ML16120A044
| ML16120A044 | |
| Person / Time | |
|---|---|
| Site: | 05200045 |
| Issue date: | 04/29/2016 |
| From: | Hitachi-GE Nuclear Energy, Ltd |
| To: | Office of New Reactors |
| Shared Package | |
| ML16120A041 | List: |
| References | |
| MFN 15-069, Supp 1 | |
| Download: ML16120A044 (44) | |
Text
Enclosure 2 MFN 15-069, Supplement 1 GEHs Response to NRCs Request for Supplemental Information to Item #26 - Fukushima Recommendation 4.2 -
Mitigation Strategies ABWR DCD Revision 6 Markups IMPORTANT NOTICE REGARDING CONTENTS OF THIS DOCUMENT Please Read Carefully The information contained in this document is furnished solely for the purpose(s) stated in the transmittal letter. The only undertakings of GEH with respect to information in this document are contained in the contracts between GEH and its customers or participating utilities, and nothing contained in this document shall be construed as changing that contract. The use of this information by anyone for any purpose other than that for which it is intended is not authorized; and with respect to any unauthorized use, GEH makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
Response to NRC Post-Fukushima Recommendations 1D Response to NRC Post-Fukushima Recommendations 1D.1 Introduction In response to the accident at the Fukushima Daiichi Nuclear Power Plant caused by the March 2011, Magnitude 9 Tohoku earthquake and subsequent tsunami, the Nuclear Regulatory Commission (NRC) established a Near-Term Task Force (NTTF) to review NRC processes and regulations to determine if improvements to its regulatory system were needed. The NTTF developed a set of recommendations intended to clarify and strengthen the regulatory framework for protection against natural phenomena. These recommendations were issued in SECY-11-0093 (Reference 1D.5-9).
SECY-11-0124 (Reference 1D.5-13) and SECY-11-0137 (Reference 1D.5-1) provided the NRC Commissioners with the Staffs recommendations, including prioritization for implementation.
Subsequently, SECY-12-0025 (Reference 1D.5-2) was issued describing proposed Orders to be issued to licensees and a draft request for information pursuant to 10 CFR 50.54(f). SECY-12-0025 stated that combined license plants under review would address the three orders and the request for information through the review process. On March 9, 2012, the Commission issued a Staff Requirements Memorandum (SRM) for SECY-12-0025 (Reference 1D.5-4) approving issuance of the orders and request for information with some modifications.
This appendix addresses the Tier 1 recommendations (recommendations that the Staff determined should be started without unnecessary delay) and Orders contained in SECY-12-0025, the Tier 2 recommendations (recommendations which could not be initiated in the near term) contained in SECY-11-0137, and the modifications documented in the SRM consistent with the as issued Orders (EA-12-049 [Reference 1D.5-5]), 050 (Reference 1D.5-6) and 51 (Reference 1D.5-7). The NRC recommendation in each of the following subsections is a summary of the recommendation from the NRC documents. The response to each recommendation discusses how ABWR addresses the recommendation. The numbers in parentheses of subsection headings correspond to the NTTF recommendation number.
1D.2.1 Seismic and Flooding Reevaluations (2.1) 1D.2.1.1 Seismic NRC Recommendation Perform a reevaluation of the seismic hazards using present-day NRC requirements and guidance to develop a Ground Motion Response Spectrum (GMRS). The seismic source models for the actual location of the plant (per applicable NUREG guidelines) may be used to characterize the hazards.
Response
In order to confirm seismic design adequacy of the standard plant, COL applicants shall perform a reevaluation of the seismic hazards using the current (at the time of COL) NRC requirements and guidance to develop a GMRS. The latest consensus seismic source models for Central America and the Central and Eastern United States Seismic Source Characterization (CEUS), NUREG-2115 (Reference 1D.5-8), may be used to characterize the hazards. The following COL License Information items provide additional guidance:
Page 1 of 12
Response to NRC Post-Fukushima Recommendations Item No. Subject Subsection 2.2 Seismic Design Parameters 2.3.1.2 19.4 Confirmation of Seismic Capacities Beyond the Plant Design 19.9.4 Bases 19.5 Plant Walkdowns 19.9.5 1D.2.1.2 Flooding NRC Recommendation Perform a reevaluation of all appropriate external flooding sources, including the effects from local intense precipitation on the site, probable maximum flood (PMF) on streams and rivers, storm surges, seiches, tsunami, and dam failures. It is requested that the reevaluation apply present-day regulatory guidance and methodologies being used for ESP and COL reviews including current techniques, software, and methods used in present-day standard engineering practice to develop the flood hazard.
The recommendation also noted that flooding risks are of concern because the safety consequences of a flooding event may increase sharply with a small increase in flooding level.
Response
The COL applicant will use the current (at the time of COL) regulatory guidance and methodologies to evaluate flooding hazards. The site selection process will take into account the worst case predicted flood. Then grade level and flood control methods (e.g., site grading) will be determined based on this predicted flood level. The grade level floor will be at least 0.3 meters above this predicted flood level.
Therefore, external flooding should not be a major concern for the ABWR.
The following COL License Information items provide additional guidance:
Item No. Subject Subsection 2.14 Floods 2.3.2.12 2.15 Probable Maximum Flood on Streams and Rivers 2.3.2.13 2.16 Ice Effects 2.3.2.14 2.19 Flooding Protection Requirements 2.3.2.17 19.3 Event Specific Procedures for Severe External Flooding 19.9.3 19.5 Plant Walkdowns 19.9.5 Page 2 of 12
Response to NRC Post-Fukushima Recommendations 1D.2.2 Seismic and Flooding Walkdowns (2.3)
NRC Recommendations Perform seismic walkdowns in order to identify and address plant specific degraded, nonconforming, or unanalyzed conditions and verify the adequacy of strategies, monitoring, and maintenance programs such that the nuclear power plant can respond to external events. The walkdown will verify current plant configuration with the current licensing basis, verify the adequacy of current strategies, maintenance plans, and identify degraded, non-conforming, or unanalyzed conditions. The walkdown procedure should be developed and submitted to the NRC. The procedure may incorporate current plant procedures, if appropriate. Prior to the walkdown, licensees should develop acceptance criteria, collect appropriate data, and assemble a team with relevant technical skills.
The NRC also requests that each addressee confirm they will use the industry developed, NRC endorsed, flood walkdown procedures or provide a description of plant specific walkdown procedures.
