ML20063K070
| ML20063K070 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 02/15/1994 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20063K058 | List: |
| References | |
| NUDOCS 9402280162 | |
| Download: ML20063K070 (136) | |
Text
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AP600 DESIGN CHANGE DESCRIPTION REPORT February 15.1994 l
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Westinghouse Electric Corporation WB5tingt10tlSe 9402280162 940215 PDR ADOCK 05200003 A PDR
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E E AP600 DESIGN CHANGE DESCRIPTION REPORT Table of Contents 1.0 introduction .. ., 1-1 2.0 Description of Design Changes . 2-1 2.1 PRHR Heat Exchanger Actuation Logic (DM-01) . . .. . 2-1 2.2 Pressurizer Heater Control Logic (DM-02) . . 2-5 2.3 CVS Makeup Control Logic (DM-03) . . . . . . 2-8 2.4 ADS Stage 1 Actuation Setpoint (DM-04) . . . 2-11 2.5 ADS Stage 2/3 Actuation Logic (DM-05) . 2-13 2.6 Core Makeup Tank Inlet Diffuser (DM-06) . . . . . 2-16 2.7 Direct Vessel injection Nozzle Venturi (DM-07) . 2-18 2.8 PRHR HX Inlet MOV Arrangement (DM-08) . 2-20 2.9 ADS Valves (DM-09. DM-10, DM-11) . .. . 2-22 2.9.1 ADS Stage 1 Flow Capacity (DM-09) ... . . 2-22 2.9.2 ADS Stage 2/3 Valves (DM-10) . ... . . . 2-25 2.9.3 ADS Stage 4 Vatves (DM-11) . . . . . 2-27 3.0 PRA Evaluation of ADS Design Changes . ... . . 3-1 4.0 Integrated Evaluation of Safety Analysis impact . ... .. .. . 4-1 4.1 Steam Line Break . . . . . . . .... . . .. .... 4-1 4.2 Steam Generator Tube Rupture . . ..... . .. . . . . 4-19 4.3 LOCA . ..... ........ .. ..... . .... ...... . 4-31 5.0 Test Plans for ADS (ADS Phase B) ... . . .. .. ...... . .. 5-1 5.1 Test Objectives . . . . . . . . . ... ... .......... .. . .. 5-1 5.2 Test Loop Configuration . . . ..... . . .......... .... . . . 5-1 5.3 ADS Phase B Tests . . . . .... . .... . ...... ...... . .. 5-1 l
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Table of Contents (continued)- a 6.0 ADS Valve Tests . . . . . . . . . . . . ..... ... .... ........ .. . . '6-1 ;
6.1 ADS /alve Type Selection Testing . . . . . . . . .... . ..... . .
6-1_
6.2 Valve Qualification Testing . ... . . ..... .. ............... 6 6.2.1 Analytica! Qualification . . . . ............ . ....... . . 6 .
6.2.2 Functional Testing . . .......................... ..... 6-2 {
6.2.3 Operator Qualification . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . 6 ;
6.2.4 Valve Design Control . . . .................. .......... 6-3 c; 6.3 Production Testing . . . . . . . . . . .... ............... ....... 6-3 !
6.4 Pre-operational Testing .............................. .... 6-3 l 6.5 I n-service Te sting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-3 4
7.0 References . . . . . . . ................ ............ ...... ....... 7 l Appendix A PX S Descriptio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . A-1 ,
A.1 Markup of SSAR Section 6.3 ,
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- AP600 SSAR, Section 6.3, Passive Core Cooling System, Revision 2 Dran,- :
text only (nonproprietary) 3
- AP600 SSAR, Section 6.3, Passive Core Cooling System, Revision 2 Draft. .
text only (proprietary) 3 A.2 Markup of AP600 PXS and RCS P&lD- !
Piping and Instrumentation Diagram Reactor Coolant System, .;
RCS M6 001, Revision 7- !
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Piping and Instrumenation Diagram, Passive Core. Cooling System, PXS M6 001, Revision 7 1
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AP600 DESIGN CHANGE DESCRIPTION REPORT Rn:
List of Tables Number Title Page 2-1 Summary of AP600 Design Modifications . . . 2-29 2-2 Revised PRHR HX Actuation Signals . . 2-31 2-3 Revised CVS Makeup Control Signals .. . . 2-32 2-4 Revised ADS Actuation Signals . . . .. . 2-33 3-1 ADS Fault Tree Unavailabilities . . . . . 3-2 4.1 -1 Sequence of Events for Steamline Break in Loop 2, Offsite Power Avail. .. 4-4 4.2-1 Sequence of Events for SGTR Analysis with PRHR Initiated on CMT Actuation Signal Compared to Actuation on High SG Level . . . . 4-21 4.3-1 DEDVI Break Sequence of Events . . . . . 4-36 4.3-2 Inadvertent ADS Actuation Sequence of Events . . .. . . 4-37 5-1 Valve Representation in the ADS Valve Package (ADS Phase B Tests) . 5-3 5-2 ADS Phase B Test Matrix . . . . . . ... 5-4 l
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E AP600 DESIGN CHANGE DESCRIPTION REPORT List of Figures Number Title Page 2-1 CMT inlet Diffuser . . . . . . . ..... .. .. .. ... .. .. . . 2-34 22 DVI Nozzle Venturi . . .... .... . .. ... 2-35 2-3 PRHR HX inlet isolation .. . ... .. . . . . . . .. . 2-36 2-4 Revised ADS Design ..... . .. ...... ............. .. . 2-37 2-5 ADS Stage 1/2/3 Examples . ................. .. . .. ... 2-38 2-6 ADS Stage 4 Examples . . . . . . ... . ..... .... . .... ... 2-39 4.1 -1 Nuclear Power Transient, Steam System Piping Failure . . . , . . . . . . . . 4-5 ,
4.1 -2 Core Heat Flux Transient. Steam System Piping Failure . ...... .. .. 4-6 4.1 -3 Reactivity Transient, Steam System Piping Failure . . . . . . ........... 4-7 4.1 -4 Core Average Temperature Transient Steam System Piping Failure . . . . . 4-8 4.1 -5 Reactor Vessel inlet Temperature Transient, Steam System Piping Failure . 4-9 4.1 -6 Reactor Coolant System Pressure Transient, Steam System Piping Failure 4-10 4.1 -7 Pressurizer Water Volume Transient, Steam System Piping Failure . . . . 4-11 4.1 -8 Core Flow Transient. Steam System Piping Failure . . . . . . .. . . . . . . . 4-12 4.1 -9 Feedwater Flow Transient, Steam System Piping Failure ............ 4-13 4.1-10 Core Boron Transient, Steam System Piping Failure . . . . . . . . . .... . 4-14 4.1-11 Steam Pressure Transient, Steam System Piping Failure . . . . . . . . . . . . 4-15 4.1-12 Steam Flow Transient, Steam System Piping Failure . . . . . . . . . . . . . . 4-16 4.1-13 Core Makeup Tank Injection Flow, Steam System Piping Failure . . . . . . . 4-17 4.1-14 Core Makeup Tank Water Volume, Steam System Piping Failure . . . . . . 4-18 4.2-1 Pressurizer Water Volume for SGTR Analysis with PRHR initiated on a CMT Actuation Signal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-22 '
4.2-2 Primary Pressure for SGTR Analysis with PRHR Initiated on a CMT Actuation Signal . . ...............................4-23 4.2-3 Secondary Pressure for SGTR Analysis with PRHR Initiated
- on a CMT Actuation Signal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-24 4.2-4 Primary to Secondary Break Flow for SGTR Analysis with PRHR Initiated on a CMT Actuation Signal . . . . . . . . . . . . . . . . . . . . . . . 4-25 i 4.2-5 Ruptured Steam Generator Water Volume for SGTR Analysis with PRHR Initiated on a CMT Actuation Signal . . . . . . . . . . . . . . . . . . . 4-26 4.2-6 Ruptured Steam Generator Steam Release for SGTR Analysis with PRHR Initiated on a CMT Actuation Signal . . . . . . . . . . . . . . . . . . . 4-27 4.2 7 Ruptured Loop CVS and CMT Injection Flow for SGTR Analysis with PRHR Initiated on a CMT Actuation Signal . . . . . . . . . . . . . . . . . . . 4-28 iv W Westingt10lJse
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AP600 DESIGN CHANGE DESCRIPTION REPORT List of Figures (continued)
Number Title Page 4.2-8 CMT Water Volume for SGTR Analysis with PRHR Initiated on a CMT Actuation Signal . . . . . . . 4-29 4.2-9 Primary to Secondary Break Flow for SGTR Analysis with PRHR initiated on a CMT Actuation Signal Compared to initiation on High SG Level . .. . . 4-30 4.3-1 Break Liquid Flow, DEDVI Break . . . .. . . 4-38 4.3-2 Upper Plenum Pressure, DEDVI Break . . . 4-39 4.3-3 Loop 1, CMT to DVI Flow. DEDVI Break . . .. .. .. . 4-40 4.3-4 Loop 2. CMT to DVI Flow. DEDVI Break . .... . . 4-41 4.3-5 Core Stack Mixture Level. DEDVI Break . .. .. . . . 4-42 4.3-6 Upper Head Mixture Level. DEDVI Break ... . . . . 4-43 4.3-7 Loop 1 SG Downflow Side. DEDVI Break . ... . . 4-44 4.3-8 Downcomer Mixture Level. DEDVI Break . . . . . .. .. . 4-45 4.3-9 Loop 1 CMT Mixture Level. DEDVI Break . . ... . . . 4-46 4.3-10 ADS Trains 1 to 3 Vapor Flow, DEDVI Break . .... . . . . . .. 4-47 4.3-11 Loop 2 Accumulator Mass Flow Rate, DEDVI Break . ... . 4-48 4.3-12 ADS 4 Liquid Flow Rate, DEDVI Break . . . ... . . . 4-49 4.3-13 ADS 4 Vapor Flow Rate, DEDVI Break . . . . . . . . ... . 4-50 4.3-14 Primary Mass inventory, DEDVI Break . . . . . . . . ...... 4-51 4.3-15 Downcomer Pressure, inadvertent ADS Actuation . ... .. .... . 4-52 4.3-16 Pressurizer Mixture Level, Inadvertent ADS Actuation . . . . . . 4-53 4.3-17 Core Stack Mixture Level, Inadvertent ADS Actuation . .. . .. .. 4-54 4.3-18 Loop 1 CMT to DVI Flow, Inadvertent ADS Actuation . . ......... . 4-55 4.3-19 Loop 1 CMT Mixture Level, Inadvertent ADS Actuation . . . . .. 4-56 4.3-20 Downcomer Mixture Level, inadvertent ADS Actuation .... . .... 4-57 4.3-21 ADS Trains 1 to 3 Uquid Flow Rate, inadvertent ADS Actuation . . . . . . . ....................... ....... .. 4-58 4.3-22 ADS Trains 1 to 3 Vapor Flow Rate, Inadvertent ,
ADS Actuation . . . . . . . . . .. ............ .. ........ ... 4-59 l 4.3-23 Loop 1 IRWST Injection Flow, inadvertent ADS Actuation . ...... . . 4-60 1 4.3-24 ADS 4 Liquid Flow Rate, inadvertent ADS Actuation . . . . ...... . . 4-61 I
4.3 25 ADS 4 Vapor Flow Rate, inadvertent ADS Actuation .. ........ 4-62 4.3-26 Primary Mass Inventory, inadvertent ADS Actuation .... ... . 4-63 W-WC5tingh0tJSe v
N!~ 9lI' AP600 DESIGN CHANGE DESCRIPTION REPORT List of Figures (continued)
Number Title Page 5-1 ADS Phase B Test Facility Schematic 5-5 5-2 ADS Phase B Test PipingNaive Package Configuration . . 5-6 "I W Westinghouse
-.a AP600 DESIGN CHANGE DESCRIPTION REPORT ~
1.0 Introduction This report provides a summary descnption of changes to the AP600 design described in the AP600 Standard Safety Analysis Report (SSAR), Revision 1 (Reference 7.1) and AP600 Probabilistic Risk Assessment (PRA). Revision 0 (Reference 7.2). The purposes of this document are to:
Identify changes to the AP600 design, Provide an overview of the impacts resulting from the changes, Serve as interim documentation until licensing documentation is revised (e.g.,
This report cc.- e eleven design changes to the AP600.
Section 2 of this report provides a description of each of these changes, including the purpose of the change, impacts (if any) to safety analyses, PRA. SSAR documentation, and AP600 test program.
Section 3 of this report provides an evaluation of the cumulative impact of the automatic depressur;zation system (ADS) design changes on the AP600 PRA.
Section 4 of this report provides an evaluation of the cumulative impact of the design changes on the AP600 design basis safety analyses.
Section 5 of this report provides a discussion on ADS Systems tests to be performed at the VAPORE facility in Cassacia, Italy.
Section 6 of this report provides a discussion of ADS valve testing proposed for performance during the development, manufacture, installation, startup and operation of the ADS valves.
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E AP600 DESIGN CHANGE DESCRIPTION REPORT 2.0 Description Of Design Changes This section describes modifications to the AP600 design described in the AP600 Standard Safety Analysis Report (SSAR). Reference 7.1. Table 2-1 provides a summary of the design modifications desenbed in this report. These mooifications are individually discussed in the following subsections.
2.1 Passive Residual Heat Removal Heat Exchanger Actuation Logic (DM-01)
Description This design modification involves a change to the actuation logic for the passive residual heat removal heat exchanger (PRHR HX). A signal to actuate the PRh3 HX on a core makeup tank (CMT) actuation signal has been added. Tt'e signals that actuate the PRHR HX on a high pressurizer level or on a high steam generator level have been deleted. Table 2-2 shows the revised PRHR HX actuation signals.
Purpose of Modification This design modification is introduced to reduce the potential for automatic depressurization system (ADS) actuation during a steam generator tube rupture (SGTR) event by increasing the margin between the minimum CMT level and the CMT level at which ADS actuation occurs.
This change, as well as changes DM-02 and DM-03 provide increased margin to pressurizer overfill during non-LOCA events on a best estimate basis. In the SSAR non-LOCA analysis, operator action is required to prevent the pressurizer from overfilling. The operator has more than one hour to take action. The combined impact of DM-01, DM-02 and DM-03 is that operator action is not required on a best estimate basis to prevent pressurizer overfall.
Additional means to provide increased margin to overfill are currently being evaluated.
Safety Analysis impact One of the concems for a SGTR event ir, tne possibility of steam generator overfill. This overfill could potentialty result in a sign;ficant increase in the offsite radiological consequences.
Automatic protection and passive design features are incorporated irito the AP600 to automaticalty terminate the break flow to prevent overfill during an SGTR. Actuation of the PRHR system, isolation of Chemical and Volume Control System (CVS) flow, and isolation of startup feedwater on high-2 steam generator (SG) level prevent c .Mll for an SGTR.
The design basis SGTR anatysis presented in SSAR Se* .3.6.3 includes an evaluation of potential for steam generator overfill. An analysis is pet' Aed to demonstrate that the AP600 design features prevent steam generator overfill, without taking credit for operator actions that WB5tingh0tJS8 24
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i AP600 DESIGN CHANGE DESCRIPTION REPORT can also prevent overtill. Tne limiting single failure for the overfill analysis is the failure of the startup feedwater control valve to throttle flow when nominal steam generator levelis reached.
Other conservative assumptions that maximize steam generator secondary volume (such as high initial steam generator secondary level, maximum initial RCS pressure, offsite power available, maximum CVS injected flow, maximum startup feedwater flow, and minimum startup feedwater delay time) are also assumed. The results of this analysis demonstrate that the AP600 protection system and passive system design features prevent steam generator overfill. The results also show that automatic actuation of the ADS does not occur. In these analyses the PRHR HX was actuated on a high SG level.
SGTR sensitivity studies performed subsequent to the SSAR analyses indicate that operator action to actuate the PRHR HX may be necessary to prevent ADS actuation in a case where CVS fails and the startup feedwater system (SFWS) works as designed (i.e., no CVS flow, SFWS throttles back to maintain steam generator level). In this case SG level does not increase to the high-2 level setpoint and actuate the PRHR HX.
PRHR HX actuation on CMT actuation addresses this event by actuating the PRHR HX independent of SG level. The high SG level actuation of the PRHR HX is no longer required because the CMTs and, therefore, the PRHR HX would already have been actuated. The high pressurizer level PRHR HX actuation signal is also not required because the CMTs and therefore the PRHR HX would have already been actuated.
The change to PRHR HX actuation logic also reduces the potential of pressurizer overfill following an inadvertent operation of the CMTs on a best estimate basis.
PRA impact The change to the PRHR HX actuation logic requires modification of the initiating signals in four of the PRHR fault trees (PRT, PRP, PRB, PRS).
The unavailability of the PRHR HX is dominated by failure of the control or the operator. As such, this change is not expected to have a significant isnpact on the PRA results.
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AP600 DESIGN CHANGE DESCRIPTION REPORT SSAR impact This change impacts the following SSAR Chapters:
Chapter 6 Engineered Safety Features (see Appendix A-1 for markup of SSAR Section 6.3, Passive Core Cooling Systems)
Chapter 7 Instrumentation and Controls ,
Chapter 15 Accident Analyses Chapter 16 Technical Specification Test Program impact Core Makeup Tank Separate Effects Tests This design change has no impact on these tests.
