ML20073E721
| ML20073E721 | |
| Person / Time | |
|---|---|
| Site: | 05200004 |
| Issue date: | 09/21/1994 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20073E715 | List: |
| References | |
| 25A5113, 25A5113-A01-RA, 25A5113-A1-RA, NUDOCS 9409290155 | |
| Download: ML20073E721 (850) | |
Text
{{#Wiki_filter:- - - _ _ _ _ l 2SAS113 Rev. A SBWR sonderdsuretyAurysis aeron eO V _ SBWR SSAR Amendment 1 Section Change instructions Replace old sections in your SSAR with the new sections listed below. Sections will have underline to indicate text additions and striheircugh to indicate text deletions. Section Nmnber Section Title Chapter 1 Table of Contents 1.2 Introduction 1.5 Requirements for Further Technical Information 1.6 MaterialIncorporated by Reference 1.9 Conformance with Standard Review Plan and Applicability of Codes and Standards IA Response to TMI Related Matters O ( Chapter 2 Table of Contents 2.0 Site Characteristics Chapter 3 Table of Contents Chapter 3 List of Tables Chapter 3 List of Figures 3.1 Conformance with NRC General Design Criteria 3.2 Classification of Structures, Systems, and Components 3.3 Wind and Tornado Loadings 3.6 Protection Agamst Dynamic Effects Associated with the Postulated Rupture of Piping 3.7 Seismic Design 3.9 Mechanical Systems and Components O w/ 3.11 Environmental Quahfication Of Safety-Related Mechanical And Electrical Equipment Transmittal- Amendment 1 1 9409290155 940923 847M PDR AltOCK 05200004 a PDR
l 2SA5113 R2v. A SBWR sonwsannyAnnoysisneron O Section Number Section Title SA Seismic Soil-Structure Intenction Analysis 3D Equipment QualiScation Design Emironmental Conditions Chapter 4 Table of Contents Chapter 4 List of Tables Chapter 4 List of Figures 4.1 Summary Description 4.2 Fuel System Design Chapter 5 Table of Contents Chapter 5 List ofTables Chapter 5 List of Figures 5.2 Integrity of Reactor Coolant Pressure Boundary 5.3 Reactor Vessel 5.4 Component and Subsystem Design Chapter 6 Table of Contents Chapter 6 List ofTables Chapter 6 List of Figures 6.1 Engineered Safety Feature Materials 6.2 Containment Systems 6.4 Control Room Habitability Systems 6.7 Main Steam Isolation Valve Leakage Control System (BWR) Chapter 7 Table of Contents Chapter 7 List ofTables l I 2 Transmittal- Amendment 1 9721M4
l l 2SAS113 Rev. A SBWR standantsaretyAnaI Ysis Report
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(vl Section Number Section Title Chapter 7 List of Figures 7.1 Introduction 7.2 Reactor Trip System 7.3 Engineered Safety Features Systems 7.4 Shutdown Systems Chapter 9 Table of Contents Chapter 9 List ofTables Chapter 9 List of Figures 9.1 Auxiliary Systems CN (v ) 9.2 Water Systems 9.3 Process Auxiliaries 9.4 Air Conditioning, Heating, Cooling, and Ventilation Chapter 10 Table of Contents Chapter 10 List of Tables Chapter 10 List of Figures 10.2 Turbine Generator Chapter 12 Table of Contents Chapter 12 List ofTables Chapter 12 List of Figures 12.1 Ensuring that Occupational Radiation Exposures are ALARA 12.2 Plant Sources [ (/ 12.3 Radiation Protection Transmittal- Amendment 1 3 9/21/94
25A5113 Rev. A SBWR Standant Safety Analysis Report
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V Section Number Section Title 12.4 Dose Assessment 12A Calculation ofAirborne Radionuclides Chapter 13 Table of Contents Chapter 13 List of Tables 13.0 Conduct of Operations Chapter 14 Table of Contents 14.2 Initial Plant Test Program - Final Safety Analysis Report Chapter 15 Table of Contents Chapter 15 List ofTables Chapter 15 List of Figures 15.1 Decrease In Reactor Coolant Temperature 15.6 Decrease In Reactor Coolant Inventory w 15.7 Radioactive Release s Chapter 16 Table of Contents Chapter 16 List ofTables Chapter 16 List of Figures 16.3.6 Containment Systems 16B.3.6 Containment Systems Chapter 17 Table of Contents Chapter 17 List of Tables Chapter 17 List of Figures / \ (,,,) 17.0 Introduction 4 Transmittal- Amendment 1 9/21,94
25A5113 Rsv. A SBWR aandadsanyAnar sis r neport O Section Number Section Title 17.1 Quality Assurance Dudng Design and Construction 17.2 Quality Assurance During the Operations Phase 17.3 Reliability Assurance Program During Design Phase Chapter 18 Table of Contents Chapter 18 List ofTables Chapter 18 List of Figures 18.1 Introduction 18.2 Design Goals and Design Bases 18.3 Planning, Development, and Design 18.4 Control Room Standard Design Features 18.5 Remote Shutdown System 18.6 Systems Integration 18.7 - Detailed Design of the Operator Interface System 18.8 COL License Information 18A Emergency Procedure Guidelines 18E SBWR Human-System Interface Design Implementation Process 18F Emergency Operation Information and Controls Chapter 19 Table of Contents Chapter 19 List ofTables Chapter 19 List of Figures 19AE.6 Standby Liquid Control System (SLCS) Transmittal- Amendment 1 5 9/21AM
2SAS113Rev. A SBWR stamtantsarety Anarrsis neport O 1 A Response to TMI Related Matters 1A.1 Introduction The investigations and studies associated with the TMI accident produced several l documents specifying results and recommendations, which prompted the issuance by the NRC of various bulletins, letters, and NUREGs providing guidance and requiring specific actions by the nuclear power indusuy. In May 1980, the issuance of NUREG-0660 (Reference IA-1) provided a comprehensive and integrated plan and listing requirements to correct or improve the regulation and operation of nuclear facilities based on the experience from the accident at TMI and the studies and investigations of the accident. NUREG-0737 (Reference 1A-2), issued in November 1980, listed items from NUREG-0660 approved by the NRC for implementation, and included additional information concerning schedules, applicability, method of implementation review, submittal dates, and clarification of technical positions. Finally, NUREG-0718 (Reference I A-3) was issued inJune 1981 to provide guidance that the NRC staff believes should be followed to account for the lessons learned from the TMI accident. n This Appendix 1A provides GE's responses for the SBWR Standard Plant required by ( Section II of the NRC Standard Review Plan. 1 A.2 NRC Positions / Responses i 1 A2.1 Short Term Accident Analysis Procedure Revision [l.C.1(3)] NRC Position In letters of September 13 and 27, October 10 and 30, and November 9,1979, the Office of Nuclear Reactor Regulation required licensees of operating plants, applicants for operating licenses and licensees of plants under construction to perform analyses of transients and accidents, prepare emergency procedure guidelines, upgrade i emergency procedures, including procedures for operating with natural circulation l conditions, and to conduct operator retraining (see also NUREG-0737 Item I.A.2.1). Emergency procedures are required to be consistent with the actions necessary to cope with the transients and accidents analyzed. Analyses of transients and accidents were to be completed in early 1980 and implementation of procedures and retraining were to be completed 3 months after emergency procedure guidelines were establishad; however, some difficulty in completing these requirements has been experienced. Clarification of the scope of the task and appropriate schedule revisions are being developed. In the course of review of these matters on Babcock and Wilcox (B&W) - designed plants, the staff will follow up on the bulletin and order matters relating to analysis methods and results, as listed in NUREG-0660, Appendix C (Table C.1, Items 3, ( 4,16,18,24,25,26,27; Table C.2, Items 4,12,17,18,19,20; and Table C.3, Items 6,35, 37,38,39,41,47,55,57). Response to TMI Related Matters - Amendment 1 1A-1
25A5113Rsv. A SBWR standardsarery Anstrsis Report O
Response
In the clarification of the NUREG-0737 requirement for reanalysis of transients and accidents, inadequate core cooling and to prepare emergency procedures guidelines, it was stated that:
" Owners' group or vendor submittals may be referenced as appropriate to support this reanalysis. If owners' group or vendor submittals have already been forwarded to the staff for review, a brief description of the submittals andjustification of their adequacy to support guideline development is all that is required.
GE has participated, and continues to participate, in the BWR Owners' Group program to develop emergency procedure guidelines (EPGs) for General Electric BWRs. The resulting EPGs are generally applicable to the SBWR as are the transient and accident analyses. Appendix 18A contains the EPGs for the SBWR Standard Plant, and Chapter 15 contains the accident analyses. Followingis a brief description of the submittals to date, and ajustification of their adequacy to support guideline development. m Description of Submittals
- NEDO-24708, Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors, August,1979.
- NEDO-24708A, Revision 1, Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors, December,1980. This report was issued via the letter from D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC) dated March 20,1981.
- BWR Emergency Procedure Guidelines (Revision 0), submitted in prepublication formJune 30,1980.
- BWR Emergency Procedure Guidelines (Revision 1), Issued via the letter from D. B. Waters (BWR Owners' Group) to D. G. Eisenhut (NRC) datedJanuary 31, 1981. 1
. BWR Emergency Procedure Guidelines (Revision 2), submitted in prepublication formjune 1.1982, Letter BWROG-8219 from T.J. Dente (BWR Owners' Group) to D. G. Eisenhut (NRC). 1
- BWR Emergency Procedure Guidelines (Revision 3), submitted in prepublication form December 22,1982, Letter BWROG-8262 from T.J. Dente (BWR Owners' Group) to D. G. Eisenhut (NRC).
1A-2 Response to TMI Related Matters - Amendment 1
25A5113Rev. A SBWR standantsafetyAnalysis Report j O
- NEDO-31331, BWR Emergency Procedure Guidelines (Revision 4), submitted April 23,1987, Letter BWROG-8717, from T. A. Pickens (BWR Owners' Group) to T. Murley (NRC). ]
a Adequacy of Submittals The submittals described above have been discussed and reviewed extensively among the B%R Owners' Group, the General Electric Company, and the NRC Staff. The NRC has extensively reviewed the latest revision (Revision 4) of the Emergency Procedures Guidelines and issued a SER, Safety Evaluation of BWR Owners' Group Emergency Procedure Guidelines, Resision 4, NEDO-31331, March 1987, letter from A. C. Thadani, NRC Office of Nuclear Reactor Regulation, to D. Grace, Chairman of BWR
- Owners' Group, dated September 12,1988. The SER concludes that this document is acceptable for implementation. It further states that the SER closes all the open items carried from the previous revisions of the EPG.
The SBWR EPG was derived from Rev. 4 of the generic EPG prepared by the BWROG as documented in NEDO-31331. Adaptations were required to accommodate the unique design philosophy, plant configuration and specific systems and components of the SBWR. Appendix 18B provides the bases for the modified EPG as follows: a Section 18B.1 briefly describes the SBWR features which greatlyimpacted the SBWR EPG; a Section 18B.2 discusses the major EPG changes arising from these features; and a Section 18B.3 tabulates, step-by-step, the differences between EPG Rev. 4 and the SBWR EPG and provides the bases for these differences. The major strategy differences between the SBWR EPG and the BWROG EPG Rev. 4 arise from the fundamental design difference in the emergency core cooling (ECC) systems, since the SBWR incorporates many automatically initiated ECC systems. Tables 18B-1 through 18B-8 delineate all changes made to BWROG EPG Rev. 4 for adaptation to the SBWR EPG and summarize the bases for these changes. GE believes that in view of these findings, no further detailedjustification of the analyses or guidelines is necessary at this time. COL license information requirements pertaining to emergency procedures are discussed in Subsection 1A.S.I. 1 A.2.2 Control Room Design Reviews - Guidelines and Requirements [1.D.1(1)] NRC Position s In accordance with Task Action Plan I.D.1.(1), all licensees and applicants for operating licenses will be required to conduct a detailed control room design review to identify Response to TMI Related Matters - Amendment 1 1A-3
I l l 2SA5113Rev. A SBWR standardsafety Anatrsis Repoa O and correct design deficiencies. This detailed control room design review (DCRDR) is expected to take about a year. Therefore, the OfIlce of Nuclear Reactor Regulation (NRR) requires that those applicants for operating licenses who are unable to complete this review prior to issuance of a license make preliminary asses:.ments of their control rooms to identify significant human factors and instrumentation problems and establish a schedule approved by NRC for correcting deficiencies. These applicants will be required to complete the more detailed control room reviews on the same schedule as licensees with operating plants.
Response
The design of the SBWR control room utilizes accepted human factors engineering principles, incorporating the results of a full systems analysis similar to that described in Appendix B of NUREG 0700. An integrated program plan, entided " Design of Controls, Instrumentation and Man-Machine Interfaces," was prepared and implemented to incorporate human factors engineering principles and to achieve an integrated design of the control and instrumentation systems and operator interfaces of the SBWR. This plan and the associated procedures provided guidance for the conduct of the SBWR control and instrumentation and Man-Machine Interface Systems (MMIS) design development activities including definition of the standard design features of the control room MMIS described in Subsection 18.4.2. Chapter 18 describes the SBWR MMIS design goals and bases, the standard MMIS design features and the detailed MMIS design and implementation process, with embedded design acceptance criteria, for the SBWR standard plant operator interface. A DCRDR specified in NUREG-0737 is not required by SRP Section 18.1. 1 A.2.3 Control Room Design - Plant Safety Parameter Display Console [1.D.2] NRC Position In accordance with Task Action Plan I.D.2, each applicant and licensee shall install a safety parameter display system (SPDS) that will display to operating personnel a minimum set of parameters which def'me the safety status of the plant. This can be attained through continuous indication of direct and derived variables as necessary to assess plant safety status.
Response
The principal functions of the SPDS as required by Supplement I to NUREG-0737 will be integrated into the control room operator interface design, as permitted by SRP Section 18.2. The SBWR control room operator interface design incorporates the SPDS function as part of the plant status summary information which is continuously displayed on the fixed-position displays on a large display panel, and also incorporates the use of on-1A4 Response to TMI Related Matters - Amendment 1
2545113Rev. A SBWR standantsafetyAnalysis Report (x screen control video display units ('/DUs), independent of the process computer, for control and monitoring of both safety-related and non-safety-related systems. Other VOUs, driven by the process computer, are available for monitoring of safety-related systems and monitoring and control of non-safety-related systems. Descriptions of the SBWR control room standard design features can be found in Subsection 18.4.2. 1 A.2.4 Scope of Test Program - Preoperational and Low Power Testing II.G.1] NRC Position Supplement operator training by completing the special low-power test program. Tests may be obseived by other shifts or repeated on other shifts to provide training to the operators.
Response
Extensive operator training will be conducted in a control room simulation facility. Simulation of all plant transients will be performed there. The test program described : in Section 14.2 supplements the simulator training consistent with the NRC position, . I and provides additional opportunity to the permanent plant operating staff to obtain practical experience in the operation and maintenance of equipment and systems. The O degree to which the potential additional benefits of the test program are realized will depend on such plant-specific factors as the organizational makeup of the startup group and overall plant staff (see Subsection 14.2.1.4), as well as how the test program is
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conducted (see Subsection 14.2.2). The SBWR test program described in Chapter 14 is derived from the BWROG program, which was developed in response to Item I.G.1 of NUREG-0737, and documented in a letter (BWROG-8120) from D. B. Waters to D. G. Eisenhut, dated February 4,1981. Adaptations of the BWROG program were required to accommodate the unique plant configuration and specific systems and components of the SBWR. The specific training requirements for reactor operators are discussed in Section 13.2 of the SRP which is outside the scope of the SBWR Standard Plant. See Section 1.8 for COL license information requirements. 1 A.2.5 Reactor Coolant System Vents Ill.B.1] NRC Position , Each applicant and licensee shall install reactor coolant system (RCS) and reactorvessel head high pointvents remotely operated from the control room. Although the purpose of the system is to vent noncondensable gases from the RCS which may inhibit core cooling during natural circulation, the vents must not lead to an unacceptable increase in the probability of a loss-of coolant accident (LOCA) or a challenge to containment { ( integrity. Since these vents form a part of the reactor coolant pressure boundary, the ; design of the vents shall conform to the requirements of Appendix A to 10CFR50, Response to TMI Related Maners - Amendment 1 1A-6
25A5113Rsv. A SBWR standardsafety Analysis oneport O' General Design Criteria. The vent system shall be designed with sufficient redundancy that assures a low probability ofinadvertent or irreversible actuation. Each license shall provide the following information concerning the design and operation of the high point vent system: a Submit a description of the design, location, size, and power supply for the vent system along with results of analyses for lossof-coolant accidents initiated by a break in the vent pipe. The results of the analyses should demonstrate compliance with the acceptance criteria of 10CFR50.46. m Submit procedures and supporting analysis for operator use of the vents that also ! include the information available to the operator for initiating or terminating vent usage.
Response
The capability to vent the SBWR reactor coolant system is provided by the safety / relief valves and reactor vessel head vent lines. The capability of these components and their satisfaction of Item II.B.1 is discussed below. The SBWR design is provided with eight safety / relief valves (SRVs) which are mounted on top of the main steamlines in the drywell and discharge through lines routed to quenchers in the suppression pool. The SRVs can be operated individually in the power-actuated mode by remote manual controls from the main control room. Plant procedures will govern the operator's use of the relief mode for venting reactor pressure. These SRVs satisfy the intent of the NRC position. The BWR Owners' Group position is that the requirement of single-failure criteria for prevention ofinadvertent actuation of these valves, and the requirement that power be removed during normal operation, are not applicable to BWRs. These safety / relief mlves provide ASME code overpressure protection via their independent safety mode l of operation and serve an important depressurization function in mitigating the effects oflocals. Therefore, the addition of a second " block" valve to the vent lines would result l l in a less safe design and a violation of the code. Moreover, the inadvertent opening of a relief valve in a BWR is a design-basis event and results in a controllable transient. In addition to these automatic (or manual) relief valves, the SBWR design includes vent and purge lines. Durir.g reactor operation, the noncondensable gases are drawn to the steamline through the 2-inch vent li ae from the RPV head, and the 3/4-inch purge line ; i from the Isolation Condeaser (IC). The vent line is isolated from the Equipment and Floor Drainage System (EFDS) wiin two normally closed valves during reactor power , operation. These vent and purge lines are not required to assure natural circulation I 1 core cooling. 1A-6 Response to TMI Related Matters - Amendment 1 l l 1
2SA5113Rsv. A SBWR standardSafety Analysis Repen v 1 A.2.6 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation [lI.B.2] NRC Position , Each licensee shall perform a radiadon and shielding design review of the spaces around systems that may, as a result of an accident, contain highly radioactive materials. The design review should identify the location of vital areas and equipment, such as the l control room, radwaste control stations, emergency power supplies, motor control centers, and instrument areas, in which personnel occupancy may be undulylimited or safety equipment may be unduly degraded by the radiation fields during post-accident operation of these systems. Each licensee shall provide for adequate access to vital areas and protection of safety equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. The design review shall determine which types of corrective actions are needed for vital areas throughout the facility.
Response
A review of the radiation and shielding of the SBWR Standard Plant post-accident [' operations has been made. It has been found that there is adequate access to vital areas ( and that safety equipment is adequately protected. An evaluation of post-accident radioactive sources concluded that the SBWR design , limits potential radiation exposure from accidents both to plant personnel and to the public by the use of passive safety features and holdup in the containment and safety envelope. The evaluation showed that for all but the most improbable accident scenarios resulting from multiple failures of all major systems including passive systems, radioactive sources from the pressure vessel will be adequately contained in the reactor pressure vessel or containment. The turbine building contains no major sources of releasable radioactivity (discounting NMbecause of the 7.7 second half-life) and potential releases are limited to liquid releases oflow activity water from the feedwater and condensate systems. The offgas system and the condenser demineralizers contain radioactive species but in a form not amenable for release due to heavy shielding and compartmentalizing of these components. Potential releases in the radwaste building will be contained by isolating the radwaste building atmosphere and sealing any w2ter releases in the building which is seismically qualified and lined to prevent any potential water releases from high activity areas. Additional details relating to plant radiation sources can be found in Section 12.2. A radiation shielding evaluation concluded that the shielding protects operating personnel and the general public from radiation emanating from the reactor, the b power conversion systems, the radwaste process systems, and the auxiliary systems, while U maintaining appropriate access for operation and maintenance. In addition, the radiation shielding is designed to keep radiation doses to equipment below levels at Response to TML Related Matters - Amendment 1 1A-7
25AS113Riv. A SBWR stamtantsarety Anarrsis neport O which disabling radiation damage occurs. In particular, shielding limits the radiation exposure of personnel, in the unlikely event of an accident, to levels that are ALARA and which conform to the limits specified in 10CFR50, Appendix A, Criterion 19 to ensure that the plant is maintained in a safe condition during an accident. Additional details relating to shielding can be found in Subsection 12.3.2. The locadons requiring access to mitigate the consequences of an accident during the post-accident period are the control room, the technical support center, the remote shutdown panel, the primary containment sample station (post accident sample system), the health physics facility (counting room), the control room air bottles, the IC pool refill nozzles, and the nitrogen gas supply bottles. Each aru has low post-LOCA radiation levels. The dose evaluations in Subsection 15.6.5 are within regulatory guidelines. Access to vital areas throughout the reactor building / turbine building complex is controlled via the senice building. Entrance to the senice building and access to the other areas are controlled via double locked secured entry ways. Access to the reactor building is via two specific routes, one for clean access and the second for controlled access. Radiation exposure is limited to gamma shine from the reactor building, turbine building, main steam line access corridor, and sky shine. During a design basis accident event, access to the remote shutdown panel, nitrogen bottles, and the Post-Accident Sampling System (PASS) and monitor systems is controlled from the senice building via the controlled access way. These corridors are not maintained under filtered positive pressure so that personal protection equipment (radiation protection suits, breathing gear, etc.) will be required in the access corridor. Primary contamination would occur from leakage through the PASS and air infiltration from the environment. Both pathways are expected to have minor contamination under even the most adverse conditions. Access to the IC pool refill nozzles and the control room air bottles is from outside and would not require special breathing gear, except in the worst case (DBA LOCA). In the worst case, breathing apparatus would be requin:d j both to reach and operate the remote shutdown panel. The reactor building vital areas are all located off the controlled access way and contamination is limited to air infiltration from the environment and penetration leakage from the PASS. Sources of radiation in each area are limited to gamma shine from the reactor building and potentialleakage from monitoring systems such as the PASS. The post-accident adiation zone maps for the areas in the reactor building are presented in Figure 21.12.3-2. The zone maps represent the maximum gamma dose rates that exist in these areas during the post-accident period. These dose rates do not include the airborne contribution in the reactor building. The zone maps are designed ' 1A-8 Response to TMI Related Matters - Amendment 1 l l l
l 25AS113Rn. A SBWR senndedsarnyannorsesnerat ; to reflect the criteria established in Subsection 3.1.2. Additional details relating to post-accident access requirements can be found in Subsection 12.3.5. An emironmental qualification program for safety-related mechanical and electrical equipment to demonstrate their capability to perform their required functions when exposed to the emironmental conditions (including accident and post-accident conditions) in their respective locations is described in Section 3.11 1 A.2.7 Post-Accident Sampling [lI.B.31
. NRC Position A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to prompdy obtain (less than I hour) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, respectively. Accident conditions should assume a Regulatory Guide 1.3 or 1.4 release of fission products. If the review indicates that 4
personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet the criteria. \ A design and operational review of the radiological spectrum anabsis facilities shall be performed to determine the capability to promptly quantify (in less than 2 hours) certain radionuclides that are indicators of the degree of core damage. Such ; radionuclides are noble gases (which indicate cladding failure), iodines and cesiums ' (which indicate high fuel temperatures), and nonvolatile isotopes (which indicate fuel melting). The initial reactor coolant spectrum should correspond to a Regulatory l Guide 1.3 or 1.4 release. The review should also consider the effects of direct radianon from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents. If the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the j criteria. ] In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be prosided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory i Guide 1.3 or 1.4 source term). Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride sample i analysis within a shift). . l Response to TMI Related Matters - Amendment 1 1A-9 l
- 25A5113Riv. A
) SBWR standardsafety Analysis Report l 9 l Response The Post-Accident Sampling System described in Subsection 9.3.2 meets the requirements of this position with the following exception. The upper limit of activity l in the samples at the time they are taken is as follows: l l liquid sample 1 Ci/g; and gas sample 105 micro Ci/cc l The Post-Accident Sampling System (PASS) obtains reactor coolant and other samples l l following an accident. Liquid samples are taken from the Reactor Water Cleanup l (RWCU) Inlets (hi-temp) and (lo-temp) via the connection to the Process Sampling System (PSS). Gas samples are obtained from a sample line connected to the i Containment Atmosphere Monitoring System (CAMS). l The PSS collects representative liquid samples for analysis and provides the analytical information required to monitor plant and equipment performance and changes to i operating parameters. The CAMS monitors the atmosphere in the containment for ! high gross gamma radiation levels and for high concentration levels of oxygen and hydrogen during post-accident conditions. Also, these three parameters are monitored during normal reactor operations to evaluate the integrity and safe conditions of the containment. Detailed descriptions of the PSS and CAMS can be found in Subsections 9.3.2.2 and 7.5.2, respectively. Means to reduce radiation exposure are provided, such as shielding, remotely operated valves, and sample transporting casks. t ! Acceptance Criterion II.K.5 of SRP Subsection 9.3.2 requires the capability of sampling i liquids of 10 Ci/g. The SBWR design has the capability of sampling liquids of I Ci/g. l Sampling and area radiation measurement will be performed. Iflevels are above safe I limits, handling samples will be delayed. The Process Radiation Monitoring System (PRMS) identifies the various gaseous and f liquid process streams and efIluent release paths or points to be monitored and sampled, and defines the required instmmentation for detection and measurement of the radioactive contents of these streams. It alerts operating personnel to excessive radiation levels and will initiate automatically the required protection action to isolate radioactivity releases to the emirons. The ."RMS is designed for operability during and following an accident. A detailed description of the PRMS can be found in Subsection 7.5.3 and in Section 11.5. O l l 1A 10 Resporse to TMIRelated Matters - Amendment 1
a w 25A5113R:v. A SBWR standedsurnyAnalysisRowt x 1 A.2.8 Rule Making Proceeding on Degraded Core Accidents [lI.B.81 A summary of the NRC position and GE's associated response to this TMI action plan item is addressed in Subsection 19G.2.1. 1 A.2.9 Coolant System Valves-Testing Requirements [lI.D.11 NRC Position Pressurized-water reactor and boiling-water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operadng conditions for design-basis transients and accidents.
Response
The overpressure protection system, of which the SRVs are a part of,is capable of accommodating the most severe pressurization transient. The SBWR pressurization is mild relative to previous BWR designs because of the large steam volume in the chimney and vessel head, which mitigates the pressurization and does not result in opening of relief valves prior to isolation condenser initiation. A detailed description of the safety enluation of transients for the overpressure protection system can be found in ; p Subsection 5.2.2.3.2. !
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The inspection and testing of applicable SRVs utilizes a quality assurance program which complies with Appendix B of 10CFR50. 1 The SRVs are tested at a suitable test facility in accordance with quality control ] procedures to detect defects and to prove operability prior to installation. The ;
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conducted tests include hydrostatic, steam leakage, full flow pressure and blowdown, and response time testing. The valves are installed as received from the factory. The valve manufacturer certifies that design and performance requirements, including capacity and blowdown, have been met. The setpoints are adjusted, verified, and indicated on the valves by the vendor. Specified manual and automatic initiated signal for power actuation (relief mode) of each SRV is verified during the preoperational test program described in Chapter 14. It is not feasible to test the SRV setpoints while the valves are in place. The valves can be removed for maintenance or bench testing and reinstalled during normal plant shutdowns. The valves will be tested to check set pressure in accordance with the requirements of the plant Technical Specifications. The external and flange seating surfaces of all SRVs are 100% visually inspected when the valves are removed for maintenance or bench testing. b Response to TMI Related Matters - Amendment 1 1A.11
i i 25AS113 Rev. A SBWR standardsafety Anarrsts neport i l l O 1 A.2.10 Relief and Safety Valve Position Indication Ill.D.3] 1 - NRC Position i Reactor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve-position detection device or a reliable indication of flow in the discharge pipe.
Response
SRV position is indicated in the control room in full compliance with this requirement. 1 A.2.11 Systems Reliability [lI.E.3.2] This TMI action plan item is superseded by USI A-45, which is addressed in Appendix 19H. i 1 A.2.12 Coordinated Study of Shutdown Heat Removal Requirements [lI.E.3.3] This TMI action plan item is superseded by USI A-45, which is addressed in Appendix 19H.
- 1 A.2.13 Containment Design-Dedicated Penetration [lI.E.4.1]
NRC Position For plant designs with external hydrogen recombiners, provide redundant dedicated containment pencuations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere.
Response
The Flammability Control System (FCS) does not use external hydrogen recombiners i that require redundant dedicated penetrations. Therefore, this TMI requirement is not applicable to the SBWR Standard Plant design. The SBWR FCS design utilizes inerting l 4 and passive autocatahtic recombiners hydrogen ig"!!ers for the purpose of nreventme i lhe mitigating e potential buildup of combustible gases generated from the radiolytic decomposition of water and from 100% metal-water reaction of the active fuel cladding during a LOCA. 1 A.2.14 Containment Design-Isolation Dependability [lI.E.4.2] NRC Position a Containment isolation system designs shall comply with the recommendations of Standard Review Plan Subsection 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of containment isolation). a All plant personnel shall give careful consideration to the definition of essential and i non-essential systems, identify each system determined to be non-essential, describe 1A-12 Response to TMI Related Matters - Amendment 1 l l
25A5113Rev. A SBWR standantsurnyAamor sis neper s_ the basis for selection of each essential system, modify their containment isolation designs accordingly, and report the results of the reevaluadon to the NRC. m All nonessential systems shall be automatically isolated by the containment isolation signal. 1 m The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require deliberate operator action. i a The containment setpoint pressure that initiates containment isolation for non-essential penetrations must be reduced to the minimum compatible with normal i operating condit'.ons, a Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position CSB fM or the StaffInterim Position of October 23,1979 , must be sealed closed as defined in SRP 6.2.4, Item II.6.f during operational ! conditions 1,2,3, and 4. Furthermore, these valves must be verified to be closed at g least every 31 days. :
/
s Containment purge and vent isolation valves must close on a high radiation signal.
Response
a Redundancy and physical separation are required in the electrical and mechanical design of the containment isolation system to ensure that no single failure in the system prevents it from performing its intended functions. Electrical redundancy is prosided for each set ofisolation valves, eliminating dependency on one power source to attain isolation. Electrical cables for isolation valves in the same line are routed separately. Cables are selected and based on the specific emironment to which they may be subjected (e.g., magnetic fields, high radiation, high temperature and high humidity). s This is directed primarily toward operating plants. However, the classification of structures, systems and components for the SBWR Standard Plant design is addressed in Section 3.2 and identified in Table 3.2-1. The basis for classification is , also presented in Section 3.2. The SBWR Standard Plant fully conforms with the NRC position so far as it relates to the new equipment supplier. e The containment isolation system, in general, closes fluid penetrations for support systems that are not safety-related. Response to TMI Related Matters - Amendment 1 1A-13
l [ 25A5113 R1v. A SBWR standardsarery Analysis neport i a The design of the control systems for automatic containment isolation valves ensures that resetting the isolation signal shall not resultin the automatic reopening of containment isolation valves. m Actuation of the containment isolation system is automatically initiated by the Leak Detection and Isolation System (LD&lS) at specific limits defined for reactor plant l l operation. The LD&IS (described in Subsections 5.2.5 and 7.3.3) is designed to l detect, monitor and alarm leakage inside and outside the containment and will i automaticallyinitiate the appropriate protective action to isolate the source of the leak. Various plant variables are monitored, including pressure, and these are used in the logic to isolate the containment. The dnwell pressure is monitored by four divisional channe!s using pressure transmitters to sense the drywell atmospheric pressure from four separate locations. A pressure rise above the nominal level indicates a possible leak or loss of reactor coolantwithin the drywell. A high pressure indication will be alarmed in the main control room and will initiate reactor uip and closure of the containment isolation valves. The alarm setpoints of the LD&lS are determined analytically or are based on actual measurements made during startup and pre-operational tests. m All SBWR containment purge valves meet the criteria provided in BTP CSB 6-4. The main purge valves are fail-closed and are verified to be closed at a frequency interval of 31 days as defined in the plant technical specifications. All purge and vent valves are pneumatically operated, fail closed and receive containment isolation signals. Bleed valves and makeup valves can be remote manually opened in the presence of an isolation signal, by utilizing override control switch if continued inerting is necessary. e in the fBWR design, redundant primary containment isolation valves (purge and vent) close automatically upon receipt of an isolation signal from the Leak Detection and Isolation System (LD&lS). The LD&lS is a four-divisional system designed to detect and monitor leakage from the reactor coolant pressure boundary, and will, in certain cases, isolate the source of the leak by initiating closure of the appropriate containment isolation valves. Various plant variables are monitored, including radiation level, and these are used in the logic to initiate alarms and the required control signals for containment isolation. liigh radiation levels detected in the reactor building safety envelope IIVAC air exhaust or in the refueling area air exhaust will automatically isolate the containment purge and vent isolation valves. O 1A-14 Response to TMI Related Matters - Amendment 1
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25A5113Rev. A SBWR standardsafery Analysis neport b g 1 A.2.15 Additional Accident-Monitoring Instrumentation [lI.F.1(1)] NRC Position Noble gas effluent monitors shall be installed with an extended range designed to function during accident conditions as well as during normal operating conditions. Multiple monitors are considered necessary to cover the ranges ofinterest. 5 m Noble gas effluent monitors with an upper range capacity of 10 Ci/cc (Xel83) are considered to be practical and should be installed in all operating plants. s Noble gas effluent monitoring shall be provided for the total range of concentration extending from normal condition (as low as reasonably achievable (ALARA)) , concentrations to a maximum of105 Ci/cc (Xel33). Multiple monitors are considered to be necessary to cover the ranges ofinterest. The range capacity of j individual monitors should overlap by a factor of ten, l Because iodine gaseous effluent monitors for the accident condition are not considered to be practical at this time, capability for cfIluent monitoring of radioiodines for the accident condition shall be provided with sampling conducted by absorption on charcoal or other media, followed by onsite laboratory analysis. O In-containment radiation-level monitors with a maximum range of 108 rad /hr shall be installed. A minimum of two such monitors that are physically separated shall be prosided. Monitors shall be developed and qualified to function in an accident emironment. A continuous indication of containment pressure shall be provided in the control room of each operating reactor. Measurement and indication capability shall include three ]' times the design pressure of the containment for concrete, four times the design pressure for steel, and -5 psig for all containments. A continuous indication of containment water level shall be provided in the control room for all plants. A narrow range instrument shall be provided for BWRs and cover the range from the bottom to the top of the containment sump. A wide range i instrument shall also be provided for BWRs and shall cover the range from the bottom of the containment to the elevation equivalent to a 600,000 gallon capacity. For BWRs, a wide range instrument shall be provided and cover the range from the bottom to 5 feet above the normal water level of the suppression pool. A continuous indication of hydrogen concentration in the containment atmosphere shall be provided in the control room. Measurement capability shall be provided over the range of 0 to 10% hydrogen concentration under both positive and negative O ambient pressure. Response to TMI Related Matters - Amendment 1 1A-15
25AS113R2v. A l SBWR standantsarery Analysis neport O
Response
GE believes the requirements of Regulatory Guide 1.97, Revision 3, incorporate the above requirements. Section 7.5 compares the SBWR design against this Regulatory Guide. l 1 A.2.16 Identification of and Recovery From Conditions Leading to inadequate Core Cooling III.F.2] NRC Position Licensees shall provide a description of any additionalinstrumentation controls (primary or backup) proposed for the pbat to supplement existing instmmentation (including primary coolant saturation monitors) in order to provide an unambiguous, easy-to-interpret indication ofinadequate core cooling (ICC). A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in l ! developing these procedures, and a schedule for installing the equipment shall be provided.
Response
The detection of conditions indicative ofinadequate core cooling will be provided in the SBWR design by the direct water level instrumentation system. GE intends to use the reference-leg type level instrumentation design described in NEDO-24708A. Currently, there is an industry issue (documented in Letter BWROG 92074, "BWR Reactor Vessel Water Level Instrumentation," dated 8/28/92) regarding erroneously high water level indication upon vessel depressurization due to the release of dissolved non-condensable gases in the reference leg. In response to this issue, the BWR Owners' Group (BWROG) is currently evaluating the water level instrumentation design for approval by the NRC. Preliminary findings show that the error in water level indication is negligible. Modifications to the SBWR water level instrumentation system design, if any, will be implemented and consistent with the NRC-approved BWROG recommendations. The BWROG water level instrumentation design will provide an unambiguous, easy-to-interpret indication of inadequate core cooling. 1 A.2.17 Instruments for Monitoring Accident Conditions [lI.F.3] NRC Position Provide instrumentation adequate for monitoring plant conditions following an accident that includes core damage.
Response
The SBWR Standard Plant is designed in accordance with Regulatory Guide 1.97, Revision 3 (Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 1A-16 Response to TMI Related Matters - Amendment 1
P 25AS113Rev. A SBWR standantsweryAnarrsisnepar t f" ( Plant and Emirons Conditions During and Following an Accident). A detailed assessment of the Regulatory Guide, including the list ofinstruments,is found in Section 7.5 of this SSAR. 1 A.2.18 Safety-Related Valve Position Indication [lI.K.1(5)] NRC Position a Review all valve positions and positioning requirements and positive controls and all related test and maintenance procedures to assure proper ESF functioning, if required. l m Verify that AFW valves are in open position.
Response
a The SBWR Standard Plant is equipped with status monitoring that satisfies the requirements of Regulatory Guide 1.47 (Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems). Subsection 7.1.2 contains detailed information on the status monitoring equipment and capability provided in the SBWR Standard Plant design. l
- v In addition to the status monitoring, plant specific procedures (see l
Subsection 1A.S.2) will assure that independent verification of system line-ups is applied to valve and electrical line-ups for all safety-related equipment, to surveillance procedures, and to restoration following maintenance. Through these procedures, approval will be required for the performance of surveillance tests and maintenance, including equipment removal from senice and return to service, a This requirement is not applicable to the SBWR. It applies only to PWRs 1 A.2.19 Review and Modify Procedures for Removing Safety-Related Systems From l Service Ill.K.1(10)1 NRC Position : Review and modify (as required) procedures for removing safety-related systems from I service (and restoring to senice) to assure operability status is known.
Response
This is the responsibility of the COL applicant. See Subsection 1A.S.2 for COL hcense informauon requirements. l Response to TMI Related Matters - Amendment 1 1A.17
25AS113Riv. A l SBWR standardsaretr Analysis Report O 1 A.2.20 Describe Automatic and Manual Actions for Proper Functioning of Auxiliary Heat Removal Systems When FW System Not Operable [lI.K.1(22)] l NRC Position l For boiling water reactors, describe the automatic and manual actions necessary for proper functioning of the auxiliary heat removal systems that are used when the main feedwater system is not operable (see Bulletin 79-08, Item 3).
Response
There are no short term manual actions which must be taken. Sufficient systems exist to I automatically mitigate the consequences of a loss of feedwater event. An analysis was performed for a loss of feedwater event. The sequence of events is I described in Table 15.6-16, and is summarized below. In the event that the main feedwater system is not operable, reactor water level will fall l due to boil-off and absence of makeup water. When Level 3 is reached, a reactor scram will be automatically initiated. Reactor water level will condnue to decrease due to boil-off until the low-low level setpoint, Level 2,is reached. At this point, the reactorisolation occurs, and an initiation signal is received by the isolation condensers (ICs). Thirty seconds later, the ICs are in full operation, and the water level will have stabilized. If the ICs are not operable, the safety /reliefvalves (SRVs) open on high vessel pressure five and a half minutes later. The SRVs open and close to maintain vessel pressure at approximately 8.51 MPa gauge (1235 psig). When reactor low water Level 1 is reached (if ICs are not operable), the 10-second ADS timer is initiated. When the ADS timer is timed out, the ADS actuation sequence is initiated, and the 150-second GDCS timer is initiated. When the GDCS timer is timed out, the GDCS injection valves open. Vessel pressure then decreases below the shutoff head of GDCS, and the GDCS reflooding flow i into the vessel begins. The core remains covered throughout the sequence of events and no core heatup occurs. 1 A.2.21 Describe Uses and Types of RV LevelIndication for Automatic and Manual Initiation of Safety Systems [lI.K.1(23)] NRC Position For boiling water reactors, describe all uses and types of reactor vessel level indication for both automatic and manual initiation of safety systems. Describe other instrumentation that might give the operator the same information on plant status. See Bulletin 79-08, item 4. l Response l GE intends to use the reference-leg type level instmmentation design described in NEDO-24708A, Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors. Currently, there is an industry issue regarding erroneously high 1A.18 Response to TMI Related Matters - Amendment 1
2SA5113Rev. A SBWR standantsafery Anstrsis neport rm
\
(G water level indication upon vessel depressurization due to the release of dissolved non-condensable gases in the reference leg. In response to this issue, the BWR Owners' Group (BWROG) is currently evaluating the water level instrumentation design for approval by the NRC. Preliminary findings show that the error in water level indication is negligible. Modifications to the SBWR water level instrumentation system design, if any, will be implemented and consistent with the NRC-approved B% TOG recommendations. An outline of the description of the reference-leg type levelinstrumentation design as applicable to the SBWR Standard Plant is provided in the following paragraphs. Figure 7.7-1 shows the water level range and the vessel penetrations for each water level range.The instruments are differential pressure devices calibrated for the specific vessel conditions (pressure and liquid temperature) conditions for which the level range is provided. All reactor water level instrumentation is referenced to a common reference zero: the top of active fuel (TAF). Reactor water level instrumentation that initiates safety-related systems and engineered l I y. safeguards systems is discussed in Subsections 7.2.1 and 7.3.1. Reactor water level ( instrumentation that is used as part of the Feedwater Control System is discussed in l V) Subsection 7.7.3. i 1 A.2.21.1 Failure of PORV or Safety to Close [ll.K.3.(3)] 1 NRC Position 1 Assure that any failure of a PORV or safety valve to close will be reported to the NRC promptly. All challenges to the PORVs or safety valves should be documented in the annual report. This requirement is to be met before fuel load.
Response
See Subsection 1A.S.4 for COL license information requirements. 1 A.2.22 Separation of HPCI AND RCIC System Initiation Levels [II.K.3(13)] NRC Position Currently, the Reactor Core Isolation Cooling (RCIC) System and the High-Pressure Coolant Injection (HPCI) System both initiate on the same low- water-level signal and both isolate on the same high-water-level signal. The HPCI System will restart on low water level but the RCIC System will not. The RCIC System is a low-flow system when compared to the HPCI System. The initiation levels of the HPCI and RCIC Systems should be separated so that the RCIC System initiates at a higher water level than the p HPCI System. Further, the initiation logic of the RCIC System should be modified so (" ) that the RCIC System will restart on low water level. These changes have the potential to reduce the number of challenges to the HPCI System and could result in less stress Response to TM) Related Matters - Amendment 1 1A-19
l t 25A5113 Riv. A SBWh standantsarety Analysis nept i e on the vessel from cold water injection. Analyses should be performed to evaluate these j changes. The analysis should be submitted to the NRC staff and changes should be implemented ifjustified by the analysis. i :
Response
High pressure inventory control and reactor decay heat removal following reactor isolation for the SBWR is by means of the Isolation Condenser System (ICS). The SBWR ICS replaces the traditional HPCI and RCIC Systems found in most BWRs, thus eliminating concerns about cold water injection and system initiadon. The ICS inidates automatically on high reactor pressure, low reactor water level (Level 2), or on closure of the MSIVs whenever the reactor mode switch is in the RUN position. SBWR low pressure inventory control is via the Gravity-Driven Cooling System (GDCS), which initiates at a lower water level (Level 1) than the ICS. The ICS will automatically restart on Level 2 in the RUN position. 1 A.2.23 Modify Break-Detection Logic to Prevent Spurious isolation of HPCI and RCIC Systems [lI.K.3(15)] NRC Position The High-Pressure Coolant injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems use differential pressure sensors on elbow taps in the steam lines to their turbine drives to detect and isolate pipe breaks in the systems. The pipe-break-detection circuiuy has resulted in spurious isolation of the HPCI and RCIC Systems due to the pressure spike which accompanies startup of the systems. The pipebreak-detection circuiuy should be modified so that pressure spikes resulting from HPCI and RCIC System initiation will not cause inadvertent system isolation.