Response
Seismic and flooding plant walkdowns will be conducted after construction as documented in COL License Information Item 19.5 (Subsection 19.9.5).
1D.2.3 Station Blackout (SBO) Rulemaking (4.1)
NRC Recommendation Strengthen the station blackout (SBO) mitigation capability at all operating and new reactors for design-basis and beyond-design-basis events. This includes (1) a minimum coping time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for loss of all Alternating Current (AC) power, (2) establishing the equipment, procedures, and training necessary to implement an extended coping time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for core and spent fuel cooling and for reactor coolant system (RCS) and primary containment integrity, and (3) pre-planning and pre-staging offsite resources to support uninterrupted core and spent fuel cooling, and RCS and primary containment integrity under conditions involving significant degradation of offsite transportationinfrastructure associated with a significant natural disaster. This recommendation will be implemented by rulemaking.
SECY-12-0025 adds the requirement that the loss of the Ultimate Heat Sink (UHS) should be evaluated as part of the SBO evaluation.
Response
(1) ABWR design can withstand an SBO for an extended period of time using an alternate AC power source, the Combustion Turbine Generator (CTG) with the only requirement to provide fuel for the CTG. This is described in Appendix 1C (Table 1C-3). Additionally, the ABWR design can withstand a sustained loss of all AC power, including the loss of CTG, for 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> while maintaining core cooling, as shown in Figure 19E.2-6c and described in Subsection 19E.2.2.3.
Page 3 of 12
Response to NRC Post-Fukushima Recommendations The ABWR design has a number of features which mitigate an SBO and extended loss of all AC power:
x The primary mitigation for an extended SBO is provided by a CTG, which is independent from the Emergency Diesel Generators (EDGs) and can be connected to the 1E buses. This CTG has black start capability and can be available for use within 10 minutes. The CTG is housed in a structure designed to the Internal Code Council International Building Code. This structure is protected from design basis floods and adverse weather conditions. (DCD Tier 1, Subsection 2.12.11 and Appendix 1C).
x The Division I batteries have a capacity of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (Subsection 8.3.2.1.3.1). The COL applicant will develop battery loading profiles and define appropriate load shedding during station blackout to confirm actual capacity (COL License Information item 19.9 Subsection 19.9.9).
x The Reactor Core Isolation Cooling system can provide core cooling for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during SBO conditions without reliance on AC power (Appendix 19E.2.1.2.2.1).
x The Alternating Current-Independent Water Addition (ACIWA) system is a seismically qualified system with an external permanent diesel-driven pump and water supply capable of providing water to the Residual Heat Removal (RHR) system for core and containment cooling without reliance on AC power. Operation of the ACIWA system is described in Subsection 5.4.7.1.1.10.
x The ACIWA system can be used to provide makeup water to the Spent Fuel Pool (SFP).
(2) The ABWR Design has the installed equipment (e.g., ACIWA system) to implement an extended coping time in excess of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without reliance on AC power for core, spent fuel and containment cooling as documented in Subsection 19E.2.2.3. The 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of core cooling can be provided without reliance on the UHS. Relevant procedures and training will be developed per the Plant Operational Procedure Development Plan described in Section 13.5 (COL License Information item 13.3 Subsection 13.5.3.1).
(3) Pre-planning and pre-staging resources to support uninterrupted core and spent fuel pool cooling, and RCS and primary containment integrity under conditions involving significant degradation of the onsite facilities will be developed as part of the Emergency Procedure development (as required in COL License Information item 13.4, Subsection 13.5.3.2). Additionally, the COL Applicant will arrange for sufficient offsite resources to sustain core, containment, and spent fuel pool cooling for an extended period of time. These plans and resources will provide this capability under circumstances involving significant degradation of offsite transportation infrastructure associated with a significant natural disaster.
Detailed procedures and training associated with enhancing SBO mitigation capabilities in accordance with the SBO Rule will be developed during implementation of operational programs as described in Section 13.5 (COL License Information item 19.7, Subsection 19.9.7).
The following COL License information items provide additional guidance:
Item No. Subject Subsection 1.5 Emergency Procedures and Emergency Procedure Training 1A.3.1 Program 1.13 Station Blackout Procedures 1C.4.1 Page 4 of 12
Response to NRC Post-Fukushima Recommendations Item No. Subject Subsection 5.8 Analysis of Non-Design Basis Loss of AC Coping Capability 5.4.15.2 5.8 Analysis to Demonstrate the Facility has 8 Hour Non-Design 5.4.15.2.1 SBO Capability 5.8 Analysis to Demonstrate that the DC Batteries and SRV/ADS 5.4.15.2.2 Pneumatics have Sufficient Capacity 9.9 Spent Fuel Firewater Makeup Procedures and Training 9.1.6.9 9.10 Protection of RHR System Connections to FPC System 9.1.6.10 9.37 Operation Procedures for Station Blackout 9.5.13.20 19.7 Procedures and Training for Use of ACIWA System 19.9.7 19.12 Procedures for Operation of RCIC from Outside the Control 19.9.12 Room 19.19c Procedures to Assure SRV Operability During Station Blackout 19.9.22 1D.2.4 Mitigating Strategies for Beyond Design Basis Events (4.2)
NRC Recommendation NRC issued Order EA-12-049 (Reference 1D.5-5) to power reactor licensees and holders of construction permits requiring a three-phase approach for mitigating beyond-design-basis external events. The initial phase requires the use of installed equipment and resources to maintain or restore core cooling, containment and spent fuel pool (SFP) cooling capabilities. The transition phase requires providing sufficient, portable, onsite equipment and consumables to maintain or restore these functions until they can be accomplished with resources brought from offsite. The final phase requires obtaining sufficient offsite resources to sustain these functions indefinitely.
Response
The ABWR Design incorporates an ACIWA System. Operation of the ACIWA system is discussed in Subsection 5.4.7.1.1.10. The system can utilize either the diesel driven fire pump (described in Subsection 9.5.1.3.5) or an external pump (e.g. fire truck or trailer mounted pump) using one of two external Reactor Building connection points. Water can be directed to the Reactor Pressure Vessel (RPV), SFP or Containment Spray lines via the RHR system piping. ACIWA support equipment will be housed in a separate building. That building will be capable of withstanding the site specific external events such as seismic events, flooding, and other site specific external events (e.g. hurricanes). The capability of the building housing the ACIWA support equipment will be evaluated in the plant specific PRA (COL License Information item 19.30, Subsection 19.9.30).
Equipment procured for the ACIWA system will be procured in accordance with Diverse and Flexible coping Strategies (FLEX) guidance (Reference 1D.5-3). Equipment survivability was evaluated as part Page 5 of 12
Response to NRC Post-Fukushima Recommendations of the PRA and the results are summarized in Subsection 19E.2.1.2.3.4. The equipment that will be installed may be different than that assumed in the PRA and it will be evaluated as required by DCD Tier 1, Section 3.6 Design Reliability Assurance Program (DRAP) Design Requirements. The Design Reliability Acceptance Program (D-RAP) requires confirmation that the SSCs will function reliably when challenged.
The COL Applicant will identify additional equipment to be procured to provide defense in depth mitigation capability. Equipment to be considered includes:
x High capacity pumps which will be in addition to those described above x Portable diesel generators to provide power to lighting and instrumentation x Portable DC power supplies x Handheld satellite phones x Various hoses, fittings, cables, and jumpers necessary to connect the above equipment The COL applicant will prepare a FLEX Integrated Plan for mitigating a Beyond Design Basis External Event (BDBEE) (Subsection 13.5.3.2). This plan will be patterned after the industry FLEX program (Reference 1D.5-3). The FLEX Integrated Plan will provide site specific guidance and strategies to restore core, containment and spent fuel cooling following a BDBEE. The strategies will be capable of mitigating a simultaneous loss of all AC power (including the CTG) and loss of normal access to the ultimate heat sink, and will be capable of being implemented in all operating modes. Subsection 19E.2.2.3 and Appendix 19Q discuss the design capabilities to be used in implementing the plan. The following COL License Information items provide additional guidance:
Item No. Subject Subsection 1.5 Emergency Procedures and Emergency Procedure Training 1A.3.1 Program 1.13 Station Blackout Procedures 1C.4.1 9.9 Spent Fuel Firewater Makeup Procedures and Training 9.1.6.9 9.10 Protection of RHR System Connections to FPC System 9.1.6.10 9.19 Use of Communication System in Emergencies 9.5.13.2 9.31 Portable and Fixed Emergency Communication Systems 9.5.13.14 9.37 Operation Procedures for Station Blackout 9.5.13.20 13.4 Emergency Procedures Development 13.5.3.2 Page 6 of 12
Response to NRC Post-Fukushima Recommendations Item No. Subject Subsection 19.7 Procedures and Training for Use of ACIWA System 19.9.7 19.12 Procedures for Operation of RCIC from Outside the Control 19.9.12 Room 19.14 Accident Management 19.9.14 19.19c Procedures to Assure SRV Operability During Station Blackout 19.9.22 The equipment required to mitigate the BDBEE will be adequately protected from external events (COL License Information item 19.19b, Subsection 19.9.21). NRC Order EA-12-049 (Reference 1D.5-5) to power reactor licensees and holders of construction permits requires a three-phase approach for mitigating beyond-design-basis external events. Based on site specific evaluations of available consumables, this may be simplified to two phases if it can be shown that on site capability is sufficient to reach the point at which offsite supplies would be available. The site specific evaluation will also determine the durations of each phase based on predicted replenishment times for the consumables and this will be documented in the FLEX Plan.
The ABWR design does not take credit for the CTG that is part of the design bases for SBO. The design as described in Subsection 19E.2.2.3 credits the installed RCIC, ACIWA, and the Containment Overpressure Protection (COPS) systems to provide core, containment, and spent fuel cooling during all phases in accordance with the NRC Order.
To support the implementation of the FLEX Plan, the following system design requirements have been added to the original design:
x Added ACIWA subsystem to RHR Loop B, including connection to the Fire Protection system (FP) and external hose connection for fire truck.
x Made FP diesel-driven fire pump fuel capacity sufficient for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of operation; x Included severe weather/flooding protection for ACIWA and the diesel-driven fire pump.
x Analyzed RCIC system for operation with 121°C (250°F) pump suction temperature.
x Added Reactor Building (RB) external connections for FLEX diesel generators to 480VAC RB 1E power centers.
x Replaced control of SRVs G, J, K, and P with Control of ADS SRVs C, H, L and R on the Remote Shutdown Panel (RSP) x Added Wide Range RPV Water Level indication (Division I and II) (Cold Calibration) to the RSP.
x Added N2 Supply Header Pressure indication (Division I and II) to the RSP x Added CST Water Level indication (Division I) to the RSP x Added Containment Wide Range Pressure indication (Division I and II) to RSP x Added Wide Range Suppression Pool Water Level indication (Division I and II) to the RSP Page 7 of 12
Response to NRC Post-Fukushima Recommendations 1D.2.5 Reliable Hardened Vents (5.1)
NRC Recommendation NRC issued an Order to operating Boiling-Water Reactor (BWR) licensees with Mark I and Mark II containments requiring them to have a reliable hardened vent to remove decay heat and maintain control of containment pressure within acceptable limits following events that result in the loss of active containment heat removal capability or prolonged SBO. The hardened vent system is required to be accessible and operable under a range of plant conditions, including a prolonged SBO and inadequate containment cooling.
Response
This recommendation does not directly apply to the ABWR since the ABWR standard plant described in the DCD does not have a Mark I or Mark II containment. The design of the ABWR COPS vent which performs the above function was approved by the NRC in FSER Subsection 19.2.3.3.4. The design of the hardened vent that is part of the COPS is seismic category 1, qualified for accident pressures, passive, and reliable. This design is described in Subsection 6.2.5.2.6. Its use in conjunction with long term cooling without AC power to prevent fuel damage is demonstrated in Appendix 19E.2.2.
1D.2.6 Spent Fuel Pool (SFP) Instrumentation (7.1)
NRC Recommendation NRC issued an order to power reactor licensees and holders of construction permits requiring them to have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred.
Response
The ABWR design includes reliable level monitors in the SFP that provide indication and annunciation via the process computer in the Main Control Room (MCR). Additionally, the SFP level indication will be provided at the remote shutdown system panel or other appropriate and accessible location. The design details of these instruments are described in Subsections 7.5.2.1. (2) (q) and 9.1.3.2. The instruments are shown in Figure 9.1-1. This design is consistent with the guidance provided in NEI 12-02, Revision 1 (Reference 1D.5-11), and JLD-ISG-2012-03 (Reference 1D.5-12) except for the labeling of the water levels. To be consistent with the RPV water level labeling the lowest water level will be identified as Level 1 and the highest water level will be identified as Level 3.
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Response to NRC Post-Fukushima Recommendations 1D.2.7 Emergency Procedures Rulemaking (8.0)
NRC Recommendation Strengthen and integrate onsite emergency response capabilities such as emergency operating procedures (EOPs), plant specific technical guidelines (PSTGs), severe accident management guidelines (SAMGs), FLEX Support Guidelines (FSGs) and extensive damage mitigation guidelines (EDMGs). This includes modification of Technical Specifications to reference the approved EOP technical guidelines and providing more realistic, hands-on training on SAMGs and EDMGs.
Response
As described in Subsection 13.5.3.2 Emergency Procedures Development, the procedure development will integrate the EOPs, PTSGs, SAMGs, FSGs and EDMGs.
ABWR Section 13.5 describes the COL License Information associated with procedure development planning.
Technical Specifications meet the requirement to reference the approved EOP Guidelines. (Technical Specifications 5.5.1.1.b)
Training development requirements are outside the scope of the ABWR (see Section 13.2). However there is a COL License Information item associated with Emergency Procedures and Emergency Procedures Training Program (COL License Information item No. 1.5, Subsection 1A.3.1).
1D.2.8 Enhanced Emergency Plan Staffing and Communication (9.3)
NRC Recommendation Assess current communications systems and equipment used during an emergency event assuming the potential onsite and offsite damage as a result of a large scale natural event resulting in a loss of all AC power. It is also requested that consideration be given to any enhancements that may be appropriate for the emergency plan with respect to communications requirements of 10 CFR 50.47, Appendix E to 10 CFR Part 50, and the guidance in NUREG-0696 in light of the assumptions stated above. Also consider the means necessary to power the new and existing communications equipment with a loss of all AC power.
Assess current staffing levels and determine the appropriate staff to fill all necessary positions for responding to an event during a beyond design basis natural event and determine if any enhancements are appropriate given the considerations of NTTF Recommendation 9.3.
Response
The Emergency Plan will be prepared as part of the COL as required by COL License Information item 13.3.1.1 Emergency Plans. NEI 12-01 (Guidelines for Assessing Beyond Design Basis Accident Response Staffing and Communications Capabilities [Reference 1D.5-10]) will be used in assessing staff and communications capabilities necessary to respond to a beyond design basis multi-unit event by the COL applicant as required by COL License Information item 13.3.1.2 Staffing and Communications Capabilities.
Page 9 of 12
Response to NRC Post-Fukushima Recommendations 1D.3 Tier 2 NRC Recommendations/Responses 1D.3.1 Other Natural External Hazards (2.1)
NRC Recommendation Reevaluate and upgrade as necessary the design basis of structures, systems, and components important to safety for protection against natural external hazards other than seismic and flooding.
Response
Hazards and natural phenomena potentially affecting the ABWR site will be identified, screened and evaluated in accordance with the latest revisions of the Standard Review Plan as required in Section 2.3.2.
The following COL License Information items provide additional guidance:
Item No. Subject Subsection 2.3 Site Location and Description 2.3.2.1 2.6 Identification of Potential Accidents 2.3.2.4 2.7 Evaluation of Potential Accidents 2.3.2.5 2.8 External Impact Hazards 2.3.2.6 2.9 Local Meteorology 2.3.2.7 2.14 Floods 2.3.2.12 2.15 Probable Maximum Flood on Streams and Rivers 2.3.2.13 2.16 Ice Effects 2.3.2.14 2.19 Flooding Protection Requirements 2.3.2.17 3.5 Flood Elevation 3.4.3.1 3.7 Flood Protection Requirements for Other Structures 3.4.3.3 3.10 Missiles Generated by Other Natural Phenomena 3.5.4.2 3.12 Impact of Failure of Out of ABWR Standard Plant Scope Non- 3.5.4.4 Safety-Related Structures, Systems and Components due to a Design Basis Tornado 19.3 Event Specific Procedures for Severe External Flooding 19.9.3 Page 10 of 12
Response to NRC Post-Fukushima Recommendations 1D.3.2 Safety-related AC electrical power for the SFP makeup system (7.2)
NRC Recommendation The NRC Is to issue an order requiring safety related AC power for the SFP makeup system.
In accordance with SECY-11-0137, Recommendation 7.2 will be implemented by rulemaking to provide reliable SFP instrumentation and makeup capabilities.
Response
The ABWR design provides emergency makeup to the SFP using any of the three trains of the RHR system, which are powered by safety-related AC power (see DCD Tier 1 Subsection 2.6.2). Reliable Spent Fuel Pool indication is provided as discussed in item 1D.2.6 (above).
1D.3.3 Technical Specifications requirement for onsite emergency power (7.3)
NRC Recommendation The NRC is to issue an order to revise technical specifications to require that one train of onsite emergency power be operable for SFP makeup and instrumentation whenever spent fuel is in the SFP, regardless of the operational mode of the reactor.
In accordance with SECY 11-0137, Recommendation 7.3 will be implemented by rulemaking to provide reliable SFP instrumentation and makeup capabilities.
Response
The ABWR Technical Specifications require at least one Emergency Diesel Generator and one RHR pump to be operable in all modes of operation. The safety related RHR system is backed by the emergency diesel generators and can also be powered by the CTG. The RHR system is capable of providing makeup to the SFP.
1D.4 Analysis of Non-Design Basis Loss of AC Coping Capability The following COL License Information items provide guidance:
Item No. Subject Subsection 5.8 Analysis to Demonstrate the facility has 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Non-Design SBO 5.4.15.2.1 Capability 5.8 Analysis to Demonstrate That the DC Batteries and SRV/ADS 5.4.15.2.2 Pneumatics Have Sufficient Capacity 19.9 Actions to Mitigate Station Blackout Events 19.9.9 Page 11 of 12
Response to NRC Post-Fukushima Recommendations 1D.5 References 1D.5-1 SECY-11-0137, Prioritization of Recommended Actions to be taken in response to Fukushima Lessons Learned October 3, 2011.
1D.5-2 SECY-12-0025, Proposed Orders and Requests for Information in Response to Lessons Learned From Japans March 11, 2011, Great Tohoku Earthquake and Tsunami.
1D.5-3 NEI 12-06 [Revision 0] "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide" August, 2012.
1D.5-4 SRM for SECY 12-0025, Staff Requirements-SECY-12-0025 Proposed Orders and Requests for Information in Response to Lessons Learned from Japans March 11, 2011, Tohoku Earthquake and Tsunami March 9, 2012.
1D.5-5 EA-12-049, Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond Design-Basis External Events, March 12, 2012.
1D.5-6 EA-12-050, Issuance of Order to Modify Licenses with Regard to Reliable Hardened Containment Vents March 12, 2012.
1D.5-7 EA-12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, March 12, 2012.
1D.5-8 NUREG-2115, Central and Eastern United States Seismic Source Characterization.
1D.5-9 SECY-11-0093, The Near-Term Report and Recommendations for Agency Actions Following the Events in Japan, July 12, 2011.
1D.5-10 NEI 12-01, Guidelines for Assessing Beyond Design Basis Accident Response Staffing and Communications Capabilities.
1D.5-11 NEI 12-02 [Revision 1] Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation August 2012.
1D.5-12 JLD-ISG 2012-03SECY-11-0124, Recommended Actions to be taken Without Delay from the Near Term Task Force Report, September 9, 2011 1D.5-13 SECY-11-0124, Recommended Actions to be taken Without Delay from the Near Term Task Force Report, September 9, 2011 Page 12 of 12
25A5675AA Revision 6 ABWR Design Control Document/Tier 1 MAIN STEAM LINES OUTBOARD MAIN STEAM ISOLATION VALVE 2
1 CONTAINMENT WALL INBOARD MAIN STEAM DRYWELL ISOLATION VALVE Remove asterisk Remove asterisk SRV SRV P SRV U D SRV J SRV SRV T C SRV SRV N SRV FE SRV H FE S B SRV SRV SRV SRV M R A G MSL MSL SRV SRV 1 1 L REACTOR F NBS RPV RPV NBS VESSEL SRV MSL MSL SRV K 1 1 E NBS RPV RPV NBS FE FE Add asterisk Add asterisk DISPLAY AND/OR CONTROL INTERFACES WITH RSS ONE NOZZLE WITH 2 SETS OF TAPS Figure 2.1.2a NBS Safety/Relief Valves and Steamline Nuclear Boiler System 2.1-19
25A5675AA Revision 6 ABWR Design Control Document/Tier 1 RPV NBS (1 OF 2 1 2 NON-SAFETY-RELATED) RPV PRESSURE T (1 IN EACH RPV SHUTDOWN OF 4 DIVISIONS)
L RANGE (NON-SAFETY RELATED)
T *
(1 OF 2 P
NON-SAFETY-RELATED) ELEVATION Add asterick ABOVE MSL CENTERLINE 1 2 RPV NARROW RPV NBS RANGE L (1 IN EACH OF 4 DIVISIONS)
ELEVATION BELOW STEAM DRYER SKIRT 1 2 RPV NBS RPV WIDE L
L RANGE DIVISION (2 IN EACH OF 4 ELEVATION DIVISONS)
BELOW TOP OF ACTIVE FUEL 1 2 RPV FUEL ZONE RPV NBS L RANGE (1 IN DIVISION I AND 1 IN ELEVATION DIVISION II)
BELOW BOTTOM OF ACTIVE 1 2 FUEL DRYWELL RPV NBS PRESSURE P (1 IN EACH OF 4 DIVISIONS)
ATMOSPHERIC NBS T CONTROL SYSTEM (1 OF 2 NON-SAFETY-RELATED)
- DISPLAY AND/OR CONTROL INTERFACES WITH RSS (2 OF 4 DIVISIONS)
NOTES:
- 1. EACH OF FOUR DIVISIONS OF NBS INSTRUMENTATION HAS ONE TURBINE INLET PRESSURE SENSOR AND ONE MAIN CONDENSER VACUUM SENSOR.
- 2. EACH DIVISION IS POWERED BY ITS RESPECTIVE CLASS 1E DIVISION.
- 3. THE MAIN CONDENSER VACUUM SENSORS ARE LOCATED ABOVE THE TURBINE BUILDING DESIGN BASIS FLOOD LEVEL.
Figure 2.1.2e NBS Drywell Pressure and Reactor Vessel Instrumentation Nuclear Boiler System 2.1-23
25A5675AA Revision 6 ABWR Design Control Document/Tier 1 2.2.6 Remote Shutdown System Design Description The Remote Shutdown System (RSS) provides remote manual control of safety-related systems to bring the reactor to hot shutdown and subsequent cold shutdown conditions from outside the main control room (MCR). Figure 2.2.6 shows the basic system configuration and scope.
The RSS has two divisional panels and associated controls and indicators for interfacing with the following systems:
(1) Residual Heat Removal (RHR) System (2) High Pressure Core Flooder (HPCF) System (3) Nuclear Boiler System (NBS)
(4) Reactor Service Water (RSW) System (5) Reactor Building Cooling Water (RCW) System (6) Electrical Power Distribution (EPD) System (7) Atmospheric Control (AC) System (8) Emergency Diesel Generator (DG)
(9) Make-up Water System (Condensate), (MUWC)
(10) Flammability Control System (FCS)
(11) Suppression Pool Temperature Monitoring (SPTM) System RSS controls and indicators are hard-wired direct to the interfacing components and sensors.
The RSS is classified as a Class 1E safety-related system.
Operation of transfer switches on the RSS panel overrides and isolates the controls from the MCR and transfers control to the RSS. Transfer switch actuation causes alarms in the MCR.
Indications required for plant shutdown are provided on the RSS panels as shown on Figure 2.2.6.
(12) High Pressure Nitrogen Gas Supply System 2.2-30 Remote Shutdown System
HPIN A, B Add Drywell Pressure Pressure Add A 2.2-32 - RHR FLOW
- HX. INLET TEMPERATURE
- HX. OUTLET TEMPERATURE - RX. WATER LEVEL MEDIUM VOLTAGE ABWR
- PUMP DISCHARGE PRESSURE - RX. PRESSURE - CST LEVEL BUS VOLTAGE
- HPCF FLOW
- PUMP DISCHARGE - S/P PRESSURE TEMPERATURE - S/P LEVEL - RCW FLOW RUN/STOP RHR-A, B HPCF-B MUWC-B EPD DIVISION I & II NBS-A, B SPTM-A, B AC-A, B RCW-A, B DG-A, B INDICATION RSS PANELS
- CONTROL & TRANSFER SWITCHES
- VALVE POSITION INDICATION
- PUMP STOP-RUN INDICATION
- PLANT PARAMETERS INDICATORS CONTROL 25A5675AA Revision 6 RHR-A, B EPD NBS-A, B HPCF-B RCW-A, B RSW-A, B FCS-B DIVISION I & II NOTES:
- 1. RSS PANELS A AND B INTERFACE WITH SYSTEM IN DIVISIONS A AND B ( I AND II), RESPECTIVELY.
Figure 2.2.6 Remote Shutdown System Design Control Document/Tier 1 Remote Shutdown System
2.11-4 ABWR CRD MUWC MUWC MUWP NNS NNS MUWC RW/B NNS T/B R
NOTE 1 & 2 L CST 2 NNS R/B 2
CRD MUWC RCIC NNS MUWC 25A5675AA Revision 6 HPCF HPCF HPCF NOTES:
- 1. ONE SENSOR ASSIGNED TO EACH OF FOUR CLASS 1E DIVISIONS.
- 2. RSS INTERFACE IS WITH DIVISION II LEVEL SENSOR ONLY.
Figure 2.11.2 Makeup Water (Condensate) System Makeup Water (Condensate) System Design Control Document/Tier 1 for Division I and Division II Level Sensor
25A5675AA Revision 6 ABWR Design Control Document/Tier 1 The HPIN System has the following displays and controls in the main control room:
(1) Parameter displays for the sensors shown on Figure 2.11.13.
(2) Control and status indication for the active safety-related components shown on Figure 2.11.13.
The safety-related electrical equipment shown on Figure 2.11.13 located in the Reactor Building is qualified for a harsh environment.
The motor-operated valves (MOVs) shown on Figure 2.11.13 have active safety-related functions to open, close, or both open and close, and perform these functions under differential pressure, fluid flow, and temperature conditions.
The check valves (CVs) shown on Figure 2.11.13, have active safety-related functions to both open and close under system pressure, fluid flow, and temperature conditions.
Inspections, Tests, Analyses and Acceptance Criteria Table 2.11.13 provides a definition of the inspections, tests, and/or analyses, together with associated acceptance criteria, which will be undertaken for the HPIN System.
The HPIN System has Pressure indication for Division I and Division II on the Remote Shutdown Panel.
2.11-50 High Pressure Nitrogen Gas Supply System
25A5675AJ Revision 6 ABWR Design Control Document/Tier 2 (v) Reactor Water level wide range indication -
cold calibration (A,B)
(b) The following nuclear boiler instrumentation is provided on the remote shutdown control panels as indicated: narrow (i) Reactor water level wide range indication (A,B)
(ii) Reactor water level shutdown range indication (A,B)
(iii) Reactor pressure indication (A,B)
(iv) Indicate lights for four SRV valve open/close condition (three on Panel A, and one on Panel B (c) The following function has transfer and control switches located at the Division 2 remote shutdown control panel: one air-operated relief valve. (The valve is 125 volt DC solenoid pilot operated.)
(5) Reactor Building Cooling Water (RCW) System (a) The following functions have transfer and control switches located on the remote shutdown panels as indicated:
(i) RCW pumps (A,D and B,E)
(ii) RCW heat exchanger cooling water outlet valves (A,D,G and B,E,H)
(iii) RCW, RHR heat exchanger, outlet valve (A,B)
(iv) RCW, diesel generator, outlet valve (A,D and B,E)
(v) RCW separator valve between essential and non-essential loads (A,B)
(vi) RCW temperature control valves (A,B)
(b) The following RCW instrumentation is provided on the RSS control panels as indicated:
(i) RCW loop flow indication (A,B)
(ii) Indicating lights for valve positions and for pump stop/run (A,B) 7.4-14 Systems Required for Safe Shutdown
25A5675AJ Revision 6 ABWR Design Control Document/Tier 2 (b) The following functions have transfer and control switches located on the Division II remote shutdown panel:
(i) 6.9 kV feeder breaker: Unit auxiliary transformer B to M/C F (b) Containment (ii) 6.9 kV feeder breaker: Reserve auxiliary transformer A to M/C F Wide Range (iii) 6.9 kV feeder breaker: Emergency diesel generator B to M/C F Pressure on both (iv) 6.9 kV feeder breaker: Combustion turbine generator to M/C F RS panels.
(v) 6.9 kV load breaker: M/C F to P/C F20 (vi) 480V feeder breaker: TR to P/C F20 (c) A 6.9 kV M/C (E,F) voltmeter is provided on RSS panels A,B, respectively.
(8) Flammability Control System (FCS)
(a) The following FCS equipment function has transfer and control switches Narrow Range and located on both remote shutdown panels as indicated:
Wide Range (i) Valve (cooling water inlet) B both RS panels.
(9) Atmospheric Control (AC) System (13) High Pressure (a) Suppression pool level indication is provided on both RS panels.
Instrument Nitrogen (10) Makeup Water Condensate System (MUWC)
(HPIN) System (a) Nitrogen Supply (a) Condensate storage pool level indication is provided on RS panel B.
Header Pressure on both RS Panels. (11) Suppression Pool Temperature Monitoring System (SPTM)
(a) Suppression pool temperature indication is provided on both RS panels.
(12) Emergency Diesel Generator (DG) System (a) A transfer switch on each RS panel (A,B) permits DG control (start/stop) from the control room to be interrupted. There are no DG controls on the RS panels.
During remote shutdown operation, the DGs can be controlled locally.
(b) Status lights provide DG status indication (run/stop) on each RS panel (A,B).
7.4.2 Analysis 7.4.2.1 Alternate Rod Insertion Function 7.4.2.1.1 General Functional Requirements Conformance The alternate rod insertion (ARI) function is accomplished by the Rod Control and Information System (RCIS) and the Fine-Motion Control Rod Drive (FMCRD) Subsystem. This function 7.4-16 Systems Required for Safe Shutdown
25A5675AU Revision 6 Remote Shutdown System 3.3.6.2 Table 3.3.6.2-1 (page 1 of 2)
Remote Shutdown System Instrumentation REQUIRED NUMBER FUNCTION (INSTRUMENT OR CONTROL PARAMETER) OF DIVISIONS
- 1. Reactor Steam Dome Pressure 2
- 2. HPCF B Flow 1
- 3. HPCF B Controls 1(c)
- 4. HPCF B Pump Discharge Pressure 1
- 5. RHR Flow 2(a)
- 6. RHR Hx Inlet Temperature 2(a)
- 7. RHR Hx Outlet Temperature 2(a)
- 8. RHR Hx Bypass Valve Position 2(a)
- 9. RHR Hx Outlet Valve Position 2(a)
- 10. RHR Pump Discharge Pressure 2(a)
- 11. RHR Controls 2(a)(c)
- 12. RPV Wide Range Water Level 2
- 13. RPV Narrow Range Water Level 2
- 14. Reactor Building Cooling Water Flow 2
- 15. Reactor Building Cooling Water Controls 2(c)
- 16. Reactor Building Service Water System Controls 2(c)
- 17. Cooling Water Flow to Flammability Control System 1
- 18. Suppression Pool Water Level 2
- 19. Condensate Storage Tank Water Level 1 2
ABWR TS 3.3-70 Design Control Document/Tier 2
25A5675AU Revision 6 Remote Shutdown System 3.3.6.2 Table 3.3.6.2-1 (page 2 of 2)
Remote Shutdown System Instrumentation REQUIRED NUMBER FUNCTION (INSTRUMENT OR CONTROL PARAMETER) OF DIVISIONS
- 20. Suppression Pool Temperature 2
- 21. Electric Power Distribution Controls 2(c)
- 22. Diesel Generator Interlock and Monitors 2
- 23. SRV Controls (b)(c)
(a) RHR A for division I RSS panel, RHR B for division II RSS panel.
(b) Three on the Division I RSS, 1 on division II RSS.
(c) The specified number of channels are required to be OPERABLE for each device that can be controlled from the RSS panels.
- 24. N2 Header Pressure 2
- 25. Drywell Pressure - Wide Range 2
- 26. Suppression Pool Water Level 2 Wide Range
25A5675AU Revision 6 Remote Shutdown System B 3.3.6.2 BASES LCO 2, 3, and 4. HPCF B Flow/Controls/Discharge Pressure (continued)
The HPCF system can be used to provide vessel inventory make up and decay heat removal while bringing the plant to MODE 3. The HPCF, in conjunction with other instruments and controls on the division II RSS panel, is sufficient to achieve and maintain MODE 3 from the Division II panel. The HPCF flow and pressure indications provide monitoring of HPCF operation. The controls provided are as given in reference 2. One channel of each Function is required to be OPERABLE to provide redundant capability to achieve MODE 3 from both RSS panels.
5 through 11. RHR A, B Control & Indication The RHR system can be used to provide vessel inventory make up, decay heat removal, and suppression pool cooling while bringing the plant to MODE 3. The RHR, in conjunction with other instruments and controls on the RSS panels, is sufficient to achieve and maintain MODE 3 from either panel. The RHR flow indications provide monitoring of RHR operation and the heat exchanger monitors provide indication of decay heat removal. The RHR controls and monitors are adequate to place it in the shutdown cooling mode. The controls provided are as given in reference 3. Two channels of each RHR Function (RHR A on the division I panel and RHR B on the division II panel) are required to be OPERABLE to provide redundant capability to achieve MODE 3 from both RSS panels.
- 13. and 27.
of the RPV Water level conditions 12, and 13. RPV Wide Range/Narrow Range Water Level and ranges Reactor vessel water level is provided to support monitoring of core cooling, to verify operation of the make up pumps, and is needed for satisfactory operator control of the make up pumps. The wide range water level channels cover the range from the near top of the fuel to near the top of the steam separators. The narrow range provides indication from near the bottom of the separators to above the steam lines. RPV level is a necessary parameter for achieving and maintaining the reactor in MODE 3. One channel of each range is provided on each of the RSS panels. Both channels are required to be OPERABLE to provide redundant capability to achieve MODE 3 from both RSS panels.
There is an additional set of wide range instruments that have been calibrated for cold conditions and will be used when the normal instruments are off scale.
(continued)
ABWR TS B 3.3-195 Design Control Document/Tier 2
25A5675AU Revision 6 Remote Shutdown System B 3.3.6.2 BASES LCO 14, 15, and 16. Reactor Building Cooling Water Flow/Controls & Reactor (continued) Building Service Water Controls These parameters and controls are required to monitor and control the water supply for cooling the equipment needed to achieve MODE 3 and to provide containment heat removal. The Reactor Building Cooling Water controls provided are as given in reference 4 and the Reactor Building Service Water controls provided are as given in reference 5.
One channel of each Function is provided on each of the RSS panels.
Both channels of each Function are required to be OPERABLE to provide redundant capability to achieve MODE 3 from both RSS panels.
- 17. Cooling Water Flow to Flammability Control System A control for the FCS B inlet valve is provided on the division II panel only. This control is needed in order for the operator to isolate cooling and 26. water flow to FCS. One channel is required to be OPERABLE to assure that MODE 3 can be achieved from the Division II RSS panel.
Narrow and Wide Range
- 18. Suppression Pool Water Level narrow range Suppression pool water level provides information needed to assess the status of the RCPB and to assess the status of the water supply to the ECCS. The level indicators monitor the suppression pool level from the bottom of the ECCS suction lines to five feet above the normal suppression pool level. One channel of this Function is provided on each of the RSS panels. Both channels are required to be OPERABLE to provide redundant capability to achieve MODE 3 from both RSS panels.
The wide range level indicators both Functions are monitor the 19. Condensate Storage Tank Level suppression pool from the centerline Condensate Storage Tank Level provides information needed to assess the status of the water supply to the HPCF. The indication is needed in of the ECCS order to achieve and maintain MODE 3 when using HPCF. A channel of suction piping to this Function is provided on the division II RSS panel. The channel is the wetwell required to be OPERABLE to achieve MODE 3 from the Division II RSS spargers. panel.
Both channels are RCIC and RCIC and s both (continued)
ABWR TS B 3.3-196 Design Control Document/Tier 2
25A5675AU Revision 6 Remote Shutdown System B 3.3.6.2 BASES LCO 20. Suppression Pool Temperature (continued)
Suppression Pool Temperature allows the operator to detect trends in suppression pool water temperature in sufficient time to take action to prevent steam quenching vibrations in the suppression pool. This Function is required in order to maintain MODE 3. One channel of this Function is provided on each of the RSS panels. Both channels are
- 24. N2 Header Pressure required to be OPERABLE to provide redundant capability to maintain This Function is provided to permit MODE monitoring 3 from both RSS panels.
the status of the N2 Bottle Header Pressure.
These monitors are required to permit the
- 21. Electric Power Distribution Controls operator to manage the N2 supply to the SRVs. One channel of this Function TheseisFunctions are provided so the operator can select various AC provided on each RSS panel. Both power channels sources for the equipment needed to achieve and maintain of the Function are required to be MODE OPERABLE
- 3. The Electric Power Distribution Controls provided are as given to provide redundant capacity toinachieve references 6 and 7. One channel of each Function is provided on each MODE 3 from both RSS panels.of the RSS panels. Both channels of each Function are required to be 25.Drywell Pressure - Wide Range OPERABLE to provide redundant capability to achieve MODE 3 from both This function is provided to permit RSS panels.
monitoring the status of the drywell pressure. This will allow the operator to determine 22.
if there is aGenerator System Interlock and Monitors Diesel potential of operation of Containment Overpressure Protection System This(COPS).
Function is provided to permit monitoring the status of the One channel of this Function is emergency provided on DG. These monitors are required to permit the operator to manage each RSS panel. Both channels of the the electric power distribution. The interlock disables DG Function are required to be OPERABLE to the control room to assure that the event that made the start/stop from control room unavailable will not disrupt DG operation. One channel of provide redundant capacity boththis RSS panels.
Function is provided on each of the RSS panels. Both channels of the Function are required to be OPERABLE to provide redundant capability to achieve MODE 3 from both RSS panels.
- 23. SRV Controls This Function is provided to permit the operator to perform a controlled depressurization and to maintain reactor pressure within limits. Three channels are provided on the division I RSS panel and one channel is provided on the division II panel. These channels, in conjunction with other controls and indications on the panels, are sufficient to achieve and maintain MODE 3 from either panel. Three channels on the division I panel and one channel on the division II panel are required to be OPERABLE to provide redundant capability to achieve MODE 3 from both RSS panels.
(continued)
ABWR TS B 3.3-197 Design Control Document/Tier 2
ABWR RSS 13 C,H,R, VALVES RSS 25A5675BB Revision 6 13 VALVE L Figure 5.1-3 Nuclear Boiler System P&ID (Sheet 2 of 11) 21-81 Design Control Document/Tier 2
21-84 ABWR 13 RSS 13 RSS LT 25A5675BB Revision 6 004 A Figure 5.1-3 Nuclear Boiler System P&ID (Sheet 5 of 11)
Design Control Document/Tier 2
ABWR 13 RSS 25A5675BB Revision 6 13 RSS LT 004 B Figure 5.1-3 Nuclear Boiler System P&ID (Sheet 6 of 11) 21-85 Design Control Document/Tier 2
25A5675BC Revision 6 RSS 8
RSS Figure 6.2-39 Atmospheric Control System P&ID (Sheet 2 of 3)
Design Control Document/Tier 2
ABWR 8
RSS 8
RSS 25A5675BC Revision 6 Figure 6.2-39 Atmospheric Control System P&ID (Sheet 3 of 3)21-139 Design Control Document/Tier 2
ABWR 6
- 6. Remote Shutdown C61-1010 System IED 25A5675BC Revision 6 6
RSS Figure 6.7-1 High Pressure Nitrogen Gas Supply System P&ID 21-149 Design Control Document/Tier 2
ABWR 25A5675BD Revision 6
- 17. Manual operation of the Division I ADS valves (R, C and H) and (Div. II) valve (L) from the Remote Shutdown Panel energizes the valve ADS solenoid of the associated division.
Figure 7.3-2 Nuclear Boiler System IBD (Sheet 2 of 37)21-251 Design Control Document/Tier 2
25A5675BD Revision 6 17 NOTE - move to page 9 of Nuclear Boiler IBD F010C C6 18 Figure 7.3-2 Nuclear Boiler System IBD (Sheet 3 of 37)
Design Control Document/Tier 2
ABWR J
17 25A5675BD Revision 6 NOTE - move to page 10 of Nuclear F010L L6 Boiler IBD 18 Figure 7.3-2 Nuclear Boiler System IBD (Sheet 4 of 37)21-253 Design Control Document/Tier 2
ABWR G
25A5675BD Revision 6 17 NOTE - move to page 9 of Nuclear Boiler IBD F010H H6 18 Figure 7.3-2 Nuclear Boiler System IBD (Sheet 6 of 37)21-255 Design Control Document/Tier 2
25A5675BD Revision 6 NOTE - move to 17 page 10 of Nuclear Boiler IBD F010R R6 18 Figure 7.3-2 Nuclear Boiler System IBD (Sheet 7 of 37)
Design Control Document/Tier 2
21-258 ABWR C H 25A5675BD Revision 6 To RSS Ref 2 two Inserts from previous pages are placed here Figure 7.3-2 Nuclear Boiler System IBD (Sheet 9 of 37)
Design Control Document/Tier 2
ABWR L R 25A5675BD Revision 6 To RSS two Inserts from Ref 2 previous pages are placed here Figure 7.3-2 Nuclear Boiler System IBD (Sheet 10 of 37)21-259 Design Control Document/Tier 2
ABWR H6 9 L6 10 RSS SEE NOTE 17 RSS SEE NOTE 17 25A5675BD Revision 6 C6 R6 9 10 RSS RSS SEE NOTE 17 SEE NOTE 17 Figure 7.3-2 Nuclear Boiler System IBD (Sheet 18 of 37)21-267 Design Control Document/Tier 2
Wide Range SP Level CST Level T31-LT-059A RPV Wide Range Cold Wide Range Drywell Pressure P13-LT-001A B21-LT-004A N2 Header T31-PT-055A P54-PT-002A ABWR C
H R
Wide Range Drywell Pressure Wide Range SP Level N2 Header T31-PT-055B T31-LT-059B P54-PT-002B RPV Wide Range Cold B21-LT-004B 25A5675BF Revision 6 L
Figure 7.4-2 Remote Shutdown System IED 21-451 Design Control Document/Tier 2
ABWR F010R F010C F010H F010L 25A5675BF Revision 6 Figure 7.4-3 Remote Shutdown System IBD (Sheet 2 of 27)21-453 Design Control Document/Tier 2
25A5675BF Revision 6 ABWR Design Control Document/Tier 2 Figure 7.4-3 Remote Shutdown System IBD (Sheet 9 of 27)
C R L
H 21-460
21-724 ABWR added connection to RSS for B channel 25A5675BH Revision 6 Figure 9.2-4 Makeup Water System (Condensate) P&ID Design Control Document/Tier 2