The CMT tests are separate effects tests designed to investigate thermal-hydraulic phenomena within the CMT over a wide range of pressures and temperatures. The PRHR actuation logic change affects event timing but not the nature of the phenomena investigated during these tests. This change does not affect the range of conditions required to be tested.
Integral Systems Tests at Oregon State University This design change impacts control logic for these tests.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tem cooling behavior. Modifications to the facility logic and control software must be performed to incorporate the specified change. The revised logic will be incorporated into the facility control systems at OSU. These changes will be implemented for all OSU matrix tests.
Integral Systems Tests at SPES-2 Test Facility This design change impacts control logic for these tests.
The SPES-2 tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate systems interactions during high pressure transients. The first matrix test at SPES-2, the 2 inch SBLOCA SSAR reference case, has been completed using the SSAR (Reference 7.1) actuation logic. The control logic change svill be implemented l
for the remaining SPES-2 tests.
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AP600 DESIGN CHANGE DESCRIPTION REPORT The results cf this first matnx test are being analyzed with respect to the AP600 design changes now ceing implemented. In addition, pretest analyses of the remaining SPES-2 tests are being performed using the revised logic and setpoints.
ADS Phase B Tests a
This design change will have no impact on the ADS Phase B tests.
The ADS Phase B tests are performed by depressurizing the test facility through the ADS valve piping package, downstream piping and sparger. The conditions of the test are determined by the facility supply tank initial conditions (supply tank pressure and temperature) and the position of the facility control valve, which determines the fluid conditions upstream of the ADS valve piping package (see Figure 5-1).
The PRHR actuation logic is not explicitly modeled in the test facility. Pre-test analysis will be performed for the facility to ensure that the test conditions during the facility blowdown are appropriate based on AP600 plant calculations. This will determine the test facility initial conditions and the position of the facility control valves during the test. This change does not affect the range of conditions required to be tested.
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2.2 Pressurizer Heater Control Logic (DM-02)
Description This design modification involves a change to the control logic for the pressurizer heaters. A change has been made to the pressurizer heater control logic to provide an automatic block of heater operation upon receipt of a CMT actuation signal. This logic will be implemented by taking isolated CMT actuation signals from the protectio'1 and monitoring system (PMS) to the plant control system (PLS). Redundancy is employed in the control system to trip redundant nonsafety-related breakers, three in series for each heater.
Forpose of Modification This logic change provides a means for blocking the pressurizer heaters (in addition to a manual block) and reduces the potential for SG overfill and automatic ADS actuation for an SGTR accident. This logic change also reduces the potential for pressurizer safety valve actuation during Condition 2 events.
Safety Analysis impact The SSAR analysis (Reference 7.1) does not include continued pressurizer heater operation.
Long term operation of the pressurizer heaters during certain non-LOCA events acts to retard the depressurization of the RCS following 'tuation of the CMTs. Without the automatic pressurizer heater block, it may be necet ry for the operators to manually open the pressunzer heater breakers to prevent SG overfill or automatic ADS actuation for the design basis SGTR event. Continued operation of the pressurizer heaters does not adversely affect the response of the plant for several thousand seconds. As a result, the safety case is based on the operators opening breakers locally. Therefore, this control logic can be nonsafety-related since its function is to reduce unnecessary demands on the operator.
PRA impact This change does not affect the systems modeled in the PRA or their success criteria.
SSAR impact This change impacts the foilowing SSAR Chapters:
Chapter 5 Reactor Coolant System and Connected Systems Chapter 7 Instrumentation and Controls Chapter 15 Accident Analyses I
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AP600 DESIGN CHANGE DESCRIPTION REPORT
'i Test Program impact ,
Core Makeup Tank Separate Effects Tests This design change has no impact on these tests. t The CMT tests are separate effects tests designed to investigate thermal-hydrauko j phenomena within the CMT over a wide range of pressures and temperatures. The [
pressurizer heater control logic change may affect event timings but not the nature of the .
phenomena investigated in this test facility. This change does not affect the range of }
conditions required to be tested.
Integral Systems Tests at Oregon State University ,
I This design change impacts contiol logic for these tests.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tem cooling behavior. Modifications to the facility logic and l
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control software must be performed to incorporate the specified change. The revised logic will be incorporated into the facility control systems at OSU. These changes will be implemented for all OSU matrix tests. .
Integral Systems Tests at SPES-2 Test Facility This design change impacts control logic for these tests.
The SPES-2 tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate systems interactions during high pressure transients. The first matrix test at SPES-2, the 2-inch SBLOCA SSAR reference case, has been completed ,
using the SSAR (Reference 7.1) actuation logic. The control logic change will be implemented for the remaining SPES-2 tests. l!
The results of this first matrix test are being analyzed with respect to the AP600 design !
changes now being implemented. In addition, pretest analyses of the remaining SPES-2 tests ;
are being performed using the revised logic and setpoints.
ADS Phase B Tests P
This design change will have no impact on the ADS Phase B test conditions. ;
The ADS Phase B tests are performed by depressurizing the test facility through the ADS valve piping package, downstream piping and sparger. The conditions of the test are
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E AP600 DESIGN CHANGE DESCRIPTION REPORT determined by the facility supply tank initial conditions (supply tank pressure and temperature) and the position of the facility control valve. which determines the fluid conditions upstream ct the ADS valve piping package (see Figure 5-1).
The pressurizer heater control logic is not explicitly modeled in the test facility. Pre-test analysis will be performed for the facility to ensure that the test conditions during the facility blowdown are appropriate based on the AP600 plant calculations. This will determine the test facility initial conditions and the position of the facility control valves dunng the test. This change does not affect the range of conditions required to be tested.
E AP600 DESIGN CHANGE DESCRIPTION REPORT 2.3 CVS Makeup Control Logle (DM-03)
Description This design modification involves a change to the CVS control logic so that on a CMT actuation signal one CVS makeup is started on a 10 percent pressurizer level. That pump will automatically stop on a 20 percent pressurizer level. The control logic as described in the AP600 SSAR (Reference 7.1) calls for both CVS pumps to start on a CMT actuation signal.
See Table 2-3 for a descnption of the CVS makeup control logic change. This change makes the control logic following CMT actuation similar to normal operation except that following CMT actuation the level setpoints for starting and stopping the CVS injection are changed.
Purpose of Modification The change improves the expected operation of the plant during non-LOCA events. This change maintains the function to provide automatic CVS makeup to minimize the potential of automatic actuation of ADS. In addition, it reduces the chance of lifting a pressurizer safety valve or of overfilling the pressurizer for non-LOCA events. The operators are able to start the second CVS pump in case the CMT draindown is approaching ADS. Higher pressurizer level control setpoints will be used at zero power when the CMTs have not been actuated, such that operation of CVS makeup will not occur with the CMTs operating unless the pressurizer level drops to 10 percent. Note that since the CVS is a nonsafety-related system, its control logic is provided by the PLS. This change does not affect the acceptability of the Chapter 15 analyses.
Safety Analysis impact The AP600 safety analyses do not credit operation of the CVS. CVS operation is only assumed when such operation results in a more limiting transient. This change does not impact the AP600 design basis safety analyses presented in Chapter 15 of the SSAR.
PRA impact This change does not affect how the CVS is modeled in the PRA.
SSAR impact This change impacts the following SSAR Chapters:
Chapter 7 Instrumentation and Controls Chapter 9 Auxiliary Systems
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AP600 DESIGN CHANGE DESCRIPT10N REPORT Test Program impact Core Makeup Tank Separate Effects Tests This design change has no impact on these tests.
The CMT tests are separate effects tests designed to investigate thermal-hydraulic phenomena within the CMT over a wide range of pressures and temperatures. The CVS makeup control logic change does ' "fect the range of conditions required to be tested.
Integral Systems Tests at Oregc e State University Ttis design change impacts control logic for these tests.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tem cooling behavior. Modifications to the facility logic and control software must be performed to incorporate the specified change. The revised logic will be incorporated into the facility control systems at OSU. These changes will be implemented for all OSU matrix tests.
Integral Systems Tests at SPES-2 Test Facility This design change impacts control logic for these tests.
The SPES-2 tests are integral systems tests to obtain thermal hydraulic data for computer code validation and to investigate systems interactions during high pressure transients. The first matrix test at SPES-2, the 2-inch SBLOCA SSAR reference case, has been completed using the SSAR (Reference 7.1) actuation logic. The control logic change will be implemented for the remaining SPES-2 tests.
The resutts of this first matrix test are being analyzed with respect to the AP600 design changes now being implemented. In addition, pretest analyses of the remainino SPES-2 tests are being performed using the revised logic and setpoints.
ADS Phase B Tests This design change will have no impact on the ADS Phase B test conditions.
The ADS Phase B tests are performed by depressurizing the test facility through the ADS valve piping package, downstream piping and sparger. The conditions of the tests are i
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AP600 DESIGN CHANGE DESCRIPTION REPORT determined by the facility supply tank initial conditions isupply tank pressure and temperature) and the position of the facility control valve, which determines the fluid conditions upstream of the ADS valve piping package (see Figure 5-1).
The CVS makeup control logic is not explicitly modeled in the test facility. Pre-test analysis will be performed for the facility to ensure that the test conditions during the facility blowdown are appropriate based on the AP600 plant calculations This will determine the test facility initial conditions and the position of the facility control valves during the test. This change j does not affect the range of conditions required to be tested. {
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AP600 DESIGN CHANGE DESCRIPTION REPORT 2.4 ADS Stage 1 Actuation Setpoint (DM-04)
Description This design modification involves a change to the CMT level setpoint that actuates ADS 2
Stage 1. This setpoint has been changed from 1500 ft to 1350 ft'.
Purposp of Modification This modification increases the margin to automatic ADS actuation. Sensitivity studies performed for small steam line breaks and SGTR events show that a minimum CMT volume of approximately 1590 ft' occurs under conditions of minimum CMT steam condensation.
Additional margin to automatic ADS actuation is desirable. This margin is obtained by reducing the CMT ievel setpoint for ADS actuation to 1350 ft'.
Safety Analysis impact Sensitivity studies were performed to verify that lowering this setpoint did not adversely cffect LOCAs. The direct vessel injection (DVI) line break analysis, shown in Section 4.3, uses this setpoint and results in improved performance.
PRA impact This change does not affect the success criteria used in the PRA.
SSAR Impact This change impacts the following SSAR Chapters:
Chapter 6 Engineered Safety Features (see Appendix A-1)
Chapter 7 Instrumentation and Controls Chapter 15 Accident Analyses Chapter 16 Technical Specifications Test Program impact Core Makeup Tank Separate Effects Tests This design change has minimalimpact on these tests.
The CMT tests are separate effects tests designed to investigato thermal-hydraulic phenomena within the CMT over a wide range of pressures and temperatures. The ADS Stage 1 setpoint change may affect event timings but not the nature of the phenomena investigated in this test facility. This change does not affect the range of conditions required to be tested.
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AP600 DESIGN CHANGE DESCRIPTION REPORT Integral Systems Tests at Oregon State University This design change impacts control logic for these tests.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tem cooling behavior. Modifications to the facildy control software must be performed to incorporate the specified change. This change will be implemented for all OSU matrix tests.
Integral Systems Tests at SPES-2 Test Facility This design change impacts control logic for these tests.
The SPES-2 tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate systems interactions during high pressure transients. The first matrix test at SPES-2, the 2-inch SBLOCA SSAR reference case, has been completed using the SSAR (Reference 7.1) actuation logic. The control logic change will be implemented for the remaining SPES-2 tests.
The results of this first matrix test are being analyzed with respect to the AP600 design changes now being implemented. In addition, pretest analysis of the remaining SPES-2 tests are being performed using the revised setpoint.
ADS Phase B Tests This design change will have no impact on ADS Phase B test conditions.
The ADS Phase B tests are performed by depressurizing the test facility through the ADS valve piping package, downstream piping and sparger. The conditions of the tests are determined by the facility supply tank initial conditions (supply tank pressure and temperature) and the position of the facility control valve, which determines the fluid conditions upstream of the ADS valve piping package (see Figure 5-1).
The ADS Stage 1 setpoint is not explicitly modeled in the test facility. Pre-test analyses will be perforTned for the facility to ensure that the test conditions during the facility blowdown are appropriate based on the AP600 plant calculations. This will determine the test facility initial conditions and the position of the facility control valves during the test.
2-12 W Westinghouse l
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AP600 DESIGN CHANGE DESCRtPT10N REPORT 2.5 ADS Second and Third Stage Actuation Logic (DM-05)
Description This design modification involves a change to the actuation logic for ADS Stages 2 and 3.
The actuation of ADS Stages 2 and 3 has been changed so that Stages 2 and 3 actuate on a time delay after actuation of the previous stage. The actuation logic of ADS Stages 1 and 4 remains based on CMT level.
Purpose of Modification This change removes dependence on the CMT level measurement during the most rapid phase of RCS / CMT pressure reduction. This removes the reliance on the heated junction resistance temperature detector (RTD) levelinstruments during this phase of operation. When the CMT level approaches the fourth stage level setpoint, the RCS pressure is nearly constant and CMT level measurement is less challenging.
The specific timer settings are shown in Table 2-4.
Safety Analysis impact These setpoints are included in the DVI break and spurious ADS analysis shown in Section 4.3. In both accidents this logic has resulted in similar or improved overall plant performance.
PRA Impact Actuation of ADS Stages 2 and 3 on timers, instead of level, will not affect the ADS reliability.
SSAR impact This change impacts the following SSAR Chapters:
Chapter 6 Engineered Safety Features (See Appendix A-1)
Chapter 7 Instrumentation and Controls Chapter 15 Accident Analyses Chapter 16 Technical Specifications Test Proaram impact Core Makeup Tank Separate Effects Tests This design change has minimalimpact on these tests.
The CMT tests are separate effects tests designed to investigate thermal-hydraulic phenomena within the CMT over a wide range of pressures and temperatures. The changes Westingt10Use 2-13 ,
l l
==d AP600 DESIGN CHANGE DESCRIPTION REPORT to the secord and third stage ADS setpoints rnay affect event timings but not the nature of the phenomena investigated in this test facility. This change does not affect the range of conditions required to be tested.
Integral Systems Tests at Oregon State University This design change impacts control logic for these tests.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tem cooling behavior. Modifications to the facility control software must be performed to incorporate the specified change. This change will be implemented for all OSU matrix tests.
Integral Systems Tests at SPES-2 Test Facility This design change impacts control logic for these tests.
The SPES-2 tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate systems interactions during high pressure transients. The first matrix test at SPES-2. the 2-inch SBLOCA SSAR reference case, has been completed using the SSAR (Reference 7.1) actuation logic. The control logic change will be implemented for the remaining SPES-2 tests.
The results of this first matrix test are being analyzed with respect to the AP600 design changes now being implemented. In addition, pretest analyses of the remaining SPES-2 tests are being performed using the revised setpoints.
ADS Phase B Tests This design change will have minimalimpact on ADS Phase B test conditions.
The ADS Phase B tests are performed by depressurizing the test facility through the ADS valve piping package, downstream piping and sparger. The conditions of the tests are determined by the facility supp9 tank initial conditions (supply tank pressure and temperature) and the position of the facility control valve, which determines the fluid conditions upstream of the ADS vake piping package (see Figure 5-1).
The second and third stage ADS setpoints are not explicitly modeled in the test facility. Pre-test analysis will be performed for the facility to ensure that the test conditions during the facility blowdown are appropriate based on the AP600 plant calculations. This will determine 2-14 [ Westitigh00Se
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- n : l AP600 DESIGN CHANGE DESCRIPTION REPORT j 1
l the test facility initial conditions and the position of the facility control valves durire r e test.
This change does not affect the range of conditions required to be tested.
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- l E AP600 DESIGN CHANGE DESCRIPTION REPORT 2.6 Core Makeup Tank Inlet Diffuser (DM-06)
_ Description This design modification involves the addition of a diffuser to the inlet of each of the CMTs.
The diffuser is an extension of the inlet pipe with the same size and schedule. The bottom of the diffuser is plugged and holes are drilled into the side. This design forces theincoming steam flow to turn 90 degrees which effectively reduces the steam penetration into the CMT.
Figure 2-2 shows the diffuser and desenbes the arrangement of the holes.
Purpose of Modification Preliminary tests at the AP600 CMT Test Facility (Waltz Mill Test Site) have shown that rapid steam condensation can occur in the CMT during some modes of operation. The mode of operation that is affected is when the CMT contains cold water and the cold leg balance line supplies steam to the CMT. This mode only occurs dunng medium to large LOCAs. The impact of this rapid steam condensation is a period of reduced CMT injection flow and additional piping loads.
Safety Analysis impact Additional testing at the CMT test facility with a prototypic steam diffuser shows that a CMT inlet diffuser greatly reduces pressure fluctuations. The period of reduced CMT injection is also reduced. LOCA analysis has been performed to assess the sensitivity to this reduced injection. A DVI LOCA was analyzed with different volumes of the CMT heating up before drain down began. These volumes bound the maximum mixing volume. With a larger mixing volume, the calculations result in a lower RCS inventory. However, with the addition of the DVI venturi the DVI LOCA results were better than those contained in the SSAR; there is now no core uncovery. Refer to the LOCA analysis contained in Section 4.3.
PR A impact This change will have no impact on the PRA since it does not affect how the systems are modeled and has no effect on success criteria.
