Regulatory Guide 1.79
| ML12300A329 | |
| Person / Time | |
|---|---|
| Issue date: | 10/04/2013 |
| From: | Talbot F Office of Nuclear Regulatory Research |
| To: | |
| Bayssie M | |
| Shared Package | |
| ML12300A323 | List: |
| References | |
| DG-1277 RG-1.079.1 | |
| Download: ML12300A329 (40) | |
U.S. NUCLEAR REGULATORY COMMISSION October 2013 REGULATORY GUIDE Technical Lead F. X. Talbot OFFICE OF NUCLEAR REGULATORY RESEARCH
REGULATORY GUIDE 1.79.1 (Draft was Issued as DG-1277 dated June 2012)
INITIAL TEST PROGRAM OF
EMERGENCY CORE COOLING SYSTEMS FOR
NEW BOILING-WATER REACTORS
A. INTRODUCTION
Purpose This regulatory guide (RG) describes methods that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for demonstrating compliance with the NRC regulations identified below as they relate to preoperational, low power and power ascension testing features of the emergency core cooling systems (ECCS) for boiling-water reactors (BWRs) whose licenses are issued after the date of issuance of this RG. This RG also describes methods that the NRC staff finds acceptable for initial plant testing of ECCS structures, systems, and components (SSCs). Additionally, this RG
describes methods the NRC staff finds acceptable for testing of the Isolation Condenser System (ICS) and the Reactor Core Isolation Cooling (RCIC) System, which support functions for alternate water injection during station blackout.
Applicable Rules and Regulations This regulatory guide (RG) describes methods that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable to implement Title 10 of the Code of Federal Regulations, Part
50, Domestic Licensing of Production and Utilization Facilities (10 CFR Part 50) (Ref. 1), Appendix A,
General Design Criteria for Nuclear Power Plants, General Design Criteria (GDC) 4, 5, 33, 34, 35, 36,
37, and 55, as they relate to features of the ECCS for new BWRs. The RG also describes testing of the ICS and RCIC System to meet the requirements in 10 CFR 50.63, Loss of All Alternating Current Power for core cooling. This RG also describes methods that the NRC staff finds acceptable for initial plant testing of SSCs for the ECCS in accordance with the regulations in 10 CFR 50.34(b)(6)(iii), Plans for Preoperational Testing and Initial Operations, and 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants (Ref. 2), Subpart B, Standard Design Certifications, and Subpart C, Combined Licenses, 10 CFR 52.79(a)(28) Plans for Preoperational Testing and Initial Operations.
Nuclear power plant SSCs must be tested to quality standards commensurate with their importance to safety. Criterion XI, Test Control, of Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50 requires licensees to establish a Written suggestions regarding this guide or development of new guides may be submitted through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading-rm/doc-collections/reg- guides/contactus.html.
Electronic copies of this regulatory guide, previous versions of this guide, and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRC Library at http://www.nrc.gov/reading- rm/doc-collections/. The regulatory guide is also available through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, in ADAMS at Accession No. ML12300A329. The regulatory analysis may be found in ADAMS at Accession No. ML12300A328 and the staff responses to the public comments on DG-1277 may be found in ADAMS at Accession No. ML12300A330.
testing program to identify and perform all tests needed to demonstrate that SSCs will perform satisfactorily in service.
The initial test program (ITP) is to be conducted in accordance with written test procedures that incorporate the requirements and acceptance criteria in applicable design documents. RG 1.79.1, Appendix A, Design Descriptions of ECCS for New BWRs, also provides design information to support the ITP staff regulatory guidance in Section C below. The ECCS functions to be tested are those necessary to ensure that specified design functions of the ECCS are met during any condition of normal operation, including abnormal operating occurrences, or because of postulated accident conditions.
Regulatory guide 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plants, (Ref. 3)
describes a method acceptable to the NRC staff for complying with the Commissions regulations with regard to preoperational, initial criticality, low power, and power ascension testing of nuclear power plant SSCs that perform functions important to safety. This RG describes initial plant testing acceptable to the staff specifically for the ECCS in new BWRs (e.g., the Advanced BWR [ABWR] and Economic Simplified BWR [ESBWR] designs) licensed under 10 CFR Part 52. In cases in which an SSC is not part of the specific nuclear plant design, the associated testing would not apply.
Related Guidance
- RG 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plants describes a method acceptable to the NRC staff for complying with NRC regulations related to preoperational testing of nuclear power plant SSCs that perform functions important to safety.
- RG 1.79, Preoperational Testing of Emergency Core Cooling Systems for Pressurized- Water Reactors describes the general scope and depth the staff of the NRC considers acceptable for demonstrating compliance with the NRC regulations related to preoperational testing of features in the ECCS of pressurized water reactors (PWRs).
- RG 1.82, Water Sources for Long-Term Recirculation Cooling Following a Loss-of- Coolant Accident, includes regulatory positions on design criteria, performance standards, and analysis methods on ECCS water sources that relate to all water-cooled reactor types.
- RG 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, provides additional guidance for the coordination and testing of protective breakers to prevent thermal overload of electrical motors.
Purpose of Regulatory Guides The NRC issues RGs to describe to the public methods that the staff considers acceptable for use in implementing specific parts of the agencys regulations, to explain techniques that the staff uses in evaluating specific problems or postulated accidents, and to provide guidance to applicants and licensees.
Regulatory guides are not substitutes for regulations and compliance with them is not required.
Paperwork Reduction Act RG 1.79.1 contains information collection requirements covered by 10 CFR Part 50 and 10 CFR Part 52 that the Office of Management and Budget (OMB) approved under OMB control numbers RG 1.79.1, Page 2
3150-0011 and 3150-0151, respectively. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information collection request or requirement unless the requesting document displays a currently valid OMB control number.
B. DISCUSSION
Reason for Development The NRC staff recently updated the guidance in RG 1.79, Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors, (Ref.4) to revise existing ECCS tests for Pressurized Water Reactors (PWRs) and included new ECCS tests for advanced PWRs. The NRC staff developed this new guide as a companion guide to RG 1.79 to address the scope of ECCS initial plant tests for new BWRs as a result of the NRCs design certification of the ABWR and the ESBWR. The NRC staff included additional guidance for initial plant tests of the ECCS based on recent operating experience from BWRs.
Standards Endorsed in this Guide This regulatory guide endorses, in part, the use of one or more codes or standards developed by external organizations, and other third party guidance documents. These codes, standards and third party guidance documents may contain references to other codes, standards or third party guidance documents (secondary references). If a secondary reference has itself been incorporated by reference into NRC
regulations as a requirement, then licensees and applicants must comply with that standard as set forth in the regulation. If the secondary reference has been endorsed in a regulatory guide as an acceptable approach for meeting an NRC requirement, then the standard constitutes a method acceptable to the NRC
staff for meeting that regulatory requirement as described in the specific regulatory guide. If the secondary reference has neither been incorporated by reference into NRC regulations nor endorsed in a regulatory guide, then the secondary reference is neither a legally-binding requirement nor a generic NRC approval as an acceptable approach for meeting an NRC requirement. However, licensees and applicants may consider and use the information in the secondary reference, if appropriately justified and consistent with applicable NRC requirements.
Harmonization with International Standards The International Atomic Energy Agency (IAEA) has established a series of safety guides and standards constituting a high level of safety for protecting people and the environment. IAEA safety guides present international good practices and increasingly reflect best practices to help users striving to achieve high levels of safety. Pertinent to this regulatory guide, IAEA Safety Guide NS-G-1.9, Design of the Reactor Coolant System and Associated Systems in Nuclear Power Plants (Ref. 5), issued in 2004, addresses design considerations for the ECCS in Sections 4.68 through 4.91. The NRC has an interest in facilitating the harmonization of standards used domestically and internationally. In this case there are many similar elements between this regulatory guide and the corresponding section of the safety guide.
This regulatory guide consistently implements and details the principles and basic safety aspects provided in IAEA Safety Guide NS-G-1.9.
RG 1.79.1, Page 3
General Discussion of Preoperational Testing Program A comprehensive preoperational, low power and power ascension test program of the ECCS for BWRs should ensure that it will accomplish its intended functions when required. The ITP should cover all test-related activities, including the following:
1. the development of test descriptions, test objectives, and specific acceptance criteria;
2. the preparation of test procedures;
3. the conduct of the tests and acquisition of system and component performance data; and
4. the resolution of deficiencies and deviations from expected performance.
The test program should include prerequisites for completion of construction tests and preoperational tests in coordination with the startup test group approval of test procedures, test configuration, and test initiation. In accordance with RG 1.68, Staff Regulatory Guidance C.4, the applicant or licensees approved test procedures should be made available to the NRC approximately 60
days before their intended use.
NRC Regulatory Issue Summary 2013-09, NRC Endorsement of NEI 09-10, Revision 1a-A,
Guidelines for Effective Prevention and Management of System Gas Accumulation, (Ref. 6) and NEI 09-10,
Guidelines for Effective Prevention and Management of System Gas Accumulation, (Ref. 7) should be used to evaluate the applicant or licensees treatment of gas accumulation concerns. As a prerequisite to ECCS tests, verify that all types of non-condensable gases (air, hydrogen, nitrogen, oxygen, etc.) in the ECCS systems are kept to an acceptable level. This should be verified using nondestructive examination techniques, opening vent valves to remove non condensable gases, or using other methods justified through an engineering evaluation. The engineering evaluation should consider void volume, void transport to pumps, and pump void acceptance criteria and should include performance of void transport analysis. The evaluation should document the rationale and determination that gas intrusion into the ECCS system would not adversely affect the ability of the system to perform its function. If non condensable gases are vented through high-point vent valves, verify closure of the valves before starting the ECCS pumps (ABWR only). For the ESBWR passive plant ECCS designs with vent valves, if applicable, verify closure of vent valves before starting the system.
As part of the design process prior to ECCS low power and power ascension tests, the applicant or licensee should use the guidance in RG 1.82, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, (Ref. 8), to evaluate the susceptibility of active and passive ECCS to debris flow blockage.
Regulatory Guide 1.82 may be used to evaluate the susceptibility of the active BWR ECCS plant designs (e.g., ABWR) to debris flow blockage and support an engineering evaluation of data collected from ECCS preoperational tests to determine the adequacy of ECCS pumps to furnish the required flow rates and net positive suction head (NPSH) needed to maintain core cooling. The engineering evaluation should review the susceptibility of the ABWR ECCS suction strainers to debris flow blockage that can affect required ECCS pump performance and verify that the pumps can perform their intended safety function over the full range of postulated conditions up to and including design basis accident conditions.
Regulatory Guide 1.82 can also be used to evaluate debris flow blockage of passive ECCS plant designs (e.g., ESBWR), except no evaluation of ECCS pump performance is needed. The engineering evaluation of the effects of passive ECCS debris flow blockage should verify that the passive ECCS
systems can achieve adequate core cooling over the full range of postulated conditions up to and including design basis accident conditions.
RG 1.79.1, Page 4
C. STAFF REGULATORY GUIDANCE
The test program should include tests and analyses for pump and valve capabilities. This test program should determine acceptable pump NPSH and valve performance under system flow, temperature and pressure conditions as close to design as practical that allows analysis to demonstrate capability over the full range of operating conditions from normal operations to design basis accident conditions. For additional details, see Staff Regulatory Guidance C.2.b and C.2.c below.
Prerequisite verifications will need to occur prior to plant startup. For example, proper ESBWR
standby liquid control system (SLCS) tank volume, temperature control, and concentration of the neutron absorber solution, boron-10 enrichment, will need to be verified before entry into a technical specification mode in which SLCS operability is required.
1. System Testing for the ABWR and the Passive ESBWR Designs For BWR plants, initial plant tests of the following ECCS systems are described below. Staff regulatory guidance for component-specific testing is described in C.2 of this regulatory guide.
a. High-Pressure Core Flooder (HPCF) (ABWR)
(1) Preoperational Flow TestsCold Conditions. Verify proper operation of the HPCF
system, including related auxiliary equipment, pumps, valves, and instrumentation and control (I&C). The suppression pool and condensate storage tank (CST) should be available as pump suction sources, and the reactor vessel should be properly prepared to receive injection flow. The instrument air system, makeup water condensate system, residual heat removal (RHR) system, remote shutdown system, reactor building cooling water system, and appropriate electrical power sources should be available as needed to support the specified testing and appropriate system configuration. System testing should do the following:
(a) Verify the proper operation of system software-based instrumentation and control (I&C) systems. Check system functional, performance, and interface requirements as specified in design specifications and hardware/software system specifications.
(b) Verify all component alarms and proper alarm actuation by operating the detectors of the alarm or using simulated alarm signals.
(c) Verify proper operation of system valves. For additional details, see Staff Regulatory Guidance C.2.b below.
(d) Verify proper operation of pumps and motors during continuous run tests. Verify acceptable pump net positive suction head (NPSH) by monitoring acceptable pump performance under system flow, temperature, and pressure conditions as close to design as practical that allows analysis to demonstrate capability over the full range of operating conditions from normal operations to design basis accident conditions.
(e) Verify pump NPSH requirements are satisfied by monitoring acceptable pump performance under system operating conditions as close to the design as practical RG 1.79.1, Page 5
for pumps, valves, piping, and instruments in the system through the following tests.
(1) Minimum Flow Operational Test: Test the HPCF pump manually using the flow path from suppression pool to suppression pool through the minimum flow line until the temperature of each pump and motor bearing is stabilized.
(2) Rated Core Flooding Operational Tests: Check proper operation of the system at rated core flooding using the test line for flooding the suppression pool through the minimum flow line until the temperatures of the pump and motor bearing stabilize.
(3) High-Pressure Flooding Operational Test: Check proper operation of the system at rated core flooding using test lines for injecting into the suppression pool (or CST) to the suppression pool. The test should be performed continuously from the pump motor start to minimum flow with lowest acceptable suppression pool and CST levels.
(4) Reactor Injection Test: Check proper operation of the system using the core flood line to confirm that core flooding run out is performed.
(5) Alternate Source Verification Test: Verify that the water source can be transferred from the CST to the suppression pool.
(6) Automatic Start Test: Verify that the system start time is within safety injection requirements. Verify proper operation of keep fill and venting components to prevent water hammer damage.
(f) Verify proper pump and motor start sequence and actuation, using all start signals, including testing from the remote shutdown panel, operation of interlocks, protective devices, and all components subject to interlocks and protective devices. This includes proper operation of permissive, prohibit, and bypass functions.
(f) Verify proper operation while powered from the primary and alternate power sources, including power transfers.
(g) Verify acceptable pump/motor vibration levels and system piping movement during steady-state and transient operation. This test may also be performed with the expansion, vibration, and dynamics effects preoperational test. For additional guidance on adequate pump vibration levels and piping movement, see Staff Regulatory Guidance C.2.c and C.2.g below.
(2) Power Ascension Flow TestHot Conditions. Verify the HPCF system initiates automatically, at the appropriate low water level set point. Verify the minimum capacity and maximum delay time between the time the vessel water level drops below the set point and makeup water enters the vessel meets safety analysis requirements.
RG 1.79.1, Page 6
b. Automatic Depressurization System (ABWR and ESBWR)
(1) Preoperational Instrumentation and Control TestCold Conditions. This preoperational test should demonstrate proper operation of automatic depressurization system (ADS)
I&C subsystem logic functions and safety system logic and control (SSLC) functions.
(a) A number of ADS safety relief valves (SRVs) operate from either ADS loss-of- coolant accident (LOCA) initiation logic signals or safety/relief steam pressure logic signals. The ADS is initiated by high drywell pressure and/or low reactor vessel water level. This test should verify that ADS logic functions meet their design acceptance criteria.
(b) The ADS logic preoperational test verifies integrated automatic decision making and trip logic functions associated with the safety actions of the ADS.
(c) This test should verify that ADS accumulator capacity meets the required number of cycles for operating the SRVs.
(d) Verify proper operation of the depressurization valves (DPVs) (ESBWR only)
and SRV position indication.
(2) Power Ascension TestHot Conditions. The objective of this test is to demonstrate that SRVs can be manually opened and closed properly during power operation. The plant should be at the appropriate operational configuration, with prerequisite testing complete.
The applicable instructions should be checked or calibrated before testing begins.
(a) Perform a functional test of each SRV during plant heat up to the 50-percent power plateau. Open and close each valve to verify steam flow using discharge tail pipe temperature sensors. The SRV open and close indications and the tailpipe temperature/flow sensor should function as designed (b) This test should verify that there is no leakage from the SRV by monitoring tail pipe temperature to confirm SRV closure.
(c) This test should verify the steam flow through each SRV should not vary significantly from the average for each valve. Each valve should properly reseat after testing.
The operators may also verify changing indications of SRV position in comparison to changing turbine valve positions and/or generator load output. These changes may also be evaluated to detect anomalies indicating restriction or blockage in a particular SRV
tailpipe by making a valve-to-valve comparison.
c. Reactor Core Isolation Cooling (ABWR)
(1) Preoperational Flow TestCold Conditions. The purpose of these preoperational tests is to verify proper operation of the Reactor Core Isolation Cooling (RCIC) system, including the turbine, pump, valves, instrumentation and controls. The signals to automatically start the RCIC system at low reactor water level or high drywell pressure and the signal for automatic isolation of the RCIC system at low steam pressure to the RCIC pump turbine should be blocked before performing this test. In addition, measures RG 1.79.1, Page 7
should be taken to verify interfacing systems will not be affected when signals involving those systems are used throughout this test. During this test, a temporary strainer should be installed in the pump suction.
The test may be performed using temporary steam supply (e.g., auxiliary boiler),
equipment, piping, and instrumentation as necessary for the test. Because preoperational testing is performed using a temporary steam supply, RCIC pump flow may be limited.
If this is the case, document the issue and schedule completion of testing during the power ascension test phase.
The preoperational tests should include individual component and integrated system tests.
All required interfacing systems such as the reactor pressure vessel, suppression pool, condensate storage pool, instrument air system, condensate makeup water system, reactor building cooling water system and communications should be available to support specified testing and system configurations. The turbine instruction manual should be reviewed in detail so that precautions relative to turbine operations are followed. These tests should be performed using RCIC system design specifications and manufacturers recommendations to demonstrate the following tests:
(a) Verify proper operation of RCIC system software-based I&C. The test should verify system functional, performance, and interface requirements as specified in design specifications and hardware/software system specifications.
(b) Verify all RCIC component alarms and alarm actuations, using actual or simulated alarm signals.
(c) Verify alignment of RCIC system suction from the condensate storage pool (or tank) and inject water into the reactor through the reactor feedwater line with the reactor at atmospheric conditions. This test should also verify proper operation of all RCIC system valves, including operability, position indication and timing.
For additional details, see Staff Regulatory Guidance C.2.b below.
(d) Verify proper operation of the RCIC turbine and supporting sub components.
Perform a RCIC pump turbine quick start under simulated automatic initiation signal with suction from the CST. The test should demonstrate proper system flow rate and time to rated flow with no malfunctions in the system. Verify pump NPSH by monitoring acceptable pump performance under system flow, pressure and temperature conditions as close to design as practical that allows analysis to demonstrate capability over the full range of operating conditions from normal operations to design basis accident conditions. If additional testing is needed, this test may also be performed during the power ascension test phase.
(e) Verify acceptable RCIC pump/turbine vibration levels and system piping movements during both transient and steady-state operation. This test may also be performed with the expansion, vibration, and dynamic effects preoperational test. For additional details on pump vibration testing and piping movements, see Staff Regulatory Guidance C.2.c and C.2.g below.
(f) Verify RCIC system performance during the following modes of operation. The test should be performed using a temporary steam supply, spool piece, piping, other equipment and instrumentation necessary to conduct the test:
RG 1.79.1, Page 8
(1) At minimum flow mode with suction from the suppression pool or from the condensate storage pool and return to the suppression pool.
(2) At test mode with required pump flow rate and head through the system test line with suction from the suppression pool and return to the suppression pool.
(3) At injection mode with suction from the condensate storage pool and inject water into the reactor through the reactor feedwater line with the reactor at atmospheric pressure conditions.
(4) Turbine quick start in response to the simulated automatic initiation signal with suction from the condensate storage pool. This test should demonstrate proper system flow rate and time to rated flow with no malfunctions in the system.
(g) Verify proper operation of RCIC interlocks and equipment protective devices in turbine, pump and valve controls (h) Verify proper operation of RCIC permissive, prohibit and bypass functions.
(i) Verify proper RCIC system operation while powered from the primary and alternate power sources, including transfer, and in degraded modes for which the system is expected to remain operational. The RCIC system should demonstrate its ability to start without the aid of alternating current (AC), except for RCIC
direct current/AC inverters. An evaluation of RCIC operation beyond its design basis for extended loss of AC power along with RCIC support systems should be performed with non RCIC station batteries disconnected to verify RCIC direct current components will remain operable to meet 10 CFR 50.63 requirements.
(j) Verify proper operation of the RCIC system barometric condenser condensate pump and vacuum pump. For newer RCIC designs, the barometric condenser, condensate pump, and vacuum pump may have been eliminated; therefore, testing is not applicable.
(k) Verify the ability of the RCIC system to swap pump suction sources from the CST to the suppression pool without interrupting system operation.
(l) Verify that all RCIC system functions operate from redundant control locations, where appropriate.
(m) Verify proper operation of the pump discharge line keep-fill system ability to prevent water hammer damage during system startup and transients.
(2) Low Power Flow TestHot Conditions. The purpose of the RCIC low power test is verify proper operation of the RCIC system over its expected operating pressure and flow ranges, and to demonstrate reliability to automatically start from cold standby with the reactor at power. Test the RCIC system through a full flow test line to the suppression pool and by flow injection directly into the reactor vessel.
RG 1.79.1, Page 9
After all controller and RCIC system adjustments have been made, two consecutive successful reactor vessel injections, by automatic initiation from the cold standby condition, should be conducted to demonstrate system reliability. A cold standby condition is defined as a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any RCIC operation. Following these tests, system data will be collected while operating in the full flow mode to provide bench mark information for future Technical Specifications (TS) surveillance tests. The RCIC system should also be tested for two hours of continuous operation at rated flow conditions. All RCIC system and related auxiliaries tests should be evaluated to verify proper operation of the system.
(a) Verify proper operation of the RCIC turbine and supporting sub components.
Perform a RCIC pump turbine quick start test at low power in the manual and automatic start mode with suction from the CST. The test should demonstrate proper system flow rate and time to rated flow with no malfunctions in the system. Verify acceptable pump NPSH by monitoring acceptable pump performance under system flow, pressure and temperature conditions as close to design as practical that allows analysis to demonstrate capability over the full range of operating conditions from normal operations to design basis accident conditions.
(b) Test RCIC capability to inject water into the suppression pool in the manual and automatic start mode and steady-state operation at near rated reactor pressure in the full flow test mode. During this test, throttle pump discharge pressure in order to simulate reactor pressure and the expected pipeline pressure drop. The RCIC turbine speed control loop will be adjusted at near rated reactor pressure conditions.
(c) Test RCIC capability to inject water into the reactor vessel at rated reactor pressure to complete controller adjustments to demonstrate automatic starting from hot standby condition. The test should demonstrate automatic injection into the reactor vessel at rated reactor pressure. Verify satisfactory RCIC system performance under the final set of controller settings after controller adjustment are made by changes in speed and flow demand and then verify system response at both low and near rated RCIC pump flow conditions.
(d) After completing RCIC system controller adjustments, test automatic initiation of the RCIC system from standby conditions to demonstrate RCIC system reliability. Collect system data in the full flow test mode to provide benchmark data for future surveillance tests. For these tests, evaluate RCIC system and related auxiliary system data. For the RCIC system steam line flow trip setting in the leak detection and isolation system trip logic, collect sufficient operating data to make adjustment and verify correct trip set points.
(e) During low power testing, the RCIC turbine should not trip or isolate during manual or automatic start tests.
(f) Testing should indicate the average RCIC pump discharge flow equal to or greater than the 100% rated value specified in the RCIC system process flow diagram for all operating modes.
RG 1.79.1, Page 10
(g) Testing should indicate the start time for the RCIC system from receipt of signal to delivering design flow within the limit specified in the applicable RCIC
system design specification from low to rated reactor pressure.
(h) During low power testing, the RCIC turbine and pump flow control loops should be adjusted so that the RCIC flow related to variable responses to test inputs are as stated in the RCIC design specifications.
(i) During low power testing, the RCIC Turbine Gland Seal System should be capable of preventing significant steam leakage to the atmosphere.
(j) For automatic RCIC start tests, the transient first start followed by turbine speed peaks should not exceed the requirements specified in the RCIC vendor Startup Test Specifications document.
(k) The RCIC Turbine Steam Supply line high flow isolation trip should be calibrated to actuate at the value specified in plant TS.
d. Gravity-Driven Cooling System (ESBWR)
(1) Preoperational Instrumentation and Flow TestsCold Conditions. The gravity-driven cooling system (GDCS) is a unique ESBWR passive cooling system to provide gravity- driven flow into the reactor vessel for emergency core cooling during LOCA conditions.
The objective of this test is to verify the operation of all four trains of the GDCS,
including valves, logic, and instrumentation. The reactor vessel should be ready to accept GDCS flow. Instrument calibration and checks have been completed. To prevent actuation of single-use squib valves during the logic portion of testing, it may be necessary to remove the valves and install test spool pieces in the flow path with test fixtures used to evaluate the explosive charges. For additional details on squib valves testing, see Staff Regulatory Guidance C.2.b, Item (3). The tests should achieve the following:
(a) Verify operation of all instruments and equipment to appropriate design logic combinations and instrument channel trips. Verify all GDCS functions from redundant control locations, where appropriate.
(b) Verify instrumentation and alarm functions used to monitor system operation and availability.
(c) Verify proper operation of system valves. For additional details, see Staff Regulatory Guidance C.2.b below.
(d) Verify that GDCS flow from the GDCS pool and suppression pool through the reactor vessel is not obstructed by measuring test flow rates consistent with a clear flow path in GDCS piping.
(e) Verify that flow passages to the drywell are not obstructed.
(f) Verify the required GDCS design flow rate under the lowest possible driving head provided by GDCS pool level.
RG 1.79.1, Page 11
(g) Verify the adequacy of the instrument channel response times, as measured from process variable input signals to the applicable process actuator confirmation signal.
e. Isolation Condenser System (ESBWR)
(1) Preoperational TestCold Conditions. The objective of this test is to verify operation of the isolation condenser system (ICS) loops, including valves, logic, and instrumentation.
High-pressure nitrogen must be available to operate the spring-loaded condensate return valves, and nitrogen-operated pneumatic rotary motor isolation valves and electrical power are available to operate valves and controls. Performance should be observed and recorded during a series of individual component and integrated system tests to demonstrate the following:
(a) Verify proper calibration of instrumentation and operation of instrumentation and equipment in appropriate design combinations of logic and instrument channel trip. Verify proper functioning of instrumentation and alarms used to monitor system operation and availability.
(b) Verify proper operation of system valves. Verify that the steam flow paths from the isolation condenser (IC)/passive containment cooling system (PCCS) pools to the atmosphere are not obstructed. Verify that IC steam and condensate-return piping flow passages are not obstructed by measuring test flow rates consistent with a clear flow path in IC/PCCS piping.
(c) Verify proper operation of IC/PCCS pool level control. Verify that the isolation condenser pool sub compartment valves are locked open. Verify operation of isolation of the isolation condenser containment isolation valves upon receipt of simulated isolation signal.
(d) Verify acceptable instrument channel response times, as measured from each applicable process variable input signal to the applicable process actuator confirmation signal.
(2) Low-Power Flow TestHot Conditions. The objective of this test is to demonstrate operation of the four isolation condensers when supplied with reactor steam at rated pressure. The instrumentation should be checked and calibrated. Any required expansion, vibration, and temporary flow and temperature measurement instrumentation for ICS piping must be in place.
(a) At approximately 20-percent steady-state power, initiate operation of one ICS
train by opening the condensate return valve and condensate return bypass valve.
Verify acceptable heat removal capability by measuring ICS steady-state flows, temperatures, and isolation valve/passive containment pool level changes and temperatures. Perform this heat removal capacity test on one train of ICS at one time. In accordance with TS surveillance requirements, all other trains of ICS
should be tested. Verify acceptable heat removal capability by measuring ICS
steady-state flows, temperatures, and IC/PCCS pool level changes and temperatures. These tests should confirm proper startup, operation, and shutdown of each ICS train. Determine proper operation to verify:
RG 1.79.1, Page 12
(1) Measurement of vibration, displacement, and strain on the ICS heat exchanger, piping, and tubing,
(2) Measurement of steam inlet and condensate flow return to the reactor,
(3) Change in bulk temperature and temperature profiles within the IC/PCCS
pool, and
(4) Level changes in the IC/PCCS pools.
f. Standby Liquid Control System (ESBWR)
(1) Preoperational Flow TestCold Conditions. The objective of this preoperational test is to verify that the operation of the standby liquid control system (SLCS), including accumulator, tanks, control, logic, and instrumentation, is as specified. The reactor vessel should be available to inject water. Required interfacing systems should be available, as needed, to support the specified testing and the appropriate system configurations. To prevent actuation of single-use squib valves during the logic portion of this testing process, the valves may be isolated electrically to prevent actuation. This process of isolation, verification of the firing signal during the test, and subsequent reconnection must be controlled within the test document. Performance should be observed and recorded during a series of individual component and integrated system tests to demonstrate the following:
(a) Verify the proper calibration of instruments and the operation of all instruments and equipment in the required combinations of logic and instrument channel trip.
Verify the proper functioning of instruments and alarms used to monitor system operation and availability.
(b) Verify proper functionality of redundant accumulator equipment room electric heaters.
(c) Verify proper operation of system valves. For additional details, see Staff Regulatory Guidance C.2.b below.
(d) Verify proper operation of the nitrogen pressurization system.
(e) Verify proper system flow paths and discharge. Demineralized water may be used in place of neutron absorber water mixture.
(f) Verify proper operation of interlocks and equipment protective devices in valve controls.
(g) Verify acceptable instrument channel response times, as measured from each applicable process variable input signal to the applicable process actuator confirmation signal.
(h) As a prerequisite to plant startup, verify proper SLCS tank volume, temperature control, and concentration of the neutron absorber solution, boron-10 enrichment, before entry into a technical specification mode in which SLCS operability is RG 1.79.1, Page 13
required. Verify the design flow rate of sodium pentaborate solution in each accumulator by using the test tank.
g. Low-Pressure Flooder (LPFL) mode of Residual Heat Removal (RHR) (ABWR)
(1) Preoperational Flow TestCold Conditions. The purpose of this preoperational test is to test proper operation of the RHR system in the LPFL mode. These tests should verify the following:
(a) Verify adequate NPSH to the RHR pumps in the LPFL operating mode by monitoring acceptable pump performance under system conditions as close to design as practical that allows analysis to demonstrate capability over the full range of operating conditions from normal operations to design basis accident conditions. For additional details, see regulatory guidance C.2.c.(2) below.
(b) Verify proper operation of LPFL valves and pumps within each train and cross- connecting train of the RHR system from all control locations.
(c) Verify proper operation while powered from the primary and alternate power sources, including transfer, automatic startup, timing, and sequencing. This test may also be performed with the integrated ECCS loss-of-offsite-power/LOCA
preoperational test.
(d) Verify acceptable flow-induced vibration during testing of the LPFL sparger structure and end bracket attachments.
h. Residual Heat Removal Systems (ABWR)
(1) Preoperational Flow TestsCold Conditions. The objective of these preoperational tests is to verify individual component and integrated RHR system tests for the RHR system under each mode of RHR operation. The RHR system LPFL mode is tested in Staff Regulatory Guidance C.1.g above. The RHR system specification tests to support each mode of RHR operation include the following:
(a) Verify proper operation of the RHR system in each test and design mode of operation. The modes include the test mode, shutdown cooling mode, wetwell to drywell spray cooling mode, suppression pool cooling mode, LPFL mode, minimum flow mode and the fuel pool cooling mode. Based on the flow test configuration for each mode, the preoperational tests should verify adequate flow rates and NPSH for each RHR pump. The tests should verify RHR pump performance under system conditions as close to the design as practical that allows for analysis to demonstrate RHR system capability over the full range of operating conditions from normal operation to design basis accident conditions.
For additional details on pump tests, see Staff Regulatory Guidance C.2.c. below.
(b) Verify proper operation of instrumentation and system controls in all combinations of instrument channel trip logic.
(c) Verify proper operation of the RHR water leg pump with flow through a bypass loop around the water leg pump. The water leg pump discharges to the RHR
main line to keep it and all of its branches filled with water. Any makeup water RG 1.79.1, Page 14
needed is drawn from the RHR pump suction lines open path to the suppression pool. The ability to keep the RHR line filled with water prevents damaging water hammer during system startup and transient operation. Tests should confirm the RHR water leg pump trip on startup of the RHR pump.
(d) Verify proper operation of interlocks and equipment protective devices in pump and valve controls, including those designed to protect low-pressure portions of the system from the reactor coolant system (RCS) at high pressure.
(e) Verify proper operation of permissive, prohibit, and bypass functions.
(f) Verify acceptable pump and motor vibration levels and system piping movements during transient and steady-state operation. For additional details, see the Staff Regulatory Guidance in C.2.c and C.2.g below.
(g) Verify the proper operating flow conditions of each RHR pump during continuous operation, with a flow path from the suppression pool through the RHR heat exchanger and return to the suppression pool.
(h) Verify proper operation of the RHR heat exchangers. This test may also be performed during the startup test phase when load test conditions from reactor steam are available.
(i) Verify proper operation of system valves. For additional details, see Staff Regulatory Guidance C.2.b below.
(2) Low-Power TestHot Conditions. This low power test will demonstrate the ability of the RHR system to remove decay heat from the nuclear boiling system and safely place the plant in the shutdown cooling mode of operation. Pertinent system parameters will be monitored in the suppression pool cooling and shutdown cooling modes to verify that overall system operation and heat removal capabilities are in accordance with design requirements.
(3) Low-Power TestHot Conditions. (ESBWR reactor water cleanup shutdown cooling mode). This low power test will demonstrate the ability of the reactor water cleanup (RWCU) non regenerative heat exchangers to remove decay heat from the nuclear boiling system to place the plant in a safe-shutdown cooling mode of operation.
(a) Verify that the heat removal capacity of the RWCU non-regenerative heat exchangers as determined by flow rates and the temperature differential indicated on the RWCU system process flow diagram.
2. Component Testing The components of the systems involved in the system tests described in Staff Regulatory Guidance C.1 should be tested, either in conjunction with the system tests at the appropriate test phase or by independent component tests. Components that are common to the ECCS and other systems should be tested according to whichever systems have the more stringent criteria. For the preoperational, low power and power ascension system tests noted above, the component tests (e.g., pumps, valves, piping, etc.) are normally performed at the operating temperature and RG 1.79.1, Page 15
pressure conditions noted for each system test (e.g., preoperational flow tests - cold or hot conditions, low power or power ascension tests - hot conditions). If the component is not fully tested in the preoperational system test phase, then separate component or system tests at the low power or power ascension test phase may be performed to fully test the component to demonstrate the satisfaction of the test acceptance criteria. For additional details on low power and power ascension tests, see RG 1.68. Performance data should be recorded and the following items verified:
a. Instrumentation
(1) Verify that test acceptance criteria are met for operation of initiating instrumentation in various combinations of logic and instrument channel trip.
(2) Verify that test acceptance criteria are met for functioning of instrumentation and alarms used to monitor system availability. Instruments and alarms should be calibrated and tested before plant startup.
b. Valves
(1) Verify that test acceptance criteria are met for operation of ECCS valves (e.g., power- operated and check valves), including response times with the applicable energy source (e.g., air/nitrogen supply or electric power source) at system flow, temperature and pressure conditions as close to design as practical that allows analysis to demonstrate capability over the full range of operating conditions from normal operations to design basis accident conditions. Valve operation testing will include opening and closing valves with operating switches, valve status indication, and travel timing, if applicable.
Verification of valve position will include a method that ensures the valve disk is in its proper position as well as proper control room indication. Verify valves open, close and throttle to their correct valve position and meet design, leak rate and test acceptance criteria.
(2) With the exception of pyrotechnic-actuated (squib) valves that are addressed in (3)
below, verify valve operation under maximum pressure, differential pressure, temperature and flow conditions (consistent with system test limitations) with evaluation of sufficient valve-specific diagnostic data to demonstrate that each valve is capable of performing its safety function over the full range of operating conditions from normal operations to design-basis accident conditions.
(3) Verify the capability of squib valves by initiating the actuator control circuitry for each valve to demonstrate acceptable electrical parameters with the charge removed from the valve, by performing external and internal examinations for structural integrity and presence of foreign material and fluids, and by firing a sample of pyrotechnic charges from the valve population in a test fixture to demonstrate their design-basis capability.
Verify that the squib valve receives a simulated signal at the valve electrical leads that is capable of actuating the valve. Verify, by analysis or other simulated test, that the squib valve flow resistance is consistent with the flow path resistance.
RG 1.79.1, Page 16
c. Pumps
(1) Verify proper operation of injection pumps in all design operating modes.
(2) Verify test acceptance criteria are met for available NPSH by monitoring pump performance under system flow, pressure and temperature conditions as close to design as practical that allows analysis to demonstrate capability over the full range of operating conditions from normal operations to design basis accident conditions to provide reasonable assurance that the NPSH requirements are satisfied during pump operation.
Additional staff guidance with regard to pump NPSH is included in RG 1.82. Some indications of insufficient NPSH available might include erratic or decreasing pump motor current, erratic flow or flow less than expected due to vaporization, gas intrusion or flow blockage on the suction side of the pump, or frequent adjustments to the pump discharge valves to maintain a constant flow rate.
The test should also verify, by inspection, that no foreign material has entered into the pump, to ensure that performance degradation does not occur, and it should verify that there is no debris in the sump and the pump suction strainer is not clogged with debris, so that pump failures or other system degradation does not occur. This inspection provides verification that foreign material/debris has not entered the system during construction and may involve inspecting and removing a temporary test strainer or inspecting and cleaning of a permanent pump suction strainer (if one is installed) and need not necessitate a pump disassembly.
(3) Verify that test acceptance criteria are met for individual pump capacity and discharge head for the full range of pump operation, including recirculation flow.
(4) Verify that test acceptance criteria are met for pump response time (time to reach rated flow conditions) under voltage and frequency as close to design minimum values as practical.
(5) Verify test acceptance criteria are met for vibration levels. ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) (Ref. 9), Subsection ISTB, In- service Testing of Pumps in Light-Water Reactor Nuclear Power Plants, as incorporated by reference in 10 CFR 50.55a, specifies preservice and inservice testing of pumps, including monitoring pump vibration in units of either pump displacement or pump velocity with acceptance criteria for both units of measurement. NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, (Ref. 10) also provides guidance for pump vibration monitoring.
d. Motors
(1) Verify that test acceptance criteria are met for pump motor performance under system flow, pressure and temperature conditions as close to design as practical that allows analysis to demonstrate capability over the full range of operating conditions from normal operations to design basis accident conditions to provide reasonable assurance that the NPSH requirements are satisfied by monitoring acceptable pump performance during pump operation. Some indications of insufficient NPSH available might include erratic or decreasing pump motor current.
RG 1.79.1, Page 17
(2) Verify that test acceptance criteria are met for motor start sequence, over speed protection, and adequate margins between motor running and currents and protective breaker ratings. RG 1.205, Risk-Informed, Performance Based Fire Protection for Existing Light Water Nuclear Power Plants (Ref. 11), Regulatory Position C.3.3, Circuit Analysis, provides additional guidance for the coordination and testing of protective breakers to prevent thermal overload of electrical motors.
(3) Verify that test acceptance criteria are met for pump response time (time to reach rated flow conditions) under motor voltage and frequency as close to minimum design values as practical.
e. Controls
(1) Verify that test acceptance criteria are met for operation of controls, including controls that transfer pump suction. The tests should also verify separately and independently each channel or bus to identify any failures or losses of redundancy. Testing should include all backup and redundant controls.
(2) Verify that test acceptance criteria are met for the operation of interlocks and equipment protective devices in pump and valve controls.
f. Power Supplies
(1) Verify that test acceptance criteria are met for operation of normal and all alternative electric power supplies used for system valves, pumps, and motors, with analysis to confirm capability under degraded voltage and frequency.
(2) Verify that test acceptance criteria are met for operation of automatic and manual power transfer switches.
g. System Piping and Piping Supports
(1) Verify that test acceptance criteria are met for system piping movements under system startup conditions and during steady-state operation. The ASME Boiler and Pressure Vessel Code,Section III, Subsections NB/NC/ND-3620, Design Considerations, and NB/NC/ND-3622.3, Vibration (Ref. 12), provide a methodology for testing, monitoring, evaluating, and controlling piping system vibration.
(2) Verify test and examination criteria are met for dynamic restraints in the ECCS using the provisions in Subsection ISTD, Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants, of the ASME OM Code as incorporated by reference in 10 CFR 50.55a.
(3) To meet the requirements in 10 CFR 50, Appendix A, GDC 4, ECCS piping and piping support test acceptance criteria should include the dynamic effects associated with flow instabilities and loads (e.g., water hammer). NUREG-0927, Evaluation of Water Hammer Occurrences in Nuclear Power Plants, (Ref.13), NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, (Ref.
14) SRP Sections 3.9.3, 5.4.6 and 5.4.7 provide guidance for meeting the ASME Boiler and Pressure Vessel Code,Section III design requirements for piping loads (including water hammer), and for preventing and mitigating water hammer in ECCS piping RG 1.79.1, Page 18
connected to the RCS. ECCS piping and piping support design features and operating experience should be provided to prevent water hammer damage caused by such mechanisms as voided lines.
3. Documentation The ITP for BWRs should be documented in a summary report and retained as part of the plant historical record. This summary report should include the following:
a. a listing and description of the objectives of each test;
b. a description of how each test was conducted;
c. the parameters monitored;
d. complete comparisons and evaluations against design predictions or system performance requirements for the HPCF, HPCS, HPCI, HPFI, RCIC, ICS, LPFL, LPCS, LPCI, and RHR flow and isolation tests and ADS steam flow tests.
e. any discrepancies or deficiencies noted;
f. any unplanned actuations of ECCS due to design deficiencies, human errors, operational deficiencies and lessons learned from these events;
g. system modifications and corrective actions required;
h. appropriate justification for acceptance of systems or components not in conformance with design predictions or performance requirements;
i. any unexpected or unusual conditions during test observations; and j. conclusions.
Retention of the test procedures, data, and summaries by the licensee should be consistent with paragraph 9 of Appendix C to RG 1.68 and in accordance with GDC 1, Quality Standards and Records, of Appendix A to 10 CFR Part 50, and Criteria XI and XVII, Quality Assurance Records, of Appendix B to 10 CFR Part 50.
D. IMPLEMENTATION
The purpose of this section is to provide information on how applicants and licensees 1 may use this RG and information regarding the NRCs plans for using this RG. In addition, it describes how the NRC staff complies with 10 CFR 50.109, Backfitting and any applicable finality provisions in 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.
1 In this section, licensees refers to licensees of nuclear power plants under 10 CFR Parts 50 and 52; and the term applicants, refers to applicants for licenses and permits for (or relating to) nuclear power plants under 10 CFR Parts
50 and 52, and applicants for standard design approvals and standard design certifications under 10 CFR Part 52.
RG 1.79.1, Page 19
Use by Applicants and Licensees Applicants and licensees may voluntarily 2 use the guidance in this document to demonstrate compliance with the underlying NRC regulations. Methods or solutions that differ from those described in this RG may be deemed acceptable if they provide sufficient basis and information for the NRC staff to verify that the proposed alternative demonstrates compliance with the appropriate NRC regulations.
Licensees may use the information in this RG for actions that do not require NRC review and approval, such as changes to a facility design under 10 CFR 50.59, Changes, Tests, and Experiments.
Licensees may use the information in this RG or applicable parts to resolve regulatory or inspection issues.
Use by NRC Staff The NRC staff does not intend or approve any imposition or backfitting of the guidance in this RG. The NRC staff does not expect any existing licensee to use or commit to using the guidance in this RG, unless the licensee makes a change to its licensing basis. The NRC staff does not expect or plan to request licensees to voluntarily adopt this RG to resolve a generic regulatory issue. The NRC staff does not expect or plan to initiate NRC regulatory action which would require the use of this RG. Examples of such unplanned NRC regulatory actions include issuance of an order requiring the use of the RG, requests for information under 10 CFR 50.54(f) as to whether a licensee intends to commit to use of this RG,
generic communication, or promulgation of a rule requiring the use of this RG without further backfit consideration.
During regulatory discussions on plant specific operational issues, the NRC staff may discuss with applicants or licensees various actions consistent with NRC staff positions in this RG, as one acceptable means of meeting the underlying NRC regulatory requirement. Such discussions would not ordinarily be considered backfitting. However, unless this RG is part of the licensing basis for a facility, the staff may not represent to the licensee that the licensees failure to comply with the positions in this RG constitutes a violation.
If a licensee believes that the NRC is either using this RG or requesting or requiring the licensee to implement the methods or processes in this RG in a manner inconsistent with the discussion in this Implementation section, then the licensee may file a backfit appeal with the NRC in accordance with the guidance in NUREG-1409, Backfitting Guidelines, (Ref. 15) and the NRC Management Directive 8.4, Management of Facility-Specific Backfitting and Information Collection (Ref. 16).
2 In this section, voluntary and voluntarily means that the licensee is seeking the action of its own accord, without the force of a legally binding requirement or an NRC representation of further licensing or enforcement action.
RG 1.79.1, Page 20
GLOSSARY OF ACRONYMS
ABWRadvanced boiling-water reactor ADAMSAgencywide Documents Access and Management System ACalternating current ADSautomatic depressurization system ASMEAmerican Society of Mechanical Engineers BWRboiling-water reactor CFRCode of Federal Regulations CSTcondensate storage tank DCDdesign certification document ECCSemergency core cooling system ESBWReconomic simplified boiling-water reactor GDCSgravity driven cooling system (ESBWR only)
HPCFhigh-pressure core flooder (ABWR only)
HPCIhigh-pressure coolant injection HPCShigh-pressure core spray HPFIhigh-pressure feedwater injection IAEAinternational atomic energy agency IC/PCCS - isolation condenser/passive containment cooling system I&Cinstrumentation and control ICSisolation condenser system ITP Initial Test Program LOCAloss-of-coolant accident LPFLlow-pressure flooder (ABWR only)
NPSHnet positive suction head RG 1.79.1, Page 21
NRCU.S. Nuclear Regulatory Commission OMBOffice of Management and Budget PCCSpassive containment cooling system (ESBWR only)
PWRpressurized water reactor RCICreactor core isolation cooling RGregulatory guide RHRresidual heat removal RWCUreactor water cleanup system SRVsafety relief valve SSLCsafety system logic and control SLCSstandby liquid control system SSCstructure, system, and component The following list of additional acronyms is used in Appendix A of this guide including acronyms used in Figures A-1 through A-11:
AOair operated ATWSanticipated transient without scram DIVdivision DPVdepressurization valves EH electro hydraulic ESF engineered safety feature FEflow element FP fuel pool FPCfuel pool cooling FPCUfuel pool cooling and cleanup FCSflammability control system HPCFhigh pressure core flooder system (ABWR only)
RG 1.79.1, Page 22
HX heat exchanger IC/PCCSinside containment/passive containment cooling system LPFLlow pressure flooder system (ABWR only)
M motor MOmotor operated MS main steam MUCWmakeup cooling water NBS nuclear boiler system NNSnon nuclear system P pressure RCIC reactor core isolation cooing system RCSreactor coolant system RCWreactor building cooling water system RHRresidual heat removal system RPV reactor pressure vessel Rredundant SLCstandby liquid control SO solenoid operated S/P suppression pool SPCUsuppression pool cleanup system SRPstandard review plan Ttemperature Note: Some acronyms may be defined more than once in this guide since they were used with slightly different vendor terms.
RG 1.79.1, Page 23
REFERENCES 3
1. U. S. Code of Federal Regulations (CFR) Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter 1, Title 10, Energy.
2. CFR, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Chapter 1, Title 10, Energy.
3. U.S. Nuclear Regulatory Commission (NRC), Regulatory Guide (RG) 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plants, Washington, DC.
4. NRC, RG 1.79, Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors, Washington, D.C.
5. International Atomic Energy Agency (IAEA), Safety Standard Series Safety Guide No. NS-G-
1.9, Design of the Reactor Coolant System and Associated Systems in Nuclear Power Plants, Vienna, Austria, 2004. 4
6. NRC Regulatory Issue Summary 2013-09, NRC Endorsement of NEI 09-10, Revision 1a- A,Guidelines for Effective Prevention and Management of System Gas Accumulation, issued August 23, 2013, Washington, D.C. (ADAMS Accession No. ML13178A152)
7. Nuclear Energy Institute (NEI) 09-01, Revision 1a-A, Guidelines for Effective Prevention and Management of System Gas Accumulation, April 2013 (ADAMS Accession No:
8. NRC, RG 1.82, Water Sources for Long-Term Recirculation Cooling Following a Loss-of- Coolant Accident, Washington, D.C.
9. American Society of Mechanical Engineers (ASME), Operations and Maintenance of Nuclear Power Plants, (OM) Code, Division 1, Section IST - Light Water Reactor Nuclear Power Plants, ASME, New York, NY. 5
10. NRC, NUREG-1482, Guidelines for In-service Testing at Nuclear Power Plants, Section 5.4, Monitoring Pump Vibration In Accordance with ISTB, Revision 1, dated January 1995, U.S.
Washington, D.C. (ADAMS Accession No. ML11224A036)
11. NRC, RG 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, Washington, D.C.
3 Publicly available NRC published documents are available electronically through the NRC Library on the NRCs public Web site at: http://www.nrc.gov/reading-rm/doc-collections/. The documents can also be viewed on-line or printed for a fee in the NRCs Public Document Room (PDR) at 11555 Rockville Pike, Rockville, MD; the mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548;
and e-mail pdr.resource@nrc.gov.
4 Copies of International Atomic Energy Agency (IAEA) documents may be obtained through their Web site:
WWW.IAEA.Org/ or by writing the International Atomic Energy Agency P.O. Box 100 Wagramer Strasse 5, A-1400
Vienna, Austria. Telephone (+431) 2600-0, Fax (+431) 2600-7, or E-Mail at Official.Mail@IAEA.Org
5 Copies of American Society of Mechanical Engineers (ASME) standards may be purchased from ASME, Three Park Avenue, New York, NY 10016-5990; telephone (800) 843-2763. Purchase information is available through the ASME
Web-based store at http://www.asme.org/Codes/Publications/.
RG 1.79.1, Page 24
12. ASME, Boiler and Pressure Vessel Code (B&PV),Section III, Sections NB/NC/ND-3620,
Design Considerations, and NB/NC/ND-3622.3, Vibration 2010 (with 2011 Addendum),
New York, NY.
13. NRC, NUREG-0927, Revision 1, Evaluation of Water Hammer Occurrences in Nuclear Power Plants, dated March 1984, U.S. Nuclear Regulatory Commission, Washington, D.C. (ADAMS
Accession No. ML071030267) (non-public)
14. NRC, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, D.C.
15. NRC, NUREG-1409, Backfitting Guidelines, July 1990, Washington, D.C. (ADAMS
Accession No. ML032230247)
16. NRC, Management Directive (MD) 8.4, Management of Facility-Specific Backfitting and Information Collection, October 2004, NRC, Washington, D.C
RG 1.79.1, Page 25
BIBLIOGRAPHY
U.S. Nuclear Regulatory Commission Documents Generic Letters GL 85-22, Potential for Loss of Post-LOCA Recirculation Capability Due to Insulation Debris Blockage, December 3, 1985. (ADAMS Accession No. ML031150731)
GL 98-04, Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System after a Loss-of Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment, July 14, 1998. (ADAMS Accession No. ML031110081)
GL 2004-02, Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors, September 13, 2004. (ADAMS Accession No.
GL 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, January 11, 2008. (ADAMS Accession No. ML072910759)
Information Notices IN 92-85, Potential Failures of Emergency Core Cooling Systems Caused by Foreign Material Blockage, December 23, 1992. (ADAMS Accession No. ML031190717)
IN 93-34, Potential for Loss of Emergency Cooling Function Due to a Combination of Operational and Post-LOCA Debris in Containment, April 26, 1993 (ADAMS Accession No. ML031070498)
IN 93-34, Supplement 1, Potential for Loss of Emergency Cooling Function Due to a Combination of Operational and Post-LOCA Debris in Containment, May 6, 1993. (ADAMS Accession No.
IN 94-57, Debris in Containment and the Residual Heat Removal System, August 12, 1994. (ADAMS
Accession No. ML031060503)
IN 96-10, Potential Blockage by Debris of Safety System Piping Which Is Not Used During Normal Operation or Tested during Surveillances, February 13, 1996. (ADAMS Accession No. ML031060270)
IN 97-76, Degraded Throttle Valves in Emergency Core Cooling System Resulting from Cavitation- Induced Erosion during a Loss-of-Coolant Accident, October 30, 1997. (ADAMS Accession No. ML031050058)
IN 2006-20, Foreign Material Found in the Emergency Core Cooling System, October 16, 2006.
(ADAMS Accession No. ML062440467)
IN 2006-21, Operating Experience Regarding Entrainment of Air into Emergency Core Cooling and Containment Spray Systems, September 21, 2006. (ADAMS Accession No. ML062570468)
RG 1.79.1, Page 26
Bulletins Bulletin 93-02, Supplement 1, Debris Plugging of Emergency Core Cooling Suction Strainers, May 11,
1993. (ADAMS Accession No. ML031190684)
Bulletin 96-03, Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in Boiling- Water Reactors, May 6, 1996. (ADAMS Accession No. ML993410152)
Licensee Event Reports LER 50-333/2004-001, Inadvertent Actuation of Emergency Core Cooling Systems and Emergency Diesel Generators While in Refueling Mode, January 19, 2005. (ADAMS Accession No.
BWR Technical Specification Amendments Related to ECCS
James A. Fitzpatrick Nuclear Power Plant, Improved Standard Technical Specification (ITS) Conversion, Emergency Core Cooling System (ECCS) Instrumentation, Justification for Differences from NUREG-
1433, Revision 1, April 7, 1995. (ADAMS Accession No. ML011640094)
James A. Fitzpatrick Nuclear Power Plant Improved Standard Technical Specification (ITS) Conversion, ECCS - Shutdown, Discussion of Changes to CTS. April 7, 1995 (ADAMS Accession No. ML011630179)
LaSalle County Station Units 1 and 2, Application for Amendment to Appendix A Technical Specification Section 3/4.5.1, ECCS Operating, Action C, RCIC Declared Operable when Reactor Steam Dome Pressure Greater than 150 psig, April 12, 2000, (ADAMS Accession No. ML003704192)
Cooper Nuclear Station, Revise Technical Specification 3.5.1 to Incorporate Technical Specification Task Force (TSTF) 318 for One Low Pressure Coolant Injection (LPCI) Pump Inoperable in Each of the Two ECCS Divisions, August 25, 2003 (ADAMS Accession No. ML032450233)
Columbia Generating Station, Revised TS 3.3.5.2, RCIC System Instrumentation, and TS 3.5.2, ECCS
Shutdown to Increase Storage Tank Level, Amendment 210, September 30, 2008 (ADAMS Accession No.ML082610056)
RG 1.79.1, Page 27
APPENDIX A
DESIGN DESCRIPTIONS OF EMERGENCY CORE COOLING SYSTEMS
FOR NEW BOILING WATER REACTORS
The Boiling Water Reactor (BWR) Emergency Core Cooling System (ECCS) meets acceptance criteria that are based on the relevant requirements of General Design Criteria 4, 5, 33, 34, 35, 36, 37 and
55 of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities. The ICS and RCIC support functions must also provide alternate water injection during station blackout to meet 10 CFR 50.63, Loss of All Alternating Current Power. The advanced boiling water reactor (ABWR) and the Economic Simplified Boiling Water Reactor (ESBWR) passive plant design must also meet the NRC design requirements noted above.
The design descriptions for BWR ECCS in this Appendix also support the Staff Regulatory Guidance in Section C.
A.2 ABWR
The ECCSs for the ABWR consist of the following:
- the low-pressure flooder (LPFL) mode of RHR,
- the high-pressure core flooder (HPCF),
- RCIC, and
- the ADS subsystem of the nuclear boiler system (NBS).
Residual Heat Removal (RHR) System The RHR system is a closed system consisting of three independent pump loops that inject water into the vessel and/or remove heat from the reactor core or containment. Each of the pump loops contains the necessary piping, pump valves, and heat exchangers. In the core cooling mode, each loop draws water from the suppression pool and injects the water into the vessel outside the core shroud (via the feedwater line on one loop and via the core cooling subsystem discharge return line on two loops). In the heat removal mode, pump suction can be taken from either the suppression pool or the RPV.
With the pump suction taken from the suppression pool, the pump discharge within these loops provides a flowpath to the following points:
- the RPV (via feedwater on one loop and via the core cooling subsystem return lines on the other two loops) and
- the wetwell and drywell spray spargers (on two loops only).
With the pump suction taken from the RPV via the shutdown cooling lines (in the shutdown cooling mode), the pump discharge in these loops provides a flow path back to the following points:
- the reactor vessel (via the core cooling discharge return lines and feedwater line) and the upper reactor well, via the fuel cooling system (on two loops only).
Appendix A to RG 1.79.1, Page A-1
Low-Pressure Flooder (LPFL) Mode of Residual Heat Removal Figures A-1 to A-3 show the three-train system. Each of the three RHR divisions can be initiated manually (LPFL mode). The RHR system channel measurements are provided to the safety system logic and control (SSLC) for signal processing, setpoint comparisons, and generation of trip signal
s. The RHR
system is automatically initiated when either a high drywell pressure or a low reactor water level condition exists (i.e., a LOCA signal). An RHR initiation signal is provided to the system
s. The SSLC
processors use a two-out-of-four voting logic for RHR system initiation.
After receipt of an initiation signal, the RHR system automatically initiates and operates in the LPFL mode to provide emergency makeup to the reactor vessel. The initiation signal starts the pumps, which run in the minimum flow mode until the reactor depressurizes to less than the pumps developed head pressure. A low-reactor-pressure permissive signal occurs above the pumps developed head pressure, which signals the injection valve to open. As the injection valve opens, the testable check valve contains the reactor pressure until the pressure becomes less than the pumps developed head pressure in the minimum flow mode, at which time injection flow begins. This sequence satisfies the response requirements for all potential LOCA pipe breaks when the injection valve opens within a short time period (i.e., about 35 seconds) after receiving the low-reactor-pressure permissive signa
l. The LPFL
injection flow for each division begins when the reactor vessel pressure falls to within a specific value above the drywell pressure. When the reactor vessel pressure is about 0.275 MPa (40 psig) greater than the drywell pressure, the LPFL injection flow for each division is about 950 cubic meters per hour, minimum. The flow rates, pressures and time periods from an individual plant may vary from the design certification document (DCD) values. All three divisions of the RHR system accomplish the LPFL mode by transferring water from the suppression pool to the reactor pressure vessel (RPV) via the RHR heat exchangers. The system automatically aligns to the LPFL mode of operation from the test mode, the suppression pool cooling, or wetwell spray mode upon receipt of an initiation signal. The wetwell spray mode is applicable for division B or C. If a drywell spray valve is open in division B or C, that division automatically aligns to the LPFL mode in response to the injection valve beginning to ope
n. The RPV
injection valve in each division requires a low-reactor-pressure permissive signal to open and closes automatically upon receipt of a high-reactor-vessel-pressure signal.
Figure A-1. ABWR low-pressure flooder mode of residual heat removaltrain A
Appendix A to RG 1.79.1, Page A-2
Figure A-2. ABWR low-pressure flooder mode of residual heat removaltrain B
Figure A-3. ABWR low-pressure flooder mode of residual heat removaltrain C
Appendix A to RG 1.79.1, Page A-3
High-Pressure Core Flooder(HPCF)
The HPCF system (Figure A-4) consists of two HPCF loops (B and C) flooding water to the RPV
above the core. Each of the two loops belongs to a separate division; electrical and mechanical separation between the two divisions is complete. Locating each division in a different area of the reactor building ensures physical separation. The two loops are both high-pressure pumping systems (i.e., they are capable of injecting water into the reactor vessel over the entire operating pressure range). The reference pressure for the operating performance of the system at high pressure is the lowest spring (safety) setpoint of the safety relief valves (SRVs).
Both HPCF divisions take primary suction from the condensate storage tank (CST) and secondary suction from the suppression pool. In the event that the CST water level falls below a predetermined setpoint or the suppression pool water level rises above a predetermined setpoint, the pump suction will automatically transfer from the CST to the suppression pool. Both HPCF system loops have suction lines that are separate from the RHR loops.
The HPCF pumps are at an elevation below the water level in the suppression pool. The motor- operated valve in the suction line from the suppression pool on each division is normally closed because primary suction is taken from the CST. This valve automatically opens upon receipt of either of the suction transfer signals noted above. The suppression pool suction valves on each loop can be closed from the control room if a leak develops in the system piping downstream of the isolation valves.
Figure A-4. ABWR high-pressure core flooder Appendix A to RG 1.79.1, Page A-4
Reactor Core Isolation Cooling The purpose of the RCIC system (Figure A-5) is to supply high-pressure makeup water to the RPV when the reactor is isolated from the main condenser and the condensate and feedwater system is not available. The system is started automatically upon receipt of a low-water-level signal or manually by the operator. Water is pumped to the core by a turbine pump driven by reactor steam.
The functional classification of the RCIC system is as a safety-related and engineered safety feature (ESF) system. For the ABWR design, the RCIC system is considered to be part of the ECCS.
The RCIC system function is completely backed up by HPCF (ABWR only).
The RCIC system consists of a steam-driven pump with associated valves and piping capable of delivering water flow to the RPV. The turbine is driven by steam produced from decay hea
t. The RCIC
system takes water from the CST or the suppression pool and delivers it to the RPV to maintain adequate RPV level. The turbine exhaust is directed to the suppression pool, where it is condensed. If the condensate and feedwater system is isolated from the RPV, the RCIC system will start automatically when decay heat boils coolant to low reactor water level. The RCIC system supplies sufficient inventory to allow complete shutdown without compromising fuel clad integrity.
During RCIC operation, the suppression pool acts as the heat sink for steam generated by reactor decay heat. This will result in a rise in pool water temperature. Heat exchangers in the RHR system are used to maintain pool water temperature within acceptable limits by cooling the pool water directly during normal plan operation. A design flow functional test of the RCIC system may be performed during normal plant operation by drawing suction from the suppression pool and discharging through a full flow test return line back to the suppression pool. During the test, the discharge valve to the vessel remains closed and reactor operation remains undisturbed. During test mode operation, should an initiation signal occur, flow will be automatically directed to the vessel.
Appendix A to RG 1.79.1, Page A-5
Figure A-5. ABWR reactor core isolation cooling Automatic Depressurization System If the RCIC and HPCF systems cannot maintain the reactor water level, the Automatic Depressurization System (ADS) (Figure A-6), which is independent of any other ECCS, reduces the reactor pressure so that flow from the RHR system operating in the low-pressure flooder mode enters the reactor vessel in time to cool the core and limit fuel cladding temperature. The ADS employs SRVs to relieve high-pressure steam to the suppression pool.
The NBS channel measurements are provided for the safety-related instrumentation and control system SSLC for signal processing, set point comparisons, and generation of trip signals. Except for the pump running interlock, the SSLC uses a two-out-of-four voting logic for ADS initiation. The ADS logic is automatically initiated when a low-reactor-water-level signal is present. If the RPV low-water-level signal is present concurrent with a high-drywell-pressure signal, both the main ADS timer (less than or equal to 29 seconds) and the high-drywell-pressure bypass timer (less than or equal to 8 minutes) are initiated. Since these are ABWR DCD values, the ADS timers may vary for each individual plant design.
Absent a concurrent high-drywell-pressure signal, only the ADS high-drywell-pressure bypass timer is initiated. On the timeout of the ADS high-drywell-pressure bypass timer, concurrent with the RPV low- water-level signal, the main ADS timer is initiated, if not already initiated. The main timer continues to completion and times out only in the continued presence of an RPV low-water-level signal. On timeout of the main ADS timer, concurrent with positive indication by pump discharge pressure of at least one RHR
or one HPCF pump running, the ADS function is initiated.
Appendix A to RG 1.79.1, Page A-6
Figure A-6. ABWR automatic depressurization system There are four trains with two divisions of actuation signals for low reactor water level and high drywell pressure. The Division I control logic signal actuates one set of pilots, and sensors from all four trains for low reactor water and high drywell pressure and the Division II control logic signal actuates the second set of pilots, either of which initiates the opening of the ADS SRVs. Redundant trip channels arranged in two divisionally separated logics that control two separate solenoid-operated pneumatic pilots on each ADS SRV accomplish ADS initiation. Either pilot can operate the ADS valve. These pilots control the pneumatic pressure applied by the accumulators and the high-pressure nitrogen gas supply system. The direct-current power for the logic is obtained from SSLC Divisions I and II.
For mitigation of anticipated transient without scram (ATWS), the ADS has an automatic and manual inhibit of the automatic initiation. Automatic initiation of the ADS is inhibited unless there is a coincident low-reactor-water-level signal, and an average power range monitors the ATWS permissive signal from the neutron monitoring system. There are main control room switches for the manual inhibit of automatic initiation of the ADS.
The ADS can be initiated manually. On a manual initiation signal, concurrent with positive indication that at least one RHR or one HPCF pump is running, the ADS function is initiated.
The ADS automatically actuates in response to the ECCS initiation signals. A two-out-of-four- level initiation logic is used to activate the SRVs and ADS valves. The 10-second delay to confirm the level initiation signal ensures that momentary system perturbations do not actuate the ADS when it is not required. The two-out-of-four logic ensures that a single failure does not cause spurious system actuation and that a single failure cannot prevent initiation.
Appendix A to RG 1.79.1, Page A-7
A.3 Economic Simplified Boiling-Water Reactor Systems The ECCS for the economic simplified boiling-water reactor (ESBWR) consists of the following:
- the ICS,
- the gravity-driven cooling system (GDCS),
- the ADS, and
- the standby liquid control system (SLCS).
Isolation Condenser System The ESBWR ICS (Figure A-7) is a passive plant system relying almost exclusively on natural recirculation to drive plant flow, which differs significantly from the BWR RCIC, which relies on active systems to accomplish its functions.
Figure A-7. ESBWR isolation condenser system The ESBWR passive decay heat removal systems (isolation condensers) are capable of achieving and maintaining safe, stable conditions for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without operator action after non-LOCA
events. Operator action is credited after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to refill isolation condenser pools or initiate nonsafety shutdown cooling.
Appendix A to RG 1.79.1, Page A-8
The ICS removes residual sensible and core decay heat from the reactor, in a passive way and with minimal loss of coolant inventory from the reactor, when the normal heat removal system is unavailable after any of the following events:
- station blackout (i.e., unavailability of all alternating-current power),
- ATWS, and
- LOCA.
The ICS functions to avoid unnecessary use of other ESFs for RHR and in the event of a LOCA.
The ICS also provides additional liquid inventory upon opening of the condensate return valves to initiate the system. In the event of ICS initiation by a reactor level below Level 2, the ICS removes core heat, causing initial depressurization of the reactor before the ADS initiates. Because of this vessel pressure reduction with return of condensed steam plus the additional initial ICS stored condensate inventory, the ADS can initiate from a lower reactor water level to complete the vessel depressurization.
The ICS is designed as a safety-related system to remove reactor decay heat after reactor shutdown and isolation. It also prevents unnecessary reactor depressurization and operation of other ESFs, which can also perform this function.
In the event of a LOCA, the ICS provides additional liquid inventory upon opening of the condensate return valves to initiate the system. The ICS also provides initial depressurization of the reactor before ADS in the event of loss of feedwater, such that the ADS can take place from a lower water level.
To ensure that an adequate inventory of cooling water is available for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after an accident, the ICS uses automatically opening connections between the equipment storage pool and isolation condenser/passive containment cooling system pools. The ICS uses parallel pneumatic and squib valves to cross connect the equipment storage pool to the isolation condenser/passive containment cooling system pools.
Gravity-Driven Cooling System The GDCS (Figure A-8) provides emergency core cooling after an event that threatens the reactor coolant inventory. Once the reactor has been depressurized, the GDCS is capable of injecting large volumes of water into the depressurized RPV to keep the core covered for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after a LOCA.
The GDCS also drains its pools to the lower drywell in the event of a core melt sequence that causes failure of the lower vessel head and allows the molten fuel to reach the lower drywell cavity floor.
This action is accomplished by detection of elevated temperatures registered by thermocouples in the lower drywell cavity, and by logic circuits that actuate squib-type valves on independent pipelines draining GDCS pool water to the lower drywell region. Because inadvertent actuation of the automatic logic circuits could result in loss of GDCS pool inventory and the consequent unavailability of water for injection into the reactor vessel on a valid GDCS actuation signal, a set of safety-related temperature switches are used to inhibit deluge actuation as long as the drywell temperature is less than a preset value.
The GDCS requires no external electric power source or operator interventio
n. The GDCS
initiation signal is the receipt of a confirmed ECCS initiation signal from the NBS. This signal initiates ADS and GDCS injection valve timers as well as longer equalization valve timers in the GDCS logic.
After injection valve timer duration, squib valves are activated in each of the injection lines leading from the GDCS pools to the RPV, making GDCS flow possible. The actual GDCS flow Appendix A to RG 1.79.1, Page A-9
delivered to the RPV is a function of the differential pressure between the reactor and the GDCS injection nozzles. The initiation of GDCS provides short-term post-LOCA water makeup to the annulus region of the reactor through eight injection line nozzles, and by gravity-driven flow from three separate water pools in the drywell at an elevation above the active core region. The system provides long-term post- LOCA water makeup to the annulus region of the reactor through four equalization nozzles and lines (with squib valves) connecting the suppression pool to the RPV. During severe accidents (i.e., if the core melts through the RPV), the GDCS floods the lower drywell region directly via four GDCS injection drain lines (one each from two pools and two from the third pool) through a deluge system using squib valves.
Figure A-8. ESBWR gravity-driven cooling system Automatic Depressurization System The ADS (Figures A-9 and A-10) is a part of the ECCS. It operates to depressurize the reactor so that the low-pressure GDCS is able to make up coolant to the reactor. The ADS is a function of the NBS.
The depressurization function is accomplished using SRVs and DPVs.
The ADS automatically actuates in response to the ECCS initiation signals. A two-out-of-four- level initiation logic is used to activate the SRVs and DPVs. The 10-second delay to confirm level initiation signal ensures that momentary system perturbations do not actuate the ADS when it is not required. The two-out-of-four logic ensures that a single failure does not cause spurious system actuation and that a single failure cannot prevent initiation. The ADS in the ESBWR uses squib valves as DPVs.
Appendix A to RG 1.79.1, Page A-10
Figure A-9. ESBWR automatic depressurization system Figure A-10. ESBWR automatic depressurization system Appendix A to RG 1.79.1, Page A-11
Use of Dual-Function Components The ECCS ADS and GDCS are designed to accomplish only one function, to cool the reactor core after a LOCA. The ECCS SLCS is designed for use during an ATWS, and the ECCS ICS is designed to avoid unnecessary use of other ESFs for RHR. Both the SLCS and the ICS provide additional liquid inventory on actuation. To this extent, components or portions of these systems, except for the pressure relief function of SRVs, are not required for operation of other systems. Because the SRV opens either on an ADS initiating signal or by spring-actuated pressure relief in response to an overpressure condition, no conflict exists.
Standby Liquid Control System The shutdown function of the SLCS (Figure A-11) is manually initiated. In addition, the system is automatically initiated for ATWS and LOCA events.
The SLCS contains two identical and separate trains. Each train provides 50-percent injection capacity. All components of the SLCS in contact with the boron solution are constructed of or lined with stainless steel. The safety-related portions of the SLCS are listed in the ESBWR design control document. The SLCS requires support from safety-related interfaces.
The SLCS includes a nitrogen charging subsystem that consists of a liquid nitrogen tank, a vaporizer, a high-pressure pump, and associated valves and piping. This subsystem is used for initial accumulator charging and makeup for normal system losses during normal plant operations. It is a nonsafety-related subsystem (but it has safety-related piping and valves) inside the reactor building, extending from the makeup valves downstream to the accumulators. The nonsafety-related high-pressure cryogenic nitrogen equipment is outside the reactor building at grade elevation. The SLCS uses squib valves for injection valves.
The core bypass spargers are in the reactor vessel and penetrate through the shroud to the core.
The portions of the standby liquid control injection line downstream of each squib valve contain only stagnant reactor water. The major components of the SLCS, which are necessary for injection of sodium pentaborate solution (neutron absorber) into the reactor, are in the reactor building. Reactor building heating, ventilation, and air conditioning control the temperature and humidity conditions in the SLCS
equipment rooms to prevent solute precipitation in the accumulators and injection lines, thereby ensuring proper system operation. This system readiness function is nonsafety-related.
The SLCS performs safety-related functions; therefore, it is classified as safety-related and is designed as a Seismic Category I system. The SLCS meets the following safety design bases:
- Provides a diverse backup capability for reactor shutdown that is independent of normal reactor shutdown provisions. The SLCS provides makeup water to the RPV to mitigate the consequences of a LOCA. The sodium pentaborate in the SLCS solution is credited for buffering to ensure that the iodine chemical distribution assumed in the LOCA dose consequence analysis remains valid.
- Injects a boron solution, which performs as a neutron absorber, at multiple locations into the core bypass region at high velocity. This ensures adequate mixing and total injection of the solution to accomplish reactor shutdown. The injection geometry ensures balancing of reaction forces.
Appendix A to RG 1.79.1, Page A-12
Figure A-11. ESBWR standby liquid control system Appendix A to RG 1.79.1, Page A-13