Level AFAS 44.1 --

High Pressurizer Pressure Trip Setpoint 57.8 2470 psia Reactor and Turbine Trip 58.7 --

Scram Rods Fall 59.2 --

PSV RC-200 Opens 60.7 PSV RC-201 Opens 61.2 Peak RCS Pressure 61.9 2730.7 psia MSSVs Open 63.4 --

PSV RC-201 Closes 65.9 PSV RC-200 Closes 68.3 SG Blowdown Isolation 79.1 --

• Modeled as SG-to-break flow area, not including MFW-to-break flow area.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 120 110 100 90 I-0L 80 0) 70 o 60 a- 50 0

(U 40 30 20 10 0

0 20 40 60' 80 100 Time (s)

Figure 5-25: FLB Event - Reactor Power 140.0

5. 120.0 F-ir x

100.0 ci 80.0 a)

"I" 60.0 a) 40.0 0

20.0 0.0 L 0 20 40 60 80 100 Time (s)

Figure 5-26: FLB Event - Core Average Heat Flux 44

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 100 80 k 60 I

-J 40

-j 20 F 0

20 40 60 80 100 Time (s)

Figure 5-27: FLB Event - Pressurizer Level 610 600 -- Average Tv

.....-- Average T.

• Average T.

590 580 ..... .. ....

570 ....

I--

560 3... /

550 4- -

540 530 0 20 40 60 80 100 Time (s)

Figure 5-28: FLB Event - RCS Temperatures 45

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.......... l..q .._Design..eP s ssure ............................................................................................................................................................

1100 U) 1000 0~

Co U) 0 900 ý a-

- - Maximum SG-1 Pressure

- Maximum SG-2 Pressure 800 700 0 20 40 60 80 100 Time (s)

Figure 5-29: FLB Event - Maximum SG Pressures 2800 0 % R.S De sig n P re ssu r e .11 . .

2600 P

.R..... ... ............ ..

Co 2400 F co 9) 0-o 2200 x

co 2000 1 1800 0 20 40 60 80 100 Time (s)

Figure 5-30: FLB Event - Maximum RCS Pressure 46

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY 100 80 -- SG-1 NR Level

\........ SG-2 NR Level

- -* SG-1 WR Level

- SG-2 WR Level 40 NA,>

20 -'

00-20 -- '= . N:.- : :'*- -. . . .=

0 20 40 60 80 100 Time (s)

Figure 5-31: FLB Event - SG Levels 5.4 Spectrum of RCCA Ejection Accidents (SRP 15.4.8 / UFSAR 14.13) 5.4.1 Event Description The CEA Ejection event is initiated by a postulated rupture of a control rod drive mechanism housing.

Such a rupture allows the full system pressure to act on the drive shaft, which ejects its control rod from the core. The consequences of the mechanical failure are a rapid positive reactivity insertion and an increase in radial power peaking, which could possibly lead to localized fuel rod damage.

Doppler reactivity feedback mitigates the power excursion as the fuel begins to heat up. Although the initial increase in power occurs too rapidly for the scram rods to have any effect on the power during that portion of the transient, the scram negative reactivity insertion does affect the fuel temperature and fuel rod cladding surface heat flux. The increase in rod cladding surface heat flux, adds heat to the RCS, causing a fairly rapid pressurization event which may cause the peak RCS pressure to approach the RCS overpressure limit. During the overpressure event, the PSVs are expected to actuate to relieve pressure.

The action of the VHP or the HPP trip in conjunction with the LCOs will prevent exceeding the RCS overpressure limit.

Zero Power Case A CEA Ejection event is initiated from HZP (10-9 RTP) and from within the LCOs by a rapid uncontrolled total withdrawal of a CEA within 0.10 seconds. The immediate reactor core response is an exponential increase in nuclear power. The delayed neutron fraction consistent with the time in cycle (beginning or end) is used.

At 40% (30% plus 10% uncertainty) of RTP or 2420 psia pressurizer pressure, a VHP or HPP trip, respectively, is initiated. As the fuel temperature starts increasing, negative Doppler feedback partially negates the ejected CEA reactivity worth and terminates the power excursion. After the VHP trip, RPS response, and CEA holding coil delay times have elapsed, the CEAs will insert and terminate the event.

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OF ANALYSIS SUPPORTING THE PRESSURIZER SAFETY VALVE TECHNICAL SPECIFICATION CHANGE - NON-PROPRIETARY Full Power Case A CEA Ejection event is initiated at HFP from within the LCOs by a rapid uncontrolled total withdrawal of a CEA within 0.10 seconds. The immediate reactor core response is an exponential increase in nuclear power. The delayed neutron fraction consistent with the time in cycle (beginning or end) is used.

At 110.33% (i.e., maximum analysis setpoint including uncertainties) of RTP or 2420 psia pressurizer pressure, a VHIP or HPP trip, respectively, is initiated. The negative Doppler feedback due to the increasing fuel temperature partially offsets the ejected CEA worth and terminates the power excursion.

The insertion of the CEAs will terminate the event after the RPS response time and CEA holding coil delay time have elapsed.

The peak deposited energy is a function of the initial stored energy, the amount of energy generated in the fuel rod, and the amount of energy released to the coolant during the transient. The initial stored energy is a function of initial Linear Heat Generation Rate (LHGR) and fuel-clad gap conductivity. The energy generated in the fuel rod during the transient is a function of the ejected CEA worth and the change in the radial and axial power distribution. The amount of heat transferred out of the fuel rod is a function of the fuel-clad gap conductivity and coolant-fuel rod film coefficient. To maximize the peak deposited energy during the transient, the analysis assumes the simultaneous occurrence of the most limiting combination of these parameters.

5.4.2 Input Parameters The key input parameters used in the analysis of the CEA Ejection event are consistent with or conservative relative to the approved methodology (Reference 1).

" Initial Conditions - The CEA Ejection event was analyzed at HFP and HZP initial conditions to provide a bounding fuel response to the ejected CEA.

For the HZP case, the biasing is as follows: [

For the HFP case, the biasing is as follows: [

• Reactivity Feedback - Reactivity feedbacks were modeled that are limiting for the time-in-life as chosen. Due to the rapidity of the transient, moderator feedback has a second-order impact on the consequences. The most positive MTC limits (based on TS) were modeled. The event is initially mitigated by the negative Doppler reactivity feedback. As such, the Doppler reactivity assumed in the analysis was conservatively biased to minimize the negative feedback due to increasing fuel temperatures. For the HZP initiated cases, fuel temperature dependent Doppler feedback was modeled.

" RPS Trips and Delays - The event is protected by the HPP and VHP trips. The RPS trip setpoints and response times were conservatively biased to delay the actuation of the trip function. In addition, control rod insertion was delayed to account for the CEA holding coil delay time.

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" Ejected CEA Worth - To maximize the core power response to the ejected CEA, a bounding high ejected CEA worth was assumed for cases initiated from HFP, based on inserting enough initial positive reactivity to see core power arrested by Doppler feedback prior to scram. For HZP cases, the limiting ejected CEA worth is an intermediate value at the transition point between HPP and VHP trips.

• Gap Conductance - The gap conductance was set to a conservative BOC value, adjusted to include burnup effects, to maximize the heat flux through the cladding and minimize the negative reactivity inserted due to Doppler feedback.

• Pressurizer Control - The pressurizer sprays, heaters, and PORVs are disabled for the RCS overpressure case.

• Steam Generator Tube Plugging - The RCS overpressure case uses a high level of SG tube plugging to limit primary-to-secondary heat transfer, although the event is too rapid for core parameters to be sensitive to minor differences in secondary heat transfer differences.

• Operator Actions - There are no operator actions required for the short-term overpressure event.

• Single Failure - A single failure of the Nuclear Instrumentation (NI) signal with the highest relative power signal is assumed, requiring the NI with the next-to-lowest power signal to reach the trip setpoint for a VHP trip (for 2-out-of-4 trip logic).

5.4.3 Acceptance Criteria This event is classified as a PA. These faults, at worst, may result in the failure of a small fraction of the fuel rods.

1. The radial average fuel pellet enthalpy at the hot spot must be <200 cal/g.

2. Pressures in the reactor coolant and MSSs should be maintained below 120% of the design pressures.

The change in PSV setpoints and uncertainties would not adversely alter the results from the previously performed CEA Ejection event analyzed for energy deposition in the fuel. Therefore, the CEA Ejection event is analyzed for peak RCS pressure, with a more stringent pressure limit of 110% of design pressure (UFSAR 14.1.1.2). The acceptance criteria are met if the peak RCS pressures remain below 110% of the design value.

5.4.4 Description of Analyses and Evaluations The CEA Ejection event was performed with the approved non-LOCA methodology (Reference 1), as discussed in Section 3.0. The S-RELAP5 code was used to model the key system components and calculate core response and fluid conditions (flow rates, temperatures, and pressures). The peak system pressures include appropriate elevation corrections.

I IOnly the results of the limiting case are presented.

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The transient response is shown in Figures 5-32 through 5-36. These Figures show reactor power, core heat flux, RCS temperatures, maximum RCS pressure and pressurizer pressure, and RCS flow rate, respectively.

It should be noted that the CEA Ejection RCS overpressure case has a much later reactor power peak than the CEA Ejection AOR for SAFDL evaluations. This is primarily caused by the biasing to delay the reactor trip. The delay to reactor trip increases as the ejected rod worth is decreased, until the point when a VHP trip is precluded by an HPP trip. The later trip was determined to be more limiting for overpressure, since the integrated power generation is greater than for a case with an earlier reactor trip.

5.4.6 Conclusions The CEA Ejection analyses have been analyzed with conservatively biased process variables, conservative equipment assumptions, and limiting kinetics parameters. The conclusions from these analyses are that the RPS and other safety-related systems are adequate to ensure the RCS pressure boundary limits will not be exceeded as a result of this event.

Table 5-11: Primary Side Overpressure CEA Ejection Sequence of Events Event Time (sec) Value Rod Fully Ejected 0.1 --

Neutron Power > 0.1% RTP 4.50 --

Peak Core Neutron Power 8.65 93.41% RTP Peak Core Neutron Power Indicated by next- 8.70 39.90% RTP to-lowest excore NI Pressurizer Pressure Reaches HPP Trip 9.27 2420 psia Setpoint 9.27_2420_psia Reactor Trip 10.18 --

Scram Rods Begin to Insert 10.69 --

PSV RC-200 Opens 11.15 PSV RC-201 Opens 11.65 Peak Core Thermal Power 11.70 1674.3 MW Peak RCS Pressure 12.00 2724.7 psia PSV RC-201 Closes 14.20 PSV RC-200 Closes 15.20 50

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80 70

0) 60 0

50 U

40 a) 30 20 10 n

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Time (s)

Figure 5-32: Primary Side Overpressure CEA Ejection - Reactor Power 80.0 I-

- 60.0 40.0

• 20.0 0.0 7 8 15 Time (s)

Figure 5-33: Primary Side Overpressure CEA Ejection - Core Heat Flux 51

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Es~

525 520 515 510 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Time (s)

Figure 5-34: Primary Side Overpressure CEA Ejection - RCS Temperatures 2800 110% Design Pressure (2750 psia) ......... ........ ......

2600

-240

__ Top of Pressurizer

(.

2200 . Vessel Bottom Pressure 2000 I 1800 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Time (s)

Figure 5-35: Primary Side Overpressure CEA Ejection - RCS Pressures 52

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411000 I 410000 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 Time (s)

Figure 5-36: Primary Side Overpressure CEA Ejection - RCS Flow Rate 6.0 Conclusion In order to address the changes made to the PSV setpoints and uncertainties, four pressure transients were analyzed. In addition to the overpressure aspects of these transients, they were also analyzed to determine the effects on other acceptance criteria (pressurizer overfill and long term decay heat removal), as applicable. The analyses demonstrate that sensible and decay heat can be removed and that an incident of moderate frequency does not generate a more serious plant condition. As shown in Table 6-1, both primary and secondary pressures remain below their pressure limits of 110% of design pressure for all events.

Table 6-1: Summary of Results Event Peak Primary Peak Secondary MSSV Inoperable Pressure, psia Pressure, psia Analysis LOEL 2706.6 1101.8 Yes LOFW 2658.9 1086.9 No FLB 2730.7 -- Yes (RCS)

CEA Ejection 2724.7 -- No 7.0 References

1. AREVA Document EMF-2310(P)(A), Revision 1, "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors"

2. XN-NF-81-58(P)(A), Supplements 1 and 2, Revision 2, "RODEX2 Fuel Thermal-Mechanical Response Evaluation Model, Exxon Nuclear Company"

3. ANF-81-58(P)(A), Revision 2, and Supplements 3 and 4, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, Siemens Power Corporation" 53

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4. BAW-10231P-A, Revision 1, "COPERNIC Fuel Rod Design Computer Code" (AREVA Document 43-10231PA-0 1) 54