SSAR Impact This change impacts the following SSAR Chapters:
Chapter 5 Reactor Coolant System and Connected Systems Chapter 6 Engineered Safety Features Chapter 15 Accident Analyses
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I AP600 DESIGN CHANGE DESCRIPTION REPORT Test Program Impact Core Makeup Tank Separate Effects Tests Integral Systems Tests at Oregon State University Integral Systems Tests at SPES 2 Test Facility This design change will have no impact on the above tests.
A CMT inlet diffuser has already been incorporated into each of these AP600 tests facilities.
The rationale for sizing the diffuser at each facility was based on the functional requirement to prevent rapid steam condensation in the CMT during periods of high steam flow into cold water. Each facility diffuser design was based on the scaling parameters of the individual facility.
ADS Phase B Tests This design change will have no impact on the ADS Phase B tests.
The CMT is not modeled in the tests. The ADS tests are not dependent on the CMT heatup or draindown rates. This change does not affect the range of conditions required to be tested.
i l
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AP600 DESIGN CHANGE DESCRIPTION REPORT 2.7 Direct Vessel injection Nozzle Venturl (DM-07)
Description A ventun has been incorporated into the reactor vessel DVI nozzle. Figure 2 3 shows the ventun.
Purpose of Modification This ventun reduces the severity of a DVI line LOCA by reducing the RCS blowdown rate.
Safety Analysis impact A DVI line break LOCA is a limiting design basis event because of the size of the DVI line and because the break location disables one half of the injection supplies. Incorporating a reduced ID venturi into the DVI nozzle effectively reduces the maximum DVI break size. The minimum ID throat of venturi is long enough to effectively choke the flow. There is no impact on CMT and accumulator injection flow because orifices in the discharges of these tanks can be adjusted to compensate for the venturi, therefore, the resistance from the CMT and accumulator to the RCS will be unchanged. There will be a small effect on the in-containment refueling water storage tank (IRWST) injection resistance.
PRA Impact This change will have no impact on the PRA because this break is still a medium LOCA when both reactor and CMT flow are taken into consideration. This change also does not affect the success enteria used in the PRA.
SSAR impact This change impacts the following SSAR Chapters:
Chapter 4 Reactor Chapter 15 Accident Analyses Test Program Impact Core Makeup Tank Separate Effects Tests This design change will have no impact on these tests.
In the AP600, the addition of a venturi to the DVI nozzle is accommodated by maintaining the overall pressure drop in the DVI line by reducing the size of the orifice already designed into the line.
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AP600 DESIGN CHANGE DESCRIPTION REPORT l The drainline in the CMT tests simulates the DVI line retuming to the reactor vessel. The test drainline is designed for a range of flowrates out of the CMT which bound the range of flowrates for all anticipated transients.
Integral Systems Tests at SPES-2 Test Facility This design change will have a small impact on facility hardware for this test facility.
The SPES-2 tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate systems interacuons during high pressure transients. The DVI line in SPES-2 will be modified to include a scated venturi prior to the performance of the DVI line break test. Since the overall pressure drop is the same for the other SBLOCAs, regardless of inclusion of the DVI venturi, the other tcats do not require its installation.
The nozzle will be installed for the duration of the test program, which will include SGTR and MSLB matrix tests. A cold test to characterize the pressure drop in the DVI line will be performed during the installation of the venturi.
Integral Systems Tests at Oregon State University This design change will have a small impact on facility hardware for this test facility.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tem cooling behavior. The DVI line in OSU will be modified to include a scaled venturi prior to the performance of the DVI line break test. Since the overall pressure drop is the same for the other SBLOCAs, regardless of inclusion of the DVI venturi, the other tests do not require its installation. The DVI venturi has a resistance that is less than 1 percent of the total injection line resistance for the CMT, accumulator or IRWST.
A cold flow test to characterize the pressure drop of the DVI venturi will be performed.
ADS Phase B Tests This design change will have no impact on these tests.
The installation of a venturi impacts the plant response to a DVI line break. The pretest analyses performed to establish the test conditions cover the range of expected transients and the test blowdowns are designed to provide fluid conditions over a wide range of conditions.
This change does not affect the range of conditions required to be tested.
3 Westingh0Use 2-19
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AP600 DESIGN CHANGE DESCRIPTION REPORT 2.8 PRHR HX Inlet Motor Operated Valve Arrangement (DM-08)
Description A motor-operated valve (MOV) is added in the inlet line to each PRHR HX. Two MOVs replace the two manual valves and a MOV in the common line. Figure 2-4 shows this arrangement. ,
Purpose of Modification This change improves the reliability of the PRHR HX and reduces the chances of forced shutdowns due to a valve problem occurring during at-power in-service testing (IST) of the PRHR HX control valves.
With the valve arrangement described in the SSAR (Reference 7.1), the MOV in the common line is normally open with its power removed. Its power is removed to avoid the possibility of a spurious valve closure failure which would render the PRHR HX inoperable. The MOV provides remote isolation in case of excessive PRHR HX leakage it will also be closed every three months during the at-power IST of the PRHR HX air operated valve (AOV).
If MOV problems occur during IST, there would be a limited time to fix them to avoid a Technical Specification violation. Such problems include the MOV failing to open or problems with MOV limit switches falsely indicating that the MOV has not opened. Either of these problems would force the operators to declare the PRHR HX inoperable. Because the MOV is located inside containment, access and repair is difficult.
With the revised design, the chance of this occurring is reduced, since both MOVs would have to experience a problem to render the system inoperable. With both HXs available, the power would not have to be removed to prevent a spurious failure from defeating the PRHR HXs.
This makes IST and leakage isolation of the PRHR HXs easier for the operators since they would not have to restore power locally at the motor control center before operating the MOV.
In the unlikely case that one of the PRHR HXs is isolated because of a leak, the single l operable HX would have one inlet MOV which would have its power removed. This change significantly improves the PRHR HX operability since isolating a PRHR HX is an unlikely situation.
Safety Analysis impact !
There is no impact on the safety analysis because this change has no impact on system l performance during design basis event or on the limiting single failures for these events.
1 2a W Westinghouse
AP600 DESIGN CHANGE DESCRIPTION REPORT PRA impact There is no impact on the PRHR HX reliability because these valves are not modeled in the PRHR HX fault tree.
In the event of a PRHR HX tube rupture, the operator should isolate the PRHR HX. Failures in the control system or operator action dominate the unavailability. Having to close two MOVs instead of one will have a small impact on the core melt frequency (CMF).
SSAR impact This change impacts the following SSAR Chapters:
Chapter 3 Design of Stn;ctures, Components. Equipment and Systems Chapter 6 Engineered Safety Features Chapter 7 Instrumentation and Controls Chapter 16 Technical Specifications Test Procram impact Core Makeup Tank Separate Effects Tests Integral Systems Tests at Oregon State University Integral Systems Tests at SPES-2 Test Facility ADS Phase B Tests This design change has no impact on the above test programs.
This design change improves the reliability of the PRHR HX by reducing the chance of forced shutdowns due to a valve problem that may occur during IST. This change does not impact the performance of the PRHR HX in the AP600 during a transient and has no impact on the above test programs.
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AP600 DESIGN CHANGE DESCRIP110N REPORT 2.9 ADS Valves (DM-09, DM-10, DM 11) l l During several meeting with Westinghouse the Nuclear Regulatory Commission (NRC) has expressed concerns over the use of gate valves to initiate / control ADS flow. Utilities have also expressed similar concems. Westinghouse held an industry review of the ADS design that focused on the type of valves used. This review concluded that there is uncertainty with the use of gate valves in the AP600 ADS. especially in Stages 2 and 3. It was recommended that the use of other valve types be considered.
Several changes are included to allow for the use of different types of ADS valves. The general approach is to define an ADS arrangement that can accommodate different valve types. The AP600 safety analysis will be performed using bounding ADS valve and system parameters. The LOCA analysis performed iq support of these changes used bounding valvo and system parameters. As shown in Section 4.3, these parameters result in similar or improved plant performance.
l l The SSAR ADS description and piping and instrumentation diagrams (P&lD) (see Appendices A-1 and A-2) show generic valve types. The ADS Phase B tests to support design l
certification have been restructured to provide system behavior data over a wide range of conditions, including flow areas and valve resistances, to provide bounding system parameters (see Section 5). The selection and qualification of ADS valves which meet the bounding system parameters (see Section 6) is not part of design certification. Figure 2-5 shows the revised ADS design and the limiting valve characteristics.
2.9.1 ADS Stage 1 Flow Capacity (DM-09)
Description The minimum effective critical flow area for the ADS stage 1 valves is reduced (see Table 2-1).
Purpose of Modification The reduced flow area provides flexibility in ADS Stage 1 valve selection.
Safety Analysts impact The impact of this change is included in the evaluation presented in Section 4.3 The results of analyses of the DVI break LOCA and spurious ADS analysis show similar or improved performance in comparison to the SSAR (Reference 7.1).
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PRA impact There is no impact on the PRA because these valves are smal! relative to the other ADS valves such that ADS success criteria are not changed.
SSAR impact This change impacts-the following SSAR Chapters: ,
Chapter 5 Reactor Coolant System and Connected Systems Chapter 6 Engineered Safety Features (See Appendix A-1)
Chapter 15 Accident Analyses Test Program impact Core Makeup Tank Separate Effects Tests This design change will have only minimal impact on these tests.
The CMT tests are separate effects tests designed to investigate thermal-hydraulic phenomena within the CMT over a wide range of pressures and temperatures. This change may affect the timing but not the nature of the phenomena being investigated in this test facility. This change does not affect the range of conditions required to be tested.
Integral Systems Tests at SPES-2 Test Facility This design change will have only a minor impact on these tests.
The SPES-2 tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate systems interactions during high pressure transients. The change required for ADS Stage 1 is not significant. A single vatve is used in each stage to initiate and control the blowdown and an orifice is used to represent the second valve. A new orifice will be sized and installed to represent the minimal vatve area, or maximum pressure drop in each stage. This would represent the minimum flow capability of the ADS for stage 1.
Integral Systems Tests at Oregon State University This change will have only minor impact on these tests.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tem cooling behavior. The same changes described above
[ Westingt100Se 2-23 1
AP600 DESIGN CHANGE DESCRIPTION REPORT for SPES 2 for ADS Stage 1 are required for OSU and are not significant. The minimum flow capability of the ADS Stage 1 will be simulated in the test facility.
ADS Phase B Tests The Stage 1 ADS valve procured for this test is consistent with this change. As a result there is no impact on this test.
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E AP600 DESIGN CHANGE DESCRIPTION REPORT 2.9.2 ADS Stage 2/3 Valves (DM 10)
Description The minimum effective flow area for ADS Stage 2 and Stage 3 valves is reduced (see Table 2-1). This reduction in flow area is compensated for by an increase in the fourth Stage ADS flow capability, see subsection 2.9.3. ,
Purpose of Modification This change allows the use of alternative valve types. Figure 2-5 shows two valve options being considered.
Safety Analysis impact The impact of this change is included in the evaluation presented in Section 4.3 The results of analyses of the DVI break LOCA and spurious ADS analysis show similar or improved performance in comparison to the SSAR (Reference 7.1).
PRA impact An evaluation of the impact of this change (Section 3) shows that this change, coupled with the revised ADS Stage 4 design, has only a smallimpact on the AP600 core melt frequency (CMF).
SSAR impact l This change impacts the following SSAR Chapters: j Chapter 5 Reactor Coolant System and Connected Systems Chapter 6 Engineered Safety Features (See Appendix A-1) -
Chapter 15 Accident Analyses l 1
Test Proaram impact )
l Core Makeup Tank Separate Effects Tests This design change will have only a minimal impact on these tests.
The CMT tests are separate effects tests designed to investigate thermal-hydraulic phenomena within the CMT over a wide range of pressures and temperatures. This change may affect the timing but not the nature of the phenomena being investigated in this test facility. This change does not affect the range of conditions required to be tested.
3 Westingt100se 2-25
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5 AP600 DESIGN CHANGE DESCRIPTION REPORT Integral Systems Tests at SPES-2 Test Facility This design change will nave only minor impact on these tests.
The SPES-2 tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate systems interactions during high pressure transjents. The change 'equired for ADS Stages 2 and 3 is not significant. A single valve is used in each stage to initiate and control the blowdown and an orifice is used to represent the second valve. A new orifice will be sized and installed to represent the minimal valve area, or maximum pressure drop in each stage. This would represent the minimum flow capability of the ADS for Stages 2 and 3.
Integral Systems Tests at Oregon State University This change will have only minor impact on these tests.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tem cooling behavior. The same changes described above for SPES-2 for ADS Stages 2 and 3 are required for OSU and are not significant. The minimum flow capability of the ADS for Stages 2 and 3 will be simulated in the test facility.
ADS Phase B Tests This change significantly changes test requirements but has only minor impact on facility hardware.
The ability to accommodate a range of valve sizes and types has a major impact on the test data requirements for the ADS Phase B test program. The tests will provide data for computer code validation of the ADS system. The test matrix has been restructured to provide test data over a range of thermal-hydraulic conditions representing high flow resistance (minimum venting flow) and low flow resistance (maximum loads).
The test conditions are obtained by varying the system configuration and the system conditions upstream of the valve piping package. The system configuration is varied through the use of orifices and flow nozzles to represent valves with different flow characteristics. The fluid conditions are controlled by establishing the appropriate initial conditions in the test and then performing a controlled blowdown through each stage or stages of the ADS valve package. Data will be obtained for both saturated steam and steam / saturated water conditions.
Section 5 discusses the changes to these tests in more detail.
2-26 3 Westingh0tJSe
AP600 DESIGN CHANGE DESCRIPTION REPORT 2.9.3 ADS Stage 4 Valves (DM-11)
Description The ADS Stage four arrangement is changed as shown in Figure 2-4. The possible use of smaller valves in ADS Stages 2 and 3 requires more flow capacity in the fourth stage.
Allowing for the use of squib valves limits the maximum line size. ,
The IST valves currently shown on the Passive Core Cooling System (PXS) P&lD (Appendix A-2) would be retained if air piston gate valves are used. However, if squib valves are used the IST valves would be deleted.
The fourth stage ADS valves are interlocked to preclude their opening at normal RCS operating pressures.
Purpose of Modificattun This change allows the use of attemate vatve types in ADS Stages 2 and 3 and ADS Stage 4.
Figures 2-5 and 2-6 show valve options being considered.
Safety Analysis impac_t The impact of this change is included in the evaluation presented in Section 4.3 The results of analyses of the DVI break LOCA and spurious ADS analysis show similar orimproved performance in comparison to the SSAR (Reference 7.1).
PRA impact An evaluation of the impact of this change (Section 3) shows that this change has only a smallimpact on the AP600 core melt frequency (CMF).
SSAR impact This change impacts the following SSAR Chapters:
Chapter 3 Design of Structures, Components, Equipment and Systems Chapter 5 Reactor Coolant System and Connected Systems Chapter 6 Engineered Safety Features Chapter 7 Instrumentation and Controls Chapter 15 Accident Aralyses Chapter 16 Technical Specifications
[ Westingh0USB 2-27
AP600 DESIGN CHANGE DESCRIPTION REPORT Test Program impact Core Makeup Tank Separate Effects Tests This design change will have only minimal impact on these tests.
The CMT tests are separate effects tests designed to investigate thermal-hydraulic phenomena within the CMT over a wide range of pressures and temperatures. This change may affect the timing but not the nature of the phenomena being investigated in this test facility. This change does not affect the range of conditions required to be tested.
Integral Systems Tests at SPES-2 Test Facility This design change will have only minor impact on these tests.
The SPES-2 tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate systems interactions dunng high pressure transients. With this design change the fourth stage of ADS can now can discharge from both hot legs even w sen a single failure of a fourth stage valve is assumed. This is a direct result of the two flow paths in the revised design. SPES-2 will be reconfigured to allow both fourth stages to discharge into a single header and flow measurement system. The total flow out of both fourth stage ADS lines will be measured simultaneously.
Onfices installed in each ADS fourth stage discharge line will be sized to simulate the total pressure drop and flow area in each line. For a single failure of a fourth stage valve, an orifice representing the pressure drop of a single flow path will be installed.
Integral Systems Tests at Oregon State University This design change Mll have only minor impact on thecc tests.
The OSU tests are integral systems tests to obtain thermal-hydraulic data for computer code validation and to investigate long tern cooling behavior. The same changes described above for SPES-2 for ADS Stage 4 are required for OSU and are not significant. Headering of flow from both ADS Stage 4 lines into a single break measurement system will only occur for double ended break cases. Separate flow measurements from the fourth stage lines is Jossible for the single ended breaks. ,
ADS Phase B Tests This design change has no impact on the ADS Phase B tests.
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AP600 DESIGN CHANGE DESCRIPTION REPORT Table 2-1 Summary of AP600 Design Modifications Mod.No. Description Current AP600 Revised AP600 Reason for (SSAR, Rev.1) Modification DM41 PRHR HX Actuation on High Actuation on CMT Reduces potential for ADS actuation Actuation Logic Pressurizer Pressure or actuation following a SGTR event. Reduces Low Steam Generator potential for pressurizer overfill.
l Level !
I DM-02 Pressurizer No automatic block Automatic block on CMT Reduces potential for ADS actuation i Heater Control actuation following a SGTR event. Reduces Logic potential for pressurizer overfill. __
DM-03 CVS Control Start both pumps on Start /stop 1 pump cn Reduces potential for pressurizer Logic post CMT CMT signal CMT signal at Pressurizer overfill.
actuation levels of 10/20%
DM-04 ADS Stage 1 CMT level,1500 ft' CMT level,1350 ft' Reduces potential for ADS actuation setpoint following SGTR/SLB events.
DM-05 ADS Stage 2/3 CMT level. Stage 1 actuation plus Removes reliance on CMT level setpoints 1300/1000 ft' timers instrumentation during rapid depressurization portion of ADS _
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AP600 DESIGN CHANGE DESCRIPTION REPORT Table 2-1 Summary of AP600 Design Modifications Mod.No. Description Current AP600 Revised AP600 Reason for (SSAR, Rev.1) Modification - - . . - . -
DM-06 CMT Inlet None '
Reduces rate of steam condensation Diffuser ; _ during medium to large LOCA events.
- -a, c.
DM-07 DVI nozzle None Reduces RCS blowdown rate for DVI venturi - -
LOCA DM-08 PRHR HX Inlet 1 MOV 2 manual 2 MOVs Technical Specification improvement Isolation valves
-} k .n)C.
ADS valve type fle bbity DM-09 ADS Stage 1
' Valve Effective L ' - -
Flow Area _ . . _ _ .
_ g 4 ~ ADS valve type flexibility DM-10 ADS Stage 2/3 F Valves DM-11 ADS Stage 4 ADS valve type flexibility
' -. L -! __ _ _ _ . _ . _
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AP600 DESIGN CHANGE DESCRIPTION REPORT TABLE 2-2 REVISED PRHR HX ACTUATION SIGNALS DESCRIPTION SIGNAL SETPOINT ;
PRHR HX Actuation - Low SG narrow range levelin any per SSAR '
(via PMS) SG + low SFW flow after time (165 gpm,60 sec) delay
- Low SG WR levelin any SG per SSAR
- CMT actuation NA
- ADS actuation NA 1
W Westinghouse 2-31
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R::: AP600 DESIGN CHANGE DESCRIPTION REPORT TABLE 2-3 REVISED CVS MAKEUP CONTROL SIGNALS DESCRIPTION SIGNAL SETPOINT Normal RCS Makeup (1) Low pressurizer level start - 0% (1)
(via PLS) relative to programmed level stop - 18% (1) starts makeup; higher level stops makeup Post CMT Actuation RCS CMT actuation + low start - 10% level Makeup (2) pressurizer level starts stop - 20% level (via PLS) makeup, higher level stops makeup Notes:
(1) One CVS pump starts with suction from boric acid and makeup water blended to match RCS boron concentration. Flow is controlled to a fixed flowrate, setpoint has not been selected but will be less than 100 gpm. Start / stop pressurizer levels (percent of level span) are relative to programmed level, approximately 23% / 40% at no power and 40% / 58% at full power.
(2) One CVS pump starts with suction from boric acid tank. Flowrate is not controlled; valve will be full open to provide maximum flow. Start / stop pressurizer levels are absolute values (percent of level span).
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AP600 DESIGN CHANGE DESCRIPTION REPORT TABLE 2-4 REVISED ADS ACTUATION SIGNALS DESCRIPTION f SIGNAL l SETPOINT I 1
ll First Stage ADS i
- CMT actuation signal + Low-1 CMT level in 67% CMT l l - First Stage Actuation, volume l Isolation Valve Actuation either CMT !
- First Stage Control Valve - 1st Stage actuation + time delay 20 sec delay Actuation Second Stage ADS
- Second Stage Actuation, -
1st Stage actuation + time delay 60 sec delay isolation Valve Actuation
- Second Stage Control -
2nd Stage actuation + time delay 30 sec delay Valve Actuation Third Stage ADS
- Third Stage Actuation, -
2nd Stage actuation + time delay 120 sec delay isolation Valve Actuation
- Third Stage Control Valve -
3rd Stage actuation + time delay 30 sec delay Actuation Fourth Stage ADS
- Fourth Stage A Actuation. -
3rd Stage actuation + tirne delay + 120 sec delay isolation Valve Actuation low-2 CMT level in either CMT 20% CMT vol.
- Fourth Stage A Control - 4th Stage A actuation + time delay 30 sec delay Valve Actuation
- Fourth Stage B Actuation. - 4th Stage A actuation + time delay 30 sec delay isolation Vaive Actuation
- Fourth Stage B Control -
4th Stage B actuation + time delay 30 sec delay Valve Actuation W85tiligt100S8 2-33
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 2-3 PRHR HX INLET ISOLATION 6'REMURilER . & ,
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 2-4 REVISED ADS DESIGN y
O
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AP600 DESIGN CHANGE DESCRIPTIOfJ REPORT FIGURE 2-5 ADS STAGE 1,2/3 EXAMPLES -4 C 2-38 W Westirigh0USe
et AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 2-6 ADS STAGE 4 EXAMPLES 4, c.
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iiiii AP600 DESIGN CHANGE DESCRIPTION REPORT 3.0 PRA Evaluation of ADS Design Changes The ADS modelin the AP600 PRA (Reference 7.2) uses motor-operated gate valves in ADS Stages 1,2 and 3 and air-operated gate valves in ADS Stage 4. The results of this PRA, specifically with respect to ADS reliability, were evaluated to determine the impact on core damage frequency (CDF) of ADS hardware modifications. .
This evaluation started with inspection of the list of dominant accident sequence cutsets shown in Table F-7 of the AP600 PRA Report. Table F-7 reports the top 100 core damage cutsets and many cutsets, starting with cutset number 5, include failure of the ADS. However, in all of these top 100 core damage cutsets that include ADS failure, the ADS failure is a result of failure to signal the ADS to function, which includes both failure to automatically actuate and failure of the operator to actuate the ADS. The PRA computer output files were reviewed to assess more core damage cutsets to determine where ADS hardware failures first appear. The first ADS hardware failure cutsets appeared as a set of six, starting with number 135, each with a CDF of 2.15E-10. The next hardware associated failure of the ADS is number 175 with a core damage frequency of 1.52E 10. In the top 100 cutsets, ADS failure due to not generating an ADS actuation signalis involved in cutsets that contribute approximately 4.6E-8 to the core damage frequency .
This evaluation continued by performing sensitivity studies using the ADS fault trees that had the most influence on CDF. The four fault trees selected are ADS, ADA, ADN and ADC from the PRHR HX tube rupture event. These are the most important because the PRHR HX tube rupture event was the initiating event for 10 of the first 11 core damage sequences that involved ADS hardware failures. For this sensitivity study, new reference values for system reliability for each of the ADS fault trees were calculated to account for the addition of pressure switches that prevent Stage 4 ADS valves from opening at high RCS pressure.
These fault trees were then evaluated for two different ADS configurations. These configurations are given in Table 3-1. The results of this study are also shown in Table 3-1.
On the basis of changes in unavailability, the various hardware design options evaluated have minimal effect on ADS reliability. As shown in the table, the maximum increase in fault tree unavailability, due to hardware changes, is less than a factor of two.
As indicated above, ADS hardware failures (6*2.15E 10+1.52E-10) contributed about 1.4E-9 of the total CDF while ADS actuation failures contributed about 4.6E-8. That is, hardware failures account for roughly three percent of the CDF cutsets involving the ADS. Even if the ADS hardware failure rate is doubled, the impact on the total AP600 CDF would be insignificant.
[ W85tingh00Se 3-1
AP600 DESIGN CHANGE DESCRIPTION REPORT TABLE 3-1 ADS FAULT TREES UNAVAILABILITY lj
, ~l i
SSAR ADS Configuration #1 ADS Configuration #2 l ADS Stages 2/3 MO gate MO globe MO globe l ADS Stage 4 AO gate AO gate Squib Fault Tree ADN 2.77E-3 2.77E-3 2.77E-3 Fault Tree ADC 3.32E-3 3.32E-3 3.32E-3 Fault Tree ADS 2.91 E-5 4.86E-5 4.93E-5 Fault Tree ADA 5.79E-4 6.09E-4 6.09E-4 3-2 W westmghouse
=
APoGO DESIGN CHANGE DESCRIPTION REPORT 4.0 Integrated Evaluation of Safety Analysis impact Chapter 15 of the SSAR (Reference 7.1) presents the design basis accident analyses performed for the AP600. Accidents are classified therein according to the type of event; increase in heat removal from the primary system, decrease in heat removal by the secondary system, etc. The passive safety system design changes affecting the CMTs and ADS impact some postulated accident events more than others. The SSAR Chapter 15 analyses were revi3wed with respect to the design changes, and those which were determined to be significantly affected have been reanalyzed modeling the design changes pertinent to that particular event.
Assumptions made in the SSAR to perform analyses of the selected transients were reviewed in light of the design changes. The events determined to have the most significant impact from the design changes are steami!ne break, SGTR and LOCA. These events were reanalyzed and the results are presented in the following sections.
4.1 Steamline Break introduction This section assesses the impact on the AP600 steamline break analysis of the design changes identified in Section 2. Four of the design changes have been identified as having potential impact on the safety analysis results for the steamline break accident. These are as follows:
PRHR HX Actuation Sicnal l Delete the current signals that actuate the PRHR on high pressurizer or high steam generator level. Add a new signal that initiates PRHR operation on CMT actuation.
ADS Staoe 1 Actuation i i
Reduce the ADS Staor actuation CMT level setpoint. The corresponding CMT volumes are 1E and 1350 ft' for the old and new setpoints, respectively.
CMT Inlet Diffuser install a CMT inlet diffuser where none previously existed.
I Westingh0Use 4,3
2
.. i R. -
AP600 DESIGN CHANGE DESCRIPTION REPORT i
DVI Nozzle Venturi install a DVI nozzle venturi where none previously existed. ,
in the discussion that follows, each of these changes is evaluated with regard to the steamline l break accident. .
PRHR HN Actuation Signal The steamline break accident does not rely on either the high pressurizer or high steam generator level functions to initiate operation of the PRHR. For the limiting steamline break transients, any PRHR actuation would be predicted to occur on the h steam generator water level signal. The nature of the steamline break transient is such that, at least during the initial phases for the largest breaks, the water levels in both the pressurizer and the steam generators are decreasing. The pressurizer level falls because of the RCS cooldown produced by the steamline b 3ak accident. As the transient proceeds, the addition of borated '
coolant from the CMTs and/or the accumulators acts to restore water levelin the pressurizer.
The primary effect on steam generator water level is the loss of inventory through the break.
For some small break cases, conservative analysis assumptions with regard to startup i feedwater may actually produce a slight increase in inventory for the intact steam generator.
The predicted steam generator water levels never approach high level setpoints.
In addition, since the conservative analysis assumptions for the core response steamline i break accident maximize the severity of the RCS cooldown, the cases presented in AP600 '
SSAR Sections 15.1.4 and 15.1.5 model PRHR actuation at time zero. This eliminates any analytical sensitivity to a change in the signals that might actuate the PRHR at some point later in the transient. Therefore, the proposed changes in the signals that actuate the PRHR HX are bounded by the AP600 SSAR safety analysis for the steamline break accident.
ADS Stage 1 Actuation The initiating signal for Stage 1 ADS is CMT water level. A basic design requirement for the ,
AP600 is that the ADS not actuate for Condition 2 accidents. On this basis, it would be acceptable for the ADS to actuate during Condition 3 and 4 non-LOCA events. However, it is a design goal for the AP600 that ADS not actuate during any of the non-LOCA transients. ;
The results for the limiting core response steamline break, as reported in Section 15.1.5 of the t AP600 SSAR, indicate a relatively small reduction in CMT water volume during the transient.~ l From an initial value of 2000 ft', the minimum CMT water volume predicted for the SSAR is l 1890 ft', which provides significant margin to the stage 1 ADS actuation analysis setpoint of !
1500 ft' that has been used. However, the assumptions used in the SSAR analysis did not maximize the potential for ADS actuation. As a result, supplemental analysis has been 62 Westinghouse i
55 50 AP600 DESIGN CHAtJGE DESCRIPTION REPORT performed with the spec:fic intent of addressing ADS actuation for the steamhne break accident.
The results of tnis analysis, which are summarized in Table 4.1-1 and in Figures 4.1-1 through 4.1-14, demonstrate that for a limiting steamline break the minimum predicted CMT water volume is 1590 ft) This resu!t indicates a 90 ft' margin to the SSAR Stage 1 AC1S actuation setpoint (CMT volume of 1500 it'). The setpoint change described in Section 2 (CMT volume of 1350 ft ) increases the margin to 240 ft'. The significant reduction in minimum CMT water 3
volume predicted for the current analysis relative to the referenced SSAR steamline break 2
case is primarily caused by two model changes. This analysis assumes a 0.4 ft break rather 2
than the 1.4 ft break in the SSAR analysis since sensitivity analyses show this to be more limiting with respect to minimum CMT level. Also the current analysis conservatively models condensation in the CMT that is only 10 percent of that applied in the SSAR case.
Verification of the condensation model components will occur .cnce the AP600 CMT test data is available.
In summary, the reduction in the Stage 1 ADS actuation setpoint to 1350 ft' helps to preclude ADS actuation during a limiting steamline break transient.
CMT inlet Diffuser The diffuser does not affect CMT draindown behavior when steam comes from the ;
pressurizer; as is the case for the steamline break event.
DVI Nozzle Venturl As described in Section 2.7, the installation of the DVI nozzle venturi should have no impact ,
on the steamline break analysis for the AP600. The presence of this venturi is intended to )
mitigate the effects of a DVI line LOCA. The total resistance from the CMTs and accumulators to the RCS will be unchanged because of orifices in the discharges of these tanks. As a result, the steamline break results for the AP600 will be unaffected.
Conclusions )
In summary, the AP600 design changes defined in Section 2.0 are acceptable with respect to the SSAR steamline break analysis for the AP600. Specific models and assumptions associated with the analysis for this accident will be verified against the AP600 tests.
[ Westingh0US8 4-3
~
:- =
5 AP600 DESIGN CHANGE DESCRIPTION REPORT TABLE 4.1-1 SEQUENCE OF EVENTS FOR STEAMLINE BREAK CASE BREAK IN LOOP 2 & OFFSITE POWER AVAILABLE TIME
^
EVENT (Seconds) 0.4 ft2double-ended rupture steamline break occurs O Reactor and turbine assumed tripped 0 PRHR heat exchanger assumed to actuate O Low steamline pressure SIS setpoint reached 2.5 Steamline isolation 14.5 Feedwater isolation 14.5 Reactor coolant pumps trip 17.5 CMT actuation 24.5 Startup feedwater isolated to all SGs 28.9 Accumulators actuate 1046 Fautted steam generator blowdown ends -2600 W Westinghouse
i 1
55 n=
- I I:q AP600 DESIGN CHANGE DESCRIPTION REPORT 1 FIGURE 4.1-1 NUCLEAR POWER TRANSIENT, STEAM SYSTEM PIPING FAILURE d
2 18 +
d
! 16-5 14-t
.j 12 c
x 1
x
{= .08-3 .C6 "
3 H
5 .04- i z ,
.02 - '
O 4 10 1 102 103 10 100 TiWE (SEC) ,
1 W WEStinch0USB a 4-5
= AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1-2 CORE HEAT FLUX TRANSIENT, STEAM SYSTEM PlPING FAILURE 2
, 13
=
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- 225 103 10'
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iEE AP600 UESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1-4 CORE AVERAGE TEMPERATURE TRANSIENT, STEAM SYSTEM PIPING FAILURE
'30 63C -
$0
? 500. -
E
- d 3 400 -
e) 8 320. -
~$
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$ 200. -
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1-5 REACTOR VESSEL INLET TEMPERATURE, STEAM SYSTEM PIPING FAILURE 630.
530.-
2 i Unfaulted Lop d 430.-
Unfaulted Lop z
~
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i Faulted Lay
, 2 30. +
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AP600 DEstGN CHANGE DESCRIPTION REPORT FIGURE 4.1-6 RCS PRESSURE TRANSIENT, STEAM SYSTEM PIPING FAILURE f
2750 2500 2250 i 2000.
~
.n 1750 a
s A
1500.
D d 1250 d
o 1000.
J 750 500. ,
250.
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nf AP600 DESIGN CHANGE DESCRIPTION REPORT J FIGURE 4.1-7 PRESSURIZER WATER VOLUME TR ANSlENT, STEAM SYSTEM PIPING FAILURE 4
2500 225C 3
- 200c -
$ 1750.
> 1500.
2 a
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1250 +
$ 1000 2
2 3 750.
E 500.
250.1
~
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W Westinghouse 4 33
AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1-8 CORE FLOW TRANSIENT, STEAM SYSTEM PIPING FAILURE 1 4
- 1 2+
=
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d s-0 x
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1-9 FEEDWATER FLOW TRANSIENT, STEAM SYSTEM PIPING FAILURE 16
, t 4 E
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liis AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1-10 CORE BORON TRANSIENT, STEAM SYSTEM PIPING FAILURE
'00.
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$ 500 7
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5 !
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l AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1-11 STEAM PRESSURE TRANSIENT, STEAM SYSTEM PIPING TAILURE 120C.
1000 t
$ Unfaulted Loop I i 800.
U S
$ 500.
E E
{ 400.
20C- Faulted Loop
~'00 to1 102 103 10' TruE (SEC)
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1 12 STEAM FLOW TRANSIENT, STEAM SYSTEM PlPING FAILURE 16 1 4-d Z
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TtWE (SEC) 4-16 W Westinghouse
= -
AP600 DESIGN CHANGE DESCRIPTION REPORT A*=.5 FIGURE 4.1-13 CORE MAKEUP TANK INJECTION FLOW, STEAM SYSTEM PIPING FAILURE 200.
15 0. +
'd 16 0 . +
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- d J
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l
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.1-14 CORE MAKEUP TANK WATER VOLUME, STEAM SYSTEM PIPING FAILURE 2500 .
I 2250 A i
i 2000 '
a
~ 1750 +
$ 1500.
" 1250 j 1300.
5 750.
500.
250.
0 oo 10 1 102 303 10 4 TiwE (SEC) 4-18 W Westinghouse
.5 EE AP600 DESIGN CHANGE DESCRIPTION REPORT 4.2 Steam Generator Tube Rupture introduction Section 15.6.3 of the AP600 SSAR presented the design basis SGTR event and the corresponding radiological dose releases for the AP600. The results show that the overfill protection logic and the passive system design features will initiate automatic actions that stabilize the RCS in a safe condition while preventing steam generator overfill and ADS actuation. The resulting radiological doses released to the environment for this accident are within the 10 CFR 100 limits.
The impact of the PXS design changes on the progression of a SGTR event has been evaluated. The following aspects of a SGTR event were included in this evaluation:
offsite radiological dose analysis, margin to ADS actuation, and margin to SG overfill.
SGTR Evaluation The identified design changes that impact the SGTR transient are:
- 1) initiation of the PRHR on a CMT actuation signal, rather than on a high-high steam generator level signal,
- 2) changes to the CVS control logic,
- 3) incorporation of a diffuser at the inlet to the CMT,
- 4) addition of a pressurizer heater shutoff on a CMT actuation signal, and
- 5) lower CMT level setpoint for ADS actuation.
Initiation of the PRHR on a CMT actuation signal, rather than on a high-high steam generator signal as presented in the SSAR, impacts the timing of the various phases of the transient, as well as the magnitude of break flow and steam release (the major factors used in determining the offsite radiological doses). The design basis SGTR analysis was reanalyzed with the PRHR initiated on a CMT actuation signal rather than on a high-high steam generator level signal. The system responses for this transient are shown in Figures 4.2-1 to 4.2-8. The figures parallel the figures in the SSAR, described in Section 15.6.3.2.1.3. The results of this analysis show that the overall progression of the transient is not changed significantly. This is demonstrated in Table 4.2-1, where the sequence of events for the SSAR transient (with PRHR initiated on high-high steam generator level) is compared to the same transient with the PRHR initiated on a CMT actuation signal, and in Figure 4.2-9, where break flow (the major
[ Westingh0tise gg
AP600 DESIGN CHANGE DESCRIPTION REPORT indicator of the progression of the transient) is shown for the two transients. This analpis demonstrated a reduction in break flow and steam releases, the major contnbutors to the offsite radiological dose calculation. It is concluded that the change in PRHR initiation signal will result in reduced offsite radiological doses.
This reanalysis conservatively does not include the change to the CVS. The cha,nge to the CVS makeup program, if modeled, would provide a benefit via reduced pnmary to secondary break flow resulting from reduced injection from the CVS.
For the SSAR, SGTR sensitivity analyses were performed to demonstrate that the ruptured steam generator will not overfill. This conclusion has been verified for the identified design changes. There are several reasons for this conclusion. Most significantly, the modified CVS control logic and the revised PRHR actuation logic will reduce break flow throughout the transient and cause earlier initiation of the RCS cooldown and increased margin to overfill.
Separate sensitivity analyses have been performed to verify that with the identified changes a design basis SGTR would not result in ADS actuation. These analyses are performed with conservative assumptions that reduce margin to the low CMT level ADS actuation setpoint.
The key assumptions include no CVS makeup, reduced CMT condensation, and the failure of the faulted steam generator power-operated relief valve (PORV) in the open position.
Neglecting the CVS makeup flow tends to reduce the RCS pressure which increases the CMT injection. Reducing the CMT condensation will enhance CMT injection while reducing makeup flow into the CMT from the pressurizer connection line, thereby maximizing the level reduction in the CMT. For this study the nominal CMT condensation predicted by the lot- t i R2 code model has been reduced by a factor of 10 to provide assurance that the analysis is bounding.
The failure of the faulted SG PORV .esults in further RCS depressurization, as the secondary pressure drops below the safety valve setpoint. The results show that the minimum CMT water volume predicted from the sensitivity analysis is well above the low CMT level ADS setpoint (1350 ft').
The addition of a pressurizer heater block, either automaticalty on a CMT actuation signal or manually, provides a benefit for SGTR offsite dose and overfill analyses by reducing heat addition to the pressurizer, thus allowing for quicker cooldown and depressurization.
Conclusion From the results of the SGTR analyses it is concluded that the design changes will not result in steam generator overfill or ADS actuation, and that the offsite radiological doses for the limiting design basis SGTR reported in the SSAR continue to be bounding,
& 20 Westingh0tJSe
~
AP600 DESIGN CHANGE DESCRIPTION REPORT TABLE 4.2-1 l SEQUENCE OF EVENTS SGTR WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL COMPARED TO
}
I SSAR ANALYSIS WITH PRHR INITIATED ON HIGH SG LEVEL l
SSAR PRHFIoT PRHR on CMT High SG Actuation Level Signal Time Time Event (seconds) (seconds)
Double ended SGTR 0 0 One CVS pump actuated and pressurtzer heater tumed on 0 0 Reactor trip on low pressurizer pressure 1074 1074 Main feedwater pumps assumed to trip and begin coastdown 1074 1074 Startup feedwater initiated (includes maximum delay) 1150 1150 l CMT actuation signal on low-1 pressurizer pressure 1411 1411 Additianal CVS pump initiated 1412 1412 CMT injection begins (includes maximum delay) 1433 1433 PRHR initiated on CMT actuation signal (includes maximum 1471 delay)
Startup feedwater to faulted SG throttled to maintain low SG 2526 2472 narrow range level setpoint l Fautted SG PORV falls open when secondary level approximately 3706 3742 reaches high-2 SG narrow range level setpoint Fautted SG PORV block vatve closes on low steam line pressure 4346 4000 signal (includes valve delay tine)
Steam release from faulted and intact SG PORVs terminated 4347 4001 CVS and startup feedwater pumps isolated on high-2 SG narrow 5174 4576 range level setpoint 1 PRHR initiated on high-2 SG narrow range level setpoint (includes 5234 l valve opening delay time) i Break flow terminated and stable condition reached 10000 10000
[ Weslingh00Se gg
I I
=:
=: AP600 DESIGN CHANGE DESCRIPTION REPORT I I
FIGURE 4.21 PRESSURIZER WATER VOLUME FOR SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL 1200 y 1000 d
800 +
b E
's 600 2
rg
] 400 S
E 200.
C.
D. 2000. 4000. 8000. 2000 1E+05 .12E*05 Time (Seconds) 4-22 W Westinghause
-7 5E AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.2-2 PRIMARY PRESSURE FOR SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL 2500 2250.
2000 b
y 1750 t
y 1500 b
1250 c.
g 1000.
8 750.
500 250.
D.
D. 2000. 4000. 5000. 8000. .1E+05 .12E.05 Time (Seconds) l W85tillgi100S8 4-23
- e l
AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.2-3 SECONDARY PRESSURE FOR SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL 1400 1200.
I 1000.
b E
~
800. m a
ECO.
2 -
400.
200.
O.
D. 2000. 4000. 6000 9000 1E*05 42E405 Time (Seconds) l W25tingh00Se
l I
l HEI EEI AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.2-4 PRIMARY TO SECONDARY BREAK FLOW FOR SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL f
50 i 40 b
d s 30
_2 8
.J 20 s
u C 10.
g '- ' vi m
'O . 2000. 4000. 6000. B000 .1E+05 .42E*05 Time (Seconds)
W85tingt10US8 4 25
AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.2-5 RUPTURED STEAM GENERATOR WATER VOLUME FOR SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL 8000
, 7000 b
6000 5
5000.
3 4000 t ti a 3000.
E g 200D at 1000 c.
O. 2000. 4000. 6000. 9000 .1E+05 12E405 Time (Seconds) 4-26
[ Westingh0050
l
~~
AP600 DESIGN CHANGE DESCRIPTION REPORT B.a ..
FIGURE 4.2-6 RUPTURED STEAM GENERATOR STEAM RELEASE FOR SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL 500.
500.
O d
s 400 s
d O 300.
d E
LJ
] 200-100.
D.
O, 2000. 4000. 6000. 9000. 1E+05 .12E+05 Time (Seconds)
W
=
westinghouse 4-2 ,/
f =
AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.2-7 RUPTURED LOOP CVS AND CMT INJECTION FLOW FOR SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL g 70 d
50 S,
5 50 E
U 40.
c
] 30 as O g J 20 b
tr 10.
O I j
'O. 2000. 4000 6000 9000. .1E+05 12E 05 Time (Seconds) 4-28 W Westingh0US8
AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.2-8 CMT WATER VOLUME FOR SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL 2250.
2000- v v i750,
[
~ 9503 D
d 1250.
3C d 1000 y 750.
O 500 250, 0
0 2000. 4000. 6C00. 9000. .1E+05 .1PE+05 Time (Seconds)
WE5tinghouse 4-29 l
l AP600 DESIGN CHANGE DESCRIPTION REPORT E==
FIGURE 4.2 9 ,
PRIMARY TO SECONDARY BREAK FLOW FOR l SGTR ANALYSIS WITH PRHR INITIATED ON A CMT ACTUATION SIGNAL COMPARED TO THE SSAR ANALYSIS WITH PRHR INITIATED ON HIGH SG LEVEL 1
l so l b
40 SSAR !
SSA G l i & '
- g NEW d :
.J 20 t
M g NEW ,
m 10 .
O ' - - . , _ _
-10.
O, 2000. 4000. 6000. B000. .iE+05 .12E+05 Time (Seconds) t e'
i r
4 0 W Westinghause .
I l
l l
I i E AP600 DESIGN CHANGE DESCRIPTION REPORT 4.3 Loss of Coolant Accidents introduction The AP600 SSAR LOCA analyses are a spectrum of postulated break sizes ranging from a one-inch equivalent diameter break to double-ended hot leg (DEHLG) and cold leg (DECLG) guillotine breaks. The DECLG cases exhibit the limiting calculated peak cladding, temperature (PCT) values. In these large break LOCA cases, minimal CMT injection occurs before PCT is calculated, and the ADS does not actuate until long after PCT occurrence. Thus, large break f LOCA performance is unaffected by these design changes.
l Two of the small break LOCA cases were chosen for reanalysis to determine the impact of the passive safeguards systems design changes. LOCA analyses for postulated double-ended direct vessel injection line (DEDVl) break and inadvertent ADS actuation events have been performed to investigate the CMT and ADS design changes. The DEDVI and inadvertent ADS cases represent limiting small break LOCAs with respect to providing safety injection delivery to limit core uncovery and ADS depressurization capability to achieve IRWST injection, respectively. The NOTRUMP computer code (References 7.3 and 7.4), used in the SSAR analysis, was utilized. Only safety systems were modeled.
The NOTRUMP AP600 input was set for the SSAR analyses to comply with the standard Westinghouse Small Break LOCA Evaluation Model methodology (Referete 7.4). For better representation of the AP600, the following changes were made to the SSAR model:
- 1) The top node of the CMT was increased from 10 percent to 20 percent of the CMT l volume. Since the larger top node is expected to bound the CMT inlet mixing region with a diffuser in place, the 20 percent top node volume is selected as appropriate and conservative.
- 2) The double-link horizontal stratified flow links are no longer used for the surge line connections. Rather, single links are adopted because they are more appropriate for the surge line flow path.
- 3) The design changes to safety-related passive safety features are modeled. The change to the CVS controls does not affect the SSAR LOCA analyses.
- 4) A multi-node PRHR representation of the heat exchanger (eight fluid nodes) is used.
PRHR HX actuation occurs on a Safeguards ("S*) signal. Standard condensation heat transfer correlations are applied when primary side steam condenses in the PRHR.
[ W85tingt10tJ58 4-31 )
l l
- s is AP600 DESIGN CHANGE DESCRIPTION REPORT The DEDVI case exhibits core uncovery in the SSAR analysis. The actuation of ADS depressurizes the RCS to accumulator actuation pressure, thereby increasing safety injection flow. Therefore, the lirrung single active failure for the DEDVI break is taken as failure of a set of one first stage and one third stage ADS valves to open on demand. For the inadvertent ADS case, reducing the venting capability of the RCS constitutes the limiting single failure.
Thus, in the inadvertent ADS case, failure of one of the four fourth-stage ADS valves to open on demand is postulated. Furthermore, in both of the cases the fourth stage ADS vent area was arbitranty reduced by 15 percent to provide a bounding calculation.
Double-ended Direct Vessel injection Line Break (DEDVI) Results This case models the double-ended rupture of the DVI line at the nozzle into the downcomer.
The broken loop injection system (consisting of an accumulator, a CMT, and an IRWST delivery line) is modeled to spill completely out of the break. The injection line break evaluates the ability of the plant to recover from a moderately large break with only half of the total Emergency Core Cooling System (ECCS) capacity available. The venturi installed in the AP600 DVI nozzles provides a four-inch diameter flow restriction, which has been modeled in NOTRUMP. A discussion of the results follows, including comparisons with the SSAR DEDVI case in Table 4.3-1. The AP600 system depressurization and break flow rates are significantly lower than in the SSAR case due to the smaller break area from the vessel downcomer through the venturi.
The break is assumed to open instantaneously at 0 seconds. The accumulator on the broken loop starts to discharge via the DVI line to the containment. The subcooled discharge from the downcomer nozzle (Figure 4.3-1) through a four-inch diameter venturi causes a rapid RCS depressurization (Figure 4.3-2), and a reactor trip signal is generated at 7.2 seconds. The "S" signalis generated at 8.7 seconds and following a 1.2 second delay, the isolation valves on the CMT tank delivery and cold leg balance lines begin to open. The "S" signal also causes closure of the main feedwater isolation vatves after a five second delay and trips the reactor coolant pumps after a 16.2 second delay. The opening of the PRHR isolation valve on an "S" signal starts the flow through the heat exchanger. The broken loop CMT discharges directly to the containment (Figure 4.3-3), and a small circulation flow provides injection from intact loop CMT (Figure 4.3-4).
As the pressure falls, the RCS fluid saturates and at about 23 seconds (vs. 6 seconds in the SSAR case) a mixture level forms in the upper plenum and falls to the hot leg elevation (Figure 4.3-5). The upper parts of the RCS start to drain (Figures 4.3-6 and 4.3-7), and a mixture level forms in the downcomer at about 115 seconds (vs. 20 seconds in the SSAR)
(Figure 4.3-8) and falls to the elevation of the break. Two-phase discharge then occurs from WB5tingh0Use
l l
1
= l
~
AP600 DESIGN CHANGE DESCRIPTION REPORT l l
the downcomer side of the break (Figure 4.3-1). A cornparison of the venturi and SSAR ,
DEDVI cases is provided in Table 4.3-1. ;
At about 85 seconds (vs. 20 seconds in the SSAR) the fluid at the top of the broken loop CMT saturates and a level forms and starts to fall (Figure 4.3-9). The first stage ADS setpoint is reached at 140 seconds, and after an appropriate delay. one first stage path is opened. The i ensumg. steam discharge from the top of the pressurizer (Figure 4.3-10) increases the RCS l' depressurization rate. Because this depressurization is beneficial in delivering safety injection water, the single active failure assumed for the DEDVI is the failure of a first and third stage ADS valve to open.
In the SSAR DEDVI case, with no venturi present and a much higher break flow, at about 60 seconds insufficient liquid remains in the core and upper plenum to sustain the mixture level. l The mixture level, therefore, starts to collapse as the core dries out and the mixture level falls ,
to a minimum at 100 seconds, uncovering the core. In the current " design change" case, the -
two second stage ADS valves begin to open at 234 seconds, following the timer delay '
between the actuation of the first two stages of the ADS. At 206 seconds the intact loop accumulator starts to inject into the downcomer (Figure 4.3-11) causing the mixture level in ,j the downcomer to slowly rise (Figure 4.3-8); the mixture level in the upper pienum never falls ;
below the hot leg elevation during the entire DEDVI transient. One third stage ADS valve -
opens at 354 seconds because of the time delay of 120 seconds for the actuation of this stage of the ADS. At 193 seconds the broken loop CMT level reaches the fourth stage ADS setpoint but the fourth stage ADS valves do not open until 474 seconds because the minimum '
time delay is 120 seconds between the actuation of the final two stages of the ADS. Two-phase discharge ensues through the fourth stage path (Figures 4.3-12 and 4.3-13). By 229 ,
seconds the broken loop CMT empties (Figure 4.3-9). l l
At about 304 seconds the fluid at the top of the intact loop CMT saturates, and the mixture j level in the tank starts to fall slowly (Figure 4.3-13). CMT and accumulator injection after 304 seconds causes the downcomer mixture level to rise slowly (Figure 4.3-8). The levelin the upper plenum is maintained up to at least the hot leg elevation (Figure 4.3-5) throughout the i break transient. After the accumulator empties at about 570 seconds stable, but decreasing, injection continues from the intact loop CMT as the RCS pressure declines slowly. At 1178 :
seconds, the intact loop CMT has not yet emptied (Figure 4.3-13), yet the RCS pressure has .l fallen to the point that IRWST injection begins. Stable injection from the tanks occurs at a l rate of about 120 lb/s until the CMT empties. This flow is greater than the break and ADS j flows, resulting in a slow rise in RCS inventory (Figure 4.3-14). The RCS inventory continues -
to increase when the IRWST provides the only injection. The minimum RCS mass inventory -
of 106,000 lbs is significantly greater than the corresponding SSAR DEDVI break value of 90,000 lbs.
i 3 Westingholjse 4-33
as ma AP600 DESIGN CHANGE DESCRIPTION REPO Inadvertent Actuation of Automatic Depressurization System Results An inadvertent ADS signalis spunously generated and the first stage ADS valves open. The plant, which is operating at 100 percent power. is depressurized via the ADS alone. Only safety-related systems are assumed to operate in this analysis. The second and third stage ADS valves actuate based on the design time delays. At the 20 percent tank level. the fourth stage ABS valves, which are on the hot legs. receive signals to open. Three of the four fourth stage ADS paths are assumed to open; one of the paths fails to open as the assumed single active failure.
The scenario analyzed is the same inadvertent ADS actuation that is considered in the SSAR.
The ADS stages 2 and 3 actuate according to the timer sequence. The sequence of events for the transient is given in a table of comparison with the SSAR case, Table 4.3 .
The transient is initiated by the opening of the two first stage ADS paths. The total throat area of the valves is 9.2 square inches. Reactor trip, reactor coolant pump trip and safeguards signals are generated via the pressurizer low pressure signals with appropriate delays. Upon generation of the reactor trip signal the main steam isolation valves begin to close, after a 2 second delay. Five seconds after an "S* signal the main feedwater isolation valves begin to close. The opening of the ADS valves and the reduction in core power due to reactor trip causes the primary pressure to fall rapidly (Figure 4.3-15). Flow of fluid toward the open ADS paths causes the pressurizer to fill by about 30 seconds (Figure 4.3-16), and the ADS flow becomes two-phase. A level begins to form in the upper plenum and drops to the hot leg elevation (Figure 4.3-17) at about 100 seconds. The safeguards signal opens the valves isolating the CMTs and injection of cold water begins (Figure 4.3-18); the PRHR is also actuated by the "S* signal. The mixture level in the CMTs is constant until about 440 seconds, then the tanks begin to drain (Figure 4.3-19). The reactor coolant pumps begin to coastdown due to an automatic trip signal following a 16.2 second delay. The revised design, by actuating ADS Stages 2 and 3 via timers, has an impact on this postulated event. As the table indicates, second stage ADS actuation at 70 secoads accelerates the transient; the accumulators now empty earlier than in the SSAR analysis because the earlier ADS flow causes the primary pressure to fall rapidly (Figure 4.3-15). At about 200 seconds following the ADS actuation, enough mass has been discharged that a mixture level forrns in the downcomer (Figure 4.3-20). CMT !njection flow is diminished by accumulator flow; emptying of the accumulator results in increased CMT injection flow. The mixture level in the CMT falls steadily after about 600 seconds.
The levels in the CMTs eventually reach the fourth stage ADS setpoint. Vent paths opened from the hot legs begin discharging fluid. The increased depressurization reduces the flow from ADS Stages 1,2 and 3 (Figures 4.3-21 and 22). The single active failure assumed is W Westinghouse
- .- -. -. . - - . - . .~ . .
i
. =
AP600 DESIGN CHANGE DESCRIPTION REPORT that one of the four fourth stage ADS valves fails to open, maximizing the resistance to j depressurizing the RCS to achieve IRWST injection.
i The reduced flow through ADS Stages 1-3 allows the pressurizer level to fall, and these stages begin to discharge only steam after 1900 seconds. By 1967 seconds the CMTs are ;
empty and delivery ceases, and at 2176 seconds the RCS pressure has_ fallen enough to allow gravity drain from the IRWST to begin (Figure 4.3-23). The calculation was stopped with i
a quasi-steady state condition existing in the RCS with the IRWST delivery exceeding the ADS flows (which are removing the decay heat) and the RCS inventory slowly rising. Core i uncovery does not occur, the upper plenum mixture level remaining well above the core elevation throughout (Figure 4.3-17). In the SSAR case, nine minutes elapsed after CMT ,
I empty time before IRWST injection could begin; during this time the core mixture level fell below the hot leg bottom elevation, ,
in the revised ADS design case, core mixture level remains within the hot leg boundary.
Once the ADS valves become fully effective by virtue of venting steam alone rather than a j two-phase mixture (Figures 4.3-24 and 25), IRWST injection begins within a matter of j seconds. The minimum RCS mass inventory in this revised ADS design case remains above (Figure 4.3-26) the SSAR case minimum inventory value of 100,000 lbs mass throughout.
The inadvertent opening of the ADS transient analysis confirms that the minimum venting area of AP600 capably depressurizes the RCS to the IRWST delivery pressure. The ADS area is sufficient to depressurize the RCS even assuming the failure of one fourth stage ADS valve to open. Appendix K decay heat is applied, which conservatively over estimates the core steam i generation rate. Even under these limiting conditions, IRWST injection is obtained readily,'
and the core remains covered such that no cladding heatup occurs. !
Conclusions '
The two limiting small break LOCA cases from the AP600 SSAR have been reanalyzed with NOTRUMP to investigate the ADS /CMT design changes. With the venturi modeled in the vessel injection line, no core uncovery is predicted for the DEDVI break case. The inadvertent ADS actuation case depressurizes the primary syst_em much more rapidly than does the SSAR case, due to the use of timers for the second and third stage actuation. The increased !
capacity of the ADS fourth stage valves allows IRWST injection to be achieved more readily j than in the SSAR inadvertent ADS case. Overall, these cases demonstrate that the CMT/ ADS !
design changes improve plant performance with respect to the SSAR analysis for the identified bounding small break LOCA events.
[ We51iflgh0tJSe 4-35
gn: ::-
=i AP600 DESIGN CHANGE DESCRIPTION REPORT TABLE 4.3-1 DEDVI BREAK SEQUENCE OF EVENTS TABLE I
j Case SSAR j New design with venturi Break open 0.0 seconds l 0.0 seconds ,
Reactor trip signal 3.5 seconds 7.2 seconds "S" signal 4.2 seconds 8.65 seconds Reactor coolant pumps start 20.4 seconds 24.85 seconds to coast down
! ADS stage 1 flow starts 75 seconds 164.2 seconds Accumulator injection starts 117 seconds 206 seconds l
l ADS stage 2 flow starts 135 seconds 234.2 seconds l
ADS stage 3 flow starts 255 seconds 354.2 seconds ADS stage 4 flow starts 375 seconds 474.2 seconds IRWST injection starts 1839 seconds 1178.7 seconds W85tiflgh00Se l
= u E
AP600 DESIGN CHANGE DESCRIPTION REPORT TABLE 4.3 2 INADVERTENT ADS ACTUATION
- r , d Case SSAR i New Design [
m!
ADS stage 1 flow starts 0.0 seconds 0.0 seconds j'j Reactor trip signal 28.5 seconds 25.6 seconds "S signal 33.5 seconds 28.97 seconds Reactor coolant pumps start to 49.7 seconds 45.2 seconds coast down Accumulator injection starts 633 seconds 204 seconds PRHR Actuation 726 seconds 30.2 seconds ADS stage 2 flow starts 916 seconds 70 seconds Accumulator empty 1325 seconds 565 seconds ADS stage 3 flow starts 1390 seconds 190 seconds ADS stage 4 flow starts 1902 seconds 1617 seconds Core make up tank ernpty 2260 seconds 1967 seconds IRWST injection starts 2792 seconds 2176 seconds l
l l
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-1 BREAK LIQUID FLOW, DEDVI BREAK 1
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-3 LOOP 1, CMT TO DVI FLOW, DEDVI BREAK
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l FIGURE 4.3 5 CORE STACK MIXTURE LEVEL, DEDVI BREAK
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a AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-7 LOOP 1 SG DOWNFLOW SIDE, DEDVI BREAK 9:
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FIGURE 4.3-8 DOWNCOMER MIXTURE LEVEL, DEDVI BREAK 35 3-C
,25-C 20-1 4 15-4 s i 10 0 250 500 750 1000 '250 TIME (SEC)
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!NE id E AP600 DESIGN CHANGE DESCRIPT10N REPORT FIGURE 4.3-9 LOOP 1 CMT MIXTURE LEVEL, DEDVI BREAK 5:
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-10 ADS TRAINS 1 TO 3 VAPOR FLOW, DEDVI BREAK
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FIGURE 4.3-12 ADS 4 LIQUID FLOW RATE, DEDVI BREAK 5::
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9 AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-14 PRIMARY MASS INVENTORY, DEDVI BREAK 4
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=c sm AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-15 DOWNCOMER PRESSURE, INADVERTENT ADS ACTUATION 2::::
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AP600 DESIGN CHANGE DESCRIPTION REPORT 5 FIGURE 4.3-16 PRESSURIZER MIXTURE LEVEL, INADVERTENT ADS ACTUATION I
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EE AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-17 CORE STACK MIXTURE LEVEL, INADVERTENT ADS ACTUATION
- f
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0 500 1000 '500 2000 :5::
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-18 LOOP 1 CMT TO DVI FLOW, INADVERTENT ADS ACTUATION 4
. l' 4
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-19 LOOP 1 CMT MIXTURE LEVEL, INADVERTENT ADS ACTUATION l
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ME =i AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-20 DOWNCOMER MIXTURE LEVEL, INADVERTENT ADS ACTUATION l
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FIGURE 4.3-21 ADS TRAINS 1 TO 3 LIQUID FLOW RATE, INADVERTENT ADS ACTUATION i
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-22 ADS TRAINS 1 TO 3 VAPOR FLOW RATE, INADVERTENT ADS ACTUATION I'
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FIGURE 4.3-23 LOOP 1 lRWST INJECTION FLOW, INADVERTENT ADS ACTUATION h
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g AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-24 ADS 4 LIQUID FLOW RATE, INADVERTENT ADS ACTUATION I
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AP600 DESIGN CHANGE DESCRIPTION REPORT FIGURE 4.3-25 ADS 4 VAPOR FLOW RATE, INADVERTENT ADS ACTUATION t
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- 4-63
s: E AP600 DESIGN CHANGE DESCRIPTION REPORT 5.0 Test Plans for Automatic Depressurization System (ADS Phase B)
This section discusses test plans for system tests to be performed for the AP600 Automatic Depressurization System (ADS) at the VAPORE facility in Cassacia, Italy. These tests, designated as ADS Phase B, have been restructured to accommodate the ADS design changes discussed in Section 2.9. ,
5.1 Test Objectives The overall objective of the ADS Phase B tests is to collect thermal hydraulic performance data on the ADS that will be used to verify the analytical models used in design basis safety analyses. Specifically, the tests will be used to obtain fluid temperature, pressure and pressure drop data over a range of mass flow rates through a full scale simulation of the AP600 ADS. The tests will also be used to obtain additional sparger and quench tank loading data over a range of flow conditions.
5.2 Test Loop Configuration Figure 5-1 shows the test loop configuration for the Phase B tests. The test facility provides a full scale simulation of the ADS System, including:
ADS valve piping package, downstream piping, and sparger.
The ADS valve piping package for the tests is shown in Figure 5-2. As shown in this figure and in Table 5-1, a flanged spool piece will be installed at one of the two valve locations in each ADS stage to allow different valve types and sizes to be simulated.
5.3 ADS Phase B Tests Tests will be performed to represent high flow resistance (minimum venting flow) and low flow resistance (maximum loads). These conditions will be obtained by varying the system configuration and by varying the system conditions at ADS valve piping package manifold.
The system configuration will be varied by changing the flow path (i.e., stage 1 only, stage 1 and 2, etc.) and by changing the flow area of each flow path. The system conditions at the ADS valve piping package manifold (e.g., pressure and fluid quality) will be controlled via initial conditions of the supply tank and via the system control valve.
W85tingh0tjse 5-1
AP600 DESIGN CHANGE DESCRIPTION REPORT Three types of tests will be performed as part of the ADS Phase B tests: 1) saturated steam blowdowns. 2) steam and saturated water blowdowns for minimum ADS venting, and 3)
Steam and Saturated Water Blowdowns for Maximum Loads.
Saturated Steam Blowdowns These tests will use steam from the top of the supply tank and will cover a range of pressures for each ADS flow path. Only maximum flow resistance will be simulated for these tests since minimum flow resistance information was previously obtained from the Phase A portion of the ADS tests.
Steam and Saturated Water Blowdowns for Minimum ADS Venting These tests will use saturated water from the bottom of the supply tank and will cover a range of pressure and two-phase conditions for each ADS flow path. Maximum flow resistance will be simulated. The facility supply valve (12 inch gate) will be positioned to obtain a range of qualities entering the valve piping package.
Steam and Saturated Water Blowdowns for Maximum Loads These tests will use saturated water from the bottom of the supply tank and will cover a range of two phase flow and pressure conditions for each ADS flow path. Minimum flow resistance will be simulated to provide sparger loading data. The facility supply valve (12 inch gate) will be fully opened to obtain maximum mass flow entering the valve piping package. Tests will be performed at different quench tank temperatures to obtain loads over a range of conditions.
Table 5-2 shows the ADS Phase B test matrix.
W85tingh00S8
'h' AP600 DESIGN CHANGE DESCRIPTION REPORT !
e g. . , , .
TABLE 5-1 VALVE REPRESENTATION IN THE ADS VALVE PACKAGE (ADS PHASE B TESTS).
_ 4,c M
WBStingtlause 53
llIIl!l
- ill AP600 DESIGN CHANGE DESCRIPTION REPORT n p .w < _____
TABLE 5-2 ADS PHASE B TEST MATRIX (preliminaty)
Blowdown Fluid ADS Simulation AP600 Pressure C omment s !
Simulated l
__ .__ _ . _ . _ j Saturated Steam from top of Stage 1 open - 2250 to 400 psig Maumum Nw tesestarme sunulat..d Stages 1 and 2 open - 800 to 100 psg Stages 1 and 3 open - 500 to 50 psg Stages 1,2, and 3 open - 500 to 50 psg Saturated water from bottom Stage 1 open - 2250 to 400 psg Maumum tiow rosestance of supply tank simulated. 12 uwh gate valvo Stages 1 and 2 open - 1200 to 100 ps'9 possr umed to ot; tan a anje of 2+
T/H data Stages 1 and 3 open - 500 to 50 pssg Stages 1.2 and 3 open - 500 to 50 pssg l
Stage 2 open (inadvertent - 2235 psig operung at full power) i Saturated water from bottom Stages 1.2 and 3 openi " ~ $00 to 50 psig Minimum flow resrstance sinuAated Maximum flow /muumum quahty for maa
, of supply tank Stages 1.2 and 3 open - 500 to 50 psag loads on sparger and quench tarA.
l l
Stages 1 and 2 open "8 ~1200 to 100 psag Stage 2 open "' - 2235 psig
._====..=== .u.===
l (1) - Quench tank water temperature initially at 212*F -
l l
l l
t 54 W Westinghouse
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__c____ g FIGURE 5-1 ADS PHASE B TEST FACILITY SCHEMATIC
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gm Wi; E 45 AP600 DESIGtJ CHANGE DESCRIPTION REPORT FIGURE 5 2 ADS PHASE B TEST PIPING / VALVE PACKAGE CONFIGURATION
- 4, c.
5-6 W851 ngt100S8 l
e AP600 DESIGN CHANGE DESCRIP110N REPoln 6.0 ADS Valve Tests There will be a number of tests performed outside af the Design Certification process during ,
the development. manufacture, installation and operation of the ADS valves. These tests include:
,. Testing to support ADS valve type selection and to support valve qualification testing
- Valve qualification testing
- Production testing
- Pre-operational testing ;
- In-service testing Each of these tests is further discussed in the following sections.
6.1 ADS Valve Type Selection Testing :
For this testing, full sized prototypic Stage 1/2/3 ADS valves will be installed in a ADS valve package piping simulation. Different valve designs will be tested for the ADS Stage 2 and 3 function, to provide a basis for performance comparisons. Testing will be performed over a range of flow conditions that bounds the actual ADS operation. ;
The overall objective of this testing is to characterize valve performance in this application, as an aid to the final selection of a valve type. This information is not required for design certification. Data obtained on valve performance willinclude the measurement of valve operator mechanical and electrical performance; required thrust to unseat, open, close, a'nd seat the valves. The valve seats and disks will be visually examined after test runs to -
observe and document valve wear / damage.
These, tests will provide input to the valve specification including fluid conditions, flow, temperature / pressure, dp conditions and IST conditions. They will also provide input to the EO testing to determine the limiting test conditions (flow, fluid, temperature / pressure, dP).-
6.2 Valve Qualffication Testing Valve qualification will occur after vendor selection and is a three step process of analytical '
qualification, functional testing of the valve / operator assembly, and IEEE qualification of the operator.
T Westinghouse 6-1 i
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AP600 DESIGN CHANGE DESCRIPTION REPORT l 6.2.1 Analytical Qualification Prior to qua!ification testing, vendors will produce calculations to confirm that their designs are acceptable for the ADS application. These calculations include:
An ASME Class 1 design report to verify valve integrity under the design conditions in the equipment specification. including specified nozzle loads and design transients A seismic analys;s to verify operability for the :>pecified seismic accelerations at the maximum operating load A weak link analysis to provide the maximum loads the valve components can withstand for both test and operating conditions Operator sizing calculations using both the manufacturer's and EPRI's methodology (when applicable) 6.2.2 Functional Testing Based on acceptable analytical qualification, valves will be manufactured for testing. The valve / actuator assembly to be tested will be identical to that used in the plant with respect to configuration, materials and dimensions. Prior to assembly, each test valve will be dimensionally inspected in accordance with EPRI guidelines. Critical dimensions will be recorded, along with any special features that are entical to operation. Testing is performed with instrumentation to measure stem thrust, torque, switch actuation, travel, motor speed, motor temperature, accelerations and fluid temperature, pressure and differential pressure.
The test conditions will be determined with input from the type selection testing (Section 6.1).
6.2.3 Operator Qualification The operator will require a separate qualification based on IEEE-382 which will include cyclic aging, vibration aging, seismic testing and environmental aging including LOCA/HELB.
Supplemental testing may be required to address separate issues related to the electric actuators in relation to effects of operating time on motor temperature and speed, motor temperature effect on motor output torque and motor speed effect on torque capability.
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AP600 DESIGN CHANGE DESCRIPTIOt1 REPORT 6.2.4 Valve Design Control An auditable link will be developed between the tested valves and those subsequently shipped to plants. Production valves will be purchased to the same equipment specification requirements as the tested valves. Dimensionalinspection of cntical parts will be performed and recorded to venty that they are within required tolerances. The entical dimertsions will be determined based on a review of the test results and design configuration by the valve manufacturer and Westinghouse.
Any modifications to the original design must be reviewed, evaluated and approved by Westinghouse prior to implementation by the manufacturer.
6.3 Production Testing Each valve will be subjected to production testing including a hydrostatic shell test of the body / bonnet, hydrostatic disc test, leakage tests on the seat, backseat and packing, and a functional test. The latter is performed at nominal and reduced voMage, without flow but with the valve closed and at the design pressure differential. Finally, a cyclic test is pedormed for
[TBD] cycles with no pressure or flow.
6.4 Pre-Operational Testing Static baseline testing will be performed prior to startup, to verify that the valve is set up and functioning correctly. Following the static test, the stage 1,2 and 3 valves will be subjected to individual vatve tests with low flow and with design differential pressure (initial). Finally, an integrated blowdown test will be performed on the first AP600 from intermediate RCS pressure and temperature which will actuate ADS Stages 1,2,3 and 4.
6.5 in-service Testing An in-service stroke time test will be performed every [TBD] months at zero differential pressure, ambient temperature. A pre-refueling test will be performed at low flow (a pressure differential between 400 and 1200 psi, and a temperature of 300 degrees). If squib valves are used for the 4th stage they would not be subject to these tests but would have IST requirements, such as periodically actuating their charges outside the ADS valves.
Every 5 years, verification testing will be performed on valve set-up ar'd operator capabilities, in accordance with GL-89-10.
I WestingholJSe 6-3
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AP600 DESIGN CHANGE DESCRIPTION REPORT Post-maintenance testing of the operator and valve will be required whenever changes are made that may affect the required operating loads or operator output.
W WestinghotJse
AP600 DESIGN CHANGE DESCRIPTION REPORT 7.0 References 7.1 Document No. GWGLO21. " Westinghouse AP600 Standard Safety Analysis Report" Revision 1.1/13/94.
7.2 Document No. GWGLO22, " Westinghouse AP600 Probabilistic Risk Assespment" l Revision 0, 6/26/92.
l l 7.3 Meyer, P.E., "NOTRUMP - A Nodal Transient Small Break and General Network Code," WCAP-10079-P-A, (Proprietary) and WCAP-10080-A (Nonproprietary), August 1985.
7.4 Lee, N., Rupprecht, S.D., Schwartz, W.R., and Tauche, W.D., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Codo," WCAP-10054-P-A (Proprietary) and WCAP-10081 A (Nonproprietary), August 1985.
[ WC51ingh0Use 7
L=E AP600 DESIGN CHANGE DESCRIPTION REPORT APPENDIX A-1 MARKUP OF AP600 SSAR SECTION 6.3 -
PASSIVE CORE COOLING SYSTEMS W westingnouse i
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i 6.3 PASSIVE CORE COOLING SYSTEM The pnm.uy tuncuon of the passne sore coolmg The passive core cooltng astem n designed to operate i system ts to provide emergency core coohng f ollowmg without the use of actne equipment such as pumps and i postulated design basis events To accomplish this ac power sources. The passne core coohng system pnmarv funcuon. the passne core cochng system ts depends on reliable passive componenp and processes designed to perform the followmg funcuons: such as gravity injecuon and expansion of compressed gases. The passive core cochng system does require a
. Emergency core decay beat removal one-time abgnment of valves upon actuanon of the specific components.
Provide core decay heat removal dunng transients, accidents, or whenever the normal beat removal 6.3.1. Design Basis paths are lost. Ttus heat removal function is available at reactor coolant system cond2tions The passive core cochng system is designed to includmg shutdowns and refuehngs. perform its safety-related funcuons based on the follow-mg considerations:
- Reactor coolant system emergency makeup and borauon . It has component redundancy to provide confidence that its safety-related funcuons are performed, even Provide reactor coolant system makeup and boranon in the unhkely event of the most liminng single dunng transients or accideras when the normal failure occurnng coincident with postulated design reactor coolant system makeup supply from the basis events.
chermcal and volume control system is unavailable or ts insuf6cient. . Components are designed arxi fabncated according to industry standard quality groups commensurate
. Safety injection aith its intended safety-related functions.
Provide safety injecnon to the reactor coolant . It is tested and inspected at appropnate intervals, as system to provide adequate core cooling for the denned by the ASME Code.Section XI, and by complete range of loss of coolant accidents, up to technical spect5 cations.
armi including the double-ended rupmre of the largest pnmary loop reactor coolant system piping. . It performs its intended safety-related functions following events such as fire, internal missiles or
. Contamment sump pH contml pipe breaks.
Provide for efemical addition to the cormnment . It is protected from the effects of extemal events sump durmg post-acciders conditions to establish such as earthquakes, tomadoes, and floods.
Goodup chetmstry corrhtions that support radionuclide resention with high radioactivity in . It is designed to be sufSciently reliable, considering consmnment and to prevent corrosion of redorxiancy and diversity, to support the plant core contamment eqmpment dunng long-term floodup melt frequency and significant release frequency corxhuous. goals.
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- 6. ENGINEERED SAFETY FEATURES Revision: 2 DRAFT Effective: 02/15/94 l
6.3.1.1 Safety Design Basis reactor coolant spiem m be depresunted and the stress m the reactor , utant a sicm .md connecting The.pwe core woobng sprem is designed to pipe to be reduced to low M eh This .tho allow s prm ide emergency core cochng dunng esents ins olving plant condinens to be estabbshed ter trutiauon of m.reases and decreases in secondary side heat removal normal residual heat temmal ss stem operauon and decreases in reactor coolant system insentory.
Sutwecuan 6.3 3 provides a desenpuon of the design Dunng a steam generator tube mpture event, the bass esents The performance entena are provided in passive residual heat removal heat exchangers Subsection 6.31 and also desenbed in Chapter 15. under iemove core decay tnt and suf ficiently reduce reactor coolant system temperature and pressure, the respecuse event secnons.
equahzmg with stearn generator pressure and terminatmg break flow, without over611mg the -
6.3.1.1.1 Emergency Core Decay Heat I steam generator or actuaung the ADS.
Removal 6.3.1.1.2 Reactor Coolant System For postulated nonLOCA events, where a loss of capabihty to remove core decay heat via the steam Emergency Makeup and generators occurs. the passive core coohng system is BOration designed to perform the following funcuons:
For postulated nonLOCA events, suf6ctent core
+ The passive residual beat removal beat exchangers makeup water mventory is automancally provided to automancally actuate to provide reactor coolant keep the core covered and to allow for decay heat system cochng and to prevent water rebef through removal. In addition, ttus makeup prevents actuanon of the pressunzer safety valves. the automanc depressunzation system for a sagru5 cant time.
- The passive residual beat removal beni exchangers For postulated events resulting in an inadvenent are capable of automatically removmg core decay cooldown of the reactor coolant system, such as a steam Seat following such an event, assurrung the steam hoe break, suf6cient borated water shall be auto-generated in the in-contamment refueling water matically provided to makeup for reactor coolant system storage tank ts condensed on the containtnent vessel shnntage. The borated water also counteracts the and retumed by gravity via the in-containment reacnvity increase caused by the resulting system refuehng water storage tank condensate retum cooldown.
gutter. The passive beat removal beat exchangers For a Condnion 11 steam line break desenbed in should provide decay beat removal for at least 72 Chapter 15, retum to power is acceptable if there is no bours tf no enade=te is recovered. core damage. For tins event, the automatic depressunzanon rystem is not actuated to avoid the
- The passive residual beat removal beat exchangers, unnecessary loss of an activity barner.
m conjunazon with the passive contamment cooling For a large steam line break, the peak retum to system, are designed to remove decay beat for an power is limited so that the offsite dose limits are mde6rnte ame in a closed-loop mode of operanon. satis 5ed. Following either of these events, the reactor In addinon, the passive residual best removal beat coolant system is automatically brought to a subentical exchanfers are designed to cool the reactor coolant condition, consistent with the passive containment system to 400*F to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, with or without cooling capabthties. l reactor coolant pumps operanng. T1us allows the 6.3-2 l
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- 6. ENGINEERED SAFETY FEATURES Revision: 2 fil' AFT E Effective: 02/15/94 =ij For safe shutdown, the passne core coebng sptem 6.3.1.1.4 Safe Shutdown n designed to supply cutticient tvron to the reactor wolant systern to maintam the techrucal specification The funcuonal requtrements for the passne core shutdow n margin for cold, post-depressunzation coolmg system specify that the plant be brought to a condiconL with the most reactive rod fully withdrawn stable condiuon usmg the passive residual heat removal trom the core. The automauc depressunzauon system heat exchangers for events not involvmg a loss or j is not expected to actuate for these events. coolant. For these events, the passfve core coolmg system, in conjunction with the passive containment l
6.3.1.1.3 Safety injection cooling system, has the capabthty to estabitsh safe shutdown cordmons, cooling the reactor coolant system The passne core cooling system provides sufficient to about 400*F in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, with or without the reactor
( coolant pumps operaung.
water to the reactor coolant system to mmgate the effects of a loss of coolant accident. In the event of a The core makeup tanks automaccally provide large loss of coolant acciders, up to and includmg the toyecnon to the reactor coolant system as the temperature rupture of a bot or cold leg pipe where essentially all of decreases and pressunzer level decreases, actuatmg the the reactor coolant volume is ininally displaced the core makeup tanks. The passive core cooling system passive core coobng system rapidly refills the reactor can maintam stable plant conditions for a long time in vessel reDoods the core, and continuously removes thel this mode of operanon. without ADS actuation. For core decay heat A large break ts a rupture with a totall example, with a RCS leak rate of I gpm, stable plant cross-secuonal area equal to or greater than one squareI condinons can be mamtamed for about 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> before foot. Altbough the enteria for mechantsuc pipe breakl ADS would occur.
are used to linut the size of prpe rupture considered ml For loss of coolaat accidents and for postulated the design and evaluation of piping systems, as events where ac power sources are lost, or when the descnbed in Subsecnon 3.6.3. such criteria are not used core makeup tank levels reach the automauc in the design of the passive core cooling system, depressunzanon system actuation serpoint, the automauc Suf6cient water is provided to the reactor vessel depressunzation system initiates. This results in following a postulated loss of coolant accident so that injecnon from the accumulators and subsequently from the performance entena for emergency core coolmg the in-containment refueling water storage tank, once the systems, desenbed in Chaper 15. are sansfied. reactor coolant system is nearly depressunzed. For The automanc depressunzanon system valves, these condanons, the reactor coolant system provided as part of the reactor coolant system, are depressunzes to saturated condanons at about 240 F designed so that together with the passive core coohng within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The passive core coohng system can system they: matarain this safe shutdown condition indefimtely for the plant.
- Sansfy the small loss of coolant accident The basis used to define the passive core cooling performance iw=.mcots system functional requnements are denved from Section 7.4 of the Srw*rd Review Plan. The funcnonal
= Provide effective core cooling for loss of coolant requuements are met over tic range of anticipated accderes frtun when the pasarve core coohng events and single failure assumpdons. The pnmary system is acmated through the long-term cooling function of the passive core cooling system dunng a safe mode. shutdown using only safety-related equipmern is to provide a means for boranon, injecnon, and core WBStiflgh0USB
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- 6. ENGINEERED SAFETY FEATURES Revision: 2 DRAFT j
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cochng Details ot'the safe shutdown design bases are analpts is used for both the passive core cwhng m. tem presented tn Subsection 5 4 7 and Secton 74. performance and for plant response.
In the esent of a small loss of coolant auident the 6.3.1.1.5 Containment Sump pH Control passive core coohng sptem limits the increase in peak clad temperature and core uncovery with design basis The passive core coohng system is capable of assumpuons. For pipe ruptures of less than eight.tnch mamtainmg the destred post-accident pH condinons in nominal diameter stze, the passive cort coobng spiem the recirculanon sump water after contamment floodup. is designed to prevent core uncovery with best estimate The pH adjustment tank is capable of mamtaming assumptions.
containment sump pH within a range of 7.0 to 9.5, to The passive residual beat removal heat exchangers enhance radionuclide retention in :he contamment sump and the in-containment refueling water storage tank are and to prevent stress conosion crackmg of contatament designed to delay signi5 cant steam release to the components dunng long-term contamment Goodup. contamment for at leur one hour.
Two 100 percent capacity passive residual beat removal beat exchangers are included in the passive core {
6.3.1.1.6 Reliability Requirements cooling system. They provide sufficient redundancy so that an mdividual heat exchanger can be isolated in the The passive core cooling system design sausfies a event of tube leakage and plant operation can continue <
vanery of reliabihty requirements. including redundancy
~
without having to shut down to repair the leakmg beat (such as for components, power supplies. actuation exchariger.
signals, and instrumentation), equipment tesung to The frequency of automatic depressurization system confirm operability, procurement of quahSed actuation is limited to a low probability to reduce safety components. and provisions for penodic maintenance, nsks and to mmimize plam outages. Equipment is In addition, the design provides protection in a number located so that it is not flooded or it is designed so that of areas including:
it is not damaged by the flooding. Major plant equiPment is designed for multiple occurrences without :
- Single acuve and passive component failures damage.
- Spunous failures The pH control equipment is designed to minimtze ,
- Physical damage from fires, Gooding, missiles, pipe the potential for system leakage dunng standby opera- l wtup. and accident loads tion, and also to msnmuze the potenual for and the
= Environmental conditions such as high temperature unpact of an inadvenent actuation.
steam and containment floodup.
The passive core cooling system is capable of Subsecuon 6.3.1.2 includes specific nonsafety. supporting the required testing and maintenance, including capabilities to isolate and drain equipment related design requuements that help to con 6rm sausfactory system reliability.
6.3.2 System Design 6.3.1.2 Power Generation Design Basis The passive core cooling system is a seismic Category I, safety.related system. It consists of two The passive core coohng system is designed to be sufficiently reliable to suppon the probabilistic nsk core makeup tanks, two accumulators, the in-containment refueling water storage tank, two passive ;
analysis goals for core damage frequency and severe residual beat removal beat _ exchangers, the pH release frequency. In assessmg the reliability for l adjustment tank, two ADS spargers and associated j probabihsuc nsk analysis purposes, best esumate i l
6.3-4 :
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- 6. ENGINEERIED SAFETY FEATURES Revision: 2 DMAFT ~
Effective: 02/15/94 D D pipmg. valves. m strumentanon. .md other related equip- 6.3.2.1.1 Emergency Core Decay Heat ;
ment. The autom,dic depressunzauon system valves. Removal at High Pressure and :
which me part ot' the reactor coolant system, also Temperature Conditions -!
1 provide unportant passive core coohng functions.
The passive core coobng system is designed to [Westmghouse Propnetaol provide adequate core coobag in the event of design [Provided under separate coverl basis events. The redundant onsite safety-related dc power sources provide power such that protecuon is 6.3.2.1.2 Reactor Coolant System provided for a loss of ac power sources, coincident with Emergency Makeup and !
an event, assummg a smgle failure has occuned. Boration l 6.3.2.1 Schematic Piping and [Wesunghmse ProprietaryJ Instrumentation Diagrams [Provided under separate cover)
Figures 6.31 through 6.3-4 show the piping and 6.3.2.1.3 Safety injection During Loss of l mstrumentanon drawmgs of the passive core cooling Coolant Accidents system. Process Dow diagrams are shown in Figures 6.3-5 through 6.3-7. Tables 6.31 through 6.3-3 provide [Westmghee Proprietary]
a summary of the expected Guid condicons for the [ Pron &d eder separate com]
vanous locanons shown on the process Gow diagrams, for the speciSc plant conditions idenafied - safety Containment Sump pH Control 6.3.2.1.4 mjecuon, decay beat removal, and sump pH control The passive core cooling system is desgned to
[Westmghme Proprietary]
supply the core cooling flow rates to the reactor coolant
[Provided under separate cover]
system specified in Chapter 15 for the accident analyses. '
The acculent analyses Sow rates and beat removal rases are calculated by assummg a range of component 6.3.2.1.5 Passive Core Cooling System parameters. including best estunate and conservanvely Actuation '
high and low values.
The passive cose coohng system design is based on (Westmgbouse Propnetary]
the stx major components. lisned in Subsection 6.3.2.2. [Provided under separate cover] {~
that function together in vanous different combmanons to support the four basic passtve core cooling sysiem 6.3.2.2 Equipment and Component funaims: Descriptions
- Emergency decay beat removal [Wesonghouse Propnetary]
- Emergency sanctae me-yA= don [Provided under separase cover) >
a Safety irgecano ,
- Caar==== sump pH control 4
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Core Makeup Tanks
- Stem leakage collecuon is used for sahes larger 6.3.2.2.1 than two inches in diameter.
(Wemnehouse Propnetaryl (Prouded under separate coler 6.3.2.2.7.1 Manual Globe, Gate, and ;
Check. Valves 6.3.2.2.2 Accumulators Gate valves have backseats and extemal screw and
[Westmghouse Propnetary] yoke assembbes.
[Provided under separate cover] I Globe salves, are full-ported with extemal screw l and yoke construction.
l 6.3.2.2.3 In-Containment Refueling Check valves are sprmg-loaded lift piston types for Water Storage Tank sizes 2 inches and smaller, and swing-type for sizes 2 5 mches and larger. Stamless steel check valves have rm Penetranon welds other than the inlet, outlet, and bonnet.
[ Westinghouse Propnetary] ,
The check valve hmge is sesviced through the bonnet. ('
[Provided under separate cover]
The stem packmg and gasket of the stainless steel manual globe and gate valves are sumlar to those 6.3.2.2.4 pH Adjustment Tank desenbed in Subsecuon 6.3.2.2.73 for motor-operated valves. Carton steel manual valves are empicyco to ,
[Wesunghouse Propnetary]
Pass nonradioacuve Guid only and, therefore, do not
[Provided under separate cover]
contain the double packing and seal weld provisions.
6.3.2.2.5 Passive Residual Heat Removal 6.3.2.2.7.2 Manual Valves Heat Exchangers Where manual valves are locked open.
[Wesungbotise Pmpnetary) administrauvely controlled. and provided with redundant ,
[Provided under separate cover) status in the main control room, bypass and inoperable .;
status indication is provided according to Regulatory.
6.3.2.2.6 Depressurization Spargers Guide 1.47. Comptiance with the recommendations of Regulatory Guide 1.47 is provided in Subsection 1.9.1.
[ Westinghouse Proprietary) Manual valves are generally used as maintenance
[Provided under separate cover] isolation valves or throtthog valves. When used for these functions they are under administrative control, 6.3.2.2.7 Valves which requires them to be locked in the conect position.
They are located so that no single valve can isolate Design feanmes used to minimize leakage for valves redundant passive core cooling system equipment or in the passive cose coohng system inchwie they are provided with alarms in the maan control room to mdicate misposinoning. *
- Where possible. padless valves are used. To help preclude the possibility of passive core j cooling system degradation due to valve mispositioning. I a Other valves which are normally open, except check line connections such as vent and drain lines. test i valves and those which perform control function, connecuons. pressure points. flow element test points.
are provided with back seats to limit stem leakage. Oush connecnons, local sample points, and bypass lines 6.3-6 y
b 4_.-v., __ , , . . - . _ . ~ . . .. _ . . _ ___.__m
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Revision: 2 ORAFT =
Effective: 02/15/94 are provided with double isolation or sealed barners. Valves -
The isolation is provided by one of the followmg methods: [Westmghouse Propnetary)
[Provided under separate cover] l
- Two valves m a senes
- A single valve with a screwed cap or bhnd llange 6.3.2.2.7.7 Low Differential Pressure i
- A smgle locked-closed valve Opening Check Valves
- A blind taange.
[Wesm@me Pmpnetary} ;
6.3.2.2.7.3 Motor-Operated Valves [Provided under separate cover]
De most common AP600 valve body design for 6.3.2.2.7.8 Accumulator Check Valves I motor-operated valves is a gate valve. Industry issues I will be considered in the design of these valves. The
[Wesughee Pmptie'ary] l motor operators ase conservanvely simi, considenng the [Provided under separate cover) fnctional componera of the hydraulic unbalance on the valve disc, the disc face inction, and the packing box 6.3.2.2.7.9 Relief Valves fncuon. Some special purpose valves have slow operators without motor brakes. For motor-operated Relief valves are installed for passive core cooling valves, the valve disc is guided throughout the full disc system accumulators and the pH adjustment tank to i tnvel to prevent codang and to pmvide case of gate movement. De scanng surfaces are hard-faced to Praect the tanks from overpressure.
The passive core cooling system piping is reviewed prevent galling and to reduce wear.
to identify those lengths of piping that are isolated by Where a gasket is employed for the body to bonnet i normally closed valves and that do not have pressure I joint. it is a fully trapped. controlled compression, spiral wound asbestos (or a quali5ed asbestos subantute) relief protection in the piping secnon between the valves.
i gasket with provisions for seal welding. The valve stuffing boxes are designed with a lantern ring leak.off D'se piping sections include:
I connection- - Poidons of m-containment passive core cooling system test lines that are not passive core cooling 6.3.2.2.7.4 Motor-Operated Vaive Controls system accadent mitigation flow paths and are not needed to &ve 2 sinidown
[Westmgbouse Nyu.smy]
[Provded under separase cover] = Piping vents, drains, and test connections that typically have tw ci sed valves or one closed 6.3.2.2.7.5 Accumulator Motor Operated valve and a blind flange g g
. c'heele valve test lines with sections isolated by two >
_m,
' I normally closed valves.
[Provided under separate cover]
The piping vents, drains. test connections, and check 6.3.2.2.7.6 Automatic Depressurization valve lines have design pressure /temixnture condiuons 6.3-7 3 Westinghouse
- 6. ENGINEERED SAFETY FEATURES Revision: 2 DRAFT jE M Effective: 02/15/94
. empatible with the process piping to wruch they
- Valve stem matenals are selected for thetr cormen u nnect. Therefore. v.ds e teak age does not resistance. high-tensile properties. and their resi-merpressunze the isolated piping anons and pressure tance to surf ace scenng by the packing rebet proustons are not required Sectmn 6 i summanzes the matenals used for 6.3.2.3 Appilcable Codes and passive core cochng system components.
Classifications 6.3.2.5 System Rellability Secuons 5.2 and 3.2 hst the equipment ASME Code and seismic classificanon for the passive core coohng [Wesunghouse Propnetary) system. Most of the piping and components of the [Provided under separate coverj passive core cooling system within containrnent are AP600 Eqmpment Class A. B. or C and are designed to 6.3.2.5.1 Response to Active Failure meet seismic Category I requirements. Some system piptog and components that do not perform safety- [Westmghouse Propnetary]
related funcuons are nonsafety-related. [Provided under separate cover]
The requirements for the control, actuanon. and Class IE power supplies are presented in Chapters 7 6.3.2.5.2 Response to Passive Failure and 8.
[Westinghuse Pmpnetary]
6.3.2.4 Material Specifications and [Provided under separate cover]
Cornpatibility Matenals used for engmeered safety feature components are given in Section 6.1. hhtenals for [Wesunghouse Propnetary]
I passive core coohng system components are selected to [Provided under separate cover) meet the applicable matenal requirements of the codes in Seccon 5.2. as well as the followmg additional 6.3.2.5.4 Potential Boron Precipitation regmrements:
(Wesungbase Pmpnetary]
+ Puts of components in contact with borated water
[Provided under separate cover) are fabncated of, or clad with, ansterntic stamiess steel or an equivalent corrosion-resistant matenal.
6.3.2.5.5 Safe Shutdown
- Internal parts of components in contact with contamment emergency sump solunon dunng Dunng a safe shutdown. the passive core cooling rectreulation are fabncated of austenitic stainless system provides r dna4ney for boration, makeup, and steel or an equivalent corrosion resistant matenal. beat removal functions. Details of the safe shutdown design are desenbed in Section 7.4.
. Valve seatmg surfaces are hard-faced to prevent failure and to reduce wear. Zero or low cobalt hard 6.3.2.6 Protection Provisions I facmg is used where performance is acceptable.
6.3-8
- 6. ENGINEERED SAFETY FEATURES
- Revision: 2 DflAFT Effective: 02/15/94 E n
5g The measures taken to protect the system trom t'aulted steam generator, cool down and depressunze the reactor coolant system to terminate the break Gow to the i damage that micht result l' rom vanous esents are descnbed in other ections. as listed below. steam generator, and stabihre plant conditions.
Section 7.5 desenbes the post-accident monatonag
- Protection from dynanuc effects is presented m instmmentation available to the operator m the mam Secuon 3.6. control room foHowtog an event.
9
- Protection from misstles is presented in Section 3.5. 6.3.3 Performance Evaluation
. Protection from seismic damage is presented in [Wesunghouse Proprietary]
Scaions 3.7, 3.8. 3.9. and 3.10. [Provided under separate cover]
- Protection from fire is presented Subsecuon 9.5.1. 6.3.3.1 increase in Heat Removal by
- Environmental quahficanon of equipment is Presented in Section 3.11. A number of events that could result in an increase in beat removal frone the reactor coolant system by the
+ Thermal stresses on the reactor coolam system are secondary sysicm have been postulated. For cada event. r presented in Secuon 5.2. consideration bas been given to operation of nonsafety-relaaed systems that could affect the event results. The 6.3.2.7 Provisions for Performance operation of the startup feedwater system and the Testing chemical and volume control system makeup pumps can affect these events. Analyses of these events, both with
[Westingbouse Proprietary) and without these nonsafety.related systems operaung.
[Provided under separate cover] are prewntad in Section 15.1. For those events resulting ;
in passive core coohng system actuation. the fonowing 1 6.3.2.8 Manual Actions summarizes passive core cooling system performance: i The passive core cooling system is automaticauy 6.3.3.1.1 Inedvertent Opening of a' actumed for those events as presensed in Steam Generator Rollef or )
Subsection 6.3.3. Following actuation, the passrve core Safety Valve coohng system connoues to operase in the injection mode until the tranation to recarretanno trnnate' [Wesingbouse @tary]
automaticauy foRowng "m==near floodup. [Provuled under separate cover]
Although the passrve core coohng system operates i automaticaUy, operator actions would be beeficial, in 6.3.3.1.2 Steam System Pipe Failure some cases, in se& acing the evn of an event.
For exasapie,in a neem genermor tube rupane with no [Wesungbouse Propnetary) oPesamt ma the proesction and safety monisming [Provuled under separate cover]
I system automancany termmmee the leak, prevents steam genermor overfill, and limits the offsite doses. 6.3.3.2 Decrease in Heat Removal by However, the operator can initiate actions, sunilar t the Secondary System aose taken to cunent plants, to idennfy and isolate the W westinghouse
i i
, 6. ENGINEERED SAFETY FEATURES Revision: 2 DRAFT EEEMEE Effective: 02/15/94
- -- y A number of events hat e been postulated that could [ Westinghouse Propnetaryl result in a decrease m heat removal from the reactor {Provided under separate coser]
c ool. int system by the secondary system. For each event, considerauon has been given to operation of 6.3.3.3.2 Loss of Coolant Accident nonsafety.related systems that could affect the consequences of an event. The operation of the startup [ Westinghouse Propnetary]
feedwater system and the chemical and volume control [Provided under separate cover) system makeup pumps can affect these events. Analyses of these events, both with and without these nonsafety- 6.3.3.3.3 Passive Residual Heat Removal related systems operanng, are presented in Secnon 15.2. Heat Exchanger Tube Rupture For those events resulung in passive core cooling system actuanon, the following summanzes passive core coohng
[Wesunghouse Propnetary]
system perfonnanx [Provided under separate cover]
6.3.3.2.1 Loss of Main Feedwater 6.3.3.4 Shutdown Events (Westmgbouse Propnetary]
Ibe passive core cooling system components are
[Provided under separate cover) available whenever the reactor is enocal and when reactor coolant energy is sufficiently high to equire '
6.3.3.2.2 Feedwater System Pipe Failure passive safety injecnon. Dunng low. temperature physics tesung, the core decay beat levels are low and -
[ Westinghouse Propnetary] there is a negligible amount of stored energy in the 1
[Provided under separate cover] reactor coolant. Therefore, an event comparable m seventy to events occurnng at operaung condinons is 6.3.3.3 Decrease in Reector Coolant not possible and passive core cooling system equipment +
System Inventory is not required. The possibility of a loss of coolant accident dunng plant startup and shutdown has been A number of events have been posndated that could considered.
result in a decrease in reactor coolant system inventory. During shutdown conditions, some of the passive For each event. consideration has been given to core cooling system equipment is isolated. In addition, operanon of nonsafety-related systems that could affect since the normal residual beat removal system is not a the consequences of the event. 'Ibe operation of the safety-related system, its loss is considered.
stanup feedwater system and the chemical and volume As a result, gravity injection. is automatically ,
control system makeup pumps can aNect these events. actuated when required during shutdown conditions prior Analyses of these events, both with and without these to refueling cavity floodup, as discussed in nonsafety-related systems operating, are su in Subsection 6.3.3. The operator can also manually Section 15.6. Foe those events wtuch resuk in passive actuate other passive core coohng system equipment.
core cooling system actuation,the following summarues such as the passive residual beat removal beat passive core cooling syseem performanx exchangers, if required for accident mitigation during '
shutdown conditi ons when the equipment does not 6.3.3.3.1 Steam Generator Tube Rupture automatically actuate.
6.3.3.4.1 Loss of Startup Feedwater 6.3-10 j i
L
l l
- 6. ENGINEERED SAFETY FEATURES Revision: 2 DRAFT -_
Effective: 02/15/94 I
l l
During Hot Standby, [Westmghouse Propnetary]
Cooldowns, and Heatups [Provided under separate cover) 4 IWestmghouse Propnetary] 6.3.6.2 In Service Testing and
[Provided under separate cover] inspection 6.3.3.4.2 Loss of Normal Residual Heat [ Westinghouse Propnetary]
Removal Cooling With The [Provided under separate cover) i Reactor Coolant System Prese.ure Boundary intact 6.3.7 Instrumentation Requirements
[Wesunghouse Propnetary] [Westingbouse Proprietary]
[Provided under separate cover] [Provided under separate cover]
6.3.3.4.3 Loss of Normal Residual Heat 6.3.7.1 Pressure Indication j i
Removal Cooling During Midloop Operation 6.3.7.1.1 Accumulator Pressure
[Wesungbouse Propnetary] [ Westinghouse Propnetary]
[Provided under separase cover] [Provided under swarase cover]
6.3.3.4.4 Loss of Normal Residual Heat 6.3.7.1.2 Passive Core Cooling Systerr, ;
Removal Cooling During Test Heeder Pressure Indicat'.on Refueling
[Westinghoine F.y.eterj] ,
[Westingbouse Propnetary] [Provided under separase cover] l 1
[Provided under separase cover) 6.3.7.2 Temperature Indication l 6.3.4 Post-72 Hour Actions i 6.3.7.2.1 Core Makeup Tank Inlet and
[ Westinghouse Propnetary] Outlet Line Temperature
[Provided under swernae cover]
[ Westinghouse Twy.kterj] ,
6.3.5 Limits on System Parameters [Prtmded under sgerate cover] l p f - -: Pi w ;.rj] 6.3.7.2.2 Passive Residual Heat Removal I
[Provided under egernae cover] Heat Exchanger Inlet and Outlet .
Line Temperature l 6.3.6 Inspection and Testing Requirements ;
[ Westinghouse F.ymy]
6.3.6.1 Prooperational Testing [Provided under separate cover]
[ Westinghouse i
., 6. ENGINEERED SAFETY FEATURES Revision: 2 DRAFT T =:
Effective: 02/15/94 6.3.7.2.3 In-Containment Refueling IWestmghouse Propnetan i Water Storage Tank IProvided under separate coser!
Temperature 6.3.7.5 Containment Radiation Level (Westmghouse Propnetary)
[Provided under separate cover) [Westmghouse Propnetg)
[Provided under separate cover]
6.3.7.3 Flow Indication 6.3.7.6.1 Valve Position Indication 6.3.7.3.1 Passive Core Cooling System Test Header Flow indication [Westmgbouse Proprietary]
[Provided under separate cover]
[Pr e sep te )er] 6.3.7.6.2 Valve Position Control 6.3.7.3.2 Passive Residual Heat Removal 6.3.7.6.2.1 Passive Residual Heat Removal Heat Exchanger Outlet Flow Heat Exchanger Outlet Valve Position Control
[Westmgboure Propnetary]
[Provided under separate cover] [ Westinghouse Propnetary]
{Provided under separate cover]
6.3.7.4 Level Indication 6.3.7.7 Automatic Depressurization 6.3.7.4.1 Core Makeup Tank Level System Actuation at 24 Hours (Westmghouse Proprietary] [Westingbouse Propnetary]
[Provided under separate cover] [Provded eder separate cover) 6.3.7.4.2 Accumulator Level
[Westingbouse Propnetary]
[Provided under separate cover]
6.3.7.4.3 in-Containment Refueling Water Storage Tank Level
[Weninghouse Lycej]
[Provided under separate cover) 6.3.7.4.4 Containment Sump Level 6.3-12 L
- 6. ENGINEERED SAFETY FEATURES Revision: 2 DRAFT ,. g Effective: 02/15/94 j 6.3.8 References
[ Westing. house Prttnetan ]
[Provided under separate coserl 6.3-13 W Westinghouse
AP600 DESIGN CHANGE DESCRIP110N REPORT =
APPENDIX A-2 AP600 PXS AND RCS P&lD This Appendix contains information proprietary to Westinghouse Electric Corporation W Westinghouse