Response
This TMI item applies to high pressure systems with steam-driven pumps that use , differential pressure sensors on elbow taps in the steam supply line to isolate the steam supply in the event of a pipe break. The SBWR isolation Condenser System (ICS) l replaces the traditional HPCI and RCIC Systems found in most BWRs. Although the ICS uses differential pressure transmitters to detect a possible pipe break, the system does not utilize steam-driven pumps. Therefore, this TMl item is not applicable to the SBWR Standard Plant design. 1 A.2.24 Reduction of Challenges and Failures of Relief Valves- Feasibility Study and f System Modification [lI.K.3(16)] NRC Position The record of relief-valve failures to close for all boiling water reactors (BWRs) in the past 3 years of plant operation is approximately 30 in 73 reactor-years (0.41 failures per reactor-year). This has demonstrated that the failure of a relief valve to close would be the most likely cause of a small-break loss-of-coolant accident (LOCA). The high failure 1A-20 Response to TMI Related Matters - Amendment 1
l 2SA5113Rtv. A SBWR standardsafety Analysis neport l \
])
l v i rate is the result of a high relief-valve challenge rate and a relatively high failure rate per challenge (0.16 failures per challenge). Typically, five valves are challenged in each i event. This results in an equivalent failure rate per challenge of 0.03. The challenge and l failure rates can be reduced in the following ways: 1 m Additional anticipatog scram on loss of feedwater, I a Revised relief-valve actuation setpoints, a Increased emergency core cooling (ECC) flow, m Lower operating pressures, a Earlier initiation of ECC systems, a Heat removal through emergency condensers, m Offset valve setpoints to open fewer valves per challenge, m Installation of additional relief valves with a block or isolation-valve feature to
/ eliminate opening of the safety / relief valves (SRVs), consistent with the ASME
"'}
y/ Code, a Incieasing the high steam line flow setpoint for main steam line isolation valve (MSIV) closure, a Lowering the pressure setpoint for MSIV Closure, a Reducing the testing frequency of the MSIVs, a More stringent valve leakage criteria, and a Early removal ofleaking valves. An investigation of the feasibility and constraints of reducing challenges to the relief j valves by use of the aforementioned methods should be conducted. Other methods should also be included in the feasibility study. Those changes which are shown to reduce relief-valve challenges whhout compromising the performance of the relief valves or other systems should be implemented. Challenges to the relief valves should i be reduced substantially (by an order of magnitude). !
Response
General Electnc and the BWR Owners' Group responded to this requirement in Reference I A-9. This response, which is based on a review of the existing operating []
\m,/ information on the challenge rate of relief valves, concluded that the BWR/6 product line had already achieved the " order of magnitude" level of reduction in SRV challenge .
1 1A-21 Response to TMI Related Matters - Amendment 1 I
25A5113 R:v. A SBWR standardsafety Anssysis neput O rate.The principal reason for this reduction is that thc BWR/6 uses direct acting SRVs, not the pilat operated design used in some earlier BWRs. The SBWR also anticipates the use of direct acting SRVs. The SBWR relief valve system also has similar design features which also reduce the SRV challenge rate, since with regard to inadvertently opened relief valves (IORVs), the BWR/6 plant design evaluated for the Owners' Croup report reflected a reduced levelin IORVs compared with previous design because of elimination of the pilot operated relief valve type of design. Furthermore, the SBWR design incorporates isolation condensers (ICs), which eliminates SRV actuation for transients. Because of the incorporation of the ICs and solid state redundant logic in the SBWR design, the likelihood of an IORV is less than the BWR/6 design evaluated in connection with the Owners' Group report. The redundant solid state design has been selected in order that the frequency of IORV with solid state logic becomes low enough so as to achieve the order of magnitude reduction in total SRV challenge rate required by NUREG-0737. 1 A.2.25 Report on Outages of Emergency Core-Cooling Systems Licensee Report and Proposed Technical Specification Changes [ll.K.3(1711 NRC Position Several components of the Emergency Core Cooling (ECC) Systems are permitted by technical specifications to have substantial outage times (e.g., 72 hours for one diesel-generator; 14 days for the HPCI System). In addition, there are no cumulative outage time limitations for ECC Systems. Licensees should submit a report detailing ourage dates and lengths of outages for all ECC systems for the last 5 years of operation. The report should also include the causes of the outages (i.e., controller failure, spurious isolation). Clarification The present technical specifications contain limits on allowable outage times for ECC systems and components. However, there are no cumulative outage time limitations on these seme systems. It is possible that ECC equipment could meet present technical specification requirements but have a high unavailability because of frequent outages within the allowable technical specifications. < The licensees should submit a report detailing outage dates and length of outages for all ECC Systems for the last 5 years of operation, including causes of the outages. This report will provide the staff with a quantification of historical unreliability due to test l and maintenance outages, which will be requirements in the technical specifications. Based on the above guidance and clarification, a detailed report should be submitted. The report should contain (1) outage dates and duration of outages; (2) causes of the 1A-22 Response to 1MI Related Matters - Amendment 1
25AS113Rsv. A SBWR standantsafety Analysis Reput O outage; (3) ECC Systems or components involved in the outage; and (4) corrective action taken. Tests and maintenance outages should be included in the above listings t which are to cover the last 5 years of operation. The licensee should propose changes to improve the availability of ECC equipment, if needed. Applicants for an operating license shall establish a plan to meet these requirements.
Response
See Subsection 1A.S.5 for COL license information requirements. 1 A.2.26 Modification of Automatic Depressurization System Logic - Feasibility for increased Diversrty for Some Event Sequences [lI.K.3(18)] NRC Position The Automatic Depressurization System (ADS) actuation logic should be modified to eliminate the need for manual actuation to assure adequate core cooling. A feasibility and risk assessment studyis required to determine the optimum approach. One , possible scheme that should be considered is ADS actuation on low reactor-vessel water level provided no High-Pressure Coolant Injection (HPCI) or High-Pressure Core It'r*7 System (HPCS) flow exists and a low-pressure emergency core cooling (ECC) system is ; running. This logic would complement, not replace, the existing ADS actuation logic. (
Response
Actuation ofADS equipment is performed automatically upon receipt of a reactorwater Level 1 signal that persists for at least 10 seconds, and without need for operator action. Manual actuation is also possible. Automatic ADS complements manual ADS. Subsection 7.3.1.1.2 descdbes the logic and sequencing of the ADS in detail. ! For the above reasons, this TMI issue is considered resolved for the SBWR Standard Plant design. 1 A.2.27 Restart of Core Spray and LPCI Systems on Low Level Design and Modification [lI.K.3(21)] NRC Position The Core Spray and Low Pressure Coolant injection (LPCl) System flows may be l stopped by the operator. These systems will not restart automatically on loss of w2ter ) level if an initiation signal is still present. The Core Spray and LPCI System logic should be modified so that these systems will restart, if required, to assure adequate core cooling. Because this design modification affecta several core <ooling modes under accident conditions, a preliminary design should be submitted for staff review and , approval prior to making the actual modification. l J Response to TMI Related Matters - Amendment 1 1A-23
i 25AS113Ratt. A SBWR standantsarety Anatrsis Report i l O 1 Response This TMI item applies to low pressure inventory control systems (Core Spray and LPCI) that can be stopped by the operator. Once the SBWR low pressure injection system, l GDCS, is initiated, the operator does not have the ability to stop it from completing the l initiation sequence. Therefore, this TMI item does not apply to the SBWR Standard Plant. 1 A.2.28 Automatic Switchover of Reactor Core Isolation Cooling System Suetion - Verify Procedures and Modify Design Ill.K.3(22)] NRC Position The Reactor Core Isolation Cooling (RCIC) System takes suction from the condensate storage tank with manual switchover to the suppression pool when the condensate storage tank level is low. This switchover should be made automatically. Until the automatic switchover is implemented, licensees should verify that clear and cogent procedures exist for the manual switchover of the RCIC System suction from the ( condensate storage tank to the suppression pool. l l Response This TMI item applies to reactor cooling systems that can take suction from multiple water sources. The SBWR has no safety systems that take suction from multiple water sources. The ICS returns steam condensate from the isolation condensers via a condensate return pipe to the reactor vessel and does not have the capability to manually or automatically switch over to an alternate source. Neither does the GDCS, I which takes suction from the GDCS pools. Therefore, this TMI item does not apply to the SBWR Standard Plant. 1 A.2.29 Confirm Adequacy of Space Cooling for High Pressure Coolant injection and Reactor Core Isolation Cooling Systems Ill.K.3(24)) NRC Position - Long-term operation of the Reactor Core Isolation Cooling (RCIC) and High-Pressure Coolant injection (HPCI) Systems may require space cooling to maintain the pump-room temperatures within allowable limits. Licensees should verify the acceptability of the consequences of a complete loss of alternating-current power. The RCIC and HPCI Systems should be designed to withstand a complete loss of offsite alternating-current power to their support systems, including coolers, for at least 2 hours.
Response
The SBWR ICS replaces the traditional HPCI and RCIC Systems found in most BWRs. The ICS does not rely on active pumps to remove excess sensible and core decay heat. l Each isolation condenser is located in a subcompartment of the IC/PCC pool, and ' requires no additional space cooling other than that provided by the surrounding water ; in the IC/PCC pool. If all of the three safety-related power supplies used to start the ICs l Response to TMI Related Matters - Amendment 1 1A-24
1 2SA5113Rsv. A SBWR stamrantsceryansrysisneport f i
\
l were to fail, then all available ICs will automatically stan into operation because of the
" fail open" actuation of the condensate return bypass valves on loss of electrical power to the solenoids which control these air-actuated valves. Therefore, this TM1 item is considered resolved for the SBWR Standard Plant design.
1 A.2.30 Effect of Loss of Alternating-Current Power on Pump Seals [lI.K.3(25)) O t NRC Position The licensees should determine, on a plant-specific basis, by analysis or experiment, the consequences of a loss of cooling water to the reactor recirculation pump seal coolers. l The pump seals should be designed to withstand a complete loss of alternating-current (ac) power for at least 2 hours. Adequacy of the seal design should be demonstrated.
Response
The SBWR design eliminates recirculation pumps and associated piping. Circulation of the reactor coolant through the SBWR core is unaided natural circulation. Therefore, this TMI item does not apply to the SBWR Standard Plant. 1 A.2.31 Study and Verify Qualification of Accumulators on Automatic Depressurhation System Valves [lI.K.3(28)]
- NRC Position l Safety analysis reports claim that air or nitrogen accumulators for the Automatic
! Depressurization System (ADS) valves are provided with sufficient capacity to cycle the valves open five times at design pressures. GE has also stated that the Emergency Core : Cooling (ECC) Systems are designed to withstand a hostile environment and still , perform their function for 100 days following an accident. The Licensee should verify that the accumulators on the ADS valves meet these requirements, even considering normal leakage. If this cannot be demonstrated, the Licensee must show that the accumulator design is still acceptable. Rrsponse The ADS utilizes the safety / relief valves (SrWs) and the depressurization valves (DPVs) for depressurization of the reactor. l l Each SRV is equipped with a pneu matic accumulator and check valve for the ADS and the manual opening functions. These accumulators assure that the valves can be opened following failure of the gas suppy ta th: accumulators. The accumulator capacity is sufficient for one actuation at drywell design pressure, or five actuations at normal drywell pressure. The valves have been designed to achieve the maximum practical number of actuations consistent with state-of-the-art technology. !lp) The DPVs are of a non-leak /non-simmer /non-maintenance design. They are straight-l through, squit>acuated, non-reclosing valves with a metal diaphragm seal. Response to TMI Related Mamw - Amendment 1 1A-2S i l
I 15A5113Rw. A SBWR standardsareer Aastrsis Report l O The SRVs and DPVs and associated controls and actuation circuits are located or protected so that their function cannot be impaired by consequential effects of accidents. ADS components are qualified to withstand the harsh emironments l l postulated for design basis accidents inside the containment, including high temperature, pressure, and radiation emironments. l 1 A.2.32 Revised Small-Break Loss-of Coolant Accident Methods to Show Compliance with 10CFR50, Appendix K [lI.K.3(30)] NRC Position The analysis methods used by Nuclear Steam Supply System (NSSS) vendors and/or fuel suppliers for small-break loss-of-coolant accident (LOCA) analysis for compliance l with Appendix K to 10CFR50 should be revised, documented, and submitted for NRC approval. The revisions should account for comparisons with experimental data, including data from the LOFT Test and Semiscale Test facilities. l
Response
GE has evaluated the NRC request requiring that the BWR small-break LOCA analysis methods are to be demonstrated to be in compliance with Appendix K to 10CFR50 or that they be brought into compliance by analysis methods changes. CE uses the TRACG code for LOCA analyses for the SBWR. The methodology used in the analyses is described in References 1 A-9 through 1 A-11 for NRC review and approval to show wmpliance with 10CFR50.46. The references include the supporting justification and comparisons required by 10CFR50.46. 1 A.2.33 Plant-Specific Calculations to Show Compliance with 10CFR50.46 [lI.K.3(31)] NRC Position Plant-specific calculations using NRC-approved models for small-break loss-of-coolant accidents (LOCAs) as described in item II.K.S.30 to show compliance with 10CFR50.46 should be submitted for NRC approval by all licensees. Response j The SBWR standard safety small-break LOCA calculations are discussed m ; l Subsection 6.3.4.7. Subsection 6.3.4.2 lists the applicable acceptance criteria, extracted j from 10CFR50.46, and indicates the applicable parts of Subsection 6.3.4 (where conformance is demonstrated) for each criterion. As noted in Subsection 1A.2.32, GE uses the TRACC code for LOCA analyses for the SBWR. O l l I 1A-26 Response to TMI Related Matters - Amendment 1 l l [
2SAS113Rtv. A SBWR stardantsneerAnstrsisReport >Q 1 A.2.33.1 Evaluation of Anticipated Transients with Single Failure to Verify No Fuel Failure [ll.K.3 (44)] NRC Position For anticipated transients combined with the worst single failure and assuming proper operator actions, licensees should demonstrate that the core remains covered or provide analysis to show that no significant fuel damage results from core uncovery. Transients which in a stuck-open relief valve should be included in this category. The results of the evaluation are dueJanuary 1,1981.
Response
GE and the BWR Owners' Group have concluded, based on a representative BWR/6 plant study, that all anticipated transients in Regulatory Guide 1.70, Revision 3, Chapter 15, combined with the worst single failure, the reactor core remains covered with water until stable conditions are achieved. Analyses performed for the SBWR (see Chapter 15) also show that the reactor core remains covered. Chapter 15 of this SSAR examines the effects of anticipated process disturbances and postulated component failures to determine their consequences and to evaluate the capability built into the plant to control or accommodate such failures and events. The ' scope of the situations analyzed includes Anticipated Operational Occurrences, off-design abnormal transients that induce system disturbances, postulated accidents oflow probability, and finally, hypothetical events of extremely low probability. All of the situations analyzed in Chapter 15 were found to result in adequate fuel integrity. As described in Subsection 15.1.4, the transient resulting from a stuck open relief valve is a mild depressurization which is within the range of normal load following and therefore has no significant effect on reactor coolant pressure boundary and containment design pressure limits. 1 A.2.33.2 Evaluate Depressurization other than Full ADS [lI.K.3 (45)] NRC Position Provide an evaluation of depressurization methods other than by full actuation of the automatic depressurization system, that would reduce the possibility of exceeding vessel integrity limits during rapid cooldown. (Applicable to BWRs only)
Response
This response is provided in Subsection 19G.2.11. 1A-27 Response to TMI Related Matters - Amendment 1
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1 25AS193Rev. A l SBWR StandardSafety Analysis Report O 1 A.2.33.3 Responding to Michelson Concerns [lI.K.3 (46)] l NRC Position General Electric should provide a response to the Michelson concerns as they relate to boiling water reactors. Clarification General Electric provided a response to the Michelson concerns as they relate to boiling water reactors by letter dated February 21,1980. Licensees and applicants should assess applicability and adequacy of this response to their plants.
Response
All of the generic February 21.1980 GE responses were reviewed and updated for the SBWR Standard Plant. m Concern 1 - Pressurizer level is an incorrect measure of primary coolant inventory. Response 1 - BWRs do not have pressurizers. BWRs measure primary coolant inventory directly using reference-leg type level instmmentation with differential pressure devices calibrated for the specific vessel conditions for which the level range is provided. Currently, there is an industry issue (documented in Letter BWROG 92074, "BWR Reactor Vessel Water Level Instrumentation," dated 8/28/92) regarding erroneously high water levelindication upon essel depressurization due to the release of dissolved non<ondensable gases in the reference leg. In response to this issue, the BWR Owners' Group (BWROG) is - currently evaluating the water level instrumentation design for approval by the NRC. Preliminary findings show that the error in water levelindication is negligible. Wdifications to the SBWR water level instrumentation system design,if any, will be ! implemented and consistent with the NRC-approved BWROG recommendations. m Concern 2 -The isolation of small breaks (e.g., letdown line; PORV) not addressed or analyzed. Response 2 - Analyses were performed for small line breaks of the GDCS injection line and the bottom head drain line inside the containment (see Subsection 6.3.4.7.6). LOCA analyses using break areas less than the maximum values were also considered. For postulated pipe breaks, if the break is small, the ; I time frame for action is still the same order as for transients, and there is no need for emergency depressurization of the reactor. Analyses for line breaks outside the containment were also performed (see Subsection 6.3.4.7.7). m Concern 3 - Pressure boundary damage due to loadings from (1) bubble collapse in subccoled liquid and (2) injection of ECC water in stearn-filled pipes. 1A-28 Response to TMI Related Matters - Amendment 1
25A5113 Rev. A SBWR standardsafety Analysis Report n Response 3 - The BWR has no geometry equivalent to that identified in Michelson's report on B&W reactors relative to bubble collapse (steam bubbling upward through the pressurizer surge line and pressurizer). Thus, the first concern is not applicable to the SBWR design. Post-LOCA water makeup to the reactor is through ECC injection oflarge volumes of water via the GDCS. The GDCS short-term cooling subsptem provides flow from three separate water pools through six dedicated nozzles. The maximum GDCS poci and injection line water temperature is 57.2 C (135 F). The GDCS long-term cooling subsystem provides flow from the suppression pool through three dedicated nozzles. The maximum suppression pool and equalizing line water temperature is 100*C (212 F) at 379.2 kPa gauge (55 psig). The thermodynamic conditions in the GDCS lines during GDCS short-term and long-term cooling are not amenable to flashing of ECC water to steam. Furthermore, since ECC (GDCS) water injection occurs when the reactor vessel is being depressurized (ADS actuated, DPVs open), the system will not re-pressurize. The core also remains covered during the event. Therefore, the second concern is not applicable to the SBWR design. m m Concern 4 - In determining need for steam generators to remove decay heat, (U} consider that break flow enthalpy is not core exit enthalpy. ; Response 4 - BWRs do not use steam generators to remove decay heat, so this concern does not apply to the SBWR design. m Concem 5 - Are sources of auxiliary feedwater adequate in the event of a delay in cooldown subsequent to a small LOCA? Response 5 - BWRs do not need feedwater to remove heat from the reactor following a LOCA, whether the subsequent cooldown is delayed or not. Therefore, this concern is not applicable to the SBWR design. The Isolation Condenser System (ICS) provides cooling of the reactor when the l reactor coolant pressure boundary becomes isolated following a scram during power operations. The ICS automatically removes residual and decay heat to limit l reactor pressure within safety limits when the reactor isolation occurs (see l l Subsection 5.4.6). a Concern 6 -Is the recirculation mode of operation of the HPCI pumps at high pressure an established design requirement?
,- Response 6 -The SBWR ICS replaces the traditional HPCI and RCIC systems found in most BWRs for high pressure inventory control. The ICS does not rely on (S'} active pumps to remove excess sensible and core decay heat. Therefore, this concern is not applicable to the SBWR Standard Plant design.
Response to TMI Related Matters - Amendment 1 1A-29 l i
l l
- 25A5113Riv. A l
I SBWR stamtus sarery Analysis aeport l > 0 m Concern 7 - Are the IIPCI pumps and R11R pumps run simultaneously? Do they share common piping / suction? If so, is the system properly designed to accommodate this mode of operation (i.e., are any NPSII requirements violated, l etc.)? l Response 7 - The SBWR does not have the traditional Residual lleat Removal (R11R) or the liigh Pressure Coolant injection (llPCI) Systems. Reactor heat l removal is provided through the ICS and ECCS (ADS plus GDCS), as described in Subsection 5.4.6 and Section 6.3, respectively. l The ICS consists of three totally independent loops, each containing a steam isolation condenser, connected to the reactor pressure vessel. The GDCS is j mechanically separated into three identical divisions, each taking suction from independent GDCS pools and the suppression pool. The ADS consists of eight safety / relief valves (SRVs) and six depressurization valves (DPVs). The SRVs discharge steam from the main steamlines to quenchers in the suppression pool. Common piping / suction are not shared between the ICS and the ECCS. l m Concern 8- Mechanical effects of slug flow on steam generator tubes needs to be addressed (transitioning from solid natural circulation to reflux boiling and back to solid natural circulation may cause slug flow in the hot leg pipes). Response 8 - BWRs do not have steam generators so this concern does not apply to the SBWR Standard Plant design. m Concern 9 -Is there minimum flow protection for the IIPCI pumps during the recirculation mode of operation? Response 9 -The SBWR ICS replaces the traditional HPCI and RCIC Systems found in most BWRs for high pressure inventory control. The ICS does not rely on active pumps to remove excess sensible and core decay heat. Therefore, this l concern is not applicable to the SBWR Standard Plant design, a Concern 10 -The effect of the accumulators dumping during small break LOCAs is not taken into account. Response 10 - BWRs do not use accumulators to mitigate LOCAs. Therefore, this concern does not apply to the SBWR Standard Plant design. m Concern 11 -What is the impact of continued running of the RC pumps during a small LOCA? Response 11 -There is no forced recirculation system for the SBWR. The SBWR relies on buoyancy forces within the reactor vessel to cause recirculation of reactor 1A 30 Response :o TMI Related Matters - Amendment 1
25A5113Rtv. A SBWR standardsarety Analysis separt
,x
[s \ v/ coolant through the core, without reliance on active pumps. Therefore, this concern does not apply to the SBWR design. m Concern 12 - During a small break LOCA in which offsite power is lost, the possibility and impact of pump seal damage and leakage has not been evaluated. Response 12 - There is no forced recirculation system for the SBWR. The SBWR : relies on buoyancy forces within the reactor vessel to cause recirculation of reactor coolant through the core, without reliance on active pumps. Therefore, this concern does not apply to the SBWR design. , l a Concern 13 - During transitioning from solid natural circulation to reflux boiling l and back again, the vessel level will be unknown to the op'erators, and emergency I procedures and operator training may be inadequate. This needs to be addressed and evaluated. Response 13 -There is no similar transition in the BWR case. In addition, the BWR has water level measurement (via differential pressure devices) within the vessel and the indication of the water level is incorporated into the operator guidelines. f'N Consequently, this concern does not apply to the SBWR. t 4
\.._/ m Concern 14 - The effect of non-condensable gas accumulation in the steam generators and its possible disruption of decay heat removal by natural circulation needs to be addressed. l Response 14 - Although BWRs do not have steam generators, the concern is valid and applicable to BWRs in general. The reduction or loss of decay heat removal due to the accumulation of non-condensable gases is provided in the SBWR design by features of the isolation Condenser System (ICS). Vent lines are provided for each
! upper and lower headers to remove the non-condensable gases away from the IC unit during the IC operation period. A purge line is provided to assure that, during normal plant operation (ICS standby condition), non-condensable gases, hydrogen, l and air will not accumulate in the IC steam supply line. This assures that the IC tubes will not be blanketed with non-condensables when the system is first started. A-cat:St cc > crier !: predded te recembine nen conden:2b!c gace: (hydrogen and eggen) under normal p!:nt opectien (!Cc standby condition). Concern 15 - Delayed cooldown following a small break LOCA could raise the t I containment pressure and activate the containment spray system. f Response 15 - The non-safety-related containment spray is a manual function of the Fuel and Auxillag Pool Cooling System (FAPCS), and provides cooled dgwell f^}
) spray, using water removed from the suppression pool. The system cannot be activated automatically. The FAPCS is a non-safety-related system, except for its i
l Response to TMI Related Matters - Amendment 1 1A-31 i
I 2SA5113 Rw. A l SBWR stamtantsafety Analysis Report 1 0 provision for emergency makeup water to the spent fuel pool and IC/PCC pool A l detailed description of the FAPCS can be found in Subsection 9.1.3. The FAPCS operation is remote manually controlled and monitored from the main control room. l l u Concern 16* - This concern relates to the possibility that an operator may isolate, where possible, a pipe break LOCA without realizing that it might be an unsafe l l action leading to high pressure, and short-term core bakeout. For example, if a BWR should experience a LOCA from a pressure boundary failure somewhere between the pump suction and discharge valve for either reactor recirculation pump, it would be possible for the operator to close these valves following the reactor blowdown to repressurize the reactor coolant system. Before such isolation should be permitted, it i- first necessary to show by an appropriate analysis that the high pressure ECCS is adequate to reflood the uncovered core without assistance from the low pressure ECCS which can no longer deliver flow because of the repressurizaus Otherwise, such isolation action should be explicitly forbidden in the emergmcy operating instructions. Response 16 -The SBWR relies on natural circulation within the RPV, and thus, does not have recirculation lines. If a pipe break LOCA occurs in another line, the ! ECCS described in Section 6.3 is designed to provide protection against postulated LOCAs caused by ruptures in primary system piping. ECCS initiation is automatic, and no operator action is required until 72 hours after an accident. Inadvertent operator-induced vessel overpressure is not a real concern. The ADS detects reactor low water level (Level 1) and automatically actuates the SRVs and DPVs sequen tially in groups after Level 1 is reached for a prescribed time. A two-out-of-four Level 1 logic is used. The ADS may also be manually initiated from the main control room. The Dl'Vs are straight-throt gh, squib-actuated, non-reclosing valves. The GDCS is designed to respond to a need for emergency core cooling, following reactor depressurization, regardless of the physicallocation of the malfunction or i break that causes the need. The GDCS is completely automatic in operation, and manual initiation is possible at any time if automatic initiation fails, provided protective interlocks have been satisfied (i.e., the reactor is depressurized), i l l l Ol f
- Memo, C. Michelson to D. Okrent, Possible Incorrect Operator Action Such as Pipe Break Isolation, June 4,1979.
1A-32 Response to TMI Related Matters - Amendment 1 l
25A5113Rev. A SBWR standardsaretyAnarr sisneport m
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1 A.2.34 Primary Coolant Sources Outside Containment Structure [Ill.D.1.1(1)] NRC Position Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical levels. This program shall include the following: a immediate leak reduction I
- Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment.
- Measure actual leakage rates with systems in operation and report them to the NRC.
a Continuing Leak Reduction - establish and implement a program of prevenuve i maintenance to reduce leakage to as-low-as-practical levels. This progra'n shall i include periodic integrated leak tests at intervals not to exceed each i efueling cycle. i l
Response
Subsection 6.2.S describes the testing program for determining the containment integrated leakage rate (Type A tests), containment penetration leakage rates (Type B tests), and containment isolation valve leakage rates (Type C tests) that complies with AppendixJ and General Design Criteria 52,53, and 54 ofAppendix A of 10CFR50. Type A, B, and C tests are performed prior to operations and periodically thereafter to assure that leakage rates through the containment and through systems or components that penetrate the containment do not exceed maximum allowable rates specified in the plant Technical Specifications described in Chapter 16. i l The periodic leakage rate test schedule requirements for Types A, B, and C tests are ! specified in Chapter 16. Type B and C tests are conducted prior to initial criticality and j periodically thereafter during shutdown periods or normal plant operations, as long as , the time interval between tests for any individual Type B or C tests does not exceed 30 months. Type A, B, and C test results are submitted to the NRC in the summary report q i approximately three months after each test, and will include descriptions of the containment inspection method, any repairs necessary to meet the acceptance criteria, and the test results. Response to iMI Related Matters - Amendment 1 1A-33
25AS113Rsv. A SBWR standantsafety Analysis Report O 1 A.2.35 in-Plant Radiation Monitoring III.D.3.3(3)] NRC Position a Each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an accident. m Each applicant for a fuel-loading license to be issued prior toJanuary 1,1981 shall provide the equipment, training, and procedures necessary to accurately determine the presence of airborne radioiodine in areas within the plant where plant personnel may be present during an accident.
Response
a This is a COL license information requirement described in Subsection I A.3.3. The SBWR Standard Plant prmides tht ee systems to monitor area radiation and airborne radioactivity the Alca Radiadon Monitoring (ARht) System, Containment Atmospheric Stonitoring (CAM) System, and Process Radiation Monitoring (PRM) System. Detailed descriptions of each system is described in Subsection 12.3.4. s Not applicable. 1 A.2.36 Control Room Habitability [Ill.D.3.4(1)] NRC Position In accordance with Task Action Plan item Ill.D.S.4 and control room habitability, licensees shall assure that control room operators will be adequately protected against the effects of accidental release of toxic and radioactive gases and that the nuclear power plant can be safely operated or shut down under design basis accident conditions (Criterion 19, " Control Room" of Appendix A, General Design Criteria for Nuclear Power Plants, to 10CFR50).
Response
The control room design meets the requirements of Criterion 19. An evaluation against Criterion 19 can be found in Subsection 3.1.2.10, and is summarized below. Safe occupancy of the control room during abnormal conditions is provided for in the design. Adequate shielding is provided to maintain tolerable radiation levels in the control room in the event of a design basis accident for the duration of the accident. The control room ventilation system has redundant equipment and includes radiation, toxic and smoke detectors with appropriate alarms and interlocks. If any hazards exist at the normal control room ventilation intake, habitability is assured by the Emergency 1A-34 Response to TMI Related Matters - Amendment 1
25A5113Rev. A SBWR stamtardsafety Analysis Report <V Breathing Air System (EBAS), which upon isolation of the control room envelope, provides a positive air purge. , l In the unlikely event that the control room must be vacated and access is restricted, l instmmentation and controls are provided outside the control room which can be j utilized to initiate reactor shutdown, maintain a safe shutdown condition and achieve subsequent cold shutdown of the reactor. f 1A.3 COL License Information 1 A.3.1 Emergency Procedures and Emergency Procedures Training Program Emergency procedures, developed from the emergency procedures guidelines, shall be i provided and implemented prior to fuel loading (see Subsection 1A.2.1). 1 A.3.2 Review and Modify Procedures for Removing Safety-Related Systems From Service ; i Administrative control procedures governing the removal and restoration of safety- j related equipment from service and the verification of system lineups upon restoration shall be provided by the COL applicant. Procedures shall be reviewed and modified (as
- [ required) for removing safety-related systems from service (and restoring to senice) to l\ assure operability status is known (see Subsections 1 A.2.18 and 1 A.2.19).
1 A.3.3 In Plant Radiation Monitoring Equipment and training procedures shall be provided for accurately determining the ! airborne iodine concentration in areas within the facilitywhere plant personnel may be present during the accident (see Subsection IA.2.35). 1 A.3.4 Reporting Failures of Reactor System Relief Valves Failures of reactor system relief valves shall be reported in the annual report to the NRC (see Subsection 1A.2.21.1). 1 A.3.5 Report on ECCS Outages Starting from the date of commercial operations, an annual report will be submitted which includes instance of emergency core cooling system unavailability because of component failure, maintenance outage (both forced or planned), or testing, the following information shall be collected: I e Outage date a Duration of outage p) e Cause of outage 1A-35 Response to TM1 Related Matters - Amendment 1 f l
2SAS113R1v. A SBWR standantsarety Anarrsis neport O s Emergency core cooling sptem or component involved a Corrective action taken The above informadon shall be assembled into a report, which will also include a discussion of any changes, proposed or implemented, deemed appropriate, to improve the availability of the emergency core cooling equipment. 1A.4 References lA-1 U. S. Nuclear Regulatory Commission, NRC Action Plan Developed as a Result of the TMI-2 Accident, USNLC report NUREG-0660, Vols. I and 2, May,1980. lA-2 U. S. Nuclear Regulatory Commission, Clarificadon of TMI Action Plan Requirements, USNRC Report NUREG 0737, November,1980. lA-3 U. S. Nuclear Regulatory Commission, Licensing Requirements for Pending Applicadons for Constmcdon Permits and Manufacturing License, NUREG-0718, Revision 1, June,1981. lA4 Letter from R. II. Buchholz, GE, to D. F. Ross, NRC,
Subject:
Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors, November 30,1979, MFN-290-79. lA-5 Letter from D. B. Waters, Chairman, BWR Owners' Group, to NRC, dated December 29,1980, BWR Owners' Group Evaluation of NUREG-0737 Requirements. lA-6 Letter from D. B. Waters, Chairman, BWR Owners' Group, to D. G. Eisenhut, NRC, dated March 31,1981, BWR Owners' Group Evaluation of NUREG-0737 Requirements ll.K.3.16 and II.K.3.18. lA-7 Letter from D.B. Waters, Chairman, BWR Owners' Group, to R.it Vollmer, NRC, dated September 17,1980, NUREG-0578 Requirements 2.12-Performance Testing of BWR and PWR Relief and Safety Valves. IA-8 NEDE-24988-P, Analysis of Generic BWR Safety / Relief Valve OperabilityTest Results, Proprietary Document, October 1981. l i 1 A-9 Licensing Topical Report, TRACG Model Description, NEDE-32176P, Class til (Proprietary), February 1993. l 1 A-10 Licensing Topical Report, TRACG Qualification, NEDE-32177P, Class III (Proprietary), February 1993. l 1A-36 Response to TMI Refoted Matters - Amendment 1 l
25AS113Rev. A SBWR standardsafety Analysis Report O 1 A-11 Licensing Topical Report, Application of TRACG Model to SBWR Licensing Safety Analysis, NEDE-32178P, Class Ill (Proprietary), February 1993. O ct Response to TM) Related Matters - Amendment 1 1A-37/38
1 l 25A5113Rev. A ; SBWR stemtentsaeryAntysisaerat 1 p Chapter 1 Table of Contents
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t l Table of Contents..... .. ..........................................................................1.0-i List of Tables.... . ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ... .. . . . . . . . .. . . . . . . . . . . . . . . .. 1. 0-iii List of Figu res .... .... .... .... ... .... . ..... .... . . . .. ...............................................1.0-v ; 1.0 Introduction and General Descdption of Plant... .. ... . . . . . . .. ...... .... ...... ..... 1.1-1 1.1 I n trod u ction .. . ... . . .. ... . .. .. .. . . .. .. .. ... . . .. . . . . . . ... .. . . . . . . . . . . . . . . . . . . . . ............ 1.1-1 1.1.1 COL License Information... . ....... .. .. ... ........ .... ....................... .... 1.1-2 ' 1.1.2 Refe re n c es .. . . .. . . . .. . . . . . . . . . .. . . . . . . . . . . .. .. . . . . . . . . . . . . . . .. .....-.......................1.1-2 1.2 General Plan t Descriptio n . . .. .... . . ........ ............ .... ..... . . .... . . . . . . . . . . . . . . . . . . . . ..... 1.2-1 1.2.1 Principal Design Criteria..... . . . . .. .. ... ......... .. ... ..... ..................... 1.2-1 1.2.2 Plant Description. . .....................................................1.2 1.2.3 COL License Information... ... .. . . ... .......... ................. .... . . . . ..... .... . .. 1.2-9 8 1.2.4 References.... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... .... . 1.2-98 1.3 Co mparis o n Tabl es . ... ... . . .... . . . . .. . . . .. . . . ... ... .. . .. .. . . . .. . .. . ... .. . . ... .... ..._ . ..... ... 1.3-1 1.3.1 COL License Information........... ... . .. .. .. . . . . .. .... ...................... 1.3-1 1.3.2 References.. .. . ... ...... .. . . . . . . . . . . . . . . . . . . . . . . . . ...........................l.S-1 O 1.4 Identification of Agents and Contractors.. ... ........... .. . . . .. ............ .. .......... 1.4-1 1.4.1 COL License Information.. .. . ..... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1.4-1 1.4.2 References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... .. 1.4-1 1.5 Requirements for Further Technical Information . . .. . ..... .. .. ......... . ... .. .. .... ..... 1.5-1 l 1.5.1 Evolutionary Design . . . .. ....... . .... .. . ............. ... .. ....................... 1.5-1 1.5.2 Analysis and Design Tools.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . 1. 5-1 1.5.3 Testing.. . . . . . . . . . . . . . . . . . . ......................................................1.5-5 1.6 Material Incorporated by Referen ce.. .. . . .... ... ...... ............... . .................. ......... .. . ..... 1.6-1 1.6.1 COL License Information.. . ...... . . . . ... .. ...... .. . . .. .. . . . . . . . . .. . . .. . . . .. . 1. 6- 1 1.6.2 References.... .. ... .. . . . . . . . . . . . . . . . . . . . ................................1.6-1 1.7 Engineering Drawings.. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...........................1.7-1 1.7.1 COL License Inform ation ...... .. . ...... . . ....... ... ....... .... ...... . . .. .. .. .... ........... 1.7-1 , 1.7.2 References........ . . .. . . . . . . . . . ....................................................1.7-1 1.8 Sum mary of COL License Information .. .. . . . ......... ... ....... ..... ... . ..... . .... ........ ....... 1.8-1 1.8.1 COL Applicant Action Items . .. . .................. ...... ............... . .... 1.8-1 1.8.2 Bounding Site Parameter Requirements ........ ...... . . .. ... .. ... ........................ 1.8-16 1.9 Conformance with Standard Review Plan and Applicability of Codes and Standards .... 1.9-1 ! 1.9.1 Conformance with Standard Review Plan..... ..... . . . . . . . . ..........................1.9-1 1.9.2 Applicability of Codes and Standards . . . .. . . . . . . . . . . . . . . . . . . . ....... .. . 1.9-7 ^ 1.9.3 Applicability of Experience Information . . . . . . . ... ...... .............. .. .... 1.9-7 ( 1.9.4 COL License Information.. . .. . . . . . . . . . . . . . . . . . . . . . . . . . .... . ..... .. 1.9-8 1.9.5 Refere nces..... ... ... ... . ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. .... . . . 1.9-8 Table of Contents - Amendment 1 1.0-1
25A5113Rsv. A SBWR StandardSafety Analysis Report Table of Contents (Continued) 1A Response to TMI Related Matters.. . . . . . . . . . . . .. . . .. .. . . . . . . .. .. l A-1 1 A.1 Introduction.. .. . . . . . .. . ... . . . .. . . . . . . . . . . . . . . . . . .. . . ... . 1 A-1 1A.2 NRC Positions / Responses... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . l A-1 1A.3 COL License Information.... . . . . . . . . ... ...... . ... . .. . . . .. . . l A-35 1A.4 References.. . . . . .. . . . . . . .. .. . . . . . . . . . . . . l A-36 IB Failure Modes and EfTects Analysis. . . . . . . . . . . . . . .. . .. .. . . . . . . . . . . . . I B-1 IB.1 Application of the Probabilistic Risk Assessment to FMEA . . . .. .. .. .. .1 B-1 1 B.2 DPV Failure Mode and Effects Analysis .. . . . . . . . . . . . . .. . lik1 1C Conformance Assessment of the SBWR Design with ALWR URD... . . . . . . . . . .101 1C.1 SBWR Conformance Assessment Program. . . . . . . . . . . . . . .. . . 1 C-1 l t O l l l l 1.0-il Table of Contents - Amendment 1
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l l I 2SAS113Rw. A SBWR standedsdory Anarrsis neput i ~p Chapter 1 N List of Tables Table 1.S1 Comparison Of Reactor System Design Characteristics.. . . . . ... .. . .. .. . .....1.3-2 )
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Table 1.S2 Comparison Of Emergency Core Cooling Systems and Safety-Related Containment Cooling Systems.. . .. .. ... .. .. ... .. . ... . .. .1.S7 i Table 1.S3 Comparison Of Containment Design Characteristics... .... ..... . ... . ....... . . 1.S9 l l Table 1.S4 Comparison Of Structural Design Characteristics .. . ..... . . . . . .. . .. ... .....1.S11 l l. l Table 1.4-1 Commercial Nuclear Reactors Completed and Under Construction By General Electric... . .. . . . . . . . . . . . . ..... . .. . . . . .. .. .~ .. 1.4-2 Table 1.5-1 Evolution of the General Electric BWR . . . . . . . . . . . . . . . . . . . . . . . . ... . . 1.5-11 . l Table 1.5-2 SBWR Features and Related Experience....... . .. .. . . . . . . . ... ... ... . ... ... . 1.5-12 Table 1.5-Sa . . . . . . . . 1.5-13 TRACG Qualification (Section 1) .. .... .... . . . . . . . . . . . . . . .
. .. . .. . . . . .. . . . 1.5-14 Table 1.5-Sb TRACG Qualification (Section 2) ... .... . . . ...... . . . . . . . . . .
Table 1.5-Sc .....................-...... ... . ... 1.5-31 TRACG Qualification (Section 3) ...
. 1.6-2 O Table 1.61 Referenced Reports.... . . . ..... .. . ......................................... i l
Table 1.7-1 Piping and Instrumentation, Air Flow, and Process Flow Diagrams. .......... .. .1.7-2 Table 1.7-2 Instrument Electrical Diagrams and Logic Diagrams.. .. ... . . .. .... . . ... 1.7-6 j Table 1.7-3 Building Arrangement Drawings... . . ... . .. . . . . . . . . ...... ....... 1.7-8 Table 1.7-4 Radiation Zone Drawings.. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . .. ... . 1. 7-10 i Table 1.7-5 Miscellaneous Drawings. .. . . . . . . . .. . . . . . . . . . . . . . . . . . . ... .. . . . .. . 1.7-12 Table 1.7-6 Structural Design Drawings . . .. .. ........ . . . . . . . . . . . . . . .. . . ... . ........ . 1. 7-13 Table 1.7-7 Fire Protection Drawings .. . . . . . . . . . . . . . . . . . . . . . . . . . ... .. ... . . . . .. 1.7-16 1 Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR. .... ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 1.9-9 1 Table 1.9-2 NRC Regulatory Guides Applicability to SBWR ......... . . .... . . . . .. .. .. .. . . .. 1.9-24 i ! Table 1.9-3 Industrial Codes and Standards Applicable to SBWR.... ..... .. .... ......... . .. .... 1.9-36 I Table 1.9-4 Experience Information Applicable to SBWR.. .. ....... .. . .. . . ..... . . .. .. .. . . 1.9-4 6 Table IB-1 DPV FMEA .. . .. . . . . . . . . .... . . . . . . . . . . . . . . . .. ... . .. ... 1 B-3 I . . . . . . . . . . . . . . . . . 1 C-3 Table 10-1 RWCU/SDC Differential Flow Detection . . ... . . . . . . .
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1.0-lil List of Tables - Amendment 1
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1 4 4 25A5113Rev. A , 1 SBWR stansantsareryAnalysis neport l List of Tables (Continued) Table IG2 Valve Position Indication . . . . . . . . . . . . . . .... ... . . . . . . . . . . . .......... . .l G 5 Table 1G3 Location of RWCU/SDC System.. . . ... . ... . . . . . . . . . .. .. ..................l G 6 Table ICA Impact Forces on Spent Fuel Storage Racks.. . . .. . . . . . . . .. ..1G7 O l 1.0-iv List of Tables - Amendment 1
1 2SAS113Rev. A SBWR sonderdsafetyAnrysis neport i O Chapter 1 U List of Figures i Figure 1.1-1 Containment Nomenclature . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , ,1,1 3 i Figure 1.1-2a Reactor System Heat Balance at 100% Power (SI Units)..... . .. . ........... 1.1-4 l ' 1 Figure 1.1-2b Reactor System Heat Balance at 100% Power (English Units) ....... .. -. ... 1.1-5 Figure 1.5-1 Evolu tio n of th e GE BWR . .. .. . ... .... ... .... . . ... .. . ..... ..... .. ......... ...... .... ... ... ... .. 1.5-50 , Figure 1.52 Evolution of the BWR Reactor Design. .... ... .... . . . . . . . . . . ... .... .. .. . .. 1.5-51 , Figure 1.5-3 Comparison of BWR Containments... ............. .. ..... . . . . . . . . . . . . ...... .... 1.5-52 ; l 1 l i l l l l l l O g,g.yyy List of Figures - Amendment 1
25A5113Rev. A SBWR senadsrdsurery Analysis naporr
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4 l 1.2 General Plant Description l 1.2.1 Principal Design Criteria The principal design criteria governing the SBWR Standard Plant are presented in two ways. First, the criteria are classified as applicable to either a power generation function t or a safety-related funcdon. Second, they are grouped according to system. Although the distinctions between power generadon or safety functions are not always clear-cut I and are somedmes overlapping, the functional classificadon facilitates safety analysis ' ! reviews, while the grouping by system facilitates understanding both the system function and design. The principal plant structures are shown on Figure 21.1.2-2, sheets 1-20, and are listed j below:
- a Reactor building - houses all structures, components, equipment and systems providing safety-related funcdons. This includes the reactor, containment, safety i
envelope, the refueling area with spent fuel storage, the control room, and auxiliary
- equipment area.
y p a Turbine building- houses equipment associated with the main turbine and 4 generator and their auxiliary systems and equipment including the condensate
- purification system and the process offgas treatment system.
I a Radwaste building - houses equipment associated with the collection and i processing of solid and liquid radioactive waste generated by the plant. a Electrical building - houses the two non-safety-related standby diesel generators f j and their associated auxiliary equipment, and the solid-state adjustable speed drive j units powering the feedwater pump motors and others powering the Reactor Water Cleanup / Shutdown Cooling System pumps. l ] 12.1.1 General Power Generation Design Criteria $ a The plant is designed to produce ciectricity from a turbine generator unit using steam generated in the reactor, s
- m Heat removal systems are provided with sufficient capacity and operational adequacy to remove heat generated in the reactor core for the full range of normal operational conditions and abnormal operational transients.
i a Backup heat removal systems are provided to remove decay heat generated in the core under circumstances wherein the normal operational heat removal systems [O Q become inoperative. The capacity of such systems is adequate to prevent fuel cladding damage. 1.2-1 General Plant Description - Amendment 1 1
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2SAS113Rw. A SBWR studardsafety Antysis Repat O e The fuel cladding,in conjunction with other plant systems, is designed to retain integrity so that the consequences of any failures are within acceptable limits throughout the range of normal operational conditions and abnormal operational transients for the design life of the fuel. m Control equipment is provided to allow the reactor to respond automatically to load changes and abnormal operational transients. m Reactor power level is manually controllable. m Control of the reactor is provided from a single location, a Reactor controls, including status displays and alarms, are arranged to allow the operator to rapidly assess the condition of the reactor system and locate system malfunctions. m Interlocks or other automatic equipment are provided as backup to procedural control to avoid conditions requiring the functioning of safety-related systems or engineered safety features. m The station is designed for routine continuous operation whereby steam activation products, fission products, activated corrosion products and coolant dissociation products are processed to remain within acceptable limits. 1.2.1.2 General Safety Design Criteria a The station design conforms to applicable codes and standards as described in Subsection 1.8.2. m The station is designed, fabricated, erected and operated in such a way that the release of radioactive material to the emironment does not exceed the limits and guideline values of applicable government regulations pertaining to the release of ; i radioactive materials for normal operations, for abnormal transients and for accidents.
)
a The reactor core is designed so its nuclear characteristics do not contribute to a divergent power transient. e The reactor is designed so there is no tendency for divergent oscillation of any ' operating characteristic considering the interaction of the reactor with other appropriate plant systems. e The design provides means by which plant operators are alerted when limits on the release of radioactive material are approached. General Plant Description - Amendment 1 1.2-2
25AS113Rev. A SBWR standardsarety Analysis neport
,m V)
I 2 m Suflicient indications are provided to allow determination that the reactor is operating within the envelope of conditions considered safe by plant analysis. a Those portions of the nuclear system that form part of the reactor coolant pressure
' boundary are designed to retain integrity as a radioactive material containment barrier following abnormal operational transients and to assure cooling of the reactor core following accidents.
m Safety-related systems and engineered safety features function to assure that no damage to the reactor coolant pressure boundary results from internal pressures caused by abnormal operational transients and accidents. m Where positive, precise action is immediately required in response to abnormal operational transients and accidents, such action is automatic and requires no decision or manipulation of controls by plant operations personnel. s Safety-related actions are provided by equipment of sufficient redundancy and independence so that no single failure of active components, or of passive components in certain cases in the long term, will prevent the required actions. For [] systems or components to which IEEE-279 apply, single failures of either active or
) passive electrical components are considered in recognition of the higher anticipated failure rates of passive elecuical components relative to passive mechanical components.
m Provisions are made for control of active components of safety-related systems from the control room. m Safety-related systems are designed to permit demonstration of their functional performance requirements. s The design of safety-related systems, components and structures includes allowances for natural environmental disturbances such as earthquakes, floods, and storms at the station site. s Standby electrical power sources have sufIicient capacity to power all safety-related systems requiring electrical power concurrently. m Standby electrical power sources are provided to allow prompt reactor shutdowTi and removal of decay heat under circumstances where normal auxiliary power is not available. a A containment is provided that completely encloses the reactor systems, drywell,
) and suppression chambers. The containment employs the pressure suppression (d
concept. 1.2-3 General Plant Description - Amendment 1
i 25AS113Riv. A \ \ l SBWR standardsareryAnarysis neport G; a It is possible to test containment integrity and leak tightness at periodic intervals. m A safety envelope is provided that basically encloses the containment, with the exception of the areas above the containment top slab and drywell head. The areas above the containment top slab and drywell head are flooded in a pool of water during operation. The safety envelope forms an additional barrier helping to control any potential post-accident containment leakage. The water pool above the containment top slab and drywell head is effecdve in scrubbing any potential l containment leakages through that path. m The containment and safety envelope in conjunction with other safety-related features limit radiological effects of design basis accidents to less than the l prescribed acceptable lindts. l ! a Provisions are made for removing energy from the containment as necessary to maintain the integrity of the containment system following accidents that release energy to the containment. l m Piping that penetrates the containment and could serve as a path for the ! uncontrolled release of radioactive material to the emirons is automadcally isolated when necessary to limit the radiological impact from an uncontrolled release to less than acceptable limits. m Emergency core cooling is provided to limit fuel cladding temperature to less than the limits of 10CFR50.46 in the event of a design basis loss-of-coolant accident (LOCA). e The emergency core cooling provides for continuity of core cooling over the complete range of postulated break sizes in the reactor coolant pressure boundary piping. m Emergency core cooling is initiated automatically when required regardless of the availability of off-site power supplies and the normal generadng system of the station. m The control room is shielded against radiation so that condnued occupancy under design basis accident conditions is possible. m In the event that the control room becomes inaccessible,it is possible to bring the reactor from power range operation to cold shutdown conditions by utilizing i alternative controls and equipment that are available outside the control room. m Backup reactor shutdown capability independent of normal reactivity control is provided. This backup system has the capability to shut down the reactor from any ! normal operating condition and subsequently to maintain the shutdown condition. General Plant Description - Amendment 1 j 1.2-4
b 25A5113Rev. A SBWR studerdsaretyAnalysis neport i _ 1
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m Fuel handling and storage facilities are designed to prevent inadvertent criticality and to maintain shielding and cooling of spent fuel as necessary to meet operating and off-site dose constraints. j u Systems that have redundant or backup safety-related functions are physically separated, and arranged so that credible events causing damage to one region of the i reactor island complex have minimum prospects for compromising the functional ! capability of the redundant system. , i s 1.2.1.3 Nuclear System Criteria 4 , a The fuel cladding is a radioactive material barder designed to retain integrity so that 2
, failures do not result in dose consequences that exceed acceptable limits through the design power range.
m The fuel cladding in conjunction with other plant systems is designed to retain ; l integrity so that the consequences of any failures are within acceptable limits throughout any abnormal operational transient. [ < n a Those portions of the nuclear system that form part of the reactor coolant pressure boundary are designed to retain integrity as a radioactive material barrier during normal operation and following abnormal operational transients and retain integrity to assure core cooling following accidents. 4 t a The capacity of the heat removal systems provided to remove heat generated in the reactor core for the full range of normal operational transients as well as for abnormal operational transients is adequate to prevent fuel cladding damage that results in dose consequences exceeding acceptable limits. m The reactor is capable of being shut down automatically in sufficient time to permit decay heat sinks to become effective following loss of operation of normal heat removal systems. The capacity of such systems is adequate to prevent fuel cladding damage. s The reactor core and reactivity control system are designed such that control rod action is capable of making the core subcritical and maintaining it even with the rod I of highest reactivity worth fully withdrawn and unavailable for insertion. a Backup reactor shutdown capability is provided independent of normal reaci.ivity control provisions. This backup system has the capability to shut down the reactor from any operating condition and subsequently to maintain the shutdown condition. 1.2-6 General Plant Description - Amendment 1
l ' l 25AS113Rw. A l l SBWR stamiant safety Analysis neport l O' s The nuclear system is designed so there is no tendency for divergent oscillation of l l any operating characteristic, considering the interaction of the nuclear system with other appropriate plant systems. 1.2.1.4 Electrical Power Systems Criteria Sufficient normal, auxiliary, and standby sources of electrical power are prmided to l attain prompt shutdown and continued maintenance of the station in a safe condition l under all credible circumstances. The power sources are adequate to accomplish all required safety-related functions under all postulated accident conditions. 1.2.1.5 Auxiliary Systems Criteria e Other auxiliary systems, such as senice water, cooling water, fire protection, heating and ventilating, communications, and lighting, are designed to function as needed j during normal and/or accident conditions. ! e Auxiliary systems that are not required to effect safe shutdown of the reactor or maintain it in a safe condition are designed so that a failure of these systems shall not prevent the safetprelated systems from performing their design functions. 1.2.1.6 Shielding and Access Control Criteria Radiation shielding is provided and access control patterns are established to allow a properly trained operating staff to control radiation doses within the limits of applicable regulations in any normal mode of plant operation. 1.2.1.7 Power Conversion Systems Criteria Components of the power conversion systems are designed to attain the following basic objectives: a The components of the power conversion systems are designed to produce ! electrical power from the steam coming from the reactor, condense the steam into l water, and return the water to the reactor as heated feedwater with a major portion ofits gases and particulate impurities removed. m The components of the power conversion systems are designed so that any fission l products or radioactivity associated with the steam and condensate during normal operation are safely contained inside the system or are released under controlled conditions in accordance with waste disposal procedures. 1.2.1.8 Nuclear System Process Control Criteria a Control equipmentis provided to allow the reactor to respond automatically to load changes within design limits. General Plant Description - Amendment 1 1.24
i I 2SAS113Rev. A i ~ SBWR standardsarety Analysis neport l l\ a Manual control of the reactor power level is provided. ; I e Nuclear systems process displays, controls and alarms are arranged to allow the i operator to rapidly assess the condition of the nuclear system and to locate process } system malfunctions. 1.2.1.9 Electrical Power System Process Control Criteria @ l 1 The Class 1E power systems are designed with four divisions with any two divisions i m l being adequate to safely place the unit in the safe shutdown condition. a Protective relaying is used, in the event of equipment failure, to detect and isolate ' faulted equipment from the system with a minimum of disturbance to uninvolved i systems or equipment. m Two standby diesel generators are started and connected to both safety-related and non-safety-related loads if both the preferred and alternate power sources are lost. If these non{ lass IE DGs are also inoperable, all safety-related loads will be powered by the Class IE divisional batteries. [
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m Safety-related electrical systems and components are monitored in the control room. 1.2.2 Plant Description 1.2.2.1 Nuclear Steam Supply , 1.2.2.1.1 Reactor Pressure Vessel The reactor pressure vessel (RPV) assembly consists of the pressure vessel and its appurtenances, supports and insulation, and the reactor internals enclosed by the vessel (excluding the core, in-core nuclear instrumentation, neutron sources, control rods, and control rod drives). The reactor coolant pressure boundary (RCPB) of the RPV retains integrity as a radioactive material barrier during normal operation and follcwing abnormal operational transients and retains integrity to contain coolant during design basis accidents (DBAs). Certain RPV internals support the core and support instrumentation used during a DBA. Other RPV internals direct coolant flow, separate steam from the steam / water mixture leaving the core, hold material surveillance specimens, and support instrumentation used for normal operation. O The RPV, together with its internals, provides guidance and support for the fine-motion control rod drives (FMCRDs). Certain of the reactor internals distribute sodium 1.24 GeneralPlant Description - Amendment 1
i 1 25A5113Riv. A SBWR standardsafety Analysis Report O pentaborate solution delivered by the Standby Liquid Control (SLC) System when necessary to achieve core subcriticality via means other than inserting of control rods. The RPV restrains the FMCRDs to prevent ejection of a control rod connected with a drive in the event of a postulated failure of a drive housing. Reactor Pressure Vessel l The RPV consists of a vertical, cylindrical pressure vessel of welded construction, with a removable top head, and head flanges, seals and bolting. The vessel also includes l penetradons, nozzles, shroud support, and venturi shaped flow restrictors in the steam outlet nozzles. The shroud support carries the weight of peripheral fuel assemblies, neutron sources, core plate, top guide, chimney shroud and chimney head with steam i separators and dryers, and it laterally supports the fuel assemblies. An integral reactor I vessel skirt supports and anchors the vessel on the RPV support structure in the containment. The reactor vesselis 6 meters (236 in.) in diameter minimum, with a wall thickness of about 158 mm (6.2 in.) with cladding, and 24.5 m (80.4 ft) tall from the inside of the bottom head (elevation zero) to the inside of the top head. The bottom of the active fuellocation is 3750 mm (147.6 in.) from elevation zero and the active core is 2743 mm (108 in.) high. The overall RPV height of approximately 25 m (82 ft) permits natural circuladon driving forces to produce abundant core coolant flow. An increased internal flow-path length relative to prior BWRs is provided by a long " chimney" in the space which extends from the top of the core to the entrance to the steam separator assembly. The chimney and steam separator assembly are supported by a shroud assembly which extends to the top of the core. The large RPV volume provides a large reserve of water above the core, which translates directly into a much longer period of dme (compared to prior BWRs) before core uncovery can occur as a result of feedwater flow interruption or a LOCA.This gives an extended period of time during which automatic systems or plant operators can reestablish reactor inventory control using any of several normal, non-safety-related systems capable ofinjecting water into the reactor. Timely initiation of these systems precludes the need for activation of emergency safety equipment. The large RPV volume also reduces the reactor pressurization rates that develop when the reactor is suddenly isolated from the normal heat sink which eventually leads to actuation of the safety-relief valves. The FMCRDs are mounted into permanently attached CRD housings. The CRD housings extend through, and are welded to CRD penetrations (stub tubes) formed in the RPV bottom head. A flanged nozzle is provided in the top head for bolting on of the flange associated with the instrumentation for the initial vibration test ofinternals. GeneralPlant Description - Amendment 1 1.2-8
25A5113Rev. A SBWR studentsurnyAntysissepar i An integral reactor vessel skirt supports the vessel on the RPV support structure. Steel anchor bolts extend through a steel structure comprising the upper part of the RPV support structure, securing the flange of the skirt to the structure. Stabilizers help the upper portion of the RPV resist horizontal loads. Lateral support among the CRD , housings and in-core housings are provided by restraints which, at the periphery, are l 1 supported from CRD housing restraint beams. The RPV insulation is supported from the shield wall surrounding the vessel. Insulation for the upper head and flange is supported by a steel frame independent of the vessel and piping. Insulation access panels and insulation around penetrations are designed for ease ofinstallation and removal for vessel inservice Inspection and maintenance operadons. The RCPB portions of the RPV and appurtenances are classified as Quality Group A, l Seismic Category 1. The design, ruaterials, manufacturing, fabrication, testing, j examination, and inspection used in the construction meet the requirements of ASME Code, Section III, Subsection NB, Class 1 Components. The RPV support skirt, stabilizers, CRD housing restraints and in-core housing restraints are Seismic Category I l and are designed and constructed to meet the requirements ofASME Code, Section Ill, ) Subsection NF, Component Supports. The shroud support is classified as Seismic Category 1, and designed and fabricated to meet the requirements of ASME Code ; Class CS (core support structures). Hydrostatic tests of the RPV are performed in i accordance with the requirements for ASME Code Class 1 vessels. The components are l code-stamped according to their code class. 1 The RPV materials comply with the provisions of the ASME Code Section III, . ! Appendix 1, Subsection NB-2000, and meet the specification requirements of 10CFR50, Appendix G. The RPV is constructed primarily from low alloy, high strength steel plate and forgings. Plates are ordered to ASME SA-533,1YPE B, Class 1, and forgings to ASME SA-508, Class 3. These materials are melted to fine grain structure and are supplied in the quenched and tempered condition. Further restrictions include a requirement for vacuum ( pssing to lower the hydrogen level and improve the cleanliness of the low alloy steel Materials used in the core beltline region also specifylimits of 0.05% , maximum copper and 0.012% maximum phosphorous content in the base materials and a 0.08% maximum copper and 0.012% maximum phosphorous content in weld j materials. The maximum sulfur content for the base material and weld material is j 0.01%. j Studs, nuts, and washers for the head flange are ordered to ASME SA-540, Grade B23 or Grade B24 having minimum yield strength level of 893 MPa (129,500 psi). The s ) maximum measured ultimate tensile strength of the stud bolting materials do not exceed 1172 MPa (170,000 psi). General Plant Description - Amendment 1 I*EA 1
4 l l l 2SA51URev. A SBWR StandardSafety Analysis Repod i Electroslag welding is not applied for structural welds. Preheat and interpass temperature employed for welding oflow alloy steel meet or exceed the values given in ASME, Section Ill, Appendix D. Post-weld heat treatment at 593 C (1099 F) minimum and not exceeding 635'C (1175 F) is applied to all low-alloy steel welds. Welding l electrodes for low alloy steel are low hydrogen type ordered to ASME SFA-5.5. l Pressure boundary welds are given an ultrasonic examination in addition to the radiographic examination performed during fabrication. The ultrasonic examination method, including calibration, instrumentation, scanning sensitivity, and coverage, is ! based on the requirements imposed by ASME, Section XI, Appendix 1. Acceptance standards are equivalent or more restricdve than required by ASME, Section XI. A stainless steel weld overlay is applied to the intedor of the cylindrical shell and the t steam outlet nozzle. Other nozzles and the top head do not have cladding. The bottom head is clad with Ni-Cr-Fe alloy. Fracture toughness tests of pressure boundary ferritic materials, weld metal and heat-affected zone (HAZ) materials are performed in accordance with the requirements for ASME Code Class 1 vessels. Both longitudinal and transverse specimens are used to l l determine the minimum upper shelf energy (USE) level of the core beltline materials. Separate, unirradiated baseline specimens are used to determine the transition temperature cun'e of the core beltline base materials, weld metal, and HAZ materials. For the vessel material surveillance program, specimens are manufactured from the l ! material actually used in the reactor beltline region and welds typical of those in the beltline region, thus representing base metal, weld material, and the HAZ material.The plate and weld specimens are heat treated in a manner which simulates the actual heat treatment performed on the core region shell plates of the completed vessel. Each in-reactor smveillance capsule contains Charpy V-notch specimens of base metal, weld metal, and HAZ material, and tensile specimens from base metal and weld metal. Brackets are welded to the vessel cladding in the core belt region for retention of the detachable holders, each of which contains a number of the specimen capsules. Neutron dosimeters and temperature monitors are located within the capsules. l Access for examinations of the installed RPV is incorporated into the design of the vessel, reactor shield wall, and vessel insulation. j i Reactor Internals \ The reactor internals consist of core support structures and other equipment. The core support structures locate and support the fuel assemblies, form partitions within the reactor vessel to sustain pressure differentials across the partitions, and direct the flow of coolant water. The stmctures consists of a shroud, shroud support, core plate, top guide, and integral fuel support and control rod guide tubes (CRGTs). I General Plant Description - Amendment 1 1.2 10 , l l I
25AS113 Rex A SBWR standardsarety Analysis nepois V The other reactor internals consist of control rods, feedwater spargers, SLC distribution headers, in-core guide tubes, surveillance specimen holders, chimney, chimney partitions, chimney head, steam separator assembly, and the steam dryer assembly. The shroud suppon, shroud, and chimney make up a stainless steel cylindrical assembly that provides a partition to separate the upward flow of coolant through the core from the downward recirculation flow outside the core. This partition separates the core region from the downcomer annulus. The core plate consists of a circular stainless steel plate with round openings and is s:iffened with a beam structure. The core plate provides lateral suppon and guidance for the integral support and CRTGs, in< ore flux monitor guide tubes, peripheral fuel supports and startup neutron sources. The last two items are also supported venically by the core plate. The top guide consists of a circular plate with square openings for fuel and with a cylindrical side forming an upper shroud extension. Each opening provides lateral ! support and guidance for four fuel assemblies or,in the case of peripheral fuel,less than four fuel assemblics. Holes are provided in the bottom surface of the top guide where ' the sides of the openings intersect, to anchor the in-core instrumentation detectors and V start-up neutron sources. The fuel assemblies are vertically supported in two ways dependir,g upon whether they are located next to a control rod or not. The peripheral fuel assemblies which are located at the outer edge of the active core, not adjacent to a control rod, are supported by the peripheral fuel supports. The peripheral fuel supports are welded to the core plate and each suppon one assembly. The peripheral fuel supports contain flow restricting sections to provide coolant flow to the fuel assembly. The remaining fuel assemblies v;hich are adjacent to the control rods are supported by the integral fuel support a .id CRGTs. Each integral fuel suppon and CRGT supports four fuel assemblies vertically upward and provides lateral support to the bottom of the fuel. The fuel suppon fctms the top part of the integral unit with the bottom section forming the CRGT. The integral fuel support and guide tube are laterally supported by the core plate. The CRGT section is cruciform in shape and is designed as a guide for the lower end of the control rod. The lower end of the CRGT section is supported by the control rod drive (CRD) housing, which in turn transmits the weight of the integral fuel support and CRGT, and the four fuel assemblies to the reactor vessel bottom head. The lower end of the CRD housing is welded to a stub tube which is directly welded to the bottom of the vessel. Coolant flow which has entered the lower plenum of the vessel travels l g} t
'v upward, adjacent to the guide tube section and enters the fuel suppon sectionjust below the core plate. The fuel support section contains four flow restricting openings which control coolant flow to the fuel assemblies.
1 2-11 General Plant Description - Amendment 1 l
25A5113Riv. A stamtardsafetY Analysis Report SBWR O The base of the CRCT section is provided with a device for coupling to the FMCRD. The CRD is restrained from ejection,in the case of a stub tube to CRD housing weld failure, by the coupling of the drive with the guide tube base. In this event, the fuel support flange will contact the core plate and thus restrain the ejection. The coupling will also prevent ejection if the CRD housing fails below the stub tube weld. In this event, the integral guide tube and fuel support remains supported by the CRD housing left intact above the stub tube weld. The control rods are cniciform-shaped neutron absorbing members that can be insened or withdrawn from the core by the FMCRD to control reactivity and reactor power. Each of the four feedwater lines is connected to four spargers via four RPV nozzles. The feedwater spargers are stainless steel headers located in the mixing plenum above the downtomer annulus. Each sparger,in tv;o halves, with a tee connection at the middle, is fitted to the corresponding RPV feedwater nozzle. The sparger tee inlet is connected to the RPV nor.zle safe end by a double thermal sleeve arrangement. Feedwater flow enters the center of the spargers and is discharged radially inward to mix the cooler feedwater wita the downcomer flow from the steam separators and steam dryers. In-core guide tubes (ICGTs) protect the in-core flux monitoring instrumentation from flow of water in the bottom head plenum.The ICGTs extend from the top of the in-core housing to the top of the core plate. The local power range monitoring (LPRM) detectors for the Power Range Neutron Monitoring (PRNM) Subsystem and the detectors for the Startup Range Neutron Monitoring (SRNM) Subsystem are inserted through the guide tubes. Two levels of stainless steel stabilizer latticework of clamps, tie bars, and spacers give lateral support and rigidity to the ICGTs. The stabilizers are connected to the shroud and shroud support. Surveillance specimen capsules, which are held in capsule holdert mentioned earlier, are located at three azimuths at a common elevation in the core beltline region. The capsule holders are non-safety-related internals. The capsule holders are mechanically retained by capsule holder brackets welded to the vessel cladding that allow capsule removal and re-installation. As a natural circulation reactor the SBWR requires additional elevation head created by the density difference between the saturated water-steam mixture exiting the core and the subcooled water exiting the regionjust below the separators and the feedwater inlet. The chimney provides this elevation head or driving head necessary to sustain the natural circulation flow. The chimney is a long cylinder mounted to the top guide and which supports the steam separator assembly. The chimney forms the annulus separating the subcooled recirculation flow returning downward from the steam General Plant Description - Amendment 1 1.2-12
2SAS113Rev. A samtantsahnyAnorsisneport ] SBWR [ separators and feedwater, from the upward steam-water mixture flow exiting the core. Inside the chimney are partitions which separate groups of 36 fuel assemblies and thereby form smaller chimney sections limiting cross flow and flow instabilities. The BWR direct cycle requires separation of steam from the steam-water mixture i leaving the core. This is accomplished inside the RPV by passing the mixture sequentially first through an array of steam separators attached to a removable cover on the top of the chimney. assembly, and then through standard BWR steam dryers. The steart, dryer and the separator assembly are connected so they can be removed as a unit assembly to simplify refueling. The dryer arrangement has been sized to dry steam with an inlet moisture content of 20%, although inlet moisture ofless than 2% is expected during normal operations. The dryers are designed to provide outlet dry steam with a moisture content s 0.1% ; The core support structures are classified as Quality Group C, Seismic Category I. The j l ; design, materials, manufacturing, fabrication, examination, and inspection used in the constmction of the core support structures meet the requirements of ASME Code Section Ill, subsection NG, Core Support Structures. [ ( These structures are code-stamped accordingly. Other reactor internals are designed per the guidelines of ASME Code NG-3000 and are constructed so as not to adversely l affect the integrity of the core support structures as required by NG 1122. i Special controls on material fabrication processes are exercised when austenitic ' stainless steel is used for construction of RPV internals in order to avoid stress corrosion cracking during senice. Design and constmction of the RPV internals assure that the internals can withstand the i effects of flow-induced vibration (FIV). i 1.2.2.1.2 Nuclear Boiler System The primary functions of the Nuclear Boiler System (NBS) are: (1) to deliver steam from the RPV to the turbine main steam system (TMSS), (2) to deliver feedwater from the Condensate and Feedwater System (C&FS) to the RPV, (S) to provide overpressure protection of the RCPB, (4) to pro ide automatic depressurization of the RPV in the event of a LOCA where the RPV does not depressurize rapidly, and (5) with the exception of monitoring the neutron flux, to provide the instrumentation necessary for monitoring conditions in the RPV such as RPV pressure, metal temperature, and water levelinstrumentation. The main steam lines (MSLs) are designed to direct steam from the RPV to the TMSS; ; the feedwater lines (FWLs) to direct feedwater from the C&FS to the RPV; the RPV General Plant Description - Amendment 1 1.2-13 l i
25AS113 Riv. A standardsafety Analysis Report SBWR O instrumentation to monitor the conditions within the RPV over the full range of reactor power operation. The NBS contains the valves necessary for isolation of the MSLs, FWLs, and their drain lines at the containment boundary. The NBS contains the safety / relief valve discharge lines (SRVDLs), including the steam quencher located in the suppression pool at the end of each SRVDL. The NBS also contains the RPV head vent line and non-condensable gas removal line. Main Steam Unos The NBS contains the portion of the MSLs from their connection to the RPV to the boundary with the TMSS which occurs at the seismic interface located downstream of l the outboard main steamline isolation valves (MSIVs). The main steam lines are Quality Group A from the RPV out to and including the outboard MSIVs, and Quality Group B from the outboard MSIVs to the turbine stop valves. They are Seismic Category I from the RPV out to the seismic interface. Main Steamline Flow Limiter I The main steam line flow limiter is essentially a flow restricting venturi built into the RPV MSL nozzle of each of the two main steam lines. The restrictor limits the coolant blowdown rate from the reactorvessel to a (choke) flow rate equal to or less than 200% of rated steam flow at 7.07 MPa (1025 psig) upstream gauge pressure in the event a main steam-line break occurs anywhere downstream of the nozzle. The MSL flow limiters thus limit offsite dose from postulated MSL breaks outside containment, while the MSIVs are closing. They limit the two-phase depressurization level swell and liquid coolant loss from the vessel, and the rate of first-peak (vent clearing) containment pressure rise for a MSL break inside containment. The flow limiters also limit the intensity of the depressurization level swell and differential pressures momentarily developed on core internals following a MSL break. The flow restrictors are designed and fabricated in accordance with the ASME Code and , designed in accordance with ASME Fluid Meters Handbook.The flow restrictor has no moving parts. l The restrictors are also used to monitor steam flow and to initiate closure of the MSIVs when the steam flow exceeds preselected operational limits. The vessel dome pressure and the venturi throat pressure are used as the high and low pressure sensing locations. l ! Main Steamline Isolation Valves l Each MSIV assembly consists of a main steamline isolation valve, a pneumatic l accumulawr, connecting piping and associated controls. l' General Plant Description - Amendment 1 l 1.2-14 ! I i l I 1
25AS113Rev. A SBWR standantsafetyAnalysisReput j l O V There are two MSIVs welded into each of the two MSLs. On each MSL there is one MSIV in the containment and one MSIV outside the containment. Each set of two MSIVs , isolate their respective MSL upon receipt of isolation signal and will close on loss of pneumatic pressure to the valve. i The MSIVs are Y-pattern globe valves. The main disc or poppet is attached to the lower end of the stem. Normal steam flow tends to close the valve, and higher inlet pressure tends to hold the valve closed. The Y-pattern configuration permits the inlet and outlet flow passages to be streamlined; this minimizes pressure drop during normal steam flow. i The primary actuation mechanism uses a pneumatic cylinder; the speed at which the valve opens and closes can be adjusted. Helical springs around the spring guide shafts will close the valve if gas pressure in the actuating cylinder is reduced. The MSIV quick-closing speed is 2 3 and 5; 4.5 seconds when N2 or air pressure is j ; admitted to the upper piston compartment. The valve can be test closed with a 45-60 l l second slow closing speed by admitting N 2 or air to both the upper and lower piston compartments. O Feedwater Lines The feedwater piping consists of two FWLs connecting to a feedwater supply header. Isolation of each FWLis accomplished by two containment isolation valves consisting of one check valve inside the drywell and one positive closing check valve outside the containment. Also included in this portion of the FWL is a manual maintenance valve l located between the inboard isolation valve and the reactor nozzle. The feedwater line I upstream of the outboard isolation valve contains an additional check valve, a remote manual motor-opeteted (MO) gate valve, and a seismic interface restraint. The outboard isolation valve ad the MO gate valve provide a quality group transitional l
- . point in the FWLs. .
The feedwater piping is Quality Group A from the RPV out to and including the outboard isolation valve, Quality Group B from the outboard isolation valve to and including the MO gate valve, and Quality Group D upstream of the MOJate valve. The feedwater piping and all connected piping 21/2-inch orlarger nominal size are Seismic Category I from the RPV to the seismic interface. Safety / Relief Valves The nuclear pressure relief system consists of safety / relief valves (SRVs) located on the MSLs between the RPV and the inboard MSIV. There are four SRVs per MSL. SRVs provide three main protection functions: iO l V (1) Overpressure Safety Operation: The valves function as safety valves and open to prevent nuclear system overpressurization. They are self-actuating by inlet 1.2-16 General Plant Description - Amendment 1
25A5113Rw. A SBWR standardsafety AnsIrsis Report O steam pressure. 4 The safety mode of operation is initiated when direct and increasing static inlet steam pressure overcomes the restraining spring and frictional forces acting against the inlet steam pressure at the valve disc. This then moves the disc in the opening direction. The condition at which this actuation is inidated corresponds to the set-pressure value stamped on the nameplate of the SRV. (2) Overpressure Relief Operation: The SRVs can be operated individually in the power-actuated mode by remote manual switches located in the main control room (MCR). The valves are opened using a pneumatic actuator to reduce pressure or to limit pressure rise. This mode of operation is initiated when an electrical signal is received at any of the solenoid valves located on the pneumatic actuator assembly. The solenoid valve (s) open, allowing pressurized air to enter the lower side of the pneumatic cylinder which pushes the piston and rod upwards. This action pulls the valve disc lifting mechanism to allow steam to discharge through the SRV. When the solenoids are deenergized, the piston and rod fall downward which causes the valve to rescat and stop SRV steam flow. The SRV pneumatic operator is so arranged that, ifit malfunctions, it will not pierent the SRV from opening when steam inlet pressure reaches the spring lift setpoint. (3) Depressurization Operation: This is discussed separately, below. The SRVs meet the requirements of ASME Code Section Ill. The power supply is 125 Vdc, Class IE for the system. The SRV controls are classified as Class IE. Each SRV has one dedicated, independent pneumatic accumulator which provides the l safety-related, assured nitrogen supply for opening the valve. l The SRVs are flange mounted onto forged outiet fittings located on the top of the main steamline piping in the drywell. The SRVs discharge through lines routed to quenchers in the suppression pool. Automatic Depressurization Subsystem The Automatic Depressurization Subsystem (ADS) quickly depressurizes the RPV in sufficient time for the Gravity-Driven Cooling System (GDCS) injection flow to replenish core coolant to maintain core temperature below design limits in the event of a LOCA. It also maintains the reactor depressurized for continued operation of GDCS after an accident without need for power. 1.2 16 General Plant Description - Amendment 1
l 2SA5113Rev. A SBWR sandardsareryAnorses nepar The ADS consists of the eight SRVs and six depressurization valves (DFVs) and their associated instrumentation and controls. Four DPVs are flange-mounted on horizontal stub lines connected to the RPV at about the elevation of the MSLs. The other two DPVs are flange-mounted on horizontallines branching from each MSL The DPVs discharge into the drywell. The SRVs and DPVs are actuated in groups of valves at staggered times as the reactor undergoes a relatively slow depressurization. This minimizes reactor level swell during the depressurization, thereby enhancing the passive resupply of coolant by the GDCS. The staggered opening of the valves is achieved by delay timers. The use of a combination of SRVs and DPVs to accomplish the ADS function provides l an improvement in ADS reliability against hypothetical common-mode failures of otherwise non-diverse ADS components. It also minimizes components and maintenance as compared to using only SRVs or only DPVs for this function. By using ; the SRVs for two different purposes, the number of DPVs required is minimized. By using DPVs, which have about twice the steam relieving capacity of the SRVs, for the additional depressurization capability needed beyond what the SRVs can provide, the total number of SRVs, SRV discharge lines, and quenchers in the suppression poolis N minimized. The need for SRV maintenance, periodic calibration and testing, and the potential for simmering are minimized with this arrangement. j The ADS automatically actuates on a reactor Level 1 signal that persists for a preset time. A two-out-of-four Level 1 logic is used to activate the SRVs and DPVs. The persistence , requirement for the Level 1 signal ensures that momentary system perturbations do not actuate ADS when it is not required. The two-out-of-four logic assures that a single failure will not cause spurious system actuation while also assuring that a single failure cannot prevent initiation. The ADS may also be manually initiated from the main control room. Depressurization Valvas The DPVs are of a non-leak /non-simmer /non-maintenance design. They are straight-through, squip-actuated, non-reciosing valves with a metal diaphragm seal. The valves are connected to an 8 in. inlet pipe and a 12 in outlet pipe. Each valve provides about twice the depressurization capacity as an SRV. The DPV is closed with a cap covering the inlet chamber. The cap will readily shear off when pushed by a valve plunger which is actuated by the explosive initiator-booster. This opens the inlet hole through the plug. The sheared cap is hinged such that it drops out of the flow path and will not block the valve. The DFVs are designed so that there is no leakage across the cap throughout the hic of the valve. t v V Two initiator-boosters (squibs) actuate the shearing plunger, which are,in turn, initiated by any one of, or any combination of, three battery-powered, independent General Plant Description - Amendment 1 1.2-17
r ! l l 25AS113 Rev. A l SBWR standardsafety Anstrsis neport l r l firing circuits. One initiator-booster has two pairs of pins connected through a wire biidge, the other has one pair of pins connected through a bridge wire. The firing of l one initiator-booster is adequate to activate the plunger. Nominal firing voltage is 125 Vdc, however the initiator-boosters are designed to function with any applied l voltage between 90 and 155 Vdc. The valve design and initiator-booster design is such that there is substantial thermal margin between operating temperature and the self-ignition point of the initiator-booster. The DPVs form a part of the reactor coolant pressure boundary (RPCB) and are l therefore Quality Group A, ASME Section III, Class 1, and Seismic Category I. NBSInstrumentation The NBS RPV instrumentation monitors and provides control inputs for operational l variables during plant operation. The NBS contains the instrumentation for monitoring the reactor pressure, metal temperature, and water level. The reactor pressure and water level instruments are used I by multiple systems, both safety-related and non-safety-related. Pressure indicators and transmitters detect reactor vesselinternal pressure from the same instrument lines used for measuring reactor vessel water level. RPV coolant temperatures are determined by measuring saturation pressure (which gives the saturation temperature), outlet flow temperature to the RWCU/SDC System, and RPV bottom head drain line temperature. Reactor vessel outside surface (metal) temperatures are measured at the head flange and the bottom head locations. Temperatures needed for operation and for operating limits are obtained from these measurements. During normal operation, either reactor steam saturation temperature and/or inlet temperatures of the reactor coolant to the RWCU/SDC System and the RPV bottom head drain can be used to determine the RPV coolant temperature. The instruments that sense the water level are differential pressure devices calibrated for a specific RPV pressure (and corresponding liquid temperature). The water level measurement instmmentation is the condensate reference chamber type. Instrument reference zero for all the RPV water level ranges is the top of the active fuel. The l following is a description of each water level range. i ! (1) Shutdown Range Water Level This range is used to monitor the t eactor water level during shutdown l conditions when the reactor system is flooded for maintenance and head i removal. The two RPV instrument taps used for this water level measurement are located at the top of the RPV head, andjust below the dryer skirt. 1.2-18 General Plant Description - Amendment 1
l l 1 l 25AS113 Rzv. A SBWR standardsafety Analysis acport \ l ,f"\ l ll 1 (2) Narrow Range Water Level l This range is used to monitor reactor water level during normal power operation. This range uses the RPV taps near the top of the steam outlet nozzles and near the bottom of the dryer skirt. The Feedwater Control l System (FWCS) uses this range for its water level control and indication l inputs. The RPS also uses this range for scram initiation. (3) Wide Range Water Level This range is used to monitor reactor water level for events where the water level exceeds the range of the narrow range water levelinstrumentation, and is used to generate the low reactor water level trip signals which indicate a potential LOCA. This range uses the RPV taps at the elevations near the top l l of the steam outlet nozzles and the nearest tap above the top guide. l l (4) Fuel Zone Range Water Level This range is provided for post-accident monitoring and provides the ' (gv) capability to monitor the reactor water level below the wide range water level instrumentation. This range uses the RPV taps at the elevations near the top of the steam outlet nozzles and the taps below the bottom of active fuel. Thermocouples are located in the discharge exhaust pipes of the SRVs. The temperature signals go to a multipoint recorder with an alarm and will be activated by any temperature in excess of a set temperature, signaling that one of the SRV seats has started to leak. Control room indication and alarms are provided for the important plant parameters { monitored by the NBS. i l l l l i l rx i l
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General Plant Description - Amendmert 1 1.2-19
25AS113Rev. A SBWR stamtantsarety Analysis neput O NBS ASME Code Requirements The major NBS mechanical components are designed to meet ASME Code Requirements as shown below: Design Conditions ASME Code Component Class Gauge Pressure Temperatum FWLs from the MOVs to the 2 8.62 MPa 302*C outboard containmentisolation (1250 psig) (575 F) check valves FWLs from the outboard 1 8.62 MPa 302 C containment isolation check valve to (1250 psig) (575 F) the RPV FWL line outboard containment 1 8.62 MPa 302 C isolation check valve (1250 psig) (575 F) Main steam isolation valves (MSIVs) 1 9.48 MPa 308*C l (1375 psig) (586*F) Safety / relief valves (SRVs) 1 9.48 MPa 308*C (1375 psig) (586 F) Main steam lines (MSLs), from RPV 1 8.62 MPa 302 C to outboard MSIVs (1250 psig) (575*F) MSLs from the outboard MSIVs to 2 8.62 MPa 302*C the seismic interface restraint (1250 psig) (575 F) SRV discharge line piping, from the 3 3.72 MPa 250 C SRVs to the vent wall penetration (540 psig) (482 F) SRV discharge line piping, from the 2 3.72 MPa 250 C vent wall penetration to the (540 psig) (482 F) suppression pool surface i l l l l O i 1.2-20 General Plant Description - Amendment 1 i
2SA5113Rw. A SBWR standardsateer Ansorsis neport O v , 1.2.2.2 Controls and instrumentation 1.2.2.2.1 Rod Control and Information System The Rod Control and Information System (RC&lS) is to safely and reliably provide: a The capability to control reactor power level by controlling the movement of control rods in reactor core in manual, semiautomatic, and automated modes of plant operations. s Controls for some RC&lS bypass and sun'eillance test functions, and summary , information of control rod positions and status in the main control room. e Transmission of fine motion control rod drive (FMCRD) status and control rod positions and status data to other plant systems (e.g., the Process Computer System). s Automatic control rod run-in function of all operable control rods following a scram (scram follow function). s Automatic enforcement of rod movement blocks to prevent potentially undesirable rod movements (these blocks do not have an effect on scram insertion function). A e Control capability for insertion of all control rods by an alternate and diverse t method [ alternate rod insertion (ARI) function]. l e The capability to enforce a preestablished sequence for control rod movement l when reactor power is below the low power setpoint. s The capability to enforce fuel operating thermal limits when reactor power is above the low power setpoint. ! i l s The capability to provide for Selected Control Rod Run In 'SCRRI) function for mitigating a loss of feedwater heating event. i l The RC&IS is classified as a non-safety-related system, it has a control design basis only, and is not required for the safe shutdown of the plant. A failure of the RC&lS will not result in gross fuel damage. However, the rod block function of RC&IS is important in l limiting the consequences of a rod withdrawal error, and prevention oflocal fuel operating thermal limits violations during normal plant operations. Therefore, the RC&lS is designed to be single-failure proof and highly reliable. l l The RC&lS consists of several different types of cabinets (or panels), which contain l , special electronic / electrical equipment modules, and a dedicated operator interface i on the main control panelin the MCR. s I ! GeneralPlant Description - Amendment 1 1.2-21 i
25AS113Rav. A SBWR standardsafetyAnalysis Report O The RC&lS is a redundant system consisting of two independent channels for normal control rod position monitoring and control rod movements. The two channels receive the same but separate input signals and perform the same exact functions. For normal functions of the RC&lS, the two channels must always be in agreement and any disagreement between the two channels results in rod block. However, the protective function logic of the RC&lS (i.e., rod block) is designed such that the detection of a rod block condition in only one channel of RC&lS would result in a rod block. There are four types of electronic / electrical cabinets that make up the RC&IS. They are: a Rod acdon control cabinets (RACC) m Remote communication cabinets (RCCs) m Fine motion driver cabinets (FMDCs) m Rod brake controller cabinets (RBCCs) In addition, the RC&lS includes a fiber-optic dual channel multiplexing network that is used for transmission of rod position and status data from RCCs to the Rod Action and Position Information (RAPI), and rod block / movement command from RAPI to RCCs. A summary descdption of each of the above functions is provided below.
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Rod Action Control Cabinets (RACC) There are two RACCs in the control room; RACC Channel A and RACC Channel B that provide for a dual redundant architecture. Each RACC consists of three main functional , subsystems, as follows: e Automated Thermal Limit Monitor (ATLM) l u Rod Worth Minimizer (RWM) m Rod Action and Position Information (RAPI) Remote Communication Cabinets (RCC) The remote communication cabinets (RCCs) contain a dual channel file control module (FCM) and several dual channel rod server modules (RSMs). The FCM interfaces with the RSMs and RAPI. 1 Fine Motion Driver Cabinets (FMDC) The fine motion ddver cabinets (FMDCs) consist of several stepping motor driver l modules. Each stepping motor driver module contains an electronic converter / inverter that converts the incoming three-phase ac power into dc and then inverts the dc power l to variable voltage / frequency ac power that is supplied to F'MCRD stepping motors. For l l 1.2-22 General Plant Description - Amendment 1 l l
2SA5113Rsv. A SBWR staatentsarnyAnsorsisnepar 1 i o I each converter / inverter, there exists an inverter controller (IC) that controls the duration of power supplied to the stepping motors under the command of RSMs. ) Rod Breke Controller Cebinets (RBCC) I The rod brake controller cabinets (RBCCs) contain electrical power supplies, electronic (or relay) logic, and other associated electrical equipment for the proper operation of the FMCRD brakes. Signals for brake disengagement / engagement are received from the associated rod server modules. The brake controller logic prosides two separate (Channel A and Channel B) brake status signals to the associated rod sener module. RC&lS Multiplexing Network The RC&IS multiplexing network consists of two independent channels (Channel A and Channel B) of fiber-optic communication links between the RACCs (Channel A and Channel B), and the dual channel file control modules located in the remote communication cabinets. The plant essential multiplexing network interfaces with FMCRD redundant separation l switches (A/B) and provides the appropriate status signals to the RACC that is used in t the RC&IS logic for initiating rod block signals if a separation occurs. The essential ( multiplexing network is not part of the RC&IS scope. i i RC&lS Power Sources l RC&IS equipment derives its power from two different sources. FMDCs and RBCCs derive their power from the plant divisional power sources that are backed up by plant - diesel generators. All other RC&lS equipment derive their power from the plant uninte:Tuptible ac power system. 1.2.2.2.2 Control Rod Drive System The Control Rod Drive (CRD) System is composed of three major elements: a the fine motion control rod drive (FMCRD) mechanisms, a the hydraulic control unit (HCU) assemblies, and a the Control Rod Drive Hydraulic (CRDH) Subsystem. l The FMCRDs provide electric-motor-driven positioning for normalinsertion and - withdrawal of the control rods and hydraulic-powered rapid control rod insertion (scram) for abnormal operating conditions. Simultaneous with scram, the FMCRDs also provide electric-motor-driven run-in of all control rods as a path to rod insertion that is diverse from the hydraulic-powered scram. The hydraulic power required for scram is provided by high pressure water stored in the individual HCUs. Each HCU is designed to scram up to two FMCRDs. The HCUs also provide the flow path for purge water to GeneralPlant Description - Amendment 1 1 2-23
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25A5113Rzv. A SBWR standardsareryAnalysis neport l 1 the associated drives during normal operation. The CRD11 Subsystem supplies high pressure demineralized water which is regulated and distributed to provide charging of the HCU scram accumulators, purge water flow to the RfCRDs, and backup makeup water to the RPV when the feedwater flow is not available. During power operation, the CRD System controls changes in core reacthity by movement and positioning of the neutron absorbing control rods within the core in fine increments via the FMCRD electric motors, which are operated in response to control signals from the RC&lS. The CRD System provides rapid control rod insertion (scram) in response to manual or automatic signals from the Reactor Protection System (RPS), so that no fuel damage results from any plant transient. There are 177 FMCRDs mounted in housings welded into the RPV bottom head. Each FMCRD has a movable hollow piston tube that is coupled at its upper end, inside the reactor vessel, to the bottom of a control rod. The piston is designed such that it can be moved up or down, both in fine increments and continuously over its entire range, by a ball nut and ball screw driven at a nominal speed of 30 mm/sec by the electric stepper motor. In response to a scram signal, the piston rapidly inserts the control rod into the core hydraulically using stored energy in the HCU scram accumulator. The scram water is introduced into the drive through a scram inlet connection on the B1CRD housing, and is then discharged directly into the reactor vessel via clearances between FMCRD parts. The B1CRD scram time requirements with the reactor gauge pressure of 7.481 MPa (gauge) (1085 psig) as measured at the vessel bottom are: Percent Insertion Time (sec) 10 s 0.42 40 s1.00 60 s 1.44 100 s 2.80 The FMCRD design includes an electro-mechanical brake on the motor drive shaft and a ball check valve at the point of connection with the scram inlet line. These features prevent control rod ejection in the event of a failure of the scram insert line. An internal housing support is provided to prevent ejection of the FMCRD and its attached control rod in the event of a housing failure. It uses the outer tube of the drive to provide support. The outer tube, which is welded to the drive middle flange, attaches by a bayonet lock to the base of the control rod guide tube section of the integral fuel 1.2-24 General Plant Description - Amendment 1
2SAS113 Rev. A SBWR standarasarety Anarysis neport [%. (
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L! support and control rod guide tube. The fuel support section, being supponed by the lower core plate, in turn, prevents any downward movement of the drive. The Bf CRD is designed to detect separation of the control rod from the drive mechanism. Two redundant and separate Class IE switches detect separation of either the control rod from the hollow piston or the hollow piston from the ball nut. Actuation of either switch will cause an immediate rod block and initiate an alarm in the MCR, thereby preventing the occurrence of a rod drop accident. Therc ate 89 HCUs, each of which provides sufficient volume of water stored at high pressure in a pre-charged accumulator to scram two BiCRDs at any reactor pressure. Each accumulator is connected to its associated BiCRDs by a hydraulic line that includes a normally-closed scram valve. The scram valve opens by spring action but is normally held closed by pressurized control air. To cause scram, the RPS prosides a de-energizing reactor trip signal to the solenoid-operated pilot valve that vents the control air from the scram valve. The system is " fail safe" in that loss of either elecuical power to the solenoid pilot valve or loss of control air pressure causes scram. The HCUs are housed in the safety envelope at the basemat elevation. This is a Seismic Categon I structure, and the HCUs are protected from external natural phenomena such as canhquakes, tornados, hurricanes and floods, as well as from internal postulated (V) accident phenomena. In this area, the HCUs are not subject to conditions such as missiles, pipe whip, or discharging fluids. The CRDH Subsystem design provides the pumps, valves, filters, instrumentation, and piping to supply the high pressure water for charging the HCUs and purging the BfCRDs. Two 100% capacity pumps (one on standby) supply the HCUs with water from the condensate treatment system and/or condensate storage tank for charging the accumulators and for supplying BICRD purge water. The CRDH Subsystem equipment is housed in the Seismic Category I portion of the reactor building to protect the system from floods, tornadoes, and other natural phenomena. The CRDH Subsystem also has the capability to provide makeup water to the RPV while at high pressure as long as ac power is available. The CRD System includes MCR indication and alarms to allow for monitoring and control during design basis operational conditions, including system flows, temperatures and pressures, as well as valve position indication and pump on/off status. Class 1E pressure instrumentation is provided on the HCU charging water header to monitor header performance. The pressure signals from this instrumentation are provided to the RPS, which willinitiate a scram if the header pressure degrades to a low pressure serpoint. This feature assures the capability to scram and safely shut down the (7 reactor before HCU accumulator pressure can degrade to the level where scram (u,) performance is adversely affected following the loss of charging header pressure. GeneralPlant Description - Amendment 1 1.2-2S
25A5113 Riv. A SBWR standardsafety Anstrsis Report O Components of the system that are required for scram (FMCRDs, HCUs and scram piping), are classified Seismic Categog I. The balance of the system equipment (pumps, valves, filten, piping, etc.) is classified as Seismic Category NS (non-seismic), with the exception of the Class IE charging water header pressure instrumentation, which is Seismic Category I. The major mechanical components are designed to meet ASME Code requirements as shown below: Design Conditions ASME Code Component Class Gauge Pressure Temperature FMCRD (RCPB parts) 1 8.62 MPa 302*C (1250 psig) (545*F) Scram piping 2 18.6 MPa 66*C (2700 psig) (150*F) HCU (scram related parts) 2 18.6 MPa 66*C (2700 psig) (150*F) CRD pumps Non-Code 18.6 MPa 66*C (2700 psig) (150*F) CRDHS piping, valves Non-Code 18.6 MPa 66*C (2700 psig) (150*F) The CRD System is separated both physically and elecuically from the Standby Liquid Control System (SLCS). 1.2.2.2.3 Feedwater Control System The Feedwater Control System (RVCS) controls the flow of feedwater into the RPV to maintain the water level in the vessel within predetermined limits during all plant operating modes. The FWCS may operate in either single- or three-element control modes. At low reactor powers (when steam flow is either negligible or else measurement is below scale), the FWCS uses only water level measurement in single-element control mode. When steam flow is negligible, the Reactor Water Cleanup / Shutdown Cooling (RWCU/SDC) System overboard control valve can be controlled by the FWCS System in single-element mode in order to counter the effects ofdensity changes during heatup l and purge flows into the reactor. At higher powers, the FWCS in three-element control l ! mode uses water level, main steamline flow, main feedwater line flow, and feedpump suction flow measurements for water level control. The FWCS is a power generation (control) system with operation range between high ; j water level and low water level tnp setpomts. It is classified as nonsafety-related. This ; l 1.2-26 Geners! Plant Description - Amendment 1 l l l l
2SA5113Riv. A SBWR stamianisatery Analysis neport i o 3 D system is not required for safety purposes, nor is it required to operate after the design basis accident. This system is only required to operate in the normal plant environment, and for power generation purposes only. Reactor vessel narrow range water level is measured by three identical, independent sensing systems. For each level measurement channel, a differential pressure transmitter senses the difference between the pressure caused by a constant reference column of water and the pressure caused by the variable height of water in the reactor vessel. The FWCS uses microprocessor-based fault tolerant digital controllers (FTDCs) which will determine one validated narrow range level signal using the three level measurements as inputs to a signal validation algorithm. The validated narrow range water level is indicated on the main control console in the MCR. Steam flow is sensed at the RPV MSL nozzle venturi in each of the two main steamlines. The Multiplexing System signal conditioning algorithms process the venturi differential pressures and provide steam flow rate signals to the FTDCs for validation. These validated measurements are summed in the FTDCs to give the total steam flow rate out of the vessel. The total steam flow rate is indicated on the main control console in the MCR.
,l Feedwater flow is sensed at a single flow elementin each of the two feedwater lines. The Multiplexing System signal conditioning algorithms process the flow element differential pressure and provide feedwater flow rate signals to the FTDCs. These validated measurements are summed in the FTDCs to give the total feedwater flow rate into the vessel. The total feedwater flow rate is indicated on the main control console in the MCR.
Feedpump suction flow is sensed at a single flow element upstream of each feedpump. The Multiplexing System signal conditioning algorithms process the flow element differential pressure and provide the suction flow rate measurements to the FTDCs. The feedpump suction flow rate is compared to the demand flow for that pump, and the resulting error is used to adjust the actuator in the direction necessary to reduce that error. Feedpump speed change and low flow control valve position control are the flow adjustment techniques involved. Three modes of feedwater flow control (and, thus, level control) are provided: (1) single-element control; (2) three-element control; and (3) manual control. Each FTDC will execute the control software for all three of the control modes. Actuator demands from the redundant FTDCs will be sent over the Multiplexing System to field voters which will determine a single demand to be sent to each actuator. Each
,.s feedpump speed or control valve position demand may be controlled either automatically by the control algorithms in the FTDCs or manually from the main (L- )
control room through the FTDCs. I GeneralPlant Description - Amendment 1 1.2-27
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1 25AS113Riv. A SBWR standantsareryAnarrsis aeroa O Three-element automatic control is provided for normal operation. Three-element control uses water level, feedwater flow, steam flow, and feedpump flow signals to determine the feedpump demands. The total feedwacer flowis subtracted from the total steam flow signal, yielding the vessel flow mismatch. The flow mismatch, summed with the conditioned level error from the master level controller, provides the demand for the master flow controller. The master flow controller output provides the demand signal to the adjustable speed dtives (ASD) for each feedpump. In the single-element control mode, only conditioned level error is used to determine the feedpump demand. The master level controller condidons the level error and sends it directly to the feedpump ASDs, and/or low flow control valve actuator. When the reactor water inventory must be decreased (e.g., during very low steam flow rate conditions), the RWCU/SDC System overboard control valve is controlled by the RVCS in single-element control. Reactor water is discharged through the RWCU/SDC System to the condenser. Each feedpump flow control actuator can be controlled ' manually' from the main control panel by selecting the manual mode for that feedpump. In manual mode, the operator may increase or decrease the demand that is sent directly to the ASD of the chosen feedpump. The RVCS also provides interlocks and control fimctions to other systems. When the reactor water level reaches the high level trip setpoint, the RVCS simultaneously annunciates an alarm in the MCR, sends a trip signal to the turbine control system to trip the turbine generator, sends trip signals to all feedpumps, and closes the main feedwater discharge valves. This interlock is enacted to protect the turbine from damage from high moisture content in the steam caused by excessive carryover, while preventing water level from rising any higher. The FWCS sends a signal to the main steamline condensate drain valves to open when steam flow rate is below a pre-determined setpoint. This also protects the turbine from damage caused by excessive moisture in the steamline. Feedwater flow is delinred to the reactor vessel through a combination of three adjustable speed motor-driven feedpumps and a low flow control valve. The low flow control valve (L1 CV) is provided in the high-pressure feedwater heater bypass line. The LFCV can also be controlled by the manual / automatic transfer station which is part of the Condensate and Feedwater System. 1 The FWCS is powered by redundant unintermptible power supplies (UPS). No single j power failure will result in the loss of any FWCS functions. l Controllers to be used for the FWCS are triplicated, fault tolerant digital type with self-test and diagnostic capabilities. 1.2-28 GeneralPlant Description - Amendment 1
I 25A5113Rsv. A , SBWR standardsneerAnar r sisnerat s' ,( 1 1.2.2.2.4 Standby Liquid Control System ! The Standby Liquid Control System (SLCS) provides an alternate method of reactor ; shutdown from full power to cold subcritical by the injection of a neutron absorbing solution into the RPV. The SLCS interfaces with Class IE 125 Vdc divisional power for the squib-type injection valves; for the valve which isolates the accumulator after injection; for accumulator solution level measurement, trip, and alarm functions; and for the particular NBS instrumentation and SSLC control logic which generates the anticipated transient without scram (ATWS) signal for automatic SLCS initiation. The SLCS includes piping, valves, accumulator, and instrumentation designed to inject a neutron absorber solution into the reactor. The system is designed to operate over the ; mnge of reactor pressure conditions up to the elevated pressur es of an ATWS event, and to inject sufficient neutron absorber solution to reach hot subcritical conditions after ; system initiation. t Instrumentation is provided to the operator for monitoring the status of the SLCS, and ip for alarming any off standard condition.
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1.2.2.2.5 Neutron Monitoring System The Neutron Monitoring System (NMS) provides indication of neutron flux in the core , in all modes of reactor operation. The safety-related NMS functions are the startup range neutron monitor (SRNM), the local power range monitor (LPRM), and the average power range monitor (APRM). The non-safety-related subsystem is the automated fixed in-core probe (AFIP). The LPRMs and APRMs make up the power i range neutron monitor (PRNM) subsystem. The safety-related portions of the NMS are classified, Seismic Category I, and IEEE Class 1E. l l The NMS provides signals to the RPS, the RC&lS, and the Process Computer System. [ The NMS provides trip signals to the RPS for reactor scram on rising excessive neuuon l flux or too short a period for flux generation. 1 The NMS consists of four divisions which correspond and interface with those of the RPS, and this independence and redundancy assure that no single failure will interfere l i l with the system operation. l The SRNM Subsystem is comprised of eight SRNM channels which are divided into four divisions and independently assigned to three bypass groups such that up to three SRNM channels are allowed to be bypassed at any time while still providing the required hG monitoring and protection capability. l General Plant Description - Amendment 1 1.2-29
I l 25Ab113 Rxv. A SBWR standardsareer Anstrsis Report i ! The LPRM function of the PRNM Subsystem is comprised of 21 LPRM assemblies evenly distributed throughout the cross-section of the core. There are four LPRM detectors within each LPRM assembly, evenly spaced from near the bottom of the fuel region to i near the top of the fuel region. These 84 detectors are assigned to four sets of 21 l detectors each. The signals from each set of 21 LPRM detectors are assigned to one APRM channel, with these signals summed and avenged to form a partial APRM signal. , This partial APRM signalis transmitted to the other three APRM channels through ! electrical isolation. Within each of the four APRM channels, all four partial APRM l signals are then averaged to form a final APRM signal. The partial APRM signal transmission between divisions is carried out through fiber optic pathways which sen'e as effective electrical isolation devices. Electrical and physical separation of the division I is thus maintained and optimized to satisfy the safety-related system requirement. With l the four divisions, redundancy criteria are met since a scram signal can still be initiated ! with a postulated single failure under allowed APRM bypass conditions. All the NMS instruments are primarily based on the digital measurement and control l (DMC) design practices that use digital design concepts. All NMS DMC instruments follow a modular design concept such that each modular unit or its subunit is l replaceable upon repair senice. 1 3 ( The SRNM Subsystem covers the lower power range from the source range (1 x 10 nv) to 15% of rated reactor power. The PRNM Subsystem overlaps the SRNM, covering the range from approximately 1% to 125% of rated reactor power. The AFIP Subsystem is comprised of sensors and their associated cables, as well as the signal processing electronic unit. The AFIP sensors are the gamma thermometer type. There are four AFIP gamma thermometer sensors evenly distributed across each LPRM assembly, with one gamma thermometer installed next to each LPRM detector. Consequently, there are AFIP sensors at all LPRM locations. The AFIP sensor cables are routed within the LPRM assembly and then out of the RPV through the LPRM assembly penetration to the vessel. The AFIP Subsystem generates signals proportional to the axial power distribution at the radial core locations of the LPRM detector assemblies. The AFIP signal range is sufficiently wide to accommodate the corresponding local , l power range that covers from 0% to 125% of reactor rated power. l The AFIP gamma thermometer sensor has a constant or very stable detector sensitivity ; that will not significantly change due to radiation exposure or other reactor conditions. I l The AFIP gamma thermometer, however, can be calibated by using a built-in calibration device inside the gamma thermometer /LPRM assembly. Due to its stable 1 l sensitivity and mgged hardware design, the AFIP sensor has a lifetime much longer than I that of the LPRM detectors. The AFIP sensors in an LPRM assembly are replaced together with the LPRM detectors when the whole LPRM assembly is replaced. 1.2-30 General Plant Description - Amendment 1 l l I i \
I 25A5113 Rsv. A , SBWR standantsarayAnsorsesnerat
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1.2.2.2.6 Remote Shutdown System The Remote Shutdown System (RSS) prmides the means to safely shut down the reactor from outside the main control room. The RSS pro ides remote manual control of the systems necessary to: (a) achieve prompt hot shutdown of the reactor after a scram, (b) achieve subsequent cold shutdown of the reactor, and (c) maintain safe conditions during shutdown. , The Remote Shutdown System is classified as a non-safety-related system. The RSS does not include control interfaces with nuclear safety-related equipment. To achieve a safe and orderly plant shutdown from outside the main control room, l controls and indicators necessary for operation of the following system and equipment , are provided on the remote shutdown panel. , a Reactor Water Cleanup / Shutdown Cooling (RWCU/SDC) System a Control Rod Drive (CRD) System (makeup function) a Reactor Component Cooling Water (RCCW) System ( a Plant Senice Water (PSW) System s Electrical Power Distribution System r a Nuclear Boiler System (NBS) instrumentation a Reactor Building HVAC 1.2.2.2.7 Reactor Protection System The Reactor Protection System (RPS) initiates an automatic and prompt reactor trip (scram) by means of rapid hydraulic insertion of all control rods whenever selected plant variables exceed preset limits. The primary function is to effect a reactor shutdown before fuel damage occurs. The RPS also provides reactor status information to other systems and will cause an alarm annunciation in the MCR whenever selected plant variables approach the preset limits. The RPS is a safety protection system, differing from a reactor control system or a power generation system. The RPS and its components are safety-related. The RPS and the system electrical equipment are classified as Seismic Category I and IEEE Class 1E. ( General Plant Description - Amendment 1 1.2-31
1 25A5113 Rov. A SBWR stamtardsarery Anarrsis neport l 1 . l Basic system parameters are: Number ofindependent divisions of equipment 4 Minimum number of sensors per trip variable (at least one per division) 4 Number of automatic trip systems (one per division) 4 Automatic trip logic used for plant sensor inputs (per division) 2-out-of-4 Separate automatic uip logic used for division trip outputs 2-outof4 Number of separate manual trip systems 2 Manual trip logic 2-out-of-2 The RPS initiates reactor trip signals within individual sensor channels when any one or more of the conditions listed below exists during reactor operation. Reactor scram will result if system logic is satisfied. a Drywell pressure high a Reactor power (neutron flux or simulated thermal power) exceeds limits for operating mode a Reactor power rapid increase a Reactor vessel pressure high a Reactor water level low (Level 3) a Reactor water level high (Level 8) m Main steam isolation valves closed (Run mode only) a Control rod drive charging header pressure low a Suppression pool temperature high n Operator-initiated manual scram The RPS is an overall complex ofinstrument channels, trip logic, trip actuators, manual controls, and scram logic circuiuy that initiates the rapid insertion of control rods by hydraulic force to scram the reactor when unsafe conditions are detected. The RPS uses the functions of the essential multiplexing subsystem (EMS) and the SSLC system to perform its functions. 1.2-32 General Plant Description - Amendment 1
25AS113Rsv. A ! i SBWR standardsurnyannorsessepar i .O The RPS is divided into four redundant divisions ofsensor channels, trip logics, and trip actuators, and two divi &c of manual scram controls and logic circuitry. Each division has a separate IEEE Class IE power supply taken from the safety-related UPS 120 Vac l power supply. The automatic and manual scram initiation logic systems are independent of each other and use diverse methods and equipment to initiate a reactor scram. The RPS design is such that, once a full reactor scram has been initiated automatically or manually, this scram condition seals-in such that the intended fast insenion of control rods into the reactor core can continue to completion. After a time , delay, the design requires operator action to reset the scram logic to the untripped state. The RPS scram logic circuits are arranged so that coincident trips in two of the four divisions (2-out-of-4 logic) of sensor channels and in two of the four tdp system outputs to the actuating devices are required to effect a scram. This arrangement permits a single failure in one division to occur without either causing a scram or preventing the other three divisions from causing a scram. For example, the single failure may be in either system logic or the individual power supply for that division. Each logic division and its associated power supply is separated both physically and electdcally from the other divisions. This arrangement permits one dinsion at a time to , i be taken out of senice (bypassed) for testing dudng reactor operation. The other divisions then perform the RPS function with system logic in a 2-out-of-3 arrangement. 1.2.2.2.8 Automatic Power Regulator System The Automatic Power Regulator (APR) System is classified as a power generation system and is not required for safety. Events requiring control rod scram are sensed and ; controlled by the safety-related RPS, which is completelyindependent of the APR System. The APR System controls reactor power during reactor startup, power generation, and reactor shutdown by appropriate commands to change rod positions. The APR system also controls the pressure setpoint or turbine bypass valve position during reactor heatup and depressurization (e.g., to control the reactor cooldown rate). The APR System consists of redundant process controllers. Automatic power regulation is l achieved by appropriate control algorithms for different phases of reactor operation I which include approach to criticality, heatup, reactor power increase, automadc load l following, reactor power decrease, and reactor depressurization and cooldown. The APR System receives input from the Neutron Monitoring System, the Process Computer System, the Power Generation Control Subsystem, the Steam Bypass and Pressure Control System, and the operator's control console. The output demand signals from the APR System are sent to the RC&lS to position the control rods, and to the Steam g
! Bypass and Pressure Control System for automatic load following operations. The \ power generation control subsystem performs the overall plant startup, power operation, and shutdown functions. The APR System performs those functions General Plant Description - Amendment 1 1.2-33 I
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2SA5113Riv. A SBWR standardsaferyAnarrsisaeport O associated with reactor power changes and with reactor pressure controller setpoint (or turbine bypass valve position) changes during reactor heatup or depressurization. The automatic power regulation system control functional logic is performed by i redundant, microprocessor-based fault-tolerant digital controllers (FTDC). The FTDC performs many functions. It reads and validates inputs from the non essential multiplexing system (NEMS). It performs the specific power control calculations and processes the pertinent alarm and interlock functions, then updates all system outputs to the NEMS. To prevent computational divergence among the redundant processing channels, each channel performs a comparison check ofits calculated results with other redundant channels. The internal FTDC architecture features redundant multiplexing interfacing units for communications between the NEMS and the FTDC processing channels. During normal operation, the APR System interfaces with the operator's console to perform its desired functions. The operator's control panel for automatic plant startup, power operation, and shutdown functions is part of the power generation control subsystem. The power generation control subsystem initiates demand signals to various controllers to carry out the pre-defined control functions. The functions associated with reactor power control are performed by the APR System. For reactor power control, the APR System contains algorithms that can change reactor power by control rod motions. During automatic load following operation, the APR System interfaces with the Steam Bypass and Pressure Control System to coordinate main turbine and reactor power changes to accomplish load following. l The normal mode of operation for the APR System is automatic. If any system or component conditions are abnormal during execution of the prescribed sequences of operation, the Power Generation Control Subsystem will be automatically switched into the manual mode and the operator can manipulate control rods using the normal controls. A failure of the APR System will not prevent manual control of the reactor, nor will it prevent safe shutdown of the reactor. The APR System digital controllers are powered by redundant uninterruptible non-Class IE power supplies and sources, No single power failure will result in the loss of any APR System function. 1.2.2.2.9 Steam Bypass & Pressure Control System The Steam Bypass & Pressure Control (SB&PC) System is a non-safety-related system whose design objective is to enable a fast and stable response to pressure and system disturbances, and to setpoint changes, over the operating range using turbine control valves and turbine bypass valves for controlling pressure. In addition, the design objective of the SB&PC System is to discharge reactor steam directly to the main 1.2-34 General Plant Description - Amendment 1
25A5113Riv. A SBWR StandardSafety Analysis Report g i 4 condenser to regulate reactor pressure whenever the turbine cannot use all of the steam generated by the reactor. The SB&PC System is designed to control reactor pressure during plant startup, power generation and shutdewn modes of operation. This is accomplished through control of the turbine control valves and/or turbine bypass valves, such that susceptibility to reactor trip, turbine-generator trip, MSIV closure and SRV opening is minimized. Command signals for the turbine control valves and the turbine bypass valves are generated by a triplicated FTDC using feedback signals from RPV pressure signals. For normal operation, the turbine control valves regulate steam pressure. However, whenever the total steam flow demand from the pressure controller exceeds the effective turbine control valve steam flow capability, the SB&PC System sends the excess steam flow directly to the main condenser, through the turbine bypass valves. The SB&PC System functional logic and process control functions are performed by triplicated microprocessor-based FTDC similar to controllers used in FWCS. Because of the triple redundancy, it is possible to lose one complete processing chant.el without impacting the system function. This also facilitates taking one channel out of service for /^'s maintenance or repair while the system is on-line. The SB&PC System receives input V signals from other systems and sensors as follows: m turbine bypass valve position switches; a turbine bypass valve servo current sensors; l a Turbine Control System (TCS) turbine trip sensors; i a TCS power / load unbalance relay operation; l l l m Turbine Bypass System (TBS) hydraulic power supply trouble sensors; a NBS MSIV position switches; a NBS narrow and wide range dome pressure transmitters; a main condenser low vacuum sensors; and a operator manual commands and manual switch positions. The SB&PC system provides output signals to: m turbine bypass valves; fh ( ,) i m turbine controlvalves; General Plant Description - Amendment 1 1.2-35
l 2545113R:v. A l SBWR stamtantsafety Anatrsis Report l I O l s APR System; l a various related control room indicators and alarms; and ) e process computer, i i At steady-state plant operation, the SB&PC System maintains reactor vessel pressure at l a nearly constant value, to ensure optimum plant performance. During normal operational plant maneuvers (pressure setpoint changes, level setpoint changes), the SB&PC System provides responsive, stable performance to minimize vessel water level and neutron flux transients. During plant startup and heatup, the SB&PC System j provides for automatic control of the reactor pressure. Independent control of reactor
- pressure and power is permitted during reactor-vessel heatup, by varying turbine bypass j flow as the main turbine is brought up to speed and syncbronized.
Additional reactor system pressure control functions are provided by other systems l when the MSIVs are closed. I 1.2.2.2.10 Process Computer System The Process Computer System (PCS) is a non-safety-related system. Its purpose is to promote efficient plant operation by: a Performing the functions and calculations necessary for the evaluation of plant operation. s Providing a permanent historical record for plant operating activities and abnormal events. 1 l s Providing analysis, evaluation and recommendation capabilities for start-up, normal j operation, safe plant shutdown and abnormal operating and emergency conditions. s Providing control and display capability on the main control room video display l units. a Providing the ability to directly control certain non-safety-related plant equipment through on-screen technology. All division to division and safety-related to non-safety-related interfacing circuits are ! made up of fiber optic cables, which act as opticalisolators for electrical separation. All power to the PCS is supplied by a non-safety-related redundant, uninterruptible power supply. No single power failure will cause the loss of any PCS function. l The PCS has self-checking provisions. It perfucras diagnostic checks to determine the operability of certain portions of the system hardware and performs internal 1 l 1.2-36 General Plant Description - Amendment 1 l
25AS113 Rsv. A SBWR stamtantsainyAnnor sisnopet l(~ lis i programming checks to verify that input signals and selec:ad program computations are either within specific limits or within reasonable bounds. The PCS is composed of two subsystems; the Performance Monitoring and Control , Subsystem (PMCS) and the Power Generation Control Subsystem (PGCS). Performance Monitoring and Control System The PMCS is a set of software routines for the PCS input / output modules and various CPUs to supply various functions and calculations. The basic input types are as follows: l m Various analog pressure signals from sensors on or in the RPV, the drywell, individual equipment and the various plant buildings. m Various analog temperature signals from sensors on or in the RPV, the drywell, individual equipment and the various plant buildings. ( m Various analog coolant and steam flow signals from sensors on or in the various pumps and pipes throughout the plant. l a Various digital "on/of1" and "open/ closed" signals from various switches and valve n { controllers throughout the plant a Various operator requests as input through the various consoles. The basic output types are as follows: l i e plant operating conditions; a process trends; j i I e alarms; i l m results of performance calculations; r a operator requests; and a switchyard operating conditions j The types of calculations performed include but are not limited to the following: l l e reactor core performance calculation; e plant performance calculation; is h a plant efficiency; a turbine generator efficiency; General Plant Description - Amendment 1 1.2-37 l l - . ,
{ l \ 25A5113R1v. A l SBWR stamtantsuretyAnalysisReport l 9 l e condenser performance (thermal load and cleanliness); l l m feedwater heater performance; ) l m moisture separator performance; and l l a condensate demineralizer performance. l The function types performed in addition to the calculations include but are not limited l to the following: l m data accumulation; e indication of control rod position; and a surveillance test guide. Power Generation ControlSubsystem The PGCS is a a set of software routines residing on the Process Computer System which j produce control outputs for the automated control sequences associated with plant l start-up, shutdown, and normal power generation. The PGCS receives the same type inputs as described for the PMCS control commands and sends system mode change and set-point change commands to subloop controllers to support the plant automation l features. The automation process is divided into phases corresponding to plant start-up, shutdown, and normal power generation. Each phase is then divided into several break-points, or logical steps in plant opention. Automation proceeds under PGCS control until the end of a break-point division is reached, at which time the operator must confirm that conditions are acceptable before automation sequence can continue. 1.2.2.2.11 Refueling Machine Computer i l The Refueling Machine is designed for automatic operation by a programn.ed l computer operated from a console above the refueling floor. I The computer will control all direct refueling machine movements to any selected core l location through the established XYZ coordinate system.
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( l 1.2.2.2.12 Leak Detection and Isolation System , i r l The Leak Detection and Isolation System (LD&lS) detects and monitors leakage from l the containment, preventing the release of radiologicalleakage from the reactor ; coolant boundary. The system initiates sa r ety isolation functions by closure ofinboard ) and outboard containment isolation valves. I I O 1.2-38 General Plant Description - Amendment 1 l i i
25A5113 Rev. A l SBWR standardsarety Analysis Report l j l [m\ l\v) l i The following functions are provided by the LD&lS: a containment isolation following a LOCA event; a main steamline isolation; e isolation condenser system process lines isolation; a RWCU/SDC System process lines isolation; a Fuel and Auxiliary Pools Cooling System process lines isolation; a reactor component cooling water lines to DW c.oolers isolation; e drywell sumps liquid drain lines isolation; a containment purge and vent lines isolation; a reactor building HVAC air exhaust ducts isolation; a CRD charging and purge water lines isolation; t ' ( )
'v' a fission products sampling line isolation; a momtonng ofidentified and unidentified leakages in the dowell; a monitoring of condensate flow from the drywell air coolers; j u monitoring of the vessel head flange seal leakage; and I m monitoring of valve stems leakages in the containment.
The following leakage detection functions are provided by other plant systems: a monitoring of fission products in the dowell; a monitoring of FMCRD leakage; l m monitoring of plant sump levels and flow rates; and m SRV Steam Discharge. l The LD&IS monitors plant parameters such as flow, temperature, pressure, water level, etc., which are used to alarm and initiate the isolation functions. m At least two parameters are monitored for an isolation function. The signal parameters (V ) are processed by the Safety System and Logic Control system (SSLC) which generates the trip signals for initiation ofisolation functions. l General Plant Description - Amendment 1 1.2-39 i
I 1 25A5113Rcv. A I SBWR standardsarery Analysis neport i l O The LD&lS safety-related functions have four divisional channels of sensors for each parameter. Two+ut-of-four coincidence voting within a channel is required for l initiation of the isolation function. The control and decision logic are of fail-safe design l which assures isolation on loss of power. The logic is energized at all times and de-energizes to trip for isolation function. , Loss of one divisional power or one monitoring channel will not cause inadvertent i:clation of the containment. Different divisional isolation signals are provided to the inboard and outboard isolation valves. The LD&lS is designed to allow periodic testing of each channel to verify it is capable l to perform the intended function. LD&IS is a safety-related system and is classified Seismic Category 1. i I The LD&lS initiates isolation functions automatically. All isolation valves have individual manual control switches and valve position indication in the MCR. However, i the isolation signal overrides any manual control to close the isoladon valves. Manual control switches in the control logic provide a backup to automatic initiation of I isolation as well as capability for reset, bypass and test of functions. The monitored plant parameters are measured and recorded by the Process Corpputer System, and are displayed on demand. The abnormal indications and initiated isolation functions are alarmed in the MCR. 1.2.2.2.13 Safety System Logic and Control System The Safety System Logic and Control (SSLC) System provides the decision logic facility for implementing safety-related logic functions. These functions enable the safety-related systems to perform their plant protection tasks. The SSLC performs the following functione a Sensor channel trip decisions a System coincidence trip decisions (2-out of-4 logic or 2-out-of-3 logic) e Control and interlock logic a ATWS prevention and mitigation j I a Manual division trip and isolation a Division-of-sensors bypass l 1.240 General Plant Description - Amendment 1 l
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25A5113Rev. A
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SBWR stancuesafetyAnalysisReput i LJ l e Division maintenance bypass (division out-of-senice) l a Calibration and self-diagnosis f The SSLC System is configured as a four-division data acquisition and control system, with each division containing an independent set of microprocessor-based, softw2re-controlled logic processors. The four divisions exchange data via fiber optic data links to implement cross-channel data comparison. l The SSLC System acquires data from redundant sets of sensors of the interfacing safety-l related systems and provides control outputs to the final component actuators. Data is i received from the Essential Multiplexing System (EMS) or directly hardwired from transmitters or sensors. 1.2.2.3 Radiation Monitoring Systems , 1.2.2.3.1 Process Radiation Monitoring System The primary functions of the Process Radiation Monitoring System (PRMS) are to: l a monitor and record the various gaseous and liquid process streams and efIluent I( f releases; e initiate alarms in the main control room to warn operating personnel of high radiation activity; and f a initiate the appropriate safety cctions and controls to prevent further radioactisity : releases to the emironment. This PRMS provides both safety-related and non-safety-related instrumentation for radiological monitoring, sampling and analysis ofidentified process and efIluents streams throughout the plant. The process and effluent paths and/or areas as described herein are monitored for potential high radioactivity releases. The radiation monitors of the first six items below j are safety-related Class IE instrumentation, while the remaining of the PRMS monitors are considered non-safety-related provided to monitor plant operations. l 1 m Main steam line (MSL) tunnel area - 2 divisional channels The MSL tunnel area is continuously monitored for high gross gamma radioactisity l in the steam flow to the turbine. Shutdown of the main condenser vacuum pump is s automatically initiated on any 1-out-of-2 channel trip. s i 9eneral Plant Description - Amendment 1 1.2-41
25AS113Riv. A SBWR standardsafety Anarrsis neport 9 m Reactor building safety envelope ventilation exhaust - 4 divisional channels The air vent exhaust from the safety envelope is continuously monitored for gross gamma radioactivity. On high level the containment ventilation ducts are isolated on any 2-out-of-4 channel trip. m Containment purge exhaust - 4 divisional channels The radiation levelin the purge exhaust from the containment is monitored for gross gamma radioactivity. On a high level, the ventilation ducts to the containment are automatically isolated on any 2-out-of-4 channel trip. m Refueling area ventilation exhaust - 4 divisional channels The air vent exhaust from the refueling area is continuously monitored for gross gamma radioactivity. On high level, the ventilation ducts in this area are isolated on any 2-out-of-4 channel trip. m Control room envelope air intake supply- 4 divisional channels The air intake to the MCR envelope area is continuously monitored for gross gamma radioactivity. On high level, the MCR ventilation ducts are isolated and the emergency air circulation system is activated on any 2-out-of-4 channel trip. m Isolation condenser vent exhaust - 4 divisional channels The atmospheric pool area within the confines of each isolation condenser is continuously monitored for gross gamma radioactivity. On high radiation level, the affected isolation condenser is automatically isolated through closure of the steam line and condensate return line isolation valves. m Turbine building ventilation exhaust - I channel The air vent exhaust from the turbine building is continuously sampled through an isokinetic probe and monitored for airborne radioactivity by a beta / gamma sensitive detector and filters for collecting air particulates and iodine. A tritium monitor is also provided for sample collection. Alarms are initiated on high radiation and on abnormal sampling flow. m Charcoal vault ventilation exhaust - I channel The vent exhaust from the charcoal vault is continuously monitored for gross gamma radioactivity that may result from leaks in the charcoal beds. An alarm is initiated on high radiation. 1.2-42 General Plant Description - Amendment 1
25AS113Rzv. A SBWR standardsarety Analysis neport ( a Pre-treated main condenser offgases - 1 channel The pre-treated main condenser offgases are continuously sampled and monitored for gross gamma radioactivity. Alarms are initiated on high radiation and on abnormal sampling flow. Vial sampling is provided for periodic isotopic analysis. m Post treated main condenser offgases - 2 channels The treated offgases are continuously sampled and monitored for airborne radioactivity by two gas samplers and filters for collecting air particulates and halogens. Each gas sampler consists of a beta / gamma sensitive detector and a source check for periodic testing. On high radiation, the offgases are routed through the entire charcoal bed for holdup. On extremely high radiation, the offgas discharge to the stack is automatically isolated. Alarms are initiated on high radiation levels and on abnormal sampling flow. Vial sampling is provided for periodic isotopic analysis. m Plant stack discharge-2 channels
) The discharge through the stack is continuously sampled through an isokinetic (d probe and monitored for airborne radioactivity by two separate channels. Each channel consists of a beta / gamma sensitive detector with a source check and a high-range ion chamber. In addition, filters are provided for collecting air particulate and halogens, and components are provided for collecting and sampling tritium.
Alarms are initiated on high radiation levels and on abnormal sampling flow. m Radwaste building ventilation exhaust-1 channel The air vent exhaust from the radwaste building is continuously sampled through an isokinetic probe and monitored for airborne radioactivity by a beta / gamma sensitive detector with a source check and filters for collecting air particulates and iodine. A tritium monitor is also provided for sample collection. Alarms are initiated on high radiation and on abnormal sampling flow. m Radwaste liquid discharge - 1 channel The liquid waste discharge from the plant is continuously sampled and monitored by a liquid sampler consisting of a scintillation detector and a source check. Alarms are initiated on high radiation levels and on abnormal sampling flow. On extremely high radiation in the discharged waste, the flow is automatically terminated and A inlated. GeneralPlant Description - Amendment 1 1.243
i 25AS113Rcv. A ! SBWR standardsafety Anatrsis Report 9 m Drywell sump liquid discharge - 2 channels,1 channel per sump The liquid discharge from each of the two diywell sumps is monitored by an in-line ion chaniber. On high radiation, the discharge to the radwaste building is terminated and alarmed. m Turbine gland steam condenser discharge - I channel The discharge from the main turbine gland steam condenser is continuously monitored for airborne radioactivity by a digital gamma ventilation detector. An alarms is initiated on high radiation level. m Intersystem radiation leakage - 2 channels, I channel per RCCW System loop t Intersystem leakage into each loop of the Reactor Closed Cooling Water System is monitored by an in-line scintillation detector for gross gamma radioactivity. An alarm is initiated on high radiation. m Reactor building ventilation exhaust -1 channel The air vent exhaust from the reactor building is continuously sampled through an isokinetic probe and monitored for airborne radioactivity by a beta / gamma sensitive detector with a source check and filters for collecting air particulates and iodine. A tritium monitor is also provided forsample collection. Alarms are initiated on high radiation and on abnormal sampling flow. m Fission Products Releases -3 channels The atmosphere in the drywell is sampled and monitored for gross gamma radioactivity resulting from fission products releases. One channel monitors for noble gases, another channel monitors for air particulates, and the third channel monitors for halogens. Alarms are activated in the main control room on high l radiation levels. 1.2.2.3.2 Area Radiation Monitoring System . l The Area Radiation Monitoring (ARM) System continuously monitors the gamma i l radiation levels within the various areas of the plant and provides an early warning to ! operating personnel when high mdiation levels are detected so the appropriate action can be taken to minimize occupational exposure. The ARM System consists of multiple channels which utilize gamma sensitive detectors, l associated digital radiation monitors, auxilian units, and local audible warning devices. ] Each monitor has two adjustable trip circuits for alarm initiation, one high radiation I
- 1.2-44 General Plant Description - Amendment 1 i l
1
25A5113Rev. A SBWR stadudsarerAar sis r nepet a level trip and one downscale trip. Also, each radiation monitor will actuate an alarm on loss of power or when gross equipment failure occurs. The gross gamma radiation levels are monitored on a continuous basis, to signal any change in exposure rates which may be caused by operational transients, maintenance activities, or inadvertent release of radioactivity. Plant operating personnel are warned of any high radiation level by MCR alarms as well as audible area alarms. 4 2 4 The system monitoring range covers a span from 10 Gy to 10 Gy (10-2 mR/hr to 10 R/hr). i This system is non-safety-related. The radiation monitors are powered from the non- l' Class 1E vital 120 Vac source which is available continuously and during loss of site power. The trip alarm setpoints will be established in the field following equipment installation at the site. The exact settings will be based on sensor location, background radiation levels, expected radiation levels, and low occupational radiation exposures. 1.2.2.3.3 Containment Atmospheric Monitoring System p) 1 U The primary function of the Containment Atmospheric Monitoring System (CAMS) is to monitor the atmosphere in the containment for high gross gamma radiation levels and for high concentration levels of oxygen and hydrogen during post-accident conditions. These three parameters are also monitored during normal reactor operations. The atmosphere in the drywell and in the suppression chamber is monitored and sampled by two independent, redundant CAMS subsystems. The CAMS is manually activated during normal plant operation to start the radiation monitoring and gas sampling process. For post-accident monitoring, the CAMS is automatically activated to perform its monitoring functions. The area of sampling can be manually selected or sequentially controlled between the drywell and the wetwell. The CAMS is a two<livision monitoring system comprising two radiation monitoring channels per division and a gas sampling and analyzer rack per division. Radiation , monitoring and gas sampling are provided for the drywell and for the airspace above , the suppression pool. One gamma sensitive ion chamber and one digitallog radiation monitor are used by each radiation monitoring channel. Two channels each for CAMS A & B are provided to monitor radiation levels in the containment. The radiation monitoring range is 10-2 Gy/hr to 10 5 Gy/hr (1 R/hr to 10 7 R/hr). In the post-accident operational mode, the safety function of the the CAMS is to cordinuously sample the oxygen and hydrogen contents in the containment, and { j O' display the results in the main control room. This information is then used by the t l GeneralPlant Description - Amendment 1 1.245-
1 2SAS113R:v. A SBWR standardsafetyAntysisaeron O operator to assess containment integrity and initiate flammability controlif the CAMS indicates the presence of a potentially explosive gas mixture in the containment. Alarms and digital readouts are provided in the MCR for indications of high radiation dosage rates, inoperative radiation moniton, high oxygen levels, high hydrogen levels and of abnormal sampiing for each subsystem. Each gas sampling rack is provided with its own gas calibration sources of known concentration levels to calibrate periodically the oxygen and hydrogen analyzers and senson. Each oxygen and hydrogen gas sampling channelis checked for proper calibration and response at two or more input gas levels, one at the zero gas concentration level and the other at a nominallevel from the calibrated gas sources. CAMS is classified as a safety-related system and Seismic Category 1. Power to each subsystem is provided from uninterruptible Class 1E 120 Vac divisional sources. 1.2.2.4 Core Cooling Systems 1.2.2.4.1 Reactor Water Cleanup / Shutdown Cooling System See discussion in Subsecdon ' .2.2.41 Subsection 1.2.2.6.1. 1.2.2.4.2 Isolation Condenser System The Isolation Condenser System (ICS) removes decay heat after any reactor isolation during power operations. Decay heat removal limits further pressure rises and keeps the RPV pressure below the SRV pressure setpoint. It consists of three independent loops, each containing a heat exchanger that condenses steam on the tube side and transfen heat by heating / evaporating water in the IC/PCC pool which is vented to the atmosphere. The ICS is initiated automatically on either a high reactor pressure, or MSIV closure, or J a Level 2 signal. To start an IC into operation, the me:cr cpected a condensate return l valve is opened whereupon the standing condensate drains into the reactor and the steam-water interface in the IC tube bundle moves downward below the lower headers to a point in the main condensate return line. The ICS can also be initiated manually , by the operator from the MCR. A pneumadc cpected fail-ooen nitrogen oiston- ) onerated condensate return bypass valve is provided for each IC which opens if the ! ! 125 Vdc power is lost, or on reactor water level signal (below L2L l l The ICS is isolated automatically when either a high radiation level or excess flow is I detected in the steam supply line or condensate return line. The IC/PCC pool is divided into ubpect: subcomnartments which are interconnected at their lower ends to provide full use of the water inventory for heat removal by any IC. The IC/PCC pee! is ne: nally coc!ed by 6e F^.PCS. Cooling and cleanun ofIC/PCC 1.246 General Plant Description - Amendment 1 l 1
25AS113Rsv. A SBWR senadsedsareerAn*Irsisnever \ l O) N / , 1 nool water is nerformed by the Fuel and Auxiliary Pools Cooline System (FAPCSL During IC operation, IC/PCC pool water will boil, and the steam produced will be vented to the atmosphere. This boil +ff action of nonradioactive water is a safe means for removing and rejecting all reactor decay heat. The IC/PCC pool has an installed capacity that provides at least 72 hours of reactor decay heat removal canability. The heat rejection process can be continued indefinitely by replenishing the IC/PCC poolinventory. !r _e-ma! make up rf::ent are unav2!!25!e, make up can be pre ided in per: L^CA pee! ezater make up tennecdonc !cratedjust abcve grade !cre! cutide We reac cr bui!d:ng.Th=e Fne are c! =in.ed Q22!!rj C cup C and Se!-!: O:rgerj !. A safety-related indeoendent FAPCS makeuoline is orovided to convev emercency makeuo water into the IC/PCS nool. from ninine connections located at crade level in the reactor vard external to the reactor buildinc. This makeup can be accomplished without any valving changes in the reactor building no matter what the prior operating mode of the FAPCS might have been. The ICS passively removes sensible and core decay heat from the reactor (i.e., heat transfer from the IC tubes to the surrounding IC/PCC pool water is accomplished by , natural convection, and no forced circulation equipment is required) when the normal heat removal system is unavailable following any of the following events: x a Sudden reactor isolation at power operating condidons a Re2cier 'c: Standbj cde a During station blackout (i.e., unavailability of all ac power) for 72 hours a Anticioated Transient Without Scram (ATWS) The ICs are sized to remove post-reactor isolation decay heat with two out of three ICs operating and to reduce reactor pressure and temperature to safe shutdown conditions, in 36 hours. with occasional venting of radiolytically generated noncondensable gases to the suppression pool. Since the heat exchangers (ICs) are independent of station ac power, they will function whenever normal heat removal systems are unavailable, to maintain reactor pressure and temperature below limits. The heat removal capacity of the ICS (with two of three IC loops in service) is at least 60 MWt (each ICS is designed for 30 MWt capacity and is ccmpied of comorises two I identical modules), at a reactor gage pressure of 7.240 MPa (1050 psig) with saturated steam. The portions of the ICS (including isolation valves) which are located inside the 3 containment and on the steam lines out to the IC flow restrictors are designed to ASME 1 l J Code Section Ill, Class 1. Other portions of the ICS are ASME Code Section Ill, Class 2. j The IC pool is safety-related and Seismic Category I. GeneralPlant Description - Amendment 1 1.2-47 1
25A5113R;v. A SBWR standardsafety Analysis Report Periodic surveillance testing of the ICS valves can be performed by the control room ei j operator via remote manual switches that actuate the isolation valves and the condensate return valves. The opening and closure of the valves is verified by their status lights. 1.2.2.4.3 Emergency Core Cooling System - Gravity Driven Cooling System Emergency core cooling is provided by the Gravity-Driven Cooling System (GDCS) in conjunction with the ADS in case of a LOCA. When a Level 1 signal is received, the ADS will depressurize the reactor vessel and the GDCS will inject sufficient cooling water to maintain the fuel cladding temperatures below temperature limits defined in 10CFR50.46. In the event of a severe accident that results in a core melt with the molten core in the lower drywell region, GDCS will flood the lower drywell cavity region with the water inventory of the three GDCS pools and the suppression pool (SP). The GDCS is an Engineered Safety Feature (ESP) System. It is classified as safety-related and Seismic Category 1. GDCS instrumentation and dc power supply are IEEE Class IE. Basic system parameters are: a Number ofindependent divisions: 3 m Initiation signal: confirmed Level 1 signal from NBS
- Type: Sealed-in NBS divisional Level 1 signal
- Number of channels: 4 m Time delay between inidadon and actuation for short-term water injection:
- 150 seconds a Time delay between initiation and actuation for long-term water injection:
- 30 minutes
- Permissive: Interlocked to RPV water level s (TAF + 1.0m) e Squib valve firing logic: 2-out-of-3 m Manual actuation:
- No. of channels: 4
! - Permissive: Interlocked to RPV low pressure signal l 1.248 General Plant Description - Amendment 1 l i l
25AS113Rw. A SBWR standedsarnyAnalysis nepar O
- Logic: Simultaneous operation of two switches of the same disision The GDCS injects water into the downcomer annulus region of the reactor after a LOCA and reactor vessel depressurization. It provides short-term gravity-driven water makeup from three separate water pools located within the upper drywell at an elevation above the active core region. The system also provides long-term post-LOCA makeup from the suppression pool to meet long-term core decay heat boil-off requirements. During severe accidents the system floods the lower drywell region with water if the core melts through the RPV. ,
The GDCS is completely automatic in actuation and operation. A backup to automatic actuation is the ability to actuate by operator action. The GDCS is composed of three identical divisions completely independent of each other both electrically and mechanically. A confirmed RPV Level 1 signal will actuate the ADS to reduce RPV pressure. Simultaneously,150-second short-term system timers, [ and 30-minute long-term system timers in the GDCS logic are started, which, after time-out, actuate squib valves providing an open flow path from the respective water sources to the vessel. 0\ The short-term system supplies gravity-driven flow to six separate nozzles on the vessel (/ with suction flow from three separate GDCS pools. The long-term system supplies gratity-driven flow to three other nozzles with suction flow from the suppression pool through equalizing lines. t Both the short-term and long-term systems are designed to ensure that adequate reactor vessel inventory is provided assuming a LOCA in one division and failure of one squib valve to actuate in the second division. Three GDCS deluge lines, each having one squib actuated valve, provide a means of flooding the lower drywell cavity in the event of a core melt sequence which causes failure of the lower vessel head and allows molten fuel to reach the lower drywell casity floor. These squib activated valves are driven by logics receiving input signals from an array of temperature sensors located in the lower drywell. 1.2.2.5 Reactor Servicing Equipment 1.2.2.5.1 Fuel Service Equipment The refueling and fuel handling platform are also included and are outlined in . Subsection 1.2.2.5.5. Senicing tools and equipment are not safety-related. ) i [h U GeneralPlant Description - Amendment 1 1.2-49
2SAS113Rsv. A SBWR standantsareryAurysis neport O FuelPrep Machine One fuel prep machine is mounted against the west wall of the spent fuel storage pool. Its primary use is to inspect spent fuel when submerged in the storage pool and to aid in reconstitution of fuel found to be defective. New FuelInspection Stand The inspection stand is mounted in a pit next to the new fuel storage vault. The pit allows inspection of the two fuel bundles over their full length. Channeling is also performed with the aid of the channel handling tool. ChannelBolt Wrench A long handled socket-end wrench used in the assembly or disassembly of the channel from the fuel bundle, by insertion or removal of the attaching bolt, while channeling new fuel or reconstituting spent fuel. ChannelHandling Tool A long handled clamping tool used to engage the channel for removal. It is manually operated and suspended from the auxiliary hoist orjib crane. Vacuum Sipper Used in the spent fuel pool to detect gasses from defective fuel. GeneralPurpose Grapple A general use grapple primarily for handling fuel when using any fuel handling equipment 1.2.2.5.2 Miscellaneous Service Equipment This equipment is generally used independently of other senicing equipment. Equipment requirements are that they operate underwater to a depth of 33 meters. The equipment is designed to be quickly decontaminated and can be stored with a minimum of manpower. Underwater Lights Three types oflights are used: a general area light, a local area light, and a drop-type light. Viewing Aids Three types of viewing aids are used. A floating type viewing aid is the simplest. Another aid features an under water viewing tube with a 15-60 power telescope. The last is an j underwater, remotely controlled television camera with an internal light source. j Underwater Vacuum Cleaner The underwater vacuum cleaner is used to clean any pool Door underwater and is remotely seniceable while submerged. 1.2-50 General Plant Description - Amendment 1
25A5113Rsv. A SBWR standardsafery Analysis neport
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1.2.2.5.3 Reactor Pressure Vessel Servicing Equipment These tools are used when the reactor is shut down and the RPV head is being removed or installed. Tools used consist of stroncbacks. nut racks. stud tensioners. orotectors. wrenches. etc. Lifting tools are designed for a safety factor of 10 or better with respect to the ultimate strength of the material used. Carbon steel equipment must be either hard chrome plated, parkerized or coated. Tools are designed for 60-year life in the working environment. General-Tools C a . . A 11F,n n e b %,.D..nnn C uA Tb,n,A T. L.:."e b m.m . n : n cl . . A n, C +. . A I I, n A 1: n ~ Tm m1
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.. A :, b General Plant Description - Amendment 1 1.2-51
i 25A5113R;v. A SBWR standardsafety Analysis Report O r Pe 'ir"ng lug: c- tN c=! cre engaged The tendone- has the capabi!!:y e ret 253;3g. l 4.e ve=! :tuds and :s:ts, i l 1.2.2.5.4 RPV Internals Servicing Equipment i instrument Strongback The instrument strongback is used to aid in handling and replacement of Power Range Neutron Monitoring (PRNM) aM Startup Range Neutron Monitoring (SRNM) dry l tubes, in conjunction with support from the instrument handling tool. Instrument Handling Tool The instrument handling tool is connected to the wire terminal of the auxiliary hoist of j the refueling platform and receives LPRMs or dry tubes from the strongback. I 1.2.2.5.5 Refueling Equipment The reactor building fuel handling floor is serviced with a fuel handling platform, Refueling machine, and an auxiliary platform. Refueling Machine
- The refueling machine is a gantry-type crane which spans the reactor vessel cavity and l fuel and storage pools to handle fuel and perform other ancillary tasks. It is equipped l with a traversing trolley on which is mounted a telescoping tubular mast and integral l fuel grapple. An auxiliary hoist is also provided. The machine is a rigid structure built to precise engineering standards to ensure accurate and repeatable positioning during the refueling process. A programmed computer located above the refueling floor controls the operational movements.
FuelHandling Platform Although similar in appearance and size to that ofits counterpart, the fuel handling platform is only used for fuel senicing and transporting tasks. It is equipped with a trolley and telesming grapple and is manually operated. Mechanical stops and J l interlocks provide the necessary operational limits. Auxillary Platform The auxiliary platform is a low-profile structure having its own track located on the fuel l handling area floor. A removable section with mounted wheels is loweitd to the reactor [ vessel flange level on which a special portable track is installed. Its primary purpose is to l aid in open vessel senicing. 1.2.2.5.6 Fuel Storage Facility
- New and spent fuel storage facilities are required for fuel and associated equipment.
Storage in wet or dry conditions depends on the item in storage. 1.2-52 General Plant Description ~ Amendment 1
i l 25AS113Rtv. A SBWR stamtardswery Aurysis neport a New FuelStorage New fuel storage racks are aluminum and are constructed for floor mounting. For dry vault storage the racks are loaded from the top,while those in pools are side loaded. The l storage vault capacity is 19% of core load while the storage pool is 39% core load. Spent FuelStorage Spent fuel storage racks are of stainless steel laminate construction with neutron absorbing material. This ensures that a full array (285% of full core) or loaded spent ! fuel will remain subcritical by 5% of Ak, under all conditions. Adequate water shielding is always maintained in storage pools by the use oflevel sensors. Dry vaults on the other hand have drains to assure they are maintained dry. All storage pools are constructed with stainless steel liners to form a leak-tight barrier. A leak detection system monitors liner integrity. The thermal-hydraulic design of the rack provides sufficient natural convection cooling flow to remove 19,9T "'/ bund!r (68,000 B:u/'/ bund!r) er decay heat without exceeding 100'C. [] 1.2.2.5.7 Under-Vessel Servicing Equipment b The primary functions of the under-vessel senicing equipment are to: a install and remove fine motion control rod drives (FMCRDs); e install and remove FMCRD packing sections and motors; a make connections to neutron detectors; a provide senicing tools; and a p nide a work platform and CRD handling equipment Under-VesselPlatform The under-vessel platform provides a working surface for personnel and equipment to the entire under-vessel area. This requires 360 rotational capability. The platform also provides the facility for operation of the FMCRD handling machine for the automatic removal of the FMCRDs. 1.2.2.5.8 FMCRD Maintenance Area The FMCRD maintenance area is designed and equipped to perform FMCRD maintenance related activities, including decontamination of the FMCRD components, [ acceptance testing, and storing spare drives. Maintenance tasks use a combination of manual and remote operations to reduce radiation exposure to plant personnel and to reduce contamination of surrounding equipment during operation. GeneralPlant Description - Amendment 1 1.2-53
25A5113Rxv. A SBWR standardsareryAnalysis neport 1 O The FMCRD maintenance area is located in a shielded room near the dr)well
- equipment entry door.The layout of the room permits a convenient and eflicient sequencing of work while reducing exposure to personnel.
1.2.2.5.9 Fuel Cask Cleaning Spent fuel cask cleaning is performed in two different areas of the plant. Spent fuel cask cleaning is performed at the receiving area in the reactor building if required to remove surface dirt accumulated during transportation. It is also performed in the cask pit following loading of spent fuel, under thejurisdiction of health physics personnel. The receiving area of the plant has facilities for: I a checking the cask for contamination; j s cleaning the cask of road dirt; e inspection of the cask for damage; a attachment of the casklifting yoke: a removal of head bolts and attachment of head lifting cables; and a raising the cask to the refueling floor using the main building crane. The cask pit area includes: m- A deep drainable pit with gate access to the storage pool for underwater cask l loading. m An underwater area for the storage of the cask head and lifting yoke, a An area for high pressure cleaning and decontamination. This area is accessible for chemical and hand scrubbing, refastening the head, and for smear tests. 1.2.2.5.10 Fuel Transfer System The fuel is removed from the reactor and transported through the pool gate into the transfer pool where it is seated in the fuel basket of the transfer machine. The basket is then conveyed along the transfer pool west wall. The fuel handling platform will then grapple the fuel and place it in the spent fuel storage pool. From vessel removal to storage in the spent fuel pool, the fuel bundle is handled in the vertical position. O: 1.2-54 GeneralPlant Description - Amendment 1
25AS113Rw. A SBWR standetsweryAnalysisurat V 1.2.2.5.11 Inservice inspection Equipment The SBWR typically uses a wide range ofinservice inspection equipment much of which . is equipment and materials used in performance of visual, surface and volumeuic examinations required by the ASME Code, Section XI. m Automated ultrasonic scanning equipment using multiple angle beam and straight beam transducers may be employed for volumetric examination of areas such as reactor pressure vessel welds and nozzle inner radii. The data from the automated i I examination is typically stored on optical disk or other appropriate recording media for subsequent computer-assisted data analysis. m Manual ultrasonic examinadon equipment may be employed to supplement the l automated exnnination if necessary or to perform the volumetric examination of areas such as ASME Class 2 vessel welds and nozzle inner radii. Manual ultrasonic examination equipment consists of an ultrasonic instrument containing analog or digital oscilloscope-style display and hand-held transducers. Where more than one angle beam examination is required due to the Class 2 vessel wall thickness, l additional manual scans may be performed using ultrasonic transducers adjusted 9 for the required angles of examination. Class 1 and 2 piping welds may be examined volumetrically using either computerized, automated ultrasonic scanning i equipment or using manual ultrasonic examination equipment. s Surface examinations of ferritic vessels and piping may be performed using the magnetic particle examination method with either prod or yoke type equipment. The magnetic particles may be either dry or may be in a wet suspension and may be either fluorescent or colored for viewing in visible light. Surface examinations of non-magnetic vessel and piping welds may be performed using either fluorescent or visible dye liquid penetrant materials.When fluorescent magnetic particles or liquid I penetrant materials are used, portable ultraviolet lights are used for siewing. l i a Eddy-current probe coils driven by automated scanning devices with computerized ! data acquisition systems may be substituted for surface examinations where the component configuration or radiation conditions render other surface ! examination techniques impractical or undesirable, a Visual examinations of Class 1 and 2 bolting and component supports and attachments on Class 1,2 and 3 piping and components may be conducted directly using simple aids such as mirrors and magnifying glasses. m Remote visual examination equipment may be used for examination ofinterior smfaces of the reactor vessel and other components. Rigid fixtures are sometirnes used as an aid in performance of the remote reactor visual examinations. General Plant Description - Amendment 1 1.2-55 l
1 2SAS113Rev. A SBWR standardsafety Anarrsis Reput 1 O it is anticipated there will be continuing technological advances in inservice inspection. 1 As these improved technologies become available and proven, they will be applied (as : appropriate) to inspection of the certified design. l l 1.2.2.6 Reactor Auxiliary Systems i 1.2.2.6.1 Reactor Water Cleanup / Shutdown Cooling System The Reactor Water Cleanup / Shutdown Cooling (RWCU/SDC) System has the following primary functions: a Purifies reactor coolant during normal operation and shutdown. m Transfers sensible and core decay heat produced when the reactor is being shutdown oris in the shutdown condition. m Provides decay heat removal and high pressure cooling of the primary coolant during periods of reactor isolation (hot standby). m Implements the overboarding of excess reactor coolant during startup and hot standby. m Maintains coolant flow from the reactor vessel bottom head to reduce thermal stratification. m Warms the reactor coolant prior to startup and hydrotesting. The system consists of two redundant tntins. Each train includes a pump powered and controlled by an adjustable speed ddve (ASD), two regenerative heat exchangers (RHX), one single-shell non-regenerative heat exchanger (NRHX), one radial-bed-type low-pressure-drop resin-bed demineralizer, an electric heater, and associated valves and pipes. The RWCU/SDC System is classified as a non-safety-related system except for its RCPB j and containment isolation functions which are safety-related and is thus Seismic j Category I and Class IE. The electrical power supplies to the two trains are from l separate electrical divisions. The system can be connected to non-safety-related standby l ac power (diesel generators). l l During normal plant operation, the system operates at reduced flow in the cleanup l i mode continuously withdrawing water from RPV. The water is cooled through the heat exchangers and is circulated by the pump to the demineralizer for removal of impurities. Purified water returns to the RHX where it is reheated, and then flows into the feedwater lines and is returned to the RPV. One train is in operation while the other is in standby. 1,2-56 General Plant Description - Amendment 1
25AS113 Rsv. A SBWR sisadardsainyAnso rsis neport
!a Redundant trains permit shutdown cooling if only one train is operable. The cooldown time will be extended when using only one train. In the event ofloss of preferred power and the most limiting single active failure, this mode of operation brings the RPV to a
$100 C (s212 F) cold shutdown condition in 36 hours in conjunction with operation of the Isolation Condensers. The RWCU/SDC provides the shutdown cooling capability to satisfy the following reactor coolant temperature reduction schedule:
a 60 C (140 F) in 24 hours a 54.4 C (130 F) in 40 hours a 48.9 C (120 F) at the completion of flooding the reactor well from sS5 C (s95*F) water sources During hot standby and startup, excess water resulting from CRD System purge water , injection and expansion during plant heatup is dumped, or overboarded, to the main condenser or the radwaste system to control reactor water level. l The RWCU/SDC System maintains the temperature difference between the reactor dome and the bottom head drain to less than 80.6 C (145 F) to preclude excessive ( thermal stratification. Flow rate, pressure, temperature and conductivity are measured, recorded or indicated, and alarmed if appropriate,in the MCR. l Pumps are provided with interlocks for the automatic operation and with switch and , status indication for manual operation from the MCR. Motor-operated isolation valves , are automatically and manually actuated with automatic closure overriding manual opening signals. 1.2.2.6.2 Fuel and Auxiliary Pools Cooling System The FAPCS performs pool water cooling, purification, and distribution (i.e., pool filling i and, where applicable, draining) for the fc!!c c? ng fuel and auxiliary pools in the reactor building. m Sper: fue! c:cmge peel; a Fue! cd pit; a Fue! tm=fer pee!; I s Ne<' fuel :tomge poc!; t
\ m D_ enter .e!!;
i GeneralPlant Description - Amendment 1 1 2-57 l
25AS113Riv. A SBWR standardsuretyAnarrsis neport O
. em= =-
m Ec!aden condenser (!C/PCC) peels; a C=ity DrNen Cee!!ng Sy-ter (CDCS) (3 pee!:); and a Supprezien pec!. In addition, the FAPCS performs the following system functions: a Reacter cce!! drai?ng and '"Fng; a Drywell spray; a Suppression chamber spray; and a Low Pressure Coolant Injection (LPCI) of suppression pool water into the RIN. The FAPCS is a low gauge pressure system which has two trains of components. Each train includes a pump, a heat exchanger and a filter-and-demineralizer water treatment unit. At each end, the trains are tied together by a four-valve bridge of motor-operated 4 valves, which are aligned to perform the system functions. The FAPCS has features to ' prevent radioactive contamination of the IC/PCC pool with untreated water from other pools. One train of the FAPCS normally operates continuously to cool, clean and clarify the water of the spent fuel storage pool. The other train is in standby or may be pe rforming periodic cooling / cleaning of one of the other pools. FAPCS provides sufIicient flowmte ! I and cooling capability to keep the spent fuel pool bulk water temperature at or below 48.9 C (120 F) for normal plant operations and normal spent fuel pool heat load conditions. With conditions associated with a full core off-load and irradiated fuel in the spent fuel pool for 10 years of plant operations, the FAPCS maintains the bulk temperature at or below 60 C (140 F). The same capability remains if there is a single failure in the FAPCS. FAPCS operation is manually controlled and monitored from the MCR. The water treatment is locally controlled with status feedback to the MCR. Automatic operation applies to sequential logic for start /stop of pumps and lineup of the valves to assure the selection of cooling / cleanup train configuration and system operating modes. The operator is able to override the automatic control and take manual control. Containment isolation initiation has priority over normal operation of the system and is controlled by the LD&lS. O 1.2-58 GeneralPlant Description - Amendment 1
25A5113R:v. A SBWR standantsareryAnalysis neport b The system contains instrumentation for sensing and transmitting water levels, water temperatures, water flow and water pressure. The FAPCS is a non-safety-related system with the exception of containment isolation and the independent cafety-related makeup water piping providing makeup water to the isolation conder.ser (IC/PCC) pools and independently to the spent fuel pool. The piping and compenents directly interacting with safety-related systems meet the classification of safety and seismic class required by these other systems. A detailed descrintion of the FAPCS. including a listine of all nools seniced bv FAPCS as well as an exolanation of the conducts of system onerations. is civen in Section 9.11 1.2.2.7 Controf Panels 1.2.2.7.1 Main Control Room Panels The main control room panel (MCR) is comprised of an integrated set of operator interface panels (e.g., main control console, large display panel). The safety-related panels are seismically qualified and provide grounding, electrical independence and
, phyncal separation between safety divisions and between safety divisions and non-safety- ;
related components and wiring. lG} The main control room panels and other MCR operator interfaces are designed to provide the operator with information and controls needed to safely operate the plant in all operating modes, including startup, refueling, safe shutdown, and maintaining the plant in a safe shutdown condition. Human factors engineering principles have been incorporated into all aspects of the SBWR MCR design. 1.2.2.7.2 Radwaste Control Room Panels The liquid and solid radwaste systems are operated from control panels in the radwaste control room. Programmable controllers are used in this application. They are not safety-related. 1.2.2.7.3 Local Control Panels and Racks Local panels, control boxes, and instrument racks are provided as protective housings and/or support structures for electrical and electronic equipment to facilitate system operations at the local level. They are designed for uniformity using rigid steel structures capable of maintaining structural integrity as required under seismic and plant dynamic conditions. The term " local panels" includes local control boxes.
,-s Local panels and racks used for plant protection systems are classified as safety-related.
They are located in areas in which there are no potential sources of missiles or pipe (V) breaks that couldjeopardize modules from more than one division. Each safety-related General Plant Description - Amendment 1 1.2-59
I 25AS113RGv. A l SBWR standardsafety Analysis neport l l panel / rack is Seismic Category 1, qualified, and provides grounding, and electrical l independence and physical separation between safety divisions and non-essential components and wiring. Electrical power to divisional panels / racks is from ac or dc power sources of the same
- division as that of each panel / rack itself. Power to the non-essential panels / racks is from the non-essential ac and/or dc sources.
1.2.2.7.4 Essential Multiplexing Subsystem The Essential Multiplexing Subsystem (EMS) provides distributed data acquisition and control networks to support the monitoring and control of the plant standby safety systems. The EMS comprises electrical devices and circuitry, such as local multiplexing units (LMUs), fiber optic transmission lines, and control room multiplexing units (CMUs), that acquire data from remote process sensors and discrete monitors located within the plant and multiplex the signals to SSLC equipment. SSLC provides decision logic that trips the final actuators of driveu equipment associated with safety systerns. The EMS is divided into four divisions of equipment, each with independent control of data acquisition, multiplexing, and control output functions. System timing is asynchronous among the four divisions. No common clock signal is transmitted among the divisions of multiplexing and no timing signals are exchanged. Both analog and discrete sensors are connected to LMUs in local areas, which perform signal conditioning, analog-to<ligital conversion for centinuous process inputs, change-of4 tate detection for discrete inputs, and message formatting prior to signal transmission. The LMUs are limited to acquisition of sensor data and the output of control signals. Trip decisions and other control logic functions are performed in SSLC processors. The LMUs transmit serial, time-multiplexed data streams representing the status of the plant variables to the SSLC logic processing equipment. Data transmission is also made over dual redundant channels to the main control room. The CMUs demultiplex the data and prepare the signals for use in interfacing monitoring systems such as the process computer or display controllers. The CMUs also receive safety-related signals from control room equipment for transmission to the LMUs and SSLC. EMS design features automatic self-test and automatic reconfiguration after failure of one channel (either a cable break or device failure). If an LMU or CMU has failed, that unit will be removed from service. Faults and their location are annunciated to the operator in the MCR. Data can be transferred to non-safety-related systems for control or display through isolating fiber-optic data links and buffering devices (gateways or bridges, if required). Data transfer is made such that failures on the non-safety side cannot inhibit operation of safety-related logic functions. Data cannot be transmitted from the non-safety side to EMS. 1.2-60 GeneralPlant Description - Amendment 1
i l l l 25A5113Rtv. A SBWR StandardSafety Analysis Report l 1 , !%J The EMS is capable of data transfer at rates suflicient to satisfy the system time response requirements of safety system functions. Data throughput capability is up to 100 megabits per second. ! The EMS starts and runs automatically upon application ofsystem power, regardless of l the sequence in which power is applied to individual controllers. EMS and SSLC automatically establish communications by detection of correct message passing. Logic is provided to prevent equipment activation outputs from occurring until stable plant sensor data and interlock permissive data are being received. Loss of power causes a controlled transition to a safe-state without transients occurring that could cause inadvertent initiation or shutdown of driven equipment. j EMS equipment is classified as safety-related, Class lE, and is Seismic Category I. The EMS includes test facilities in the MCR that will monitor data transmission to l ensure that data transport, routing, and timing specifications are accurate. Bit error rate of each EMS network shall be better than 1 error in 10'3. Out-of-tolerance parameters detected on-line for a particular input signal will result in an inoperative condition for j that input into the trip logic processors of SSLC. lQ l 1.2.2.7.5 Non-Essential Multiplexing System The Non-Essential Multiplexing System (NEMS) is the data communication portion of all control systems in the plant that are not pan of the shutdown control systems. The NEMS in non-safety-related. The NEMS equipment is designed and constructed using state-of-the-art fiber optics communications equipment and computer controls which perform the following: I s Transfer via the NEMS to control system equipment, in digital format, analog or , binary data that has been collected and digitized from remote transmitters, contact l closures, and other sensors located throughout the plant. I m Transfer from the main control room via the NEMS to control system equipment, in digital format, processed activation signals for the control of remote desices such as pumps, valves, or solenoids. l { m Exchange self-test data between local equipment and control room equipment for the reporting of NEMS and control system component malfunctions. m Communicate requests to the main control room for the reporting of NEMS and ! ('N control system equipment component malfunctions. t /
%.)
General Plant Description - Amendment 1 1.2-41
25AS113 Rev. A SBWR standardsanrryAnalysisnepois O The NEMS has no access to the safety-related data base; however safety-related data can be read by the NEMS on the optically-isolated memory portions of the Essential Multiplexing System (EMS) Local Multiplexing Units (LMU). This data can be read by any NEMS multiplexing units that is configured to do so. The NEMS cannot write data to any portion of the EMS. The NEMS consists of two types ofmultiplexing units: Local Multiplexing Units (LMU), and Control Room Multiplexing Units (CMU) connected via fiber optic cables. The NEMS also includes network gateways which allow transfer of data between data highway systems. Throughout the plant, LMUs are located in local plant areas to acquire sensor data and transmit this data to the any equipment that requires it. The LMUs also receive processed signals from the control room for command of control system actuatos. CMUs are located in the control room to transmit and receive data for the logic processing units of the plant control systems. All interconnections are fiber optic data links. Within each NEMS highway, the system uses redundant links for greater reliability. There are a number of NEMS highway systems that are routed throughout the plant. These systems all have CMUs located in the main control room. Gateways connect the multiple NEMS highway systems to allow for transfer of data between NEMS highway systems. 1.2.2.8 Nuclear Fuel 1.2.2.8.1 Nuclear Fuel Ee! design for 6e SEWR S*2ndrd-Plant-*s-not-within We ccope of Se cerdEed desigm It is intended that the specific fuel to be used in any facility which has adopted the certified desigu be in compliance with U.S. NRC approved fuel design criteria. This strategy is intended to permit future use of enhanced / improved fuel designs as they become available. However, this approach is predicated on the assumption that future fuel designs will be extensions of the basic fuel technology that has been developed for boiling water reactors. Key characteristics of this established BWR fuel technology are: m Uranium oxide based fuel pellets; a Zirconium-based (or equivalent) fuel cladding; a All material selected on the oasis of BWR operating conditions; a Multi-rod fuel bundles in an N lattice; and 9! 1.2-62 General Plant Description - Amendment 1 l
2SA5113Rsv. A SBWR standardsaferyAnairsissepar , b I ~ s Fuel bundle inlet o-ificing to control bundle flow rates, core flow distribution, and reactor coolant hydraulic characteristics. The SBWR design provides a Loose Parts Monitoring System (LPMS) aimed at ! protecting the fuel against the potential effects ofloose parts entrained in the reactor coolant flow, A discussion of the LPMS is included in this section. l The following is a summary of the principal requirements which must be met by the fuel
~
j supplied to any facility utilizing the certified design. I 1 GeneralCriteria l l a NRC-approved analytical models and analysis procedures are applied. j i e New design features are included in lead test assemblies. a The generic post-irradiation fuel examination program approved by NRC is j maintained. l Thermal-Mechanical ,
/' '
The fuel design thermal-mechanical analyses are performed for the following conditions: ; , i a Either worst tolerance assumptions are applied or probabilistic analyses are j performed to determine statistically bounding results (i.e., upper 95% confidence). ! s Operating conditions are taken to bound the conditions anticipated dudng normal steady 4 tate operation and anticipated operational occurrences. The fuel design evaluations are performed against the following criteria: a The fuel rod and fuel assembly component stresses, strains, and fatigut life usage j are evaluated to not exceed the material ultimate stress or strain and the thermal fatigue capability. e Mechanical testing is per rormed to ensure that loss of fuel rod and assembly component mechanical integrity will not occur due to fretting wear. e The fuel rod and assembly component evaluations include consideration of metal
- thinning and any associated temperature increase due to oxidation and the buildup of corrosion products to the extent that these influence the material properties and structural strength of the components.
f a The fuel rod internal hydrogen content is controlled dudng manufacture of the ( fuel rod consistent with ASTM standards. l General Plant Description - Amendment 1 1.243 l i l
2SA5113 Rtv. A SBWR standantsafety Analysis neport 9 m The fuel rod is evaluated to ensure that fuel rod bowing does not result in loss of fuel rod mechanical integnty due to boiling transition. m Loss of fuel rod mechanical integrity will not occur due to excessive cladding pressure loading. e The fuel assembly (including channel box), control rod and CRD are evaluated to assure control rods can be inserted when required. These evaluations consider the effect of combined safe shutdown earthquake (SSE) and LOCA loads. m Loss of fuel rod mechanical integrity will not occur due to cladding collapse into a fuel column axial gap. m Loss of fuel rod mechanical integrity will not occur due to pellet <ladding mechanicalinteraction. Nuclear a A negative Doppler reactivity coeflicient is maintained for any operating condition. m A negadve core moderator void reactivity coeflicient resulting from boiling in the active flow channels is maintained for any operating conditions. m A negative moderator temperature reactivity coefficient is maintained above hot standby. m For a super prompt critical reactivity insenion accident originating from any operating condition, the net prompt reactivity feedback due to prompt heating of the moderator and fuel is negative. m A negative power reactivity coefficient, as determined by calculating the reactivity change, due to an incremental power change from a steady-state base power level, is maintained for all operating power levels above hot standby. m The plant meets the cold shutdown margin requirement, a The effective multiplication factor for fuel designs stored under near.al and abnormal conditions is shown to meet fuel storage limna oy demonstrating that the peak uncontrolled lattice k-infinity calculated in a normal reactor core configurations meets the limits for the storage racks. Hydraulic Flow pressure drop characteristics are included in the calculation of the operating limit minimum critical power ratio (MCPR). I 1.2-64 General Plant Description - Amendment 1
k I 25AS113Rn. A SBWR stamtadsareryAnnar sisnepar l Because of the channeled configuration of BWR fuel assemblies, there is no bundle-to- f bundle cross-flow inside the core, and the only issue of hydraulic compatibility of various bundle types in a core is the bundle inlet flow rate variation and its impact on margin-to-thermal limits. The coupled thermal-hydraulic-nuclear analysis performed to i determine fuel bundle flow and power distribution uses the various bundle pressure loss coefficients to determine the flow distribution required to maintain a total core pressure drop boundary condition to be applied to all fuel bundles. The margin to the thermal limits of each fuel bundle is determined using this consistent set of calculated bundle flow and power. l Loose Parts Monitoring System The Loose Parts Monitoring System (LPMS) is designed to provide detection ofloose , metallic parts within the RPV. Detection ofloose parts can provide the time required to L avoid or mitigate safety-related damage to or malfunctions of primary system j components. The LPMS detects structure-borne sound that can indicate the presence j i ofloose parts impacting against the RPV internals. The system alarms when the signal amplitude exceeds preset limits. The LPMS detection system can evaluate some aspects i of selected signals. However, the system by itself will not diagnose the presence and } location of a loose part. Review of LPMS data by an experienced LPM engineer is required to confirm the presence of a loose part. ( The LPMS continuously monitors the RPV and appurtenances for indications ofloose parts. The LPMS consists of sensors, cables, signal conditioning equipment, alaiming monitor, signal analysis and data acquisition equipment, and calibration equipment. The alarm setting is setlow enough to meet the sensitivity requirements, yet is designed to discriminate between normal background noises and the loose part impact signal to minimize spurious alarms. The array of LPMS sensors consists of a set of sensor channels that are strategically mounted on the external surface of the primary pressure boundary at various elevations and azimuths at natural collection regions for potential loose parts. General mounting locations are at the (a) main steam outlet nozzle, (b) feedwater inlet nozzle, and, (c) control rod drive housings. The online system sensitivity is such that the system can detect a metallic loose pan that weighs between 0.25 lb to 30 lbs and impacts with a kinetic energy of 0.5 ft-lb on the inside surface of the RPV within 3 feet of a sensor. The LPMS frequency range of interest is typically from 1 to 10 kHz. Frequencies lower than 1 kHz are genemlly associated with flow induced vibration signals or flow noise. The LPMS includes provisions for both automatic and manual start-up of data acquisition equipment with automatic activation in the event the preset alert level is s reached or exceeded. The system also initiates an alarm to the control room personnel when an alert condition is reached. General Plant Description - Amendment 1 1.2-65
l l 25A5113 Rev. A t SBWR sondant Safety Analysis Report l @ 1.2.2.8.2 Fuel Channel Fuc! cMnne! de:ign for de SBWR is not-wiSin Se =cpe of 6e certiSed design. It is
- intended that the specific fuel channel to be used in any facility which has adopted the certified design be in compliance with U.S. NRC approved fuel channel design criteria.
This strategy is intended to pennit future use of enhanced / improved fuel channel designs as they become available. However, this approach is predicated on the l assumption that future fuel channel designs will be extensions of the basic technology that has been developed for boiling water reactors. The key characteristic of this established BWR fuel channel technology is the use of zirconium-based (or equivalent) i t fuel channels which preclude cross-flow in the core region. l l The following is a summ:ny of the prinopal requirements which must be met by the fuel l channel supplied to any facility using the certified design: a The material of the fuel channel shall be shown to be compatible with the reactor environment. l a The channel will be evaluated to ensure that channel deflection does not preclude l control rod drive operation. m The effects of channel bow will be included in the fuel rod critical power enluations. 1.2.2.8.3 Control Rod Centrol red de:ig : fer ie SEMT i not -iiin Se =cpe of $e certiEed design. It is intended that the specific control rod to be used in any facility which has adopted the certified design be in compliance with U.S. NRC approved control rod design criteria. This strategy is intended to permit future use of enhanced / improved control rod designs as they become available. However, this approach is predicat'e d on the
- assumption that future control rod designs will be extensions of the baic technology
( that has been developed for boiling water reactors. Key characteristics of this established BWR control rod technology are: a Control rods perform dual functions of power distribution shaping and reactivity control. : i l l m The control rod has a cruciform cross 4ectional envelope shape. ' s The control rod has a coupling at the bottom for attachment to the CRD. m The control rod has an upper bail handle for transporting. l 1.2-66 General Plant Description - Amendment 1
25A5113Rsv. A SBWR standantsanyAnssrsis neport A V a The cruciform cross section contains neutron poison materials which are either contained within or as part of the control rod stmcture.
]
The following is a summary of the principal requirements which must be met by the control rod supplied to any facility utilizing the certified design: a The control rod stresses, strains, and cumulative fatigue shall be evaluated to not exceed the ultimate stress or strain of the material. e The control rod shall be evaluated to be capable ofinsertion into the core during all modes of plant operation within the limits assumed in the plant analyses. l s The material of the control rod shall be shown to be compatible with the reactor emironment. e The reactivity worth of the control rod shall be included in the piant core analyses. m Lead Surveillance program shall be implemented if a change in design features such as new absorber material or stomural material not previously used in reactor cores could impact the function of the coneol rod. ( 1.2.2.9 Radioactive Waste Managemem System 1.2.2.9.1 Liquid Waste Management System The Liquid Waste Management System (LWMS) collects, monitors, and treats liquid radioactive waste for plant reuse whenever practicable. The LWMS consists of the following six subsystems: a Equipment (Iow conductivity) drain subsystem; l s Floor (high conductivity) drain subsystem; e Chemical drain subsystem; l s Detergent drain subsystem; l s Mobile systems interface subsystem; and a Mixed waste subsystem. The LWMS processing equipment is located in the radwaste building. Any discharge is such that concentrations and quantities of ridioactive material and other contaminants J
) are m accord with applicable local, state, ar d federal regulations.
GeneralPlant Description - Amendment 1 1.2-67 I
BSAS113Rcv. A SBWR stamtantsarety Analysis neport ; 4 O All potentially radioactive liquid wastes are collected in sumps or drain tanks at various I locations in the plant. These wastes are transferred to collection tanks in the radwaste building. Waste processing is done on a batch basis. Each batch is sampled as necessary in the collection tanks to detcrmine concentrations of suspended solids and chemical l contaminants. Equipment drains and other low-conductivity wastes are treated by filtration, uv/ ozone, demineralization and are transferred to the condensate storage tank for reuse. Floor drains and other high conductivity wastes are treated by filtration and ion exchange prior to being either discharged or recycled for reuse. Laundry drain wastes and other detergent wastes oflow activity are treated by filtration, sampled and released via the liquid discharge pathway. Chemical wastes are treated by filtration, sampled and released from the plant on a batch basis. Protection against inadvertent release ofliquid radioactive waste is prosided by design redundancy, instrumentation for the detection and alarm of abnormal conditions, automatic isolation, and administrative controls. Connections are provided for mobile processing systems that could be brought in to augment the installed waste processing capability. Connections for addition of a permanent evaporation subsystem are provided in the event that site conditions warrant. Mixed waste will be segregated from the other types of radioactive waste for packaging. If the liquid is returned to the plant, it meets the purity requirements for condensate makeup. If the liquid is discharged, the activity concentration is consistentwith the discharge cdteria of 10CFR20 and dose commiunent in 10CFR50, Appendix 1. 1.2.2.9.2 Solid Waste Management System The Solid Waste Management System (SWMS) is designed to control, collect, handle, process, package, and temporadly store prior to shipment solid radioactive waste generated as a result of normal operation, including anticipated operational occurrences, that includes filter backwash sludges and bead resins generated by the LWMS, RWCU/SDC, FAPCS, and condensate system. Contaminated solids such as High Efficiency Particulate Air and cartridge filters, rags, plastic, paper, clothing, tools, and equipment are also processed in the SWMS. There is no liquid plant discharge from the SWMS. The SWMS consists of the following four subsystems: a Wet solid waste collection subsystem; a Wet solid waste processing subsystem; e Dry solid waste processing subsystem; and l 1.248 General Plant Description - Amendment 1 I
2SAS113Rsv. A l l ( SBWR standantsehny Ansor sesnepar l e Mobile systems interface subsystem. Spent bead resin sluiced from the RWCU/SDC System, FAPCS, condensate and LWMS are transferred by the wet solid waste collection subsystem to one of two spent resin tanks for decay and storage. t The wet solid waste processing subsystem consists of a built-in dewatering station. A l High Integrity Container (HIC) is filled with either sludges from the phase separator or ] bead resin from the spent resin tanks. Spent cartridge filters may also be placed in the l ! HIC. Dry wastes consist of air filters, miscellaneous paper, rags, etc., from contaminated , areas; contaminated clothing, tools, and equipment parts that cannot be effectively j decontaminated; and solid laboratory wastes. The activity of much of this waste is low enough to permit handling by contact. These wastes are collected in containers located in appropriate areas throughout the plant. The filled containers are sealed and moved to controlled-access enclosed area for temporary storage. i Connections are prosided for mobile processing systems that could be brou ght ill to augment the installed waste processing capability. j Connections for addition of a permanent solidification subsystem are p ~.Wd m the , event that site conditions warrant. ! Temporary storage for over one month's volume of packaged waste is provided in the radwaste building. Packaged waste includes high integrity containers, compactor boxes, l shielded filter containers, and 55-gallon drums as necessary. The SWMS is designed to package the radioactive solid waste for off-site shipment and ! burial, in accordance with the requirements of applicable NRC and DOT regulations, including Regulatory Guide 1.143,10CFR61,10CFR71, and 49CFR170 through 178. ! 1.2.2.9.3 Gaseous Waste Management System The function of gaseous waste management system is to minimize and control the ;
- release of radioactive material into the atmosphere by delaying, filtering, or diluting )
l various offgas process and leakage gaseous releases which may contain the radioactive 4 l isotopes of krypton, xenon, iodine, and nitrogen. The Offgas System (OGS) is the principal gaseous waste management subsystem. The various bui lding HVAC systems perform other gaseous waste functions. ! The OGS provides for holdup and decay of radioactive gases in the offgas from the i steamjet air ejector (SJAE) and consists of process equipment along *f;ith monitoring ] I instrumentation and control components. 1 1 i General Plant Descriptian - Amendment 1 1.2-69 l
2SA5113 R1v. A SBWR standardsarety Anarrsis neport O The OGS design minimizes the explosion potential in the offgas process stream through recombination of radiolytic hydrogen and oxygen under controlled conditions. Although the OGS is non-safety-related, it is capable of withstanding an internal hydrogen explosion and is designed to ASME Code Section Vill-Division I and the ANSI B31.1 Piping Code. The OCS includes redundant hydrogen / oxygen catalytic recombiners and ambient temperature charcoal beds to provide for process gas volume reduction and radionuclide retention / decay. The system processes the SJAE discharge during plant startup and normal operation before discharging the air flow to the plant stack. The charcoal beds can operate in three different modes: a Bypass - all flow bypasses the beds (used during startup); e Guard bed -all flow passes through the guard bed only; and e Adsorber beds- all flow passes through the guard bed and then through parallel pairs of adsorber beds. 1.2.2.10 Power Cycle 1.2.2.10.1 Turbine Main Steam System The turbine Main Steam (MS) System conveys steam generated in the reactor to the turbine. It also prosides steam to the steamjet air ejectors, the turbine gland seals, the deaeration section of the main condenser, and the turbine bypass system. System boundaries are from after the outermost containmentisolation valves up to the turbine stop valves. The MS System is not required to effect or support safe shutdown of the reactor or to perform in the operation of reactor safety features; however, the MS System is designed: a to comply with applicable codes and standards to accommodate operational stresses , such as internal pressure and dynamic loads without risk of failures and I consequential releases of radioactivity in excess of the established regulatory limits; a to accommodate normal and abnormal environmental limits; l l m to assure that failures of non-Seismic Category I equipment or structures, or pipe l cracks or breaks in high or moderate piping in the MS with.ot preclude ftmctioning
- of safety-related equipment or structures in the plant; and i
n with access to permit insenice testing and inspections. l l 1.2 70 GeneralPlant Description - Amendment 1
25AS113Rsv. A SBWR standedsenyAnsIrsisaeron The TMSS main steam piping consists of two lines.The header arrangement upstream of the turbine stop valves allows them to be tested on-line with minimum load reduction ; and also supplies steam to the power cycle auxiliaries, as required, j 1.2.2.10.2 Condensate and Feedwater System The Condensate and Feedwater System (C&FS) consists of the piping, valves, pumps, heat exchangers, controls and instrumentation and the associated equipment and subsystems which supply the reactor with heated feedwater in a closed steam cycle utilizing regenerative feedwater heating. The C&FS extends from the main condenser outlet to the second feedwater isolation valve outside of containment. l The C&FS provides a dependable supply of high quality feedwater to the reactor at the required flow, pressure and temperature. The condensate pumps take the deaerated l condensate from the condenser hotwell and deliver it through the SJAE condenser, the ! ! gland steam condenser, the condensate demineralizer and through a string of four low. ; pressure feedwater heaters to the reactor feed pump suction. The reactor feed pumps discharge through two high pressure feedwater heaters to the reactor. Turbine extraction steam is used for a total of six stages of closed feedwater heating. The drains lQ from each stage of the feedwater heaters are cascaded through successivelylower lQ pressure feedwater heaters to the main condenser. The condensate portion of the C&FS has three motor-driven, constant speed centrifugal pumps, each rated at 33% to 60% of total system-rated flow, i The feedwater portion of the C&FS has three pumps operating in parallel, each rated I
- at 33% to 60% of total system rated flow with adjustable speed motor drives with variable frequency power supplies.
- The C&FS does not serve or support any safety function and has no safety design basis.
l Failure of this system cannot compromise any safety-related systems or prevent safe shutdown. Portions of the system that are radioactive during operation are shielded with access control forinspections. Leakage is minimized with welded construction used wherever practicable. ! Relief discharges and operating vents are channeled through closed systems. The majority of the C&FS piping is located within the turbine building which contains j no safety-related equipment or systems. The portion which connects to the second l h V isolation valve outside the containment is located in the steam tunnel in the reactor building. This portion of the piping is analyzed for dynamic effects from postulated events and SRV discharges. GeneralPlant Description - Amendment 1 1.2-71 l l
2SAS113Rxv. A SBWR sandardsarety Analysis neport O The entire system piping is analyzed for waterhammerloads that could potentially result from anticipated flow transients. 1.2.2.10.3 Condensate Purification System The Condensate Purification System (CPS) continuously purifies and treats the condensate as required to maintain reactor feedwater purity, using filtration to remove solid corrosion products, ion exchange to remove condenser leakage and other dissolved impurities, and water treatment additions to minimize corrosion / erosion product releases in the power cycle. The CPS does not serve or support any safety function and has no safety design basis. It is designed to Quality Group D standards. Vent gases and other wastes from the CPS are collected in controlled areas and sent to the radwaste system for treaunent and/or disposal. The CPS is located in the turbine building, and piping or equipment failures will not affect plant safety. 1.2.2.10.4 Main Turbine The main turbine is a tandem compound, two flow,52-inch last-stage bucket with one high pressure (HP) turbine and one low pressure (LP) turbine. The steam passes through an in-line 'nigh velocity moisture separator (HVS) prior to entering the LP turbine. Steam exhausted from the LP turbine is condensed and degassed in the condenser. The turbine uses steam at an atmospheric pressure of 6.79 MPa (985 psia) from the reactor and rotates at 1800 RPM. Steam is bled off from each turbine and is used to heat the feedwater. The steam and power conversion system is designed to operate at 105% of maximum guaranteed turbine throttle flow for transients and short-term loading conditions. Turbine Overspeed Protection System In addition to the normal speed control function provided by the turbine control system, a separate turbine overspeed protection system is included to minimize the possibility of turbine failure and high energy missile damage. The following component redundancies are employed to guard against overspeed: l s Main stop valves /controlvalves; i s Intermediate stop valves / intercept valves (CIVs); a Primary speed control / backup speed control; s Fast acting solenoid valves / emergency trip fluid system (ETS); and l 1,2-72 General Plant Description - Amendment 1 l f
2SAS113 Rn. A SBWR standardSafdy Analysis Repon O U s Speed control /overspeed uip/ backup overspeed trip. . The TG System is enclosed within the turbine building, which contains no safety-related equipment or structures. The turbine generator is orientated within the turbine building to be inline with the reactor building to minimize the potential for any high energy TG System generated missiles from damaging any safety-related equipment or structures. 1.2.2.10.5 Turbine Gland Seal System The Turbine Gland Seal System (TGSS) provides steam and prevents the escape of radioactive steam from the turbine shaft / casing penetrations and valve stems and , prevents air in-leakage through subaunospheric turbine glands. The TGSS consists of a sealing steam pressure regulator, sealing steam header, a gland steam condenser, two full capacity exhaust blowers and associated piping, valves and instrumentation. The TGSS is non-safety-related system and is designed to Quality Group D standards. The HP turbine shaft seals must accommodate a range of turbine shell pressure from full vacuum to approximately 220 psia. The LP turbide shaft seals operate against a i vacuum at all times. The g!and seal outer portion steam air mixture is exhausted to the gland steam condenser via the seal vent annulus (i.e.., end glands), which is maintained at a slight vacuum. The radioactive content of the naling steam, which eventually exhausts to the plant vent and the atmosphere, makes a negligible contribution to overall plant radiation release. In addition, the auxiliary steam system is designed to provide a 100% backup to the normal gland seal process steam supply. A full capacity gland steam condenser is provided and equipped with two 100% capacity blowers. The TGSS effluents are first monitored by a system dedicated continuous radiation monitor installed on the gland steam condenser exhauster blower discharge. High monitor readings are alarmed in the MCR. 1.2.2.10.6 Turbine Bype .s System A Turbine 1spass System (TBS) is pro ided which passes steam directly to the main condenser under the control of the pressure regulator. Steam is bypassed to the condenser whenever the reactor steaming rate exceeds theload permitted to pass to the turbine generator. The TBS has the capability to shed 40% of the turbine generator rated load without reactor trip or operation of a SRV. The pressure regulation system provides main turbine control valve and bypass valve flow demands so as to maintain a nearly constant reactor pressure during normal plant operation. The TBS does not serve or support any safety-related function and has no safety design. General Plant Description - Amendment 1 1.2 73
1 25A5113 R:v. A SBWR standardsarery Analysis neport O Both automatic and manual control of the turbine bypass valves are provided. The turbine bypass valves are opened by a signal received from the SB&PC System whenever the actual steam pressure exceeds the preset steam pressure by a small margin. This occurs when the amount of steam generated by the reactor cannot be entirely used by the turbine. This bypass demand signal opens the first of the individual valves. As the bypass demand increases, additional bypass valves are opened, dumping the steam to the condenser. Pressure-reducing odfices are located at the condenser connections, and sparger piping distributes the steam within the condenser. The bypass valves are equipped with fast-acting servo valves to allow rapid opening of bypass valves upon turbine trip or generator load rejection. The bypass valves automatically trip open upon load rejection or turbine trip. The bypass valves automatically trip closed whenever the vacuum in the main condenser falls below a preset value and/or insufficient circulating water flow exists. The bypass valves also fail closed on loss of electrical power or hydraulic system pressure. 1.2.2.10.7 Main Condenser The main condenser is designed to condense and deaerate the exhaust steam from the main turbine and provide a heat sink for the Turbine Bypass (TB) System. The main condenser dc= ne: =ce e Suppen ary :9:y fumden and has no safety design basis. It is, however, designed with necessary shielding and controlled access to protect plant personnel from radiation. The main condenser is a single-shell type deaerating unit with this shell located directly beneath the low pressure turbine. The shell has tube bundles through which circulating water flows. The condensing steam is collected in the condenser hotwells (the lower shell portion) which provide suction to the condensate pumps. Since the main condenser operates at a vacuum, any leakage is into the shell side of the main condenser. Tubeside or circulating water inleakage is detected by measuring the conductivity of sample water extracted beneath the tube bundles. In addition, conductivity is continuously monitored at the discharge of the condensate pumps and alarms are provided in the MCR. In all operational modes, the condenser is at vacuum and consequently no radioactive releases can occur. Loss of vacuum sequentially leads to a control room alarm, turbine trip and eventually bypass and MSIV closure to prevent condenser overpre'.sudzation. Ultimate overprotection is provided by rupture diaphragms on the turbine exhaust hoods. i l l l 1.2-74 General Plant Description - Amendment 1 l l l
i l 25A5113 R5v. A SBWR standardsafety Analysis Report l l l The instrumentation and control features that monitor the performance to ensure that ! the condenser is in the correct operating mode include: l a Hotwell water level- Automatically controlled within preset limits. Dming normal ! full load operation with nominal hotwell levels, the main condenser provides a four-minute active condensate storage volume and has a two-minute surge capacity. At minimum normal operating hotwell water level, and normal full load condensate
- flow rate, the condenser provides a two minute minimum holdup time for N-16 decay.
m Condenser pressure - Key overall performance indicator that initiates alarms and trips at preset levels. m LP turbine exhaust hood temperature - Automatically initiates turbine exhaust water sprays to protect the turbine. m inlet and outlet circulating water temperature -Monitors performance only. m Conductivity within the condenser and at the discharge of the condensate pumps n - Initiates alarms at preset levels.
)
- V The potential for flooding from the main condenser is less than that from the Circulating Water System (CWS) so only the CWS flooding protection is needed. The Condenser pressure indicators are located above any potential flood level.
Spray pipes and baflies are designed to protect the main condenser internals from ingh energy flow inputs. Hydrogen buildup during operation is prevented by continuous evacuation of the main condenser. Hydrogen sources are excluded during shutdown. Noncondensable gases are removed from the power cycle by the Main Condenser Evacuation System (MCES). The MCES removes power cycle noncondensable gases including the hydrogen and oxygen produced by radiolysis of water in the reactor and exhausts them to the OCS during plant power operation, or to the turbine building ventilation system exhaust during early plant startup. The MCES establishes and maintains a vacuum in the condenser by the use of steamjet air ejectors during power operation, and by a mechanical vacuum pump during early startup. The system consists of two 100% capacity, double stage, steamjet air ejector (SJAE) units complete with intercondenser, for power plant operation, and a mechanical vacuum pump for use during startup. The last stage of the SJAE is a noncondensing stage. One SJAE unit is normally in operation and the other is on standby. NJ General Plant Description - Amendment 1 1.2-75
l 2SAS113 R2. A SBWR standardsarery Analysis neport l 9 The steamjet air ejector is placed in senice to remove the gases from the main condenser after a pressure of about 11 to 22 cm IIg (5 to 10 inches Hg) absolute is established in the main condenser by the mechanical vacuum pump and when sufficient nuclear steam pressure is available. Steam supply to the second stage ejector is maintained at a minimum specified flow to ensure adequate dilution of the hydrogen to prevent the offgas from reaching the flammable limit of hydrogen. 1.2.2.10.8 Circulating Water System The Circulating Water System cooling towers are not part of the SBWR standard scope. A conceptual design is used for reference. The conceptual SBWR design uses a hyperbolic natural draft cooling tower. The Circulating Water System (CWS) provides cooling water for removal of the power cycle waste heat from the main condensers and transfers this heat to the power cycle heat sink, which is the cooling tower. The tower has a basin underneath it to collect the cooled water. The circulating water pumps are in the intake structure adjacent to the tower, and takes suction from the basin. The CWS does not serve or support any safety function and has no safety design basis. To prevent flooding of the turbine building, the CWS will automatically isolate in the event of gross system leakage. The circulating water pumps are tripped and the pump and condenser valves are closed in the event of a system isolation signal from the condenser area high-high level switches. A condenser area high level alarm is prosided in the MCR. A reliable logic scheme is used (e.g.,2-out-of-3 logic) to minimize potential for spurious isolation trips. The CWS consists of the following components: s Intake screens located in a screen house; a Pumps; a Condenser water boxes; a Piping and valves; a Tube-side of the main condenser; and a Water-box fill and drain subsystem. O 1.2-76 General Plant Description - Amendment 1
25AS113Rsv. A SBWR standantsafetyAnalysis naron (3 b , 1.2.2.11 Station Auxiliaries 1.2.2.11.1 Makeup Water System The Makeup Water System (MWS) demineralizes water from the station water system, stores it, and transfers it to plant water systems and supply points. The demineralization subsptem consists of cartridge filters, and a reverse osmosis and filter package. The storage and transfer system includes an outdoor makeup water , storage tank, and two redundant transfer pumps. The system is housed in and ; controlled from the water treatment building. System components in contact with the demineralized water are stainless steel. The storage tank is freeze protected. The MWS is non-safety-related. 1.2.2.11.2 Condensate Storage and Transfer System The Condensate Storage and Transfer System (CS&TS) stores condensate grade water and transfers it to plant water systems and supply points. End users include the main condenser hotwell, CRD system, RWCU/SDC system fill, FAPCS fill, suppression and GDCS pools fill, C&FS fill, and liquid and solid radwaste system flushing. ( '( The CS&TS includes a storage tank and transfer pumps. Components in contact with the condensate are stainless steel. The storage tank has a floadng stainless steel cover and is freeze protected. A wall is built around the tank to ensure the entire tank contents is contained if there is a leak.The system is non-safety-related. 1.2.2.11.3 Reactor Component Cooling Water System The Reactor Component Cooling Water System (RCCWS) cools reactor auxiliary equipment including the Reactor Building Chilled Water System, the Dzywell Cooling System, the RWCU/SDC non-regenerative heat exchangers, the FAPCS heat exchangers, and several local air coolers. , The RCCWS has two trains. Each train has two pumps, a heat exchanger, a head tank, and a chemical addition tank. The RCCWS heat exchangers are cooled by the Plant l Service Water System. Except for containment isolation, the RCCWS is non-safety-related and Seismic Category NS. 1.2.2.11.4 Turbine Component Cooling Water System The Turbine Component Cooling Water System (TCCWS) cools turbine building tO auxiliary equipment including turbine lube oil coolers, offgas condensers, generator l
- U stator and hydrogen coolers, and the instrument and service air compressors. The l TCCWS is non-safety-related. l l
GeneralPlant Description ~ Amendment 1 1.2-77 \ l
25A5113Rev. A SBWR stamtantsafety Analysis Report O 1.2.2.11.5 Chilled Water System The Chilled Water System (CWS) is made up of the Reactor Building Chilled Water System (RBCWS) and the Main Control Room Chilled Water System (MCRCWS). The RBCWS provides chilled water to the air handling units in the clean area, controlled area, and refueling area ventilation systems, the access area and change room recirculation air conditioning units, and is a backup to the RCCWS for the drywell air coolers. The MCRCWS provides chilled water to the main control room air handling units. The RBCWS and MCRCWS each have two trains. Each train has a packaged water chiller unit with local control panel, pump, head tank, air separator, and shared chemical feed tank. The RBCWS condensers are cooled by the RCCWS and the MCRCWS condensers are air cooled by electric fans. A chilled water system provides chilled water for the turbine and radwaste buildings. The CWS is non-safety-related and Seismic Category NS. 1.2.2.11.6 Oxygen injection System The Oxygen Injection System (OIS) adds oxygen to the condensate to suppress , corrosion and corrosion product release in the C&FS. The oxygen supply consists of high pressure gas cylinders or a liquid tank. A condensate injection module is provided with pressure regulators, piping, valves and controls to depressurize the gaseous oxygen and route it to the injection modules. 1.2.2.11.7 Plant Service Water System The Plant Service Water System (PSWS) cools the RCCWS and TCCWS heat exchangers. The PSWS cooling towers are not in the SBWR standard scope. A conceptual design using cooling towers for the auxiliary heat sink is used for reference purposes. The reference design for the PSWS consists of two mechanical draft cooling towers, basins, and two 100% capacity trains (50% capacity during shutdown cooling operation). Each train has two 50% capacity vertical wet pit pumps, and duplex strainers. A drain pump is also included for draining the RCCWS heat exchangers to the PSWS basin. The towers are the multiple cell type with a twmpeed reversible fan. Mechanical and electrical isolation of the cooling towers allows maintenance during full power operation. Makeup to the basins is from the station water system. The basins are normally interconnected but can be separated from each other for maintenance by using gates. Blowdown is by gravity to the natural draft cooling tower basin. The PSWS is non-safety-related and Seismic Category NS. 1.2-78 GeneralPlant Duription - Amendment 1
25A5113Rsv. A
- SBWR standardSafety Analysis Report o
1.2.2.11.8 Service Air System The Senice Air System (SAS) provides air for general plant use via senice outlets, filter backwashing, tank sparging, and the plant breathing air system. It also sen'es as a backup to the Instrument Air System (IAS). The system consists of two 50% (maximum) capacity trains each with an intake air filter, compressor, aftercooler, moisture separator, and an air receiver. The breathing air subsystem includes a breathing air purifier package and an air receiver. The system is non-safety-related and Seismic Category NS. 1.2.2.11.9 Instrument Air System The Instrument Air System (IAS) supplies dry, oil-free compressed air for plant instrumentation, control systems, and pneumatic valve actuators in the various plant buildings. It consists of two 100% capacity trains each with an intake air filter, compressor, aftercooler, moisture separator, air receiver, and air dryer package. The system is non-safety-related and Seismic Category NS. 1.2.2.11.10 High Pressure Nitrogen Supply System i v The High Pressure Nitrogen Supply System (HPNSS) supplies clean dry, oil-free high pressure nitrogen gas through piping from the Containment Atmospheric Control System (CACS) to meet the requirements of the main steam system SRVs, ADS accumulators, and MSIVs, instruments and pneumatic valves using nitrogen in the containment. Normally the CACS supplies nitrogen gas; however, when this pressure is l lost, the CACS is isolated and the HPNSS then supplies nitrogen from its bottle racks. This system is non- safety-related and Seismic Category NS except for safety-related penetrations, and isolation valves. These components are safety-related, and Seismic l l Category I. The SRV ADS accumulators and piping are part of the Nuclear Boiler System. L 1.2.2.11.11 Auxiliary Boiler System The Auxiliary Boiler System (ABS) supplies steam for heating of the Hot Water System l (HWS) when extraction steam is not available, turbine gland sealing during startup and as a backup during normal operation, warming of the offgas preheater, and evaporation l l ofliquid nitrogen for containment inerting. The system consists of a package boiler and steam distribution piping and salves. It is non-safety-related. I i\ General Plant Description - Amendment 1 1.2-79 l
2SA5113Riv. A i SBWR standardsafety Analysis Report 9 1.2.2.11.12 Hot Water System The Hot Water System (HWS) supplies hot water for building heating. The system has two heat exchangers, two circulating pumps, and a head / surge tank. The auxiliary boiler is used to heat the water. The system supplies ventilating systems in the reactor building, turbine building, and radwaste building. It is non-safety-related. 1.2.2.11.13 Hydrogen Water Chemistry System The Hydrogen Water Chemisuy (HWC) System is used, along with other measures, to reduce the likelihood of corrosion failures which would adversely affect plant availability. The function of the HWC System is to reduce the dissolved oxygen in the reactor water by the addition of hydrogen to the feedwater. This reduction has been demonstrated to be highly effective in the mitigation of the potential for intergranular stress corrosion cracking (IGSCC) of sensitized austenitic stainless steels. The concentration of hydrogen and oxygen in the main steam line and eventually in the main condenser is altered during HWC System operation. This leaves an excess of hydrogen in the main condenser that would not have equivalent oxygen to combine with in the OGS. To maintain the process offgas nearer its normal constituent balance, the IIWC injects a flow of oxygen upstream of the recombiner. The HWC System is composed of hydrogen and oxygen supply systems, systems to inject hydrogen into the C&FS and oxygen into the OGS and subsystems to monitor the effectiveness of system operation. The HWC System is non-safety-related. It is required to be safe and reliable, consistent with the requirement of using hydrogen gas. 1.2.2.11.14 Post-Accident Sampling System The post-accident sampling system (PASS) consists of sample holding rack, sampling rack, sample conditioning rack, local control panel, and shielding casks. All valves for PASS operation are operated remotely. The sampling sptem isolation valves are operated fr om the main control room and all other valves are operated from the local control panel. After the sample vessel has been isolated and removed, the piping is flushed with demineralized water. The sample holding rack has an enclosure around the sample vessel to contain any leaks ofliquids or gases. The liquids drain to the radwaste system and the gases go to the reactor building exhaust system. The PASS isolation valves are connected to a reliable source of power that will be available starting at least one hour after a LOCA or ATWS event. The isolation valves 1.2-80 General Plant Description - Amendment 1
1 25A5113 Rsv. A sawa ,t -s ,t l q i b have Class IE power and the panels and other equipment are powered with two offsite power supplies. Gas samples are obtained from a sample line connected to the containment atmosphere monitoring system. A vacuum pump is provided to transfer the gas sample from a l sample holding rack to a sampling rack. ; The upper limit for activity levels in liquid and gas samples are: 4 a liquid samples 3.7 x 10 MBq/g (1 Ci/g) a gas samples 0.37 MBq/g (10 Ci/g) Means to reduce radiation exposure are provided such as, shielding, remotely operated j valves, and sample transporting casks. 1.2.2.11.15 Process Sampling System The Process Sampling System (PSS) collects liquid and gas samples for analysis and provides the information required to mcnitor plant and equipment performance and p changes to operating parameten. The system samples all principal fluid process streams and consists of permanently installed sampling nozzles and sample lines, sampling l \j i panels with analyzers and associated equipment, and provisions for local grab sampling. l l The system is non-safety-related. 1.2.2.11.16 Freeze Protection The Freeze Protection System provides insulation, steam, and electrical heating for all l external tanks and piping that may freeze during winter weather. This system is not part l of the SBWR standard design and is provided here for reference purposes. l l l 1.2.2.11.17 Iron injection System ; l The Iron Injection System consists of an electrolytic iron ion solution generator and i equipment to inject the iror. solution into the feedwater system in controlled amounts. 1 l 1.2.2.12 Station Electrical System 1.2.2.12.1 Electrical Power Distribution System Onsite power is supplied from either the plant turbine generator, utility power grid, or ; an off-site power source depending on the plant operating status. During normal operation, plant loads are supplied from the main generator through the unit auxiliary n transformers. A generator breaker allows the unit auxiliary transformers to stay ( connected to the grid to supply loads by backfeeding from the switchyard when the turbine is not online. l l l i GeneralPlant Description - Amendment 1 1.2-81 l
i 25A5113Rcv. A SBWR standardsafery Analysis neport l l 9 The isolated phase bus connects the main generator to the generator breaker, on to the main transformer, and over to the unit auxilia y transformers. The unit auxiliary transformers power the metal clad 6.9 kV switchgear via the non-segregated phase bus. This switchgear powers some large loads and load centers consisting of 6.9 kV/480 V transformers and associated metal clad switchgear. The design includes four Class IE 480 Vac motor control centers (MSCCs)that supply the Class 1E battery chargers and provide backup power to the Vital ac power supply. l Six individual voltage regulating transformers supply 120 Vac non-safety-related control and instrument power. Grounding The electrical grounding system is comprised of: a an instrument grounding network for grounding ofinstrumentation and computer systems; e an equipment grounding network for grounding electrical equipment (e.g., switchgear, motors, distribution panels, cables, etc.) and selected mechanical components (e.g., fuel tanks, chemical tanks, etc.); e a lightning protection network for protection of structures, transformers and other equipment located outside buildings; and a a plant grounding grid. All grounding networks are insulated from each other and separately grounded to the plant grounding grid outside the structures. All grounding networks and equipment are low resistance grounded except the main generator, the emergency diesel generators, and the CTG,which are high resistance grounded to maximize availability. All components requiring grounding are identified and provided with grounding connections. 1.2.2.12.2 Direct Current Power Supply The Class IE dc power supply provides power to the Class 1E vital ac buses through inverters, and to 125 Vdc loads required for safe shutdown. Each of the four divisions of Class IE dc power is separate and independent. The dc systems operate ungrounded (with ground detection circuitry) for increased reliability. Each division has a 125 Vdc battery and a battery charger fed from its divisional 480 Vac Motor Control Center (MCC). This system is designed so that no single failure in any division of the 125 Vdc system will prevent safe shutdown of the plant. I 1.242 General Plant Description - Amendment 1 l
i 25A5113Rev. A SBWR StandardSafety Analysis Report ,
/N )
i :V During a total loss of off-site power, the Class 1E system is powered automatically from two non-Class IE standby diesel genemtors. If these are not available, each division of Class 1E isolates itself from the non-Class 1E system, and power to safety-related loads is provided unintermpted by the Class 1E batteries. The batteries are sized to power safety-related loads for a 72 hour period. The Class 1E dc power supply is designed to permit periodic testing for operability and functional performance to ensure that the full operational sequence transfers power and brings the system into operation. i j Non-Class 1E dc power is supplied through four non-Class 1E 480 Vac MCCs in the same manner as the Class 1E dc power (Subsection 1.2.2.3.1). Each of the two load groups receives power from two of the non-Class IE MCCs. One MCC in each group provides power to a 125 Vdc bus through a batten charger. A 125 Vdc stadon battery prosides backup to the supply from the battery charger. The second MCC in each, group provides power to a 250 Vdc bus through a battery charger. A 250 Vdc station battery provides backup to the supply from the battey l charger. ,f I l The two non-safety-related dc buses also supply power to the non-safety-related dc-to-ac inverter discussed in Subsecdon 1.2.2.12.2. i 1.2.2.12.3 Standby AC Power Supply ( The non-safety-related Standby ac Power Supply consists of two diesel generators. Each ! diesel generator (DG) provides 6.9 kVac power to one of the two load groups whenever the main turbine generator and the normal preferred off-site power source are not l operating. When operating, the standby ac power supply provides power to safety-i related loads and to non-safety-related investment protection loads. Other non-safety-related loads are not powered from the standby power source. The 6.9 kV permanent plant buses are normally energized by either the main generator or the normal preferred offsite power source. Should these power supplies fail, their supply breakers will trip and the standby power supply (diesel generators) will be automatically signalled to start. After the standby voltage and frequency reach normal values, the standby supply breakers will close. After bus voltage is reestablished, large l motor loads will be sequentially started. Each of the two DGs will start and reach full speed and voltage within two minutes after receiving a start signal. In addition, the DGs will sustain full loads within another 65 l, seconds. These delays are acceptable since most loads are non-safety-related. Vital (% safety-related loads are powered by the Standby ac Power Supply; however, these loads l General Plant Description - Amendment 1 1.2-83 l
2SAS123Ry. A SBWR standardsarety Anstrsis neport O are powered by UPS (for ac loads) or safety-related dc power from Gass 1E stadon batteries when normal, preferred, or standby power is not available. 1.2.2.12.4 Vital (Uninterruptible) Power Supply The Gass IE vital ac power supply provides redundant, reliable power to the safety logic and control functions during normal, upset and accident conditions. Each of the four divisions of this Qass IE vital ac power is separate and independent. Each division is powered from an inverter supplied from a Class 1E dc bus. The dc bus receives its power from a divisional battery charger and battery. Provision is made for automatic switching to an alternate Class 1E non-vital supply in case of failure of the l inverter. 1.2.2.12.5 Instrument and Control Power Supply The Instrument and Control Power Supply provides 120 Vac single phase power to instrument and controlloads that do not require an uninterruptible power source. 1.2.2.12.6 Communication System The Communications System includes a dial telephone system, a power-actuated paging facility, a sound-powered telephone system, and an in-plant radio system. 1.2.2.12.7 Lighting Power Suppiv The lighting systems inclu& the normal, standby, emergency, and security lighting l systems. The normal lighting system provides illumination under all normal plant j conditions, including maintenance, testing, and refueling operations. It is powered by preferred ac from the unit auxiliary non-safety-related buses. The standby lighting system supplements the normallighting system and also supplements the emergency lighting system in selected area of the plant. The standby lighting system is normally supplied power from preferred ac power or, alternately, from the on-site standby diesel-l generators. Both lighting systems are non-safety-related. Upon loss of the normal lighting system, the emergency lighting system prosides illumination throughout the plant and, particularly, areas where emergency operations are performed (e.g., main control room, battery rooms, local control stations, ingress / egress routes). It includes self-contained dc battery-operated units for exit and stair lighting. The system supplies at least 108 lux (10 foot-candles) oflighting in those areas of the plant where emergency operations could require reading printed materials or instrument scales. In other area it provides illumination levels adequate for safe ingress or egress. Inside the main control room, emergency lighting is integrated with standby lighting. l l General Plant Description - Amendment 1 1.2-84
25AS113Rsv. A SBWR studardsareryAnssysis nepar V The emergency lighting is normally supplied by preferred ac powered or, alternately, the onsite standby diesel-generators. If these sources are not available, the system (excluding self-contained battery units) is supplied by Class 1E battedes through Class 1E inveners. Excluding the self-contained battery lighting units, the emergency lighting system is safety-related. The security lighting system provides lighting for the secudty center, selected security , areas, and the outdoor plant perimeter. The system is normally supplied power by 7 preferred ac or, alternately, by the on-site standby diesel-generators. The security lighting system is further backed up by a dedicated security standby diesel-generator and a dedicated uninterruptible power supply. The security lighting system is non- ; safety-related. 1.2.2.13 Power Transmission This is not part of the reference SBWR scope. Interface requirements are established i for off-site power transmission. 1.2.2.14 Containment and Environmental Control Systems 1.2.2.14.1 Containment System lQ The SBWR containment, centrally located in the reactor building, features the same basic pressure suppression design concept previously applied in over three decades of BWR power generating reactor plants.The containment consists of a steel lined reinforced concrete containment structure fulfilling its design basis as a fission product barrier even at the increased pressure associated with a postulated pipe rupture. l Main features include the upper and lower drywell surrounding the RPV and a ! suppression chamber containing the suppression pool that serves as a heat sink during abnormal operations and accidents. The containment is constructed as a stepped right cylinder set on the reactor building's l reinforced concrete base mat. The drywell design conditions are 379 kPa gauge (55 psig) and 171*C (340 F). The suppression chamber design conditions are 379 kPa j i gauge (55 psig) and 121 C (250 F). l l The drywell comprises two volumes: an upper drywell volume surrounding the upper portion of the RPV and housing the steam and feedwater piping, the SRVs, GDCS pools, main steam drain piping and upper drywell coolers; and a lower drywell volume surrounding the lower portion of the RPV, housing the FMCRDs, neutron monitodng system, equipment platform, lower drywell coolers and two drywell sumps. The drywell top opening is enclosed with a steel head removable for refueling operations. l NY General Plant Description - Amendment 1 1.245
2SAS113Riv. A ! SBWR stamtantsaretyAntysis anat i 9 l The gas space above the suppression pool serves as the LOCA blowdown reservoir for the upper and lower drywell nitrogen and noncondensables which pass through the eight dr>well-to-supression chamber vertical vents, each with 3 horizontal vents located below the suppression pool surface. The suppression pool water serves as the heat sink to condense steam released into the dowell during a LOCA or steam from SRV actuations. Access into the upper and lower dowells is provided through a double scaled personnel lock and also an equipment hatch. The equipment hatch is removable only during ! refueling or maintenance outages. Access into the suppression chamber is prosided by a hatch located in the safety envelope. l Prior to reactor operation, the containment atmospheric control system in conjunction l ! with the containment purge system and the dowell cooling fans are utilized to establish l an inert gas emironment in the containment with nitrogen to limit the oxygen concentration. This precludes combustion of any hydrogen which might be released subsequent to a LOCA. After the containment is inerted and sealed for plant power operation, small flows of nitrogen gas are added to the dowell and the suppression chamber as necessary to keep oxygen concentrations below 4% and to maintain a positive pressure for preventing air inleakage. High pressure nitrogen is also used for l pneumatic controls inside the containment to preclude adding air to the inert atmosphere. The containment structure has the capability to maintain its functionalintegrity du-ing an'! fe!!cHng esen at the increased pe2h ande-: pressures and temperatures c2ured S; the e :t that could follow a LOCA pipe break postulated to occur simultaneously with loss of off-site power. The containment structure is designed to accommodate the full range ofloading conditions associated with normal and abnormal operations including LOCA related design loads in and above the suppression pool (including negative differential pressure between the drpeell, wetwell and reactor building), and safe shutdown earthquake (SSE) loads. The containment structure is protected from or designed to withstand fluidJet forces l associated with outflow from the postulated rupture of any pipe within the l containment. The containment design considers and utilizes leak-before-break (LBB) applicability only in regard to protection against dynamic effects associated with a postulation of rupture in high energy piping. Subsection 3.6.3 and Appendix 3C describe the implementation of the LBB approach for excluding design against the dynamic effects from postulation of breaks in high energy piping. Protection against the dynamic efTects l from the piping systems not qualified by the exclusion from the dynamic effects caused by their failure is provided for the drywell structure. The drywell structure is prosided l protection against the dynamic effects of plant-generated missiles (Section 3.5). 1.2-86 General Plant Description ~ Amendment 1 l
25A5113Rev. A SBWR stadudsanyAndyds Reput l The containment structure has design features to accommodate flooding to sufficient depth above active fuel to permit safe removal of fuel assemblies from the reactor core after a postulated design basis accident. The containment structure is configured to channel flow from postulated pipe ruptures in the drywell to the suppression pool through vents submerged in the suppression pool , which are designed to accommodate the energ'; of the blowdown fluid. l The containment structure and penetration isolation system with concurrent operation of other accident mitigation systems, are de.tigned to limit fission product leakage during and following a postulated design basis accident (DBA) to values well below leakage calculated for allowable off-site doses. In accordance with AppendixJ to 10CFR50, periodic leak rate tests conducted at a i reduced pressure below the peak calculated DBA LOCA pressure are performed to confirm containmentleakage is below the design limit of 0.5% by weight per day of the containment free air volume. Special testing capabilities are provided during outages to measure local leakage, such as individual air locks, hatches, drywell head, piping, . l electrical and instrument penetrations. Other features are provided to measure ! isolation valve leakage and to measure the integrated containment leak rate. Results ; l lv from the individual and integrated preoperational leak rates are recorded for companson with subsequent periodic leak rate test results. The design value for a maximum steam bypass leakage between the drywell and the suppression chamber through the diaphragm floor including any leakage through the suppression chamber-to-daywell vacuum breakers is limited. Satisfying this limit is confirmed by initial preoperational tests as well as by periodic tests conducted during refueling outages. These tests are conducted at differential pressure conditions between the drywell and suppression chamber that do not clear the drywell-to-suppression chamber horizontal vents. Equipment is provided to obtain a water tight barrier between the open reactor and the drywell during refueling. This enables the reactor well to be flooded prior to removal of the reactor steam separator, dryer assembly and to facilitate underwater fuel handling ; operations. Piping, cooling air ducts and return air vent openings in the reactor well ! l platform must be removed, vents closed and sealed watertight before filling the reactor well with water. The refueling bellows assembly is provided to accommodate the i movement of the vessel caused by operating temperature variations and seismic actisity. Containment isolation is accomplished with inboard and outboard isolation valves on each piping penetration which are signaled to close on predefined plant parameters. 'O Systems performing a post LOCA function are capable of having their isolation valves reopened as needed. General Plant Description - Amendment 1 12-87 l
25A5113 Rev. A SBWR stamtantsarety Analysis Report O Dgwell coolers are provided to remove heat released into the drywell atmosphere during normal reactor operations. The Hammability Control System provides recombiners ig:ncr: located throughout both the dowell and suppression chamber to prevent any high-energy-release recombinant reactions potentially developing within the containment following a LOCA. 1.2.2.14.2 Containment Vessel The containment vessel is a reinforced stepped cylindrical concrete vessel (RCCV). The RCCV supports the upper pools whose walls are integrated into the top slab of the containment to provide structural capability for LOCA and testing pressures. , 1.2.2.14.3 Containment internal Structures The containment system's principal internal stnicture consists of the structural barrier separating the dgwell from the suppression chamber. This barrier is comprised of the suppression chamber ceiling (diaphragm floor) and the inboard wall (vertical vent wall) separating the drywell from the suppression chamber. Both of these stnictural components are designed as steel structures filled with insulating concrete to minimize long-term heat transfer from drywell to wetwell. The vertical vent wall also prosides a durable attachment point for the RPV horizontal stabilizers. An all-steel reactor shield wall of appropriate thickness is provided, which surrounds the RPV to reduce gamma shine on drywell equipment during reactor operation and protect personnel during shutdowns for maintenance and insenice inspections. The RPV insulation is supported from the internal surface of the reactor shield w211. The reactor shield wall is supponed on top of the pedestal support structure. Various drywell piping and equipment suppon structures are provided to support electric and instrument cable trays, drywell coolers, air distribution ductwork, steam and feedwater piping, and SRV discharge piping. Support is provided for isolation valves and piping of the ICS and PCCS. This steel structure also supports access stairs, walkways, railings and gratings. Monorails are suspended from the ceiling of the drywell for hoists to work on NSSS equipment. 1.2.2.14.4 Passive Containment Cooling System l l The PCCS maintains the containment within its pressure limits for design basis accidents such as a LOCA. The system is passive with no components that move. j The PCCS consists of three low pressure, totally independent loops, each containing a l steam condenser (passive containment cooling condenser) that condenses steam on 1.2-88 General Plant Description - Amendment 1 l l
25A5113Riv. A SBWR standardsafety Anarrsis aeport g i s i 1
%.J tube side and transfers heat to water in a large cooling pool (IC/PCCS pool), which is vented to atmosphere.
Each PCCS condenser is located in a subcompartment of the IC/PCCS pool, and all pool subcompartments communicate at their lower ends to enable full use of the collective water inventory, independent of the operational status of any given PCCS loop. Each loop which is open to the containment, contains a drain line to the GDCS pool, and a vent discharge line the end of which is submerged in the pressure suppression pool. The PCCS loops are driven by the pressure difference created between the containment drywell and the suppression pool during a LOCA so require no sensing, control, logic or power actuated devices for operation. The PCCS is classified as safety-related and Seismic Category I. Each of the three PCC condensers is designed for 10 MWt capacity. Together with the fm pressure suppression containment system, the three PCC condensers limit containment ( ') pressure to less than its design pressure for at least 72 hours after a LOCA without make-up ta the IC/PCC pool. The PCC condensers are closed-loop extensions of the containmentpressure boundary. Therefore, there are no containment isolation valves and they are always in " ready standby". The PCCS can be periodically pressure-tested as part of overall containment pressure testing. Also, the PCC loops can be isolated for individual pressure testing during maintenance. During refueling outages,in-service inspection (ISI) of PCC condensers can be performed,if necessary, because ultrasonic testing of tube-to-heater welds and eddy current testing of tubes can be done with PCCs in place. The PCC condensers are located in the IC/PCC pool. 1 2.2.14.5 Containment Atmospheric Control System The Containment Atmospheric Control System (CACS) is designed to establish and l maintain an inert atmosphere within the containment during all plant operating modes l except during plant shutdown for refueling or equipment maintenance and during ! limited periods of time to permit access for inspection at low reactor power. The l/^\ objective of the system is to establish conditions that help preclude combustion of i U hydrogen and thereby prevent damage to safety-related equipment and structures. l General Plant Description - Amendment 1 1.2-89 l i
25AS113Rev. A SBWR standardsafety Analysis neport O The CACS does not perform any safety-related functions except for its containment isolation function. Failure of the CACS does not compromise any safety-related system or component nor does it prevent a safe shutdown of the plant. (The inerted conditions that the CACS establishes are safety-related, however.) The CACS establishes an inert atmosphere (i.e., an oxygen concentration s 4% by volume) throughout the containment following an outage (or other occasions when the containment has become filled with air) and maintains it inert during normal conditions. The system maintains a slight positive pressure in the containment to prevent air (oxygen) in-leakage. The CACS is comprised of a pressurized liquid nitrogen storage tank, a steam heated main vaporizer for large nitrogen flow, electric heater for vaporizing makeup flow, two injection lines, an exhaust line, a bleed line, associated valves, controls, and instrumentation. All CACS components are located inside the reactor building except the liquid nitrogen storage tank and the steam-heated main vaporizer which are located in the yard. The first of the injection lines is used only for makeup; it includes an electric heater to vaporize the nitrogen and to regulate the nitrogen temperature to acceptable injection temperatures. Remotely operated valves together with a pressure-reduction valve enable the operator to accomplish low rates of nitrogen injection into the drywell and suppression chamber airspace. The second injection line is used when larger inerting flow rates are required. This line takes vaporized nitrogen from the steam-heated main vaporizer, uses remotely operated valves together with a pressure-reduction valve and injects nitrogen at points in common with makeup supply. The inerting and makeup lines converge to common injection points in the lower drywell and suppression chamber airspace. The CACS includes an exhaust line leading from the upper drywell at the opposite side from the injection points. The discharge line connects to the reactor building HVAC system exhaust where exit gases are served by exhaust fans, filters, and radiation monitors before being diverted to the plant stack. A small bleed line bypassing a short portion of the main exhaust line, upstream of the fans, filters, and stack monitors,is also i l provided for manual pressure control of the containment during normal reactor l heatup. Redundant containment isolation valves provided in the inerting, makeup, exhaust and l bleed lines close automatically upon receipt of an isolation signal from the LD&IS. Upstream of the pressure-reduction valve in the makeup line, a small branch line is provided and connected to the HPNSS. This line is used for the initial charging of the l 1.2-90 General Plant Description - Ame utment 1 l l
25A5113 Rsv. A SBWR standardSafety Analysis Report f (. i HPNSS and makeup to keep the HPNSS changed with nitrogen during normal plant ; operanon. l 1 During plant startup, a large flow of nitrogen from the liquid nitrogen storage tank is I vaporized by the steam-heated vaporizer and injected into the dnwell and the I suppression chamber. It is then mixed into the containment atmosphere by the drywell cooling fans. The exhaust line is kept open to displace containment resident atmosphere with nitrogen. Once the desired concentration of nitrogen is reached, the exhaust line is allowed to close. When the required inerted containment operating pressure is attained, the inerting process is terminated by the closure of the nitrogen supply shutoff valve and inerting isolation valves. The system is designed to inert the containment to s 4% oxygen by volume within four hours. Following shutdown, the containment atmosphere is de-inened to allow safe personnel access inside the containment. Breathable air from the Reactor Building HVAC System is injected to the dowell and suppression chamber air space through the inerting injection line. The incoming air displaces containment gases (mostly nitrogen) into the exhaust line. Vented gases are served by the Reactor Building HVAC system exhaust fans, filters, and radiation detectors before being diverted to the plant stack. The system (g) k/ is designed to de-inert the containment to an oxygen concentration of 218% within four hours. 1.2.2.14.6 Drywell Cooling System The Dawell Cooling System (DCS) consists of four fan coil units (FCUs), two located in the upper drywell, and two in the lower dqwell. The system uses the FCUs to deliver cooled air / nitrogen to various areas of the upper and lower drywell through ducts / diffusers. The DCS is a closed loop recirculating air / nitrogen cooling system where no outside air is introduced into the system except when the containment is open. The DCS is manually controlled from the MCR. During normal plant operation, j the DCS is cooled by the RCCWS. During shutdown operation, the DCS is cooled by the l CWS to facilitate obtaining cooler temperatures. The CWS water is also used as a backup to RCCWS water when the latter is not available. Through the entire plant operating range, from startup to fullload condition or from full load to shutdown, the DCS performs the following functions: l l e Maintains temperature and humidity in the upper and lower dowell spaces within specified limits during normal operation. l a Maintains the RPV support skirt temperature within specified limits to satisfy l (~'; structural requirements.
\ l l V' i
General Plant Description - Amendment 1 1.2-91
25A5113 Riv. A l SBWR standardsarety Anatysis neport ! O e Accelerates dnwell cooldown during the period from hot reactor shutdown to cold shutdown. l m Aids in complete purging of nitrogen from the dnwell during shutdown. m Maintains a habitable emironment for plant personnel during plant shutdowns for refueling and maintenance. l m Limits drywell temperature during loss of preferred power (LOPP). The DCS is designed to maintain the following conditions in the upper and lower
- dnwell during normal and plant shutdown modes of operation
l Normal plant operation: l e Average dry bulb temperature: 57 C (135 F) e Maximum temperature of ambient atmosphere in each dnwell zone: 66*C (150 F) Plant shutdown: m Average dry bulb temperature: 26 C (77 F) l There are two direct-drive fans in each FCU. Each FCU motor is controlled manually from the MCR. Indicator lights show the status of each unit. Failure of an FCU with consequent temperature rise in the discharge stream or loss of flow actuates an alarm in the MCR. Each upper drywell FCU has a cooling capacity of 50% of the upper drywell design cooling load under normal plant operating conditions. Likewise, each lower dowell FCU has a cooling capacity of 50% of the lower drywell design cooling load. All FCUs normally operate. Each FCU is composed of a cooling coil and two fans downstream of the coil. One FCU is supplied by RCCWS loop A and the other by RCCWS loop B. One of the fans operates while the other is on standby status and will automatically start upon loss of the lead fan. During normal operation,if both fans of an FCU are out of commission, or the unit is not in senice for some other reason, then both fans on the other unit in the area (upper orlower dnwell) operate and the cooling supply transfers to the CWS. Cooled air / nitrogen leaving the FCUs enter a common plenum and is distributed to the various zones in the dnwell through distribution ducts. Retarn ducts are not provided; the FCUs draw air / nitrogen directly from the upper or lower dawell. O 1.2 92 General Plant Description - Amendment 1
25A5113Rev. A l SBWR stamrantsareryannor sis sepat a O b i A condensate collection pan is provided with each FCU. The condensate collected from
- all FCUs in the upper and the lower dowellis piped to an LD&IS flow meter to measure the condensation rate of unidentified leakages.
1.2.2.14.7 Flammability Control System The Flammability Control System (FCS) is designed to limit the concentration of 1 oxygen in a potentially hydrogen-rich post-accident containment atmosphere by 1 controllably recombining bur":ng hydrogen at lowlevels of oxygen inside the ! containment. The FCS consists 'of nassive autocatalytic recombiners (PARS) S :e n 2!!y r c.g n ed !c" percer cer u npdcr ign::er :emb!!e:strategicallyintermixed throughout the containment including the upper and lower dnwell cavities, and wetwell air space..and- ! pt ercred by C!r IE d*:^nd per e-
*^e FCS E cer e!!rd frc- ^c '!CR. Prior to the postulated design basis LOCA, the containment is maintained inert at s 4% oxygen volumetric concentration by the CACS. i The FCS utemadca!!y :n!date: ?! 'cu ::f:er -eceipt ef c L^CA. ugni! for Fe s centre!!ed ign den of bydrc; r "dt h egger Once : :deted,ign!!e~. "i'! :endnue te Oper :e n!rz :nnnud!y :cpped by $c cpercier '!:nud F5 i-!daden E cpez!b!
frc:r S '!CR During normal plant operation, the CACS provides containment atmosphere oxygen level monitoring. During FCS operation, post-accident oxygen level monitoring is provided by the Containment Atmospheric Monitoring System (CAMS). i I The FCS is designed and qualified as a sa'.ety-related and Seismic Category I system. All required FCS components are deMgned and qualified to withstand the adverse emironmental conditions resulting from DBA LOCA. 1.2.2.15 Structures and Servicing Systems 1.2.2.15.1 Cranes, Hoists, and Elevators Large bridge cranes are provided in the turbine building and for the refueling floor. A bridge crane is also installed in the radwaste building. Miscellaneous hoists and , monorails are installed in the reactor, turbine and other buildings as necessary for i maintenance and replacement of equipment. Elevators are installed in the reactor and turbine buildings. U 1 General Plant Description - Amendment 1 1 2-93
25A5113 Rev. A SBWR standardsafety Anstrsis Report O 1.2.2.15.2 Heating Ventilating and Air Conditioning Reactor Building HVAC Systems l These systems are the Clean Area Ventilation System (CLAVS), the Controlled Area Ventilation System (CONAVS), Control Room Envelope HVAC (CREHVAC), and the Refueling and Pool Area Ventilation System (REPAVS). A common intake is used for these systems. With the exception of containmentisolation components, the systems are non-safety-related and Seismic Category NS. The CLAVS includes redundant supply fans, redundant air conditioning units (with air mixing plenum, filters, heating and cooling coils, and humidifier), dampers, and ducting. The system also includes redundant return / exhaust, battery room exhaust, and smoke removal fans. Local cooling / heating coils and fans are provided for the main entrance area and the access and change room area. The CONAVS has two main trains each including a supply fan, air conditioning units (with filters and heating and cooling coils), and an exhaust fan. Two redundant exhaust fans are provided for the safety envelope area, and local recirculating systems (including redundant fans, and cooling / heating coils) are provided for the FAPCS, RWCU/SDC, RCCW, main steam tunnel, and CRD pump rooms. CONAVS also includes a separate containment purge and exhaust subsystem with purge supply and exhaust filters, redundant supply and exhaust fans, and main stack radiation monitors. All CLAVS and CONAVS equipment is non-safety-related with the exception of the isolation dampers and ducting that penetrate the safety envelope. CREHVAC serves the MCR, technical support center (TSC), computer room, and adjacent rooms and includes redundant supply fans, air conditioning units (with air mixing plenum, filters, heating and cooling coils, and humidifier), and exhaust fans. Two utility exhaust fans are also provided as well as a supplementary filtration unit with HEPA and charcoal filters and redundant exhaust fans for removal of airborne hazardous materials. With the exception ofisolation dampers and ducting for the MCR/TSC computer areas, the system is nonsafety-related. Refueling and Pool Area Ventilation System (REPAVS) has two full capacity supply fans, air conditioning units with filters, cooling and heating coils, and exhaust fans. It is non-safety-related except for dampers and ducting associated with refueling floor isolation. Turbine Building HVAC The Turbine Building Ventilation System includes an intake plenum and dampers and two 100% capacity supply trains with an air conditioning unit (filters, heating and cooling coils, and humidifier). The Turbine Building Chilled Water System provides chilled water to local unit coolers and outside air intake coils when required. Two 1.2-94 General Plant Description - Amendment 1
l 2SA5113Rsv. A SBWR standardsafety Analysis Report b V , redundant exhaust fans are provided. Local unit coolers and fans are prosided in areas l with high local heat loads. The system is non-safety-related. l l Other Building HVAC Ventilation for other buildings includes the radwaste building, electrical building, ' i senice building, water treatment building, administration building, guard house, etc. i All these systems are non-safety-related, of conventional design and typically include redundant supply and exhaust fans, and air conditioning units. The radwaste building l and hot machine shop ventilation systems also include additional filtration and airborne radioactivity monitoring equipment. 1.2.2.15.3 Fire Protection System The Fire Protection System (FPS) includes the fire protection water supply system, yard piping, water sprinkler, standpipe and hose systems, a foam system, smoke detection and alarms systems, and fire barriers. The water supply system includes a motor-driven pump and a backup diesel-engine driven pump. Yard piping supplies fire water to all buildings. Fire hydrants are located ,p throughout the site. Standpipes are provided within buildings as well as automatic sprinkler and deluge systems. Foam fire suppression systems are provided for the l standby diesel generator and day tank rooms, outdoor diesel fuel oil storage tanks, and the turbine tube oil system and storage tanks. Smoke and heat detectors are located throughout the various buildings and are controlled by local panels and proside remote indication in the MCR. Fire barriers (typically three-hour rated), including penetration seals, doon, and fire dampers are provided wherever separation of redundant safety-l related equipment is required. The FPS is non-safev-related. The diesel-driven fire pump, its suction line, a portion of the yard piping and connecting piping serving safety-related areas are designed to l remain functional after an SSE. 1.2.2.15.4 Equipment and Floor Drainage System The Equipment and Floor Drainage System (EFDS) serves the plant building with floor and equipment drains and consists of the following drain subsystems: clean, low conductivity waste (LCW), high conductivity waste (HCW), detergent, and chemical l waste. All potentially radioactive drains are routed to the Radynste Management System for processing. Each subsystem includes sumps, sump pumps, piping and valves, and level insuumentation and controls. The EFDS is non-safety-related except for containment penetrations and isolation , i valves. I General Plant Description - Anundment 1 1.2-95 l
2SA5113Rw. A SBWR _ standardSafety Analysis Report O 1.2.2.15.5 Reactor Building The reactor building houses the reactor system, reactor support and safety systems, containment, refueling and spent fuel storage areas and equipment, main steam tunnel, MCR and other control areas, auxiliary area, liquid waste processing area, health physics, laboratories, security and access control areas. The reactor building structure is integrated with that ofa stepped cylindrical reinforced concrete containment vessel (RCCV); the RCCV is located on a common basemat and surrounded by three concentric boxes: the inner box (safety envelope), the intermediate steel frame, and the outer box. The inner and outer boxes are made of reinforced concrete shear walls and the intermediate steel frame is made of structural steel framework with non-stmetural walls as required for ra diation shielding, separation, etc. The building is partially embedded. All SBWR safety-related equipment is housed in the reactor building safety envelope, main steam tunnel, and pools located beneath the operating floor, with the non-safety-related systems and areas (including the MCR) surrounding this envelope. The safety envelope is leaktight for holdup and decay of fission products that may leak from the containment after an accident. This holdup capability decreases releases to the atmosphere. The building and systems are also arranged to separate clean and potentially contaminated areas, with separate stairway and elevator service for each area. l On the upper levels of the reactor building is the refueling area which contains the l spent and new fuel pools, cask loading area, isolation condenser / passive containment j cooling system pools, other pools and storage areas, and refueling and fuel handling l systems. A bridge crane is installed that operates along the length of the floor and senices a large equipment hatch that is provided at grade with a shaft allowing communication with all elevations up to the refueling floor. A plant stack is located on the reactor building and rises above the top of the building. The stack is of steel shell construction supported by an external steel tubular framework. The stack vents the reactor building. The reactor building is a safety-related and Seismic Category I structure. 1.2.2.15.6 Turbine Building The turbine building encloses the turbine-generator, main condenser, condensate and feedwater systems, condensate purification system, turbine-generator support systems, i and bridge crane. I 1 The turbine building is a reinforced concrete structure up to the turbine operating deck, and steel frame and metal siding thereafter. It is built at grade. Shielding is 1.2-96 GeneniPlant Description - Amendment 1
i 25A5113Rzv. A SBWR studentsaretyAarrsisneport i O !V 1 provided for the turbine on the operating deck. The turbine-generator and condenser are supported on spring type foundations. The turbine building is a non-safety-related structure. 1.2.2.15.7 Radwaste Building , The radwaste building houses tanks and processing equipment, storage areas, a laundry l room, a control room and health physics area, a truck bay, and other support facilities. l A pipe tunnel connects the radwaste building to the turbine and reactor buildings. l Space is included for storage of dry active waste. The stmeture up to grade is reinforced concrete (first story), and has a structural steel framework with metal siding and a metal roof above that. The reinforced concrete portion of radwaste building below grade is designed to the requirements of Regulatory Guide 1.143, and the balance of the structure is Seismic Category NS. 1.2.2.15.8 Other Building Structures l , l i Other facilities include the electrical building, the senice building, the service water l and fire building, mechanical draft cooling towers, the water treatinent building, gate l houses, guard house, an administration building, a training center, sewage treatment plant, warehouse, and hot and cold machine shops. These are all of conventional size l and design. The electrical building houses the two non-safety-related standby DGs. Itis a reinforced concrete structure. It is non-safety-related and Seismic Category NS. The senice water and fire building houses the PSW pumps and fire pumps, and associated water storage, piping and valves. It is a concrete foundation steel frame building with metal siding and metal roof. It is non-safety-related and Seismic Category NS. l The water treatment building is a conventionally sized and designed building. l 1.2.2.16 intake Structure and Servicing Equipment 1.2.2.16.1 Intake and Discharge Structure l l l The intake and discharge structure (which is the reference design only) is adjacent to ) ! the natural draft cooling tower and houses the circulating water pumps, isolation valves, water treatment equipment, and associated electrical power and controls equipment. The structure is of conventional reinforced concrete construction. A traveling screen and trash rake system is installed to prevent debris from entering into the circulating water system. The structure and systems are not safety-related. Blowdown from the cooling tower basin is via a blowdown line to the site water source. The intake and ' Q discharge structure are provided by the applicant. l General Plant Description - Amendment 1 1.2-97
25A5113Rw. A SBWR standedSafety Analysis Report O 1.2.2.16.2 Cooling Tower The conceptual SBWR cooling tower is a single hyperbolic natural draft cooling tower that cools circulating water. The reinforced concrete tower is located atop a cooling tower basin. The tower system is equipped with drift eliminators and a winter bypass line for cold weather operation. It is not safety-related. 1.2.2.17 Yard Structures and Equipment 1.2.2.17.1 Oil Storage and Transfer Systems The major components of this system are the fuel-oil storage tank, pumps, and day tanks. Each standby DG has its own individual supply components. Each fuel-oil pump is controlled automatically by day-tank level and feeds its day tank from the storage tank. 1.2.2.17.2 Site Security The site security system includes fencing, E-field intrusion detection systems, closed circuit television system, site access control equipment (portal monitors, identification equipment), an electronic lock /cardreader building access control system, vehicle inspection bays, and monitoring and control computers and stations. 1.2.3 COL License Information The applicant shall provide necessary design information on the cooling tower, intake structure, and discharge stmcture. 1.2.4 References None. O 1.2-98 General Plant Description - Amendment 1
2SA5113 Rev. A SBWR ssendardsanrayAnarr sisneport p 1.5 Requirements for Further Technical information None-This section presents the background from which the SBWR design has evolved, the methodology used to assess the need for further technical information, the computer code used for analysis and design, and the major SBWR Test Programs. 1.5.1 Evolutionary Design 1 The SBWR design is an evolutionary step in boiling water reactor (BWR) design which traces its commercial demonstration and operating plant history back before 1960 and represents ~- ^~~A hundreds ofreactor years of successfullicensed plant operation (Figure 1.5-1). Since its inception, the BWR has had plant simplification as a goal for each product improvement (Figure 1.5-2).The SBWR as described in this SSAR has major simplifying improvements drawn from predecessor designs such as pressure-suppression containment, natural circulation, isolation condenser handling of waste heat, and gravity <lriven makeup water systems (Table 1.5-1). The incorporation of these features from predecessor designs has been done with safety in mind and has g emphasized employment of passive means of dealing with operational transients and hypothetical lossef. coolant accidents (LOCAs).The result of this particular design assemblage of previously licensed plant features is simplified operator response to these events (most plant upset conditions are dealt with in the same manner, that typified for the hypothetical steamline break) and operator response time for all hypothetical events has been relaxed from minutes for previously licensed reactors to days for the SBWR. Most features of the SBWR have been taken directly from licensed commercial BWRs and reviewed and redesigned as appropriate for the SBWR (Table 1.5-2). This rich history of previously licensed plant features allowed the SBWR to draw together the best of previously licensed plant features to continue the simplification process as the . passive BWR concept was brought to maturity (Figure 1.5-3). I 1.5.2 Analysis and Design Tools As implied in Section 1.5.1, there is now an immense amount of data available from operating plants and from the testing and licensing efforts done to license the ! l predecessor designs and individual plants, the most recent of which is the ABWR. As a l measure of comparison, approximately 50% of the umtent of the ABWR SSAR is !. technically identical or technically similar (with minor diferences) to the SBWR SSAR l (Ref. NEDC-32231," Design and Analysis Similarities Between the ABWR and SBWR", December 1993). The vast database of feature performance in licensed reactors, combined with the recent thorough licensing review of the ABWR, provides an extremely well<lualified foundation from which to make the modest extrapolations to IOs the SBWR. To make that extrapolation, GE has developed one computer code ! (TRACG) to use for design and for three out of the four most limiting licensing I Requirements for Further Technicalinformation - Amendment 1 1.5-1 I
-,- w--, -- , , ,,- p
2SAS113 Rsv. A SBWR Standard Safety Analysis Report O analyses. GE has chosen to develop the TRACG code, validated by the operating plant experience and appropriate testing, in order to analyze the challenges to the fuel (10CFR50.46 and Appendix K, SSAR Section M M), the challenges to the containment (SSAR Section M M), and many of the operational transients (MCPR, SSAR Chapter 15). The radiological responses to hypothetical accidents (LOCAs) are presented also in SSAR Chapter 15, but do not use TRACG for analysis. Thus, TRACG draws from the very large database oflicensed BWRs which includes all features of the SBWR (albeit in various configurations) and appropriate testing, and allows direct application to SBWR design and analysis (Table 1.5-3). 1.5.2.1 TRACG The TRACG Code and its application to the SBWR is documented in a series of GE Nuclear Energy Topical Repcrts as follows: a NEDE-32176P, "TRACG Model Description", February 1993,J. G. M. Andersen, et al. m NEDE-32177P,"TRACG Qualifications", Febntary 1993,J. G. M. Andersen, et al. m NEDE-32178P, " Application of TRACG Model to SBWR Licensing Safety Analysis", February 1993, H. T. Kim, et al. TRACG is a GE proprietary version of the Transient Reactor Analysis Code (TRAC). It is a best-estimate code for analysis of BWR transients ranging from simple operational transients to design basis LOCAs, stability, and anticipated transients without scram (ATWS). 1.5.2.1.1 Background TRAC was originally developed for pressurized water reactor (PWR) analysis by Los Alamos National Laboratory (LANL), the first PWR version ofTRAC being TRAC-PIA. The development of a BWRversion ofTRAC started in 1979 in a close collaboration between GE and Idaho National Engineering Laboratory. The objective of this cooperation was the development of a version of TRAC capable of simulating BWR LOCAs. The main tasks consisted ofimproving the basic models in TRAC for BWR applications and developing models for the specific BWR components. This work culminated in the mid<ighties with the development of TRACB04 at GE and TRAC. BD1/ MOD 1 at INEL, which were the first major versions of TRAC having BWR LOCA capability. Due to thejoint development effort, these versions were very similar, having virtually identical basic and component models. The GE contributions werejointly funded by GE, the Nuclear Regulatory Commission (NRC) and Electric Power Research Institute (EPRI) under the REFILL /REFLOOD and FIST programs. 1.5-2 Requirements for Further Technicalinformation - Amendment 1
25AS113Rev. A SBWR stuentsarerAusysisaeron O l The development of the BWR version has continued at GE since 1985. The objective of j this development was to upgrade the capabilities of the code to include transient, j stability and ATWS applications. During this phase, major developments included the i implementation of a core kinetics model and addition of an implicitintegration scheme I into TRAC. The containment models were upgraded for simplified boiling water l reactor (SBWR) applicadons, and the simulation of the BWR fuel bundle was also improved. TRACG was the end result of this development. 1.5.2.1.2 Scope and Capabilities TRACG is based on a multi-dimensional two-fluid model for the reactor thermal hydraulics and a three-dimensional neutron kinetics model. The two-fluid model used for the thermal hydraulics solves the conservation equations ) for mass, momentum and energy for the gas and liquid phases. TRACG does not l include any assumptions of thermal or mechanical equilibrium between phases. The gas phase may consist of a mixture of steam and a noncondensable gas, and the liquid phase may contain dissolved boron. The thermal-hydraulic model is a multi-dimensional formulation for the vessel component and a one-dimensional formulation for all other (% components. The conservation equations for mass, momentum and energy are closed through an extensive set of basic models consisting of constitutive correlations for shear and heat transfer at the gas / liquid interface as well as at the wall. The constitutive correlations are flow regime dependent and are determined based on a single flow regime map, which is used consistently throughout the code. In addition to the basic thermal-hydraulic models, TRACG contains a set of component models for BWR components, such as channels, steam separators and dryers. TRACG also contains a control system model capable of simulating the major BWR control l systems such as RPV pressure and water level. The neutron kinetics modelis consistentwith the GE BWR core simulator PANACEA. It solves a modified one-group diffusion model with six delayed neutron precursor groups. Feedback is provided from the thermal-hydraulic model for moderator density, l fuel temperature, boron concent:7 tion :nd control rod position. The TRACG stmcture is based on a modular approach. The TRACG thermal-hydraulic model contains a set of basic components, such as pipe, valve, tee, channel, steam separator, heat exchanger and vessel. System simulations are constructed using these components as building blocks. Any number of these components may be combined. l The number of components, their interaction, and the detail in each component are specified through code input. TRACG consequently has the capability to simulate a 1.5-3 Requirements for Further Technicalinformation - Amendment 1 l
l l 25AS113Riv. A StandantSafety Analysis Repc.'t SBWR O wide range of facilities, ranging from simple separate effects tests to complete BWR plants. TRACG has been extensively qualified against separate effects tests, component pedormance data, integral system effects tests and full-scale BWR plant data. A detailed documentation of the qualification is contained in the TRACG qualification report NEDE-32178P. 1.5.2.2 Scope of Application of TRACG to SBWR The total effort and extent of qualification performed on TRACG, since its inception in 1979, now exceeds, both in extent and breadth, that for any other engineering computer program which GE has submitted to the NRC for design application approval. l 1.5.2.2.1 Transient Analysis TRACG is used to perform safety analyses of nearly all of the Anticipated Operational Occurrences (AOO) described in SSAR Chapter 15 ;nd the ASME reactor vessel overpressure protection events in Chapter 5. The loss of feedwater heating and the control rod withdrawal error events are analyzed using the GE S-D BWR core simulator model. Other exceptions are the control rod drop and the fuel-handling accidents, and radiological calculations for all postulated accidents. l The analysis determines the most limiting event for the AOOs in terms of Critical Power Ratio (CPR) and margin loss (A CPR) and establishes operating limit minimum CPR (01 MCPR). The OLMCPR includes the statistical CPR adder which accounts for uncertainty in calculated results arising from uncertainties associated with the TRACG model, initial conditions, and input parameters. Sensitivity analysis of important parameters affecting the transient results is performed using TRACG. Concepts derived i from the Code Scaling, Applicability, and Uncertainty (CSAU) methodology are utilized for quantifying the uncertainty in calculated results. 1 The analysis also determines the most limiting overpressure protection events in tenns of peak vessel pressure. The results are used to demonstrate adequate pressure margin l to the reactor vessel design limit with the SBWR design safety / relief valve capacity. The j overpressure protection analysis is performed based on conservative initial conditions l and input values. 1.5.2.2.2 A'IWS Analysis TRACG is used for evaluadon of the ATWS events in SSAR Chapter 15. The analysis determines the most limiting ATWS events in terms of reactor vessel pressure, heat flux, l neutron flux, peak cladding temperature, suppression pool temperature, and l l 1.54 Requirements for Further TechnicalInformation - Amendment 1 i
I 25A5113Rev. A SBWR standardsdety Analysis Repat (G : containment pressure. The results are used to demonstrate the capability of the SBWR mitigation design features to comply with the ATWS licensing criteria, 1.5.2.2.3 ECCS/LOCA Analysis j TRACG is used for evaluation of the complete spectrum of postulated break sizes and l locations, together with possible single active failures for Section 6.3 of the SBWR SSAR. This evaluation determines the worst case break and single failure combinations. The results are used to demonstrate the SBWR Emergency Core Cooling System (ECCS) capability to comply with the licensing acceptance criteria. A sensitivity analysis ofimportant parameters affecting LOCA results is performed using TRACG. For the SBWR, the LOCA analysis results are adjusted so that they provide 95% l probability LOCA results for use as the licensing basis. The SBWR LOCA results have l large margin with respect to the licensing acceptance criteria. 1.5.2.2.4 Containment Analysis TRACG is also used for evaluation of containment response during a LOCA. The analysis determines the most limiting LOCA for containment (or Design Basis Accident, , , DBA) in terms of containment pressure and temperature responses. The DBA is determined from consideration of a full spectrum of postulated LOCAs. The results are used to demonstrate compliance with the SBWR containment design limits. l Sensitivity of the containment response to parameters identified as important is evaluated using TRACG to assess the effect of uncertainties of these parameters on the ,
- containment responses.The procedure derived from the CSAU methodology (see !
Section 1.5.3) is used for this purpose. 1.5.3 Testing GE has used a procedure similar to Code Scaling, Applicability and Uncertainty (CSAU) methodology developed by the NRC [NUREG/CR-5249, EGG-2552," Quantifying ; Reactor Safety Margins", and " Quantifying Reactor Safety Margins", B. E. Boyack, et al, , l I i Nuclear Engineering and Design (Parts 1-4),119 (1990), Elsevier Science Publishen B. V. (North Holland)] and submitted to the NRC by GE letter MFN 042-92-92 andJSC 010," Implementation ofTRACG for Licensing Analysis",J. L Rash (GE) to R. C. Jones, Jr. (NRC), March 9,1992. This procedure developed a list of phenomena important to the SBWR behavior in a large number of anticipated and hypothetical events and - matched them againstinformation available from operating plant and/or test ; experience (Table 1.5-3). The Phenomena identification and Ranking Table (PIRT) (Tables 1.5-Sa and 1.5-Sb) identifies 124 specific governing phenomena,74 of which were concluded to be "important" in prediction of SBWR transient and LOCA performance. TRACG contains models capable of simulation of each of the 74 important phenomena, and each has been qualified by the successful predictions of at Requirements for Further Technicatinformation - Amendment 1 1.5-5
25A5113Rn A SBWR standardsafery Analysis neport O least one, and in most cases, several test data points. Section 3 of the PIRT (Table 1.5-Sc) defines more than 900 specific data points, from 42 different tests and test facilities, that compdse the TRACG qualification database. Data from separate effects tests, components tests, systems and systems interaction tests, and operating BWR experiences have been successfully predicted by TRACG. In nearly all cases, existing test data was available from several data sources to address the identified important SBWR phenomena. Only one piece ofinformation was identified as needed for the SBWR for which there was no infonnation in the database: a heat transfer correlation for steam condensation in the presence of noncondensable gases. A test program has since been conducted to secure this information and has been reported to the NRC in GE topical report NEDC-32301,
" Single Tube Condensation Test Program", March 1994, by W. R. Usry.
The Single Tube Condensation Test Program was conducted to investigate steam condensation inside of tubes in the presence of noncondensables. The work was independently conducted at the University of California at Berkeley (UCB) and at the Massachusetts Institute of Technology (MIT). The work was initiated in order to obtain a database and a correlation for heat transfer in similar conditions as would occur in the SBWR PCCS tubes during a DBA LOCA. Three researchers utilized three separate experimental configurations at UCB, while two researchers utilized one configuration at MIT. The researchers ran tests with pure steam, steam / air, and steam / helium mixtures with flow rates and noncondensable mass fractions representative and bounding of a DBA LOCA. The experimenters found the system to be well behaved for all tests, with either of the noncondensables, for forced flow conditions similar to the SBWR design. The results of the tests at UCB have become the basis for the heat transfer correlation used in the TRACG computer code. As noted previously, the SBWR design philosophy is to extrapolate from existing designs by use of TRACG. Since the CSAU methodology concluded that, with the addition of the single tube condensation data, all important phenomena were covered, the conclusion could be drawn that no additional testing is required. However, for a variety of reasons, it was deemed conservative to secure other SBWR unique data to further validate the overall predictions of the TRACG Code. The various tests are described in Subsection 1.5.3.1. However, none of these tests should be taken outside of the context of the overall program. While all SBWR features are extrapolations from current and previous designs, two features (specifically, the Passive Containment Cooling System and the Grasity-Driven Cooling System) represent the two most challenging extrapolations. Therefore, it was decided, for these two specific cases, to obtain additional test data, which could be used to demonstrate the capabilities of TRACG to successfully predict SBWR performance over a range of conditions and scales. The methodology used is to perform blind (or in 1.5-6 Requirements for Further TechnicalInformation - Amendment 1
25AS113 Rev. A SBWR standardsafety Analysis Report (Gv) some cases double blind) predictions of test facility response using only the internal correlations of TRACG. No ' tuning' of the TRACG inputs is to be performed, and no modifications to the coding are anticipated as a result of these tests. For the case of the PCCS,it is planned to predict steady state-heat exchanger pedormance in full-vertical-scale three tube (GIRAFFE),20 tube (PANDA), and 496 tube (PANTIIERS) configurations, over the entire range of SBWR expected steam and noncondensable conditions. This process addresses scale and geometry differences between the basic phenomena tests pedonned in the single tube condensation tests to near prototype conditions. Transient performance is similarly investigated at two different scales in both GIRAFFE and PANDA. TRACG GDCS performance predictions were performed against the GIST test series. Predictions were perfonned over a range that bounds expected SBWR conditions. 10CFRS2.47 10CFR52.47(b)(2)(i)(A) requires in part that: a The performance of each safety feature of the design has been demonstrated (~T t i through analysis, appropriate test programs, experience, or a combination thereof,
\"/ a Interdependent effects among the safety features of the design have been found acceptable by analysis, appropriate test programs, experience, or a combination thereof.
m Sufficient data exist on the safety features of the design to assess the analytical tools used for safety analysis over a sufficient range of operating conditions, transient conditions, and specified accident sequences, including equilibrium core conditions. , 1 The SBWR meets the above requirements, as discussed below: l m SBWR plant features have been used in earlier BWR designs and most continue in l operation today after many years and over a very large number of combined plant operatingyears of senice. While the details of the particular plant feature design for , the SBWR may differ somewhat from those in current plants, the function of each feature is substantially the same. This experience constitutes a sufIicient database to meet the requirements of 10CFR52.47(b)(2)(i)(A)(1), a In those scenarios in which SBWR safety features come into operation, no other systems are required and, therefore, system interactions are not an issue, or the system designs are similar in the SBWR and the operating plants having the feature. l(n) l The operating plant feature (s) perform under the same general conditions and for the same scenarios as are anticipated to occur in the SBWR.The operating plant database is sufficient to meet requirements of 10CFR52.47(b)(2)(i)(A)(2) and (3). Requirements for Further Technicalinformation - Amendment 1 1.5-7 l
25AS113 Rev. A SBWR Standant Safety Analysis Report O'l m Feature performance has been predicted with the TRACG computer program. TRACG has been qualified by comparison to data from experiments and operating BWRs over a wide range of reactor conditions, including temperatures and pressures during which the features are expected to operate. The TRACG analyses add to the confidence that the features will perform as expected and reinforce the GE position that the requirements of10CFR52.47(b) (2)(i)(A)(1), (2) and (3) have been met. The detailed design of specific SBWR plant equipment is, in some cases, not specified in the SBWR SSAR;in some instances, only the design requirements of the equipment are given. When this is the case, a requirement for hardware testing is not appropriate under the certification program. However, since the SBWR-specific hardware design differs from that currently in use, GE believes that testing before application of a specific equipment design in a plant should be planned. Therefore, testing of plant hardware is done prior to or during startup testing of the plant. For any SBWR constructed, equipment performance will be demonstrated. For example, overall testing of the heat rejection capability of the ICs is to be included as part of the plant startup test program. No SBWR plant will be licensed to operate until tests confirm that each IC meets the performance requirements. Experience with condensing heat exchangers in many industries gives high confidence that the requirements will be met. 1.5.3.1 Major SBWR Unique Test Programs As noted previously, the vast majority of data supporting the 1TNR design was generated using the design of the previous BWR product lines. SBWR-unique certification and confirmation tests are listed below. 1.5.3.1.1 GIST (Certification) GISTis an experimental program conducted by GE to demonstrate the Gavity-Driven Cooling System (GDCS) concept and to collect data to be used to quahfy the TRACG computer code for SBWR applications. Simulations were conducted of Design Basis Accident Loss <>fCoolant Accidents (LOCAs) representing main steamline break, j l betom dram line break, GDCS line break, and a non-LOCA loss ofinventory. Test data j have been used in the qualification of TRACG f o SBWR and documented in GEFR-l 00850, " Simplified BWR Program Gravity-Driven Cooling System (GDCS) Integrated Systems Test", October 1989, by P. F. Billig. Tests were completed in 1988 and documented by GE in 1989. 1.5.3.1.2 GIRAFFE (CcM::d=) (Confirmatorv) l GIRAFFE is an experimental program conducted by the Toshiba Corporation to l investigate thermal-hydraulic aspects of the SBWR Passive Containment Cooling System j 1.5-8 Requiremonts for Further TechnicalInformation - Amendment 1 i
25AS113 Rsv. A SBWR staodard ssrety Anairsis neport OJ (PCCS). Fundamental steady-state tests on condensation phenomena in the PCC tubes were conducted, and heat transfer coefIicients were obtained. Characteristics of nitrogen gas venting from the PCC vent to the suppression pool were obsened. Simulations were run of Design Basis Accident LOCAs; specifically, the main steamline break, bottorn drain line break, and GDCS line break. The resultant GIRAFFE containment pressures demonstrated the feasibility of PCCS to control Gli<AFFE containment pressure. Tests have been completed and results have been documented in NEDC-32215P, " GIRAFFE Passive Heat Removal Testing Program", June 1993, by K. , M. Vierow. 1.5.3.1.3 PANDA (C:r'inder;) (Certification) PANDA is an experimental program to be run by the Paul Scherrer Institute of l Switzerland to enlarge the database for TRACG qualification. PANDA is a full-vertical-scale 1/22.5 scale byvolume model of the SBWR system designed to model the thermal-hydraulic performance and post-LOCA decay heat removal of the PCCS. Both steady-l state and transient performance simulations are planned. Testing at the same thermohydraulic conditions as previously tested in GIRAFFE and PANTHERS will be I performed, so that scale specific effects may be quantified. Blind pre-test analyses using
- O
'\ TRACG will be submitted to the NRC prior to start of the testing. i 1.5.3.1.4 PANTHERS (Certification) fer "CC C:nfrmder; fer !C) l PANTHERS is an experimental program to be performed by SIET in Italy, with the dual purpose of enlarging the TRACG qualification database, as well as demonstration testing of the prototype PCCS and IC heat exchangers. Steam and noncondensables will ) be supplied to prototype heat exchangers over the complete range ofSBWR conditions to demonstrate the capability of the equipment to handle post-LOCA heat removal. ( Only steady-state testing will be performed. Testing at the same thermohydraulic conditions as performed in GIRAFFE and PANDA is planned. Blind pre-test analyses of l selected test conditions using TRACG will be submitted to the NRC prior to the start of l testing. l l In addition to thermohydraulic testing, an objective of PANTHERS is to demonstrate l the structural adequacy of the heat exchangers to exceed the SBWR expected lifetime requirement. This will be accomplished by pre- and post-test nondestructive examination, following cycling of the equipment in excess of requirements. 1.5.3.1.5 Scaling of Tests A discussion of scaling is contained in GE Topical Report NEDC-32288, " Scaling of the p SBWR Related Tests," November 1993, by G. Yadigaroglu. That report contains a Q complete discussion of the features and behavior of the SBWR during challenging events. It includes the general (top < lown approach) scaling considerations, the scaling i l Requirements forFurther Technicalinformation - Amendment 1 1.5-9
J 2SA5113Rev. A i SBWR standardsareryAnasysis aeport 9 i of specific (bottom-up approach) phenomena, and the scaling approach for the specific tests discussed above. i I O l l l 1 0 1510 Requirements for Further TechnicalInformation - Amendment 1
1 l i 25AS113Rev. A SBWR studantsatoryAasysis neport .O-h Table 1.5-1 Evolution of the General Electric BWR Product Line Year of Characteristic Plants / Features i Number introduction BWR/1 1955 Dresden 1, Big Rock Point, Humboldt Bay, KRB, Dodewaard
+ Commercial a First internal steam separation
+ lsolation condenser
- Pressure Suppression Containment BWR/2 1963 Oyster Creek
- Large direct cycle B W R/3/4 1965/1966 Dresden 2/ Browns Ferry
- First jet pump application
+ Improved ECCS: spray and flood
- Reactor core isolation cooling system (replaced isolation condenser)
BWR/5 1969 Zimmer Class
- Improved ECCS systems
- Valve recirculation flow control
/ BWR/6 1972 Grand Gulf, Perry, Clinton l + Improved jet pumps and steam separators
- Reduced fuel duty: 13.4 kW/ft (44 kW/m)
. Improved ECCS performance
+ Gravity Containment Flooder
* (option) Solid-state nuclear system protection system (Clinton only)
+ (option) Compact control room ABWR Internal Recirculation Pumps SBWR Gravity Flooder, IC, Natural Circulation I
l
/ 1 ) \] l l
i 1.5-11 j Requirements for Further TechnicalInformation - Amendment 1 l l
25AS113 R:v. A SBWR StandantSafetyAnalysis Report O Table 1.5-2 SBWR Features and Related Experience SBWR Feature Plants Testing References J IC Dodewaard , Dresden 1,2,3, Operating Plants Big Rock Pt., Tarapur 1,2, Nine Mile Pt 1, Oyster Creek, Millstone 1, Tsuruga, Nucienor, Fukushima 1 Natural Circulation Dodewaard Operating Plants Humboldt Bay Squib valves BWR1-6 and ABWR Operating Plants l SLC Injection Valves IEEE 323 Qualification Testing Gravity Flooder Perry, Clinton, Grand Gulf Operating Plants Upper Pool Dump System, Preoperational Suppression Pool Flooder Testing System Intemal Steam BWR1-6 and ABWR Operating Plants Separators Chimney (Core to Dodewaard, Humboldt Bay Operating Steam Separators) FMCRDs ABWR ABWR Test / Development Program (Demonstration at La Salle Plant) Automatic All BWRs Operating Plants Depressurization Valves (MSIVs) Pressure BWR1-6 and ABWR Mk I, Mk ll, Mk 111 and Suppression ABWR Tests Horizontal Vents BWR6 and ABWR , Perry, ABWR Testing Horizontal Vent Grand Gulf, Clinton, River Confirmatory Tests, Bend, etc. Part i NEDC-31393, CLASS 111, March 1987 Quenchers BWR2-6 and ABWR Operating Plants PCC (Dual Function Operating Plants, RHR HX Operating Plants, Heat Exchangers) Steam Condensing Mode PANDA, GIRAFFE, SIET Solid State Control ABWR, Clinton Operating Plants, System (NSPS) Clinton 1.5-12 Requirements for Further Technical Information ~ Amendment 1
25A5113Rsv. A \ SBWR standantseraryAnarysisneport l 'V Table 1.5-3a TRACG Oualification (Section 1) i l Key BWR and SBWR Phenomena identified (summary) ' + Phenomena are ranked according to importance and major phenomena identified. BWR/SBWR Region Phenomena identified Major Phenomena Lower Plenum 10 7 Bypass 12 9 Core 40 28 l Guide Tube 7 5 l Downcomer 7 6 Upper Plenum 4 2 Jet Pump 7 3 Recirculation Pump 4 1 l Steam Separator 3 3 Steam Dryer 2 0 l Steam Dome 3 0 Stesmiine 6 3 Recirculation Line 5 2 Containment 10 5 Pump Motor 1 0 l
\
Feedwater System 3 0 i Total 124 74 l C l l
\ %f Requirements for Further TechnicalInformation - Amendment 1 1.5-13 t
P
Table 1.5-3b TRACG Qualification (Section 2) g i BWRrPlflT (Phenomena identibcahon and flanking Table ) GOVERNING P'HENOMENA L0CA (FOCUS: PCT. WATER LEVEL. ENERGY CEPOSITION IN FUEL, AND CONTAINMENT PRESSURE (FOR SBWR)) l TRACG PLANT BWR/2 BWme BWR/5-4 ABWR SBWR MODEL BREAK SML BAK LRG BRK SK BRK LRG BRK SMLBRK LRG BRK SMLBRK LRG BRK SMLBAK LRG BRK PHASE BLDN l R/R BLDN l R/R BLDN l R/R BLDN l R/R BLDN l RR BLDN l R/R BLDN l R/R BLDN l RS BLDN l RR BtDN l R/R PHENOMENA R A N K I N G (0.NOT APPLICABLE.1.2.3. LOW. 4.5.6.MCD. 7.8.9.HIGH) A REGION: LOWER PLENUM (FOCUS: FLASHINGINITIATION YtME AND REFILL TIME) A1 FLASHWGr 5 5 8 7 5 5 8 7 5 5 8 7 5 5 8 7 5 5 8 7 y REDISTRIBUTK)N A2 HEAT SLAB STORED 5 7 6 6 5 7 5 7 5 7 5 7 5 7 5 7 5 8 5 e y u ENERGY RELEASE p A3 TWO-PHASE LEVEL 2 3 2 3 5 7 4 e 5 7 7 7 5 7 7 7 5 a 7 e y
- w l g (SEO UNCOVERY TIME 8 A4 TWO-PHASE LEVEL TWE) 0 0 0 0 3 5 8 6 4 5 8 5 0 0 0 0 0 0 0 0 y l,
l UP UNCOVERY TNE* b
- h A5 VOt0 DISTRIBtJTION 4 6 8 4 's 7 8 7 4 4 4 4 4 4 4 4 4 4 4 4 y k As MtxWG/ CONDENSATION 3 4 3 4 3 7 3 7 2 4 2 4 2 4 2 4 2 8 2 8 y p VOID COLLAPSE f A7 As CCR
- LP TO BYPASS CCFL LP TO GutDE TUBE 2
I 2 1 2 1 2 2 2 1 3 1 3 1 2 1 2 1 3 3 2 2 2 3 3 2 2 2 3 3 2 y y h 1 1 1 1 1 1 1 1 1 A9 PRESSURE DROP 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 y ! E !! Er B BYPASS (SUB-FOCUS: INVENTORY. LEAKAGE FLOW) S Bt FLASHING 5 5 5 4 5 5 8 6 7 5 7 4 7 5 7 4 e 8 e a y e B2 TWO-PHASE LEVEL 4 0 4 8 4 0 4 e 5 s 6 a 5- e 5 6 5 8 5 e / B3 STORED HEAT 4 8 4 8 4 8 4 8 5 4 4 3 5 4 4 3 5 4 4 3 y l (CORE PLATEIBLADES) h B4 CCFt>CCFL BREAKDOWN 3 5 4 4 3 5 4 e 7 7 7 7 7 7 7 7 7 7 7 7 y g B5 CCFL/CCFtBREAMDOWN 4 4 4 4 4 4 4 8 6 6 7 7 6 6 7 7 6 6 7 7 y W (TOP OF BYPASS)
- R o 1 - A l
O __ O O
.O O V U d g Table 1.5-3b TRACG Oualification (Section 2) (Continuad) m c I T
2 3 g BWR ElRT (Phenomena ldentification and Banking Table )(Co,4% GOVERNHG PHENOMENA
$e f
LOCA (FOCUS. PCT. WATER LEVEL. ENERGY DEPOSITION IN FUEL. AND CONTAINMENT PRESSURE (FOR SBWR)) M R TRACG
$ PLANT BWNV2 BWR/4 BWR/!Mi ABWR SBWR MODEL k BREAK SMLORK LRG BRK SK BRK LRG BRM SML BRK LRG BRK SMLBRK LRG BAK SMLBRM LRG BRM PHASE BLON l WR BLDN l R/R BLDN l NR BLDNlWR BLDN l WR BLONlNR BLDN l R/R BLDN l RR BLDNlRR BLON l R/R h
l PHENOMENA R ANKINO (0.NOT APPLICABLE 1.2.3. LOW.4.5.6-MEDtuM. 7.8.hHIGH) 3 B6 CHANNEL-BYPASS LEAKAGE FLOW 4 4 4 8 4 8 4 0 7 8 5 8 7 8 5 8 7 8 5 8 V I y B7 REFN.L t 8 3 3 1 4 1 4 1 8 8 8 1 8 1 8 8 8 1 8 g 88 LPCllNTERACTION/ 0 0 0 0 t 1 I 1 1 7 1 6 1 7 1 6 0 0 0 0 y $a { CONDENSATION
$ B9 30 EFFECTS (LPCI) t 3 1 3 I t t t t 5 1 4 1 5 1 4 0 0 0 0 v O 3 to O Bio GEYSERING DURING 2 2 3 3 1 4 1 8 1 4 1 5 1 4 1 5 1 4 1 5 / :n REFILL j Bil CONTROL RODS 1 1 1 1 1 1 I t t t t t t t t t t t t y b B12 PRESSURE DROP 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 5 y C CORE /90NDLE(SUS-FOCUS: PCT.REFILLTIME.90NDLEINVENTORY)
C1 NUCLEATE BOILING 2 4 2 4 2 4 2 4 2 2 2 2 2 2 2 2 2 2 2 2 / C2 SUSCOOLED DOILING 2 2 2 2 2 2 2 2 3 3 3 3 3 3 3 3 3 3 3 3 y C3 VARIA8tE GAP CONDUCTANCE 3 2 3 2 3 2 3 2 4 3 5 4 4 3 5 4 4 3 5 4 / h C4 FLASHING 3 1 3 1 3 3 3 t 5 5 5 5 5 5 5 5 8 8 8 8 / g C5 SEO INLET UNCOVER 1 3 1 3 1 3 1 3 7 6 8 7 7 6 8 7 7 6 8 7 y VAPOR FLOW SPLIT C8 CCFU t 3 1 3 5 3 1 5 3 3 4 8 7 4 8 8 7 6 4 8 y CCFL BREAKDOWN (SEO) f C7 CCFU t 5 1 3 1 5 1 5 6 7 7 7 5 7 4 6 5 7 4 6 y CCFL BREAKDOWN UTP l M T G A __ ______________________m_ _ __ _
;; Table 1.5-3b TRACG Qualification (Section 2) (Continued) m 3; #
own Pinnehenomenaldentification and Banking Table )iConi-.m GOVERNING PHENOMENA L0CA (FOCUS. PCT, WATER LEVEL, ENERGY DEPOSITION IN FUEL AND CONTAINMENT PRESSURE (FOR SBWR)) TRACG PLANT BWR/2 BWR/4 BWR/5-4 ABWR 58WR MODEL BREAK SML DRK LRGBRK SK BAK LRG BRK SMLBRK LRG BRK SMLBRK LRG BRK SML BRK LRGBRK PHASE SLDN l R/R BLDN l R/R BLON l RrR BLON l R/R BLON l R/R BLDN l R/R BLON l RR BLDN l RR DLON l RR BLDN l RR PHENOMENA R A N K I N O (0-NOT APPLICABLE, t.2.3. LOW,4.5.8-MEDIUM,7.8.9.HtGH) C. MULTIPLE BUNDLE i 3 1 6 1 3 1 5 4 8 8 7 4 6 6 7 4 6 6 7 y HYDRAULICS / FLOW REGEAES Ce PARALLEL CHANNEL 1 3 i 5 1 3 1 5 5 6 e e 5 5 6 6 5 5 6 6 y FLOW OfSTRIBUTION C10 VolO DISTRIBUTION 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 V 2 Ctl BUNDLE-BYPASS 4 0 4 4 4 4 4 3 7 8 5 6 7 8 5 e 7 e 5 8 y -* fF C12 LEAKAGE FLOW NATURAL 3 3 7 3 7 3 7 7 2 7 7 7 7 7 7 y U 1 1 7 7 7 g CIRCULATION
, rtow 3 C13 DRYOUT/ST 2 2 7 5 2 2 7 5 e e o e o e e e s e 9 8 y C
C14 FILM 3 5 6 9 3 5 6 9 3 4 8 2 8 4 8 2 3 4 8 2 y BOILING C15 FILM BOILtNGr 3 5 8 8 3 5 6 8 6 7 6 3 6 7 6 3 6 7 6 3 y y DISPERSED FLOW
} Cis RADIATION i 4 i e i 4 1 e i 2 1 3 i 2 i 3 i i 1 i y (F C17 STEAM COOLING t 7 8 9 1 7 8 9 2 3 2 5 2 3 2 5 t I t 1 y Ct. FUEL ROD t 6 t 8 1 4 4 8 2 2 3 3 2 2 3 3 2 2 3 3 y k PERFORATIONr E.
l M-W REACTION M i c,e C2o o m ,COOuNG REwET, i i e
, e i
i e i e e 1 2 1 e 7 i 5 e 7 i e i o t o i o i e i y J5 l 3 BLOWOOWN PHASE l j lg Os REWETCONOUCTION CONT rot t ED 1 3 1 4 1 2 1 2 3 2 3 2 3 2 3 2 1 1 1 1 y M{ Gi' l $ ! a 1 - O _ O _ - O
p ( ) (
! I
(. i i L ,/ G G p Table 1.5-3b TRACG Qualification (Section 2) (Continued) m E. a y
!B BWR :PIRTEhenomena identification and Banhing Tabte hConimu.oi lQ GOVERNING PHENOMENA y LOCA (FOCUS. PCT, WATER LEVEL. ENERGY DEPOSITION IN FUEL. AND CONTAINMENT PRESSURE (FOR SBWR)) ' TRACG
{E PLANT BWRr2 BWN4 BWRr54 SML BRK LRGBRK SMLBRK ABWR LRG BRK SBWR SML BRM LRG BRK MODEL BREAK Sna. BRK LRG BRK SML B AK LRG BRM W PHASE BLDNlNR BLDN l RtR BLON l R/R BLDN l R/R BLDNlNR BLON l NR BLDN l R,Tt BLDN l RrA BLDN l FVRBLDN l R/R PHENOMENA R A N K I N G (0.NOT APPLICABLE; 1.2.3. LOW,4.5.6. MEDIUM. 7.8,9.HIGH)
. C22 CHAWBYPAS HT 1 1 1 3 1 1 I 3 3 3 3 3 3 3 3 1 1 1 y t 2 2 2 3 2 2 2 3 2 2 2 3 /
h C23 WATER ROD HYDRAULICS 1 1 1 1 1 1 8 g C24 PRESSURE DROP 5 5 5 5 5 6 5 5 5 5 5 5 5 5 5 5 5 5 5 5 y C25 DECAY HEAT 2 9 2 0 2 9 2 9 2 9 2 9 2 9 2 9 2 9 2 9 J 4
- c. -- g g C26 STORED ENERGY 9 2 9 2 9 2 9 1 9 2 9 2 9 2 9 2 9 2 9 2 / $
a t: D GUIDE TUBE (SUB-FOCUS: REFILLt F g Dt FLASHINGI 2 2 2 2 2 2 3 6 5 7 6 6 5 7 6 6 5 A 4 7 6 / REDISTRIBUTION 02 CCFL: TOPOF GT t 2 1 2 1 2 8 8 6 6 7 7 6 6 7 7 6 6 7 7 y 03 CONDENSATION t 2 1 2 1 2 3 4 2 6 2 7 2 6 2 7 2 6 2 7 / D4 REFILL t 2 1 2 1 2 1 4 3 7 3 8 3 7 3 8 3 7 3 8 / D5 GEYSERING DURING 1 2 1 2 1 2 1 4 2 3 2 6 2 3 2 5 2 3 2 5 / REFILL De GT-LP LEAKAGE 1 1 1 1 1 1 1 I 2 2 2 4 2 2 2 4 2 2 2 4 y k 07 PRESSURE DROP 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 / k a, E DOWNCOMENANNULUS (SUB-FOCUS: UNCOVERY TIMES FOR JET PUMP SUCTION. RECIRCULATION, TWO PHASE LEVEL) EI RECIRC. SUCTION 2 2 8 3 2 2 8 3 2 2 7 2 2 2 8 2 2 2 7 2 V g BREAK /OTHER BREAK a UNCOVERY (TWO-PHASE A BREAK FLOW) % E2 votO PROFILE / 2 2 8 3 2 2 8 3 7 7 5 7 7 7 5 7 7 7 5 7 / TWO-PHASE L EVEL g 4 - R
g Table 1.5 3b TRACG Qualification (Section 2) (Continued) m m W owa c cinuchenomena identification andItanhina Tab!e )(C-am GOVERNING PHENOMENA L0CA (FOCUS' PCT, WATER LEVEL. ENERGY DEPOSITION IN FUEL, AND CONTAINMENT PRESSURE (FOR SBWR)) TRACG PLANT DWR/2 BWRr4 BWRrS-4 ABWR SBWR MODEL BREAK SMLBRK LRG BRK Sam. BRK LRGBAK SMLBRK LRG BRK SMLBRK LRG BRK SML BRK LRG BRK PHASE BLON l IVR BLDN l R/RBLDN l R/RBLDN l R/R BLDN l R/R BLDN l RrRBLDN l RtR BLDN l RR BLDN l R/RBLON l RR PHENOMENA R ANKING (0.NOT APPLICABLE; 1.2.3. LOW,4.5.8. MEDIUM,7.8.bHIGH) E3 GDCS INTERACTION / 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 8 8 8 8 y CONDENSATION E4 3-D EFFECTS 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 3 V E5 HEAT SLABS 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 8 8 8 8 y E8 FLASHING 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 8 8 8 8 / 0 0 p E7 ISOLATION CONDENSER INTERACTION 4 4 4 4 0 0 0 0 0 0 0 0 0 0 7 5 7 5 / -. G g 3 5 F UPPER PLENUM (SUS-FOCUS: etVENTORY, SUBCOOttNG TIME.CCFL BREAKDOWN TIME) b k F VolO DISTRIBUTION 1 1 I 2 1 1 1 3 3 7 3 8 3 7 3 8 7 7 7 8 y % TWO-PHASE LEVEL F2 ECC INTERACTION / 2 2 1 6 2 2 1 5 2 7 1 0 2 7 6 9 2 1 1 1 y 7 MIXtNG/SUBCOOLING DISTRIBUTION N 3 V g F3 CONDENSATIONIAIR 1 7 1 7 1 1 1 1 1 2 1 1 2 1 3 1 2 1 3
- F4 SPRAY DISTRIBUTION t 1 I O I t t 0 t 8 8 8 0 0 0 0 0 0 0 C y i
o G JET PUMP (SUB-FOCUS UNCOVERY/ RECOVERY TIMES, FLOW DISTRIBUTKNP/ CORE) NOT APPLICABLE TO ABWR/SBWR l-g GI JET PUMP O O O O 2 2 3 2 8 2 8 2 0 0 0 0 0 0 0 0 / U. CHARACTERISTICS. S STEADY STATE h G2 JP CHARACTERISTIC 0 0 0 0 2 2 5 1 4 1 7 1 0 0 0 0 0 0 0 0 y h b COASTDOWN 3 G3 JP CHARACTERISTIC 0 0 0 0 2 2 5 1 6 4 6 5 0 0 0 0 0 0 0 0 / - E REVERSE FLOW $. ! 4 E
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= a e Requirements for Further TechnicalInformation - Arnendment 1 1.5 19
g Table 1.5-3b TRACG Qualification (Section 2) (Continued) m 8 CD DWft: ElRT (Phenomena ldentification and flanhing Table ) # Cod ===d) GOVERNING PHENOMENA L0CA (FOCUS. PCT, WATER LEVEL. EhERGY DEPOSITION IN FUEL. AND CONTAINMENT PRESSURE (FOR SBWR)) TRACG PLANT BWR/2 BWR/4 BWR/S-4 ABWR SBWR MODEI . BREAK 8MLBRK LRG BRK 884 BRK l LRG BRK SMLBRK LRG BRK SMLBRK LRG BRK SML BRK LRG BRK PHASE BLDN l R/R BLON l RtRBLDN l R/R BLDN l RR BLON l R/RBLON l RR BLON l R/R BLON l RR BLDN l R/R BLON l R/R PHENOMENA R A N K 1 N G (0.NOT APPLICABLE; 1.2.3. LOW. 4.5.6. MEDIUM. 7.8,9.HIGH) J DRYER (SUS-FOCUS: NONE) J1 DRYER 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 d CHARACTERISTCS/ CARRYOVER 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 / 4 J2 PRESSURE DROP 2 2 5 8 K STEAM DOME (SUB-FOCUS: NONE) i3 5. K1 HEAT SLAB EFFECT t t t t i 1 1 1 4 2 3 2 1 4 2 3 2 1 3 4 5 2 3 2 1 3 y
/
f k R K2 CONDENSATION ON 1 1 1 1 1 t i 1 5 1 3 5 1 1 WALLS K3 PRESSURE DROP 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 / Q 5 L STEAMLWE (SUS-FOCUS: CRITEAL ROWRATE) Lt CRITICAL 1 1 5 I t t 5 1 1 1 5 1 1 1 5 t l 1 5 1 / E FLOW . k L2 DROPLET ENTRAINM t 1 5 1 1 1 5 1 4 3 5 3 4 3 5 3 4* 3 5 3 / h L3 PRESSURE DROP 1 1 5 1 1 1 5 1 8 3 6 6 4 3 6 6 6 3 6 6 / h a. il
$ M RECIRCULATION LINE (SUS-FOCUS: TIMES FOR FLASHING. CRITICAL FLOW) NOT APPLICABLE TO ABWR/SBWR Mt CRITICAL FLOW t l1 4 5 t t 8 5 8 2 9 6 0 0 0 0 0 0 0 0 V b M2 FLASHING t e 7 1 1 I 7 1 3 2 7 1 0 0 0 0 0 0 0 0 V u3 CCFL: AIR IHf 1 2 4 1 1 2 2 1 1 2 2 0 0 0 0 0 0 0 0 V [g.
( 1 l TWO-PHASE FLOW OUT u4 GEYSERING t t t i 1 1 1 1 3 1 5 1 0 0 0 0 0 0 0 0 y l
- R 1
O O O
N g Table 1.5-3b TaACG Oualification (Section 2) (Continued) e s. a W 9 .
= i a '
y BWR PIRT (Phenomena Identification innd Bankmg Table ) cceo.or ' 2 GOVERNING PNENOMENA LOCA' !
$ (FOCUS: PCT. WATER LEVEL. ENERGY DEPOSITION IN FUEL. ANO CONTAINMENT PRESSURE (FOR SeWR))
l TRACG f g- PLANT 4WRr2 eW=4 eWRtS-4 AeWR SeWR MODEL l s eReAx SML em LRG em 8.S.eRx tRG eRx .S a sRx LRG em smem tRGem SuteRx LRGem [ e, o PHENOMENA PmSE moNI R eteN I =A moNI R moNI R moNI R mon I .R mm l =R ma l =R am l =R am l =R R A N x 1 N G (0.NOT APPLICAelE; 5.2.3. LOW. 4.5,4. MEDIUM. 7.8.e.HIGH) [
- MS STRATICIATION 1- 1 1 3 1 1 1 4 6 '2 5 1 0 0 0 0 0 0 0 0 y 1
1 5 N CONTA9dMENT(SUS-FOO3S: PRESSURE) ** Nt KOWDOlflDI FLOW 5 6 5 6- y + e _ . 3, N2 . MIXedG: DRWifELL' -7 7 7 7 y C
-. MIX 80G: SUP. POOL e a a- a .g i N3 NON-CONDENSISLE INTERACTION 2 7 3 7 y ( l i.
N4 OOCS FLOW 0 s e a v NE 'PCCS S S 6 8 V , PERFORMANCE NS PRESSURE DROP SSWR FLOWS DRNEN OV SMALL DELTAP-, . 3 e 8 e / i (10.PCCS 00CS Vs.EOuALISATION LINE FLOWS-W4LL FRACTION & LOCAL LOSS) - , N7 ICtPCCS POOL SSWR ICSCCS TueE INTEmeOR-e e 9 9 e y (SECONDARY SIDE) ' SSWR FREE SURFACESfWALLS-+ 7 7 7 7 SueSCOLEOSeuCLEATE - (89. GOCS POOL. IC-PCCS WATER SOX. ETC ) ' {; eOILING NEAT TRANSFER SSWR WEWIfELL PLUME WITN STEAM & GAS SUSSLES-+ e e a e NS IC#PCCS POOL ' 6 6 6 6 / (SECONDARY SIDE) SUSCOOLEDeeUCLEATE - SOIL 90G HEAT TRANSFER 1 T 3: , m m. _ . . . _ _ . . . . . _ __m..._._._m.m____._.____.__ _ . _ _ . _ _ _ - ____.___ __ _ . - , ~ . - , - . -. ,.~..v.-. . , , , , . . . - . , . . . . . . #,_ .v
s Table 1.5-3b TRACG Qualification (Section 2) (Continued) (4 'e tg 18 BWil- ElRT (Ellenomena idenhfication and Banlong Mle } (Coanu.o> GOVERNING PHENOMENA L0CA (FOCUS: PCT. WATER LEVEL. ENERGY DEPOSITION IN FUEL. AND CONTAINMENT PRESSURE (FOR SBWR)) PLANT TRACG BWR/2 BWR/4 BWN5-6 ABWR SBWR MODEL BREAK BMLBRK LRGBRK SKBAK LRG BRK SMLBRK LRG BRK SML BRK LRG BRK SML BRM LRG BRK PHASE BLDN l NR BLDN l NR BLDN l R/R BLON l NR BLON l R/R BLDN l RfR BLON l WR BLDN l WR BLDN l RR BLDN l %R PHENOMENA R A N K I N O (0.NOT APPLICA8tE; 1.2.3. LOW. 4.5.6. MEDIUM. 7.e,9.HIGH) N9 INTERFActAL SHEAR IN SBWR (PHASE-SEPARATION OF TWO-PHASE 4 4 6 6 y VENT MIXTURE EXITlHG HORIZONTAL VENT)-e NIO CONDUCTION HEAT SBWR (SUPPRESSION CHAMBER.: TRANSFER IN COMPOSITE STEEL LINERCONCRETE WALL; 7 7 7 7 y g WALLS ORYWELL-WETWELL VENT WALL & g OtAPHRAGM FLOOR) ; 2 " s. c iP s 5 e k 2 h,
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/O L L RIT N O H P EE I
C E P O HS TR I L EA N S N l SI H C 9 OU D N IO L I S WR EG A O TOO Fl RR UO E S E O AD AE F F Cu O Xo I l H A E LE EN WE WP C C OT W O R F T S(C CG I L P R R FR HE T( S Ip V MV C C LS H T T Y L QV A t A 2 A 3 A 4 A 5 A S A 7 A A s 0 A1 8 t B 2 8 3 8 4 8 N O s . i1a E TIfhhg. E 5ggE+ 3 Ib1g . g* Eu
g Table 1.5-3b TRACG Qualification (Section 2) (Continued) m 2 % owa = einT (Pnenomena identlication and Banking Table ) TR A N S 1 E N T S GOVERNING PHENOMENA TRANSIENTS (FOCUS: DELTA (CPRylCPR ; MAX! MUM PRESSURE, WATER LEVEL; ENERGY DEPOSITION IN FUEQ PHENOMENA
- RANMiNO (0.NOT APPLICABLE; t .2.3. LOW,4.5.6. MEDIUM. 7.8.9.HtGH)
TRANSIENT TYPE PRESSURIZATION DEPRESSURIZATION FLOW / STABILITY COLD WATER ROD WITHDR AWAU TRACG MODEL TRANStENTS TRANSIENTS T RANSIENTS TRANSIENTS DROP TRANSIENTS B5 CCFUCCFL BREAKDOWN I 2 1 1 1 y (TOP OF E!YPASS) B6 CHANNEL-BYPASS 5 5 3 3 3 y LEAKAGE FLOW B7 REFILL 0 0 0 0 0 y B8 LPCI CONDENSATIONI O O o 0 0 y INTERACTION B9 3-D EFFECT3 (LPCI) 0 0 0 0 0 y BIO GEYSERINGr SBWR-+ f(SBWR) 7(SBWR) 9(SBWR) 1(SBWR) 1(SBWR) y $ STARTUP y f bit CONTROL ROD VOLUME 3 0 0 0 3 / @ g DISPLACEMENT e (SCRAM, ROD WITHORAWAQ h 3 C REGION: COREfBUNDLE Q CIX NUCLEAR MODEL C C?AX VotD COEFFICIENT 9 8 8 e s y CtBX DOPPLER COEFFICIENT S 5 7 7 8 y ( C CX SCRAM REACTIVITY e 3 5 4 3 y y
- g. CiDx 3-D KINETICS 7 5 5 7 e I y C2X THERMAL HYDRAULIC MODELS h
(3 C2AX INTERFACIAL SHEAR S 9 7 8 8 y E p, C29X SUBCOOLED VotD MODEL 8 8 7 e a y @ I C2CX TURBULENT t 1 1 1 l y uixlNG MODEL l b C3M FUEL HEAT TRANSFER C3AX PELLET HEAT 7 7 7 7 7 y g. k DIST RIBUTION 5
?,
5o
- :t
? O - O - - O
V) N p Table 1.5-3b TRACG Qualification (Section 2) (Continued) m
- s. 9
!It ewa= eini tehenomena wenecation and nanhina Taue )
T R A N S I E N T S { GOVERNING PHENOMENA h TRANSIENTS [ FOCUS. DELTA (CPR)/ICPR ; MAXIMUM PRESSURE, WATER LEVEL; ENERGY DEPOSITION IN FUEL l PHENOMENA
- R A N K 1 N G (0 NOT APPLICABLE,1,2.3. LOW,4.5.6. MEDIUM,7 S.9.HIGH)
TRANSIENT TYPE PRESSURIZATION DEPRESSURl2ATION FLOWISTABILITY COLD WATER ROD WITHDRAWAU TRACG MODEL
-2. TRANSIENTS TRANSIENTS TRANSIENTS TRANSIENTS DROP TRANSIENTS C3BX PELLET HEAT TRANSFER S S S 6 6 y g PARAMETERS l C3CX C3OX GAP CONOUCTANCE PROMPT NEUIRON HEATING 7
S 7 6 7 6 7 6 7 6 y y
.g-l ./
b Cl NUCLEATE BOILING S 5 5 6 5 y C2 SUBCOOLED SOILING S 5 6 6 6 y f 3 C3 VARIABLE CAP CONDUCTANCE e 6 5 5 6 / $ C C4 FLASHING 3 6 4 4 7 y C5 SEO INLET UNCOVERY1 0 0 0 0 0 y b VAPOR FLOW SPLIT C6 CCFU 0 0 0 0 0 / CCFL BREAMDOWN(SEC) C7 CCFU 0 0 0 0 o y CCFL GREAMDOWN UTP Ce MULTIPLE CHANNEL 7 7 7 7 7 y EFFECTS CSX VOID COLLAPSE e 3 5 7 7 y P C10 VOID DISTRIBUTION 8 8 8 8 8 y Cll SUNDLE-8YPASS 5 5 6 4 5 / LE AKAGE FLOW
~
Cl2 NATURAL CIRCULATION SWRS* 3 3 3 3 3'~ y FLOWS SBWR-+ ABWR., 8(S8WR) 3(ASWR) O(SBWR) O(ABWR) 9(SBWR) 9(ABWR) 8(SBWR) 8(ABWR4 7(SBWat g 7(ADWR) DRYOUTt9T 8 Cl3 9 9 9 9 l V k r N W
g Table 1.5-3b TRACG Qualification (Section 2) (Continued) m W W Dwa r einT (Pnenomena toontincation and nanhing Tatate ) TR A N S 1 E N T S GOVERNING PHENOMENA TRANSIENTS [ FOCUS DELTA (CPRylCPR ; MAXIMUM PRESSURE, WATER t EVEL ENERGY DEPOSITION IN FUEL] PHENOMENA
- R A N K i N G (0-NOT APPLICABLE.1.2.3. LOW. 4.5.6= MEDIUM. 7.8,9*HIGH)
TRANSIENT TYPE PRESSURIZATION DEPRESSURIZATION FLOW / STABILITY COLD WATER RODVITHDRAWAL/ TRACG MODEL TRANStENTS TRANSIENTS TRANSIENTS TRANSIENTS DROP TRANSIENTS CI4 FILM BOltINGI BWRS* 2 2 2 2 2 / LOW FLOW ABWR-e 2 2 e 2 2 C15 FILM BOILINGI BWRS-+ 2 2 P 2 2 y DISPERSED FLOW ABWR-e 2 2 4 2 2 C16 RADtATION 1 1 ""1 1 1 / Cl? STEAM COOLING 1 1 1 1 1 / Clt FUEL ROD 1 1 1 1 1 V PERFORATION / M-W REACTION y h- C19X TMIN(MWfMUM STABLE e s 6 6 1 y @ FILM 90lLING TEMP) 5 $ b g C21 REWETICONDUCTON CONTROLLED 1 1 1 1 1 / g C22 CHAN-8YPASS HT 1 1 1 1 1 / p C23 WATER ROD 3 3 3 3 & y
- a HYDRAUllCS C24 COME PRESSURE DROP 7 7 7 7 5 y f
D REGION: GUIDE TUGE / 3 D1 FLASHNGt REDISTRIBUTION 1 e 1 1 1 / g = g D2 CCFL TOP OF GT 0 1 0 0 0 y 5 @ D3 CONDENSATON 1 0 0 0 t y p
. D4 REFILL 0 0 0 0 0 / [
I DS GEYSERING DURING 0 2 0 0 0 y REFILL > t l g
'I GET L P L E AM AGE BWAS4 sewn .
1 t 1 t 1 7(SBWR) 1 1 1 1 y $, C;- 5 a O O O
l ; ' i t t !l 1 i1l,
%O4$sb 8 o
t9 .
/..
L E D O M G y y y / y y y y y V y /
] C L A E R U T .
F N I N ttS O I AT ) T I
)
H WN AI E R S G RS W . O I ON 8 P S E H. 9 HA TR 0 1 0 1 11( 0 0 0 1 D 8 IT Y WP G 7, DO . R M OR E U RD
) N I D
E d e L E
; E M. )
u V 6 R _ n _ E - . 5 W 4, RS L i R ET 9 t S( - n ) E T W TN A 0 7 0 6 8 67 7 0 5 4 7 . o le W A O WES C ( b a E; L. 3 DN LA .
) T R 2 OR .
g U 1 CT 2 n S E n i k E S L - i o nS a R P B A C Y ) R t l T c f TA M I L i t t S W _ d U P S e n N M P 8T AN S S A ( ( a E nNM X A I T O r S T S E 0 7 0 4 4 69 0 0 o 4 7 n W o O M WN i o aEN c E S; T R N. OA 0 LR FT ( t a li tI H NPC - n P EM O c e S N _ i f i dSGN NR i A P N O t I T ) l a a I R(C K 2A R -. u nNN e R TAT t W L N nS S Q m E E uT S D A SN ( oAV SE 0 7 0 4 e 87 0 0 O 4 7 . G ne O S R ES s C hHG P U C RN PA A R (T T l O F l ER DT - - T f E l N O ) I T R b : AS W 3- l f ZT I S . 5 W RN S( B UE SS I 0 7 9 4 T S8 0 0 0 4 7 1 SN EA e RR l PT - b a d - T , O P A
+- e ee I S R M SR R R U RW _
E W W N W9 M 8 E O SN 8 L P S8 P C W O I R M U N O T E W L A J F F S P S O P r G N T RN f D M N Nf) R O E C L OE U lG rN : D R YA J S I NRE E V O I R E OEN NI I TVI M O L AI I T C T S E OER E TN P N T IO f N U I I P I TVB CO SOC W A UEI B T SE R OLM TON
/
Y E L O I E A T COE ILE AI RT R/ CC RIET ABIO T I R IT RA T N - N UCS EA T ET CW TS EY DEJ . E T FS T SN RUT A S M E N 8. UH NA OA RH NN TR EN SAE E9U H L T#B S N e D PTS AO N MCY RD - N AE FI N I O CKP PP- IE WV D PU NGI E Y UAD AT N E S N O I S R IA GU CEO OO DO SD CN DO EC RAY WR O e O DOO INRT DN A O PRA I R G TAE HS CA H A R EL ERW oW l EN 8P E O IWU V R CIMS CI I O P E EHT PO I T RU RBT VT GC FU SS R (TT EMD C S R JCS JC m l 2 3 T 1 2 3 4 1 2 E E E E E F F F F F G G G pei2=E T Df j' hleg l ls; *PD
?
s' ' ' : _
g; Table 1.5-3b TRACG Quai!!icatler, (Section 2) (Continued) m 2 % own :Emi mnenomena identticauen and flanNng Table ) l T 11 A N S I E N T S GOVERNING PHENOMENA TRANSIENTS l FOCUS. DELTA (CPRk1CPR ; MAXtMUM PRESSURE; WATER LEVEL; ENERGY DEPOSITION IN FUEL] PHENOMENA
- R A N K I N O (0.NOT APPLICABLE. l.2.3. LOW,4.5.6. MEDIUM 7.8.9.HIGH)
TRANSIENT TYPE PRESSURIZATION DEPRESSURtZATION FLOW / STABILITY COLD WATER ROD WITHDRAWAU TRACG MODEL TRANSIENTS TRANSIENTS TRANSIENTS TRANSIENTS DROP TR ANSIENTS G3 JP CHARACTERISTICS / 1 1 7 7 / REVERSE FLOW G4 FLOW COASTDOWN 7 7 7 7 1 / G5 LPCI CONDENS ATION O O O O O V IN JET PUMP G6 TWO-PHASE FLOW 0 2 0 0 0 y COMBINATIONS: SUCTIOP& DRIVE 4 07 PRESSURE DROP 7 7 7 7 1 / H REGION: RECtRCtAATON V U g- PUMP p l g H1 PUMP CHARACTERISTICS STE ADY STATE 3 3 3 3 1 / (
@ H2 PUMP CHARACTER!STICSr 7 7 7 3 I / ' COASTDOWN 2 H3 PUMP TWO-PHASE o 3 0 0 0
- n /
{ p i DEGRADATION REGION: SEPARATOR V h 11 SEPARATOR 7 7 7 7 1 y { CHARACTERISTICS: CARRYUNDER k p 82 SEPARATOR UA 7 7 7 4 1 y g f G SEPARATOR PRESSURE DROP 7 7 7 7 t y J DRYER y l J1 DRYER BWRS-e 3 3 3 3 1 / h y CHARACTERISTICSI SBWR-e 3(SBWR) 3(SBWR) 3(SBWR) h 7(SBWR) f(SBWR) CARRYOVEFV g (g J2 PRESSURE DROP 1 l 1 1 1 1 y 4' w 3
; R O O O
b O a ' f Table 1.5-3b TRACG Oualification (Section 2) (Continued) g s a. W
$ BWH n PIRT (Phenomena idenHcaWon and RanNng Table )
$ TR A N S I E N T S GOVERNING PHENOMENA
{ I TRANSIENTS [ FOCUS: DELTA (CPRylCPR ; MAXIMUM PRESSURE WATER LEVEL, ENERGY DEPOSITION IN FUEll PHENOMENA
- R A N K I N G (0.NOT APPLICABLE.1.2.3. LOW. 4.5.5. MEDIUM. 7.8.9.HIGH) 4 TRANSIENT TYPE PRESSURIZATION DEPRESSURIZATION FLOWrSTABILITY COLD WATER ROD WITHORAWAU TRACG MODEL h
- TRANSIENTS TRANSIENTS TRANSIENTS TRANSIENTS DROP TRANSIENTS K REGION: STEAM DOME y K1 HEAT StA8 EFFECTS 3 3 3 3 y h 1 y
l e. K2 CONOENSATION ON WALLS 3 3 3 3 1 o K3 PRESSURE DROP 1 1 1 1 1 y b L REGION: STEAM LINE y L1X PRESSURE LOSS COEFF. 9 4 4 4 1 y u d (FORWARO 4 REVERSE) g L2X ACOUSTIC EFFECTSr S 7 1 1 1 y $
; GEOMETRY g Lt CRITICAL FLOW 6 T '1 1 1 y ,%
I t y k L2 DROPLET ENTRAINMENT 1 1 1 M REGION: RECIRCULATION y LINE MI CRITICAL FLOW 0 0 0 0 0 y . M2 FLASHING 1 5 1 1 1 y MS CCFL: AIRIPW 0 0 0 0 0 / TMPNASE FLOWOUT M4 GEYSERING 0 0 0 0 0 y M5 STRATIFICATION . 0 0 0 0 0 y M7 SYSTEM INERTIA 6 5 5 5 1 y E N REGION. CONTAINMENT y N1 BLOWOOWN FLOW 0 0 0 0 0 y N2 MINING 0 0 0 0 0 y N3 NON CONOENSIBLE o 0 0 0 0 y
- INTERACTION M
Y O i
;; Table 1.5 3b TRACG Qualification (Section 2) (Continued) m 8 03 DWil: P1fl1(Ehenomena identification and flanhing Table )
TH A N S I E N T S GOVERNING PHENOMENA TRANSIENTS [ FOCUS. DELTA (CPR)/tCPR ; MAXIMUM PRESSURE WATER LEVEL, ENERGY DEPOSITION IN FUEL] PHENOMENA
- R A N K I N G (0.NOT APPLICABLE 1.2,3. LOW,4.5.6. MEDIUM,7.s.9.HtGH)
PRESSURt2ATION DEPRESSURIZATION FLOW / STABILITY COLD WATER ROD WITHDR AWAU TRACO MODEL TRANS!ENT TYPE TRANStENTS TRANStENTS TRANSIENTS TRANSIENTS DROP TRANSIENTS N4 GDCS FLOW 0 0 0 0 0 V H5 ISOLATION CONDENSER 0 0 0 0 0 V P REGION: MOTOR GENERATOR SET V Pt MG SET RESPONSE 4 4 8 4 1 / R REGION. FEEDWATER SYSTEM V 3 3 4 5 l / 4 Rt PUMP SEtZURE 3 4 5 1 / G; R2 HEATER TRANSIENTS 3 g, RJ PIPING 2 2 2 2 1 V E @ 5 i it 2 a if ir E E. E d f , [t a. 1 e E a
' E O O O
_,...-..__~..._.~.__..mm.._.-- -
--...-~._.._.m..._.....~.-.__o__~~~.--__mm.-....__.__..~.._._
[% % f i i { {. Table 1.5-3c TRACG Oualification (Section 3) g I MATRIX OF HIGM.Y RANKED PHENOMENA VS. SEPARATE EFFECTS TESTS QUAUFICATION DATA SASE la SWR-fta Phanamana idamIEmmann and Rashing Taistan
#. MENOMENA
( 4 FACILITY -+ l i E CISE W Ceuin- M C8HT TNTF PSTF Edmed FRIGG FRIGG CSHT Man 4- N E8WN ATLAS ATLAS SPERT f At-ni
- tensen CCR.
X X Nat C. StahEgy red. hen X Benehnsi - VK60 DP CPR-h F S A2 X.
=
8' A3 X ! Y M X ; as x x x x X x x x x
.g-f As x !
a Att l ] st x X x 33 X l a M X X
~
w N y 96 * . b , SS 87 ; 38 ste CIAX CISX X C1CX X CRAX X X X X X X X X X X X X X X CSX X X X ~X X- X X CSAM X i i CSCX C4 X - C5 Y C4
-3 i
t , - , ,- , -, -- --~n,----,,---n.~,,-r- .ns- -e ,s--,- e---w-,,--,se ,,mvm--n,- .m v. - v- v ,,,,,.w,,,--.,,a~.~,---,,.--,,..,,,,--,u -,n, -n-an n,-n-,.v,,, ,. ,,.-,.,.>,e-n-v
;; Table 1.5-3c TRACG Qualification (Section 3) (Continued) m ti MATRIX OF HIGM.Y RANKED PHENOMEMA VS. SEPARATE EFFECTS TESTS OUALIFICATION DATA BASE W
BWR - PIRT (Phannmana idensecaban adn Hardag TaMan PHENOMENA 1 FACILITY -. CISE Rouhe- cme- FNGG CSHT INTF PSTF Edwans FRIGO FRIGO CSHT Man 4- Weserv E9WRI ATLAS ATLAS SPERT nt
- tensen CCFL Net C. Stabety red. kort Bartalmei VK50 DP CPR C7 X CS CSX C10 X X X X X X X X C11 C12 X X C13 X X C14 X C15 X X g
X X ui g C1s E. C17 X X O M g C1s .*
$ C19 X k C20 X f C21 X
( C4 X X X f C2s
- r 3 C26 g D1 X !
f
~
I
- e. k S De (
[ os El
~~ T g
!4 f E O X j X X k R E3 4 (E4) X O O - O
ug g$ _ D mC gig - 4 fa g4R i O T . R E
. P S
S AR LP TC A S AP RD A t R0 E W5 K X S BV E - A B A i m
) T fnm
_ d A oa sa X D e at u N O WsB a n I T ._ i t A - n C I 4 vn r e ho X X o FI L M C ( A i Ua . ) QM T 3 SaT Hd. n T g Sn Se Cr i o Eu Td s t y c Sa Ge e TR Gt Cn I b S ( Ed Fa Rat FS Fn n Eo i o Ea Ts l GC .. Oi i f i t a c A Rm Ae Pd Ei r GI t Ra FN de . l a S. a n e u Sa Vy n w X X X n d Q An Na E G EhP F ._ Ml C OT T S X X X X A NR ERP P R H T P- F c DR T EW N _ 3- KB T N . 5 A . R 1 T . Y HR e t i SC l b G CC _ a I H _ T F O G G ._ X I . I R R F T _ A M esen s
. l w n Cet
+ -
- a Y A*
T oln t I R L C A AF E N i S* I i C M O __O N Ei H P 6 E s E - 1 F 2 F S F 4 F
)
t O ( 2 a 8 0 M " 8 7 G 2 H 1 1 2 8 S I X 1 L 1 L t M M 2
- 6 N m ghg$g2gQ tE&ggY5$i2b t .b8
r Table 1.5-3c TRACG Qualification (Section 3) (Continued) m E D MATRIX OF HIGHLY RANKED PHENOMENA VS SEPARATE EFFECTS TESTS OUALIFICATION DATA BASE BWR - PIRT IPhanomana idenslAcalman adn Ranimg Tabla) ' PHENOMENA 1 FACILITY a CISE Rode- cms- FRIGG CSHT THTF PSTF Edwant FRIGG FBK3G CSHT Maru Wilsonf EBWR/ ATLAS ATLAS SPERT rd ' tensen CCR Net C. Stabaly red. len Bartoernal VK50 DP CPR NT Nio Analyucald%
Ihese tests were used for the qualification of previous versions of'IRACG, the qualification was not repeated with TRACG02
!s 3
?
!a >
k 2 a e 8-
?T E l s 5 e t l k t?
? 5 i kg h h
- 8. =
! E
- a O O O
.- . . . . . . . . . .- - . . - . . - - - . . . . . - - -_ - . . - . . - . . . - - . - - ... _ . . - = . .. . . . . .
,m fm f'
(
'v) \
x
=
Table 1.5-3c TRACG Qualification (Section 3) (Continued) (4 3 # MATRIX OF HIGHLY RANKED PHENOMENA VS. COMPONENT PERFORMANCE OUALIFICATION DATA BASE
= SWR = ERTRhenomena Iderdecakm and Ranlung Table)
C
& PHENOMINA g i FACILITY --+ ;* 1st scale Jet Fidi Scale huofhase Two-Stage Three-Stege Hortrortal UP Segment Leekage
( Pump Jet Pump Jet Pump Separator Separator Sprey Test
- Test
- Fkm **
GIRAFFE SSTF UP SSTF LP
- SSTF PC *
# A1 X
s A2 X It A3 Eir X
& A4 X
h A5 X O.
@ A6 X
[l A10 B1 X X d M X m 2 : R M X M X X s M X X X X M X X X X 97 X 98 X B10 CIAX C1BX cicx C2AX C2BX C3CX m C5 1
- X X X s !i
r Table 1.5-3c TRACG Qualification (Section 3) (Continued) m h
- MATRIX OF HIGHLY RANKED PHENOMENA VS. COMPONENT PERFORMANCE OUALIFICATION DATA BASE BWR - PIRTRhenomena kinneW6 canon and flankba_Intdal PHENOMENA 1 FACluTY -e 1Al scale Jet Fui Scale h Phase Two-Stage Tivee-Stage Hortrontal UP Segment Leakage GIRAFFE SSTF SSTF SSTF Pump Jet Pump JulPump Separator Separator Sprey Test
- Test
- Fkm " UP LP* PC*
u X X X C7 X X C8 X X X CSX C10 C11 X X X X C12 c13 C14 oi C15 b E cis [
! Ci7 5 g C18 ) Cte } Cao C2, 3
S Q4 X X X e to
- b. k r Ca a-E l ] ot x x x l y. m x x x-l t -
M f, g X X X g l . os % l & a-3 11 X to S $
; !i O _
O - O
%h5 ?h mW P5k e.h 4 F*
T r mU SC SP X X F* T X SP L X X S F TP X X X SU S E F F A E R I S G A B A T A e D g
)
d N O k a "m e I T ak eF u A L i n C I t F n I L A dr o U )e o C Q l m* gt ( E ba ee s X X X
)
3 C T ST N g P n A M i n U i o R k n a t O R
- c F lat e R d n t ne s
S ( E P n a oT ry X t T o ora r n l N la Hp i o E r N e S t . a O n c P l a e i f M k gr o O a at i l n t Sr a a C. e ep- a X x . u S m _ V o ee . Q A n e hs T r N h G E P ( C M T e A O R gar N I aa R E P 1 s x T H - 4ra - X - P R op e c D E W T wS 3- B K 5 N A 1 R e sp e Y_ an - l H hu x . b G PP - a I et se T H F TJ O X I R ep T l A M cam u x SP dtt e FJ
+
t e Y Jsp T U lam I cu x X AC oP NA 4 EF
)) M 1
r(_ O N E & H 2 3 E S E 4 E 7 E 1 F R 3 F 4 F n G = 3 4 5 G G G 7 2 G H 1 u o X 1 L X 2 L 1 L 1 M 2 N P ( 1 ps. g T yf, {5 ,S lbk- P
l g Table 1.5-3c TRACG Qualification (Section 3) (Continued) m M Cll MATRIX OF HIGH Y RANKED PHENOMENA VS. COMPONENT PERFORMANCE QUALIFICATION DATA BASE BWR - PART iPhenomena h-W and Ranlune Table) PHENOMENA 1 FACluTY -+ 14 ocale Jet Ftd Soeie hyoPhase Two-Stage Ttwee4tage Hortrontal UP Segment Leekage GIRAFFE SSTF SSTF SSTF Pump Jet Pump Jet Pump Separator Separator Spray Test
- 1est
- Fkm " UP LP* PC
- N3 N4 N5 X N6 N7 N10
- These tests were used for the qualification of previous versions of TRACG, the qualification was not repeated with 1RACG02 j ** Best estimate 6ts to leakage Gow tests consistent with GE design methods inplemented into 'IRACG 3
.T P 3
E k
?
l B-i P
~s $
!g. s e i it a- ! R
- a O O O
. . . _ . -. . . _ - . . . ~ . .. . . .-
i (3 f~) i
\j i
(,/ ) iV /
=
3 Table 1.5-3c TRACG Qualification (Section 3) (Continued) m 08 MATRIX OF HIGHLY RANKED PHENOMENA VS. SYSTEM EFFECTS OVALIFICATION DATA BASE s BWB _ElBI (PheaQmtDa idenarcahon ADI1flankloglalde) tr
& PHENOMINA 4 1 FACLITY -.
b TLTA TLTA TLTA TLTA FIST F1ST FIST FIST GIST GIST GIST GIST GIST SSTF FIX FIX GIRAFFE ARTTOMI PSTF ( BO LB LBNE LEPP LB S8 LPCl* TT* BDB SB S8 LL GB NB LB
- SB* contairt
{ A1 X X X X X X X X X X X X X X X
$ A2 X X X X X X X X X X X X X X X b M X X X X X X X r
F A4 X X X X X X k M X X X X X X X X X X X X X X X E A6 X X X 3 X X f A10 X X X X X l B1 X X X X X X X X X X X X X X X m X X X X X X X X X g g B3 X X X X X X X X X X X X X X X U fh M X X X X X X B5 X X X X X B6 X X X X X X X X X X X X X X X 87 X X X X X X X X X X X X X X B8 X X X X X BIO X CIAX C1BX CICX D CM X X X X X X X X X X X X X X X MX X X X X X X X X X X X X X X X C-C3CX C4 X X X X X X X X X X X X X X X X g. C5 X X X X X X X X e 4
r Table 1.5-3c TRACG Qualification (Section 3) (Continued) m b W MATRIX OF HIGFLY RANKED PHENOMENA VS. SYSTEM EFFECTS QUALIFICATION DATA BASE BWR - PtRL{Ehenomena ideotGcallon and Ranidng Tablel PHENOMENA 1 FACILITY -e TLTA TLTA TLTA TLTA PIST FIST FIST FIST GIST GIST GIST GIST GIST SSTF FIX FIX G!HAFFE ARITOMt PSTF BO LB LENE LEPP LB SB IPCl* TT* BDB SB SB LL GB NB LB
- SB* contain.
C6 X X X X X X X C7 X X X X X X X X C6 X C8X X C10 X X X X X X X X X X X X X X X C11 X X X X X X X X X X X X X X X X C12 X X X X X X X X X X X ' C13 X X X X X X X X X X X g 3 C14 X X X X X X g 3 C15 X X X X X X X X C C16 X X X X f b C17 X X g C18 C19 X X X h f m Q1 X X X X X X X X X X X X X X h S. C24 X X X X X X X X X X X X 4 X X Q5 X X X X X X X X X X X X X X Y C26 X X X X X X X X X X X X X X D1 X X X X X X X X X X X X X X X
- g. D2 X X X X X X X
- X X D3 X X X X X D4 X X X X X X X X M X X X X X .
Ei X X X X X X X
- s O O O
.__ . . . . _ . _ _ . _ . . . _ . . _ . . _ _ . . . . . _ . ~ . _ . _ . _ . _ . . . . . _ . . . . _ , _ _ _ . _ . . _ . _ . _ . . . _ - . _ . - _ _ - _ . . . ..
(m v J
= Table 1.5-3c TRACG Oualification (Section 3) (Continued) 3 to S. 11 0 l MATRIX OF HIGHLY RANKED PHENOMENA VS. SYSTEM EFFECTS OUALIFICATION DATA BASE
% BWR=ERT_1 Phenomena idennhcanon and flanlung.Taide) i it j p PHENOMliNA 1 FACLIP/ -,
4 1LTA TLTA TLTA TLI FBT FIST FIST FIST GIST GIST GIST GIST GIST SSTF FlX FIX GIRAFFE ARITOMI PSTF BO LB LBNE LB SB LPCl* TT* BOB SB = LL GB NB LB* SB* mntain. o E2 X X X X X X X X X X X X X X X E E3 X X X X X k E5 X X X X X X X X X X X X X X h E6 X X X X X X X X X X X X X X 1 ET f F1 X X X X X X X X X X X X X l F2 X X X X X X h X X X l F3 X X X X X X u
- 8. F4 X s -
g (G1) X X X X X X C m G3 X X X X X X X X X X X X f h G4 X X X X X X G5 G7 X X X X X X X X H2 X X X X X X X X 11
= P
= .
5 L1X X L2X X L1 X X X X X X M1 X X X X X X M2 X X X X X X N ill'
, X X X X X X X
[
e Table 1.5-3c TRACG Qualification (Section 3) (Continued) fa b III MATRIX OF HIGit.Y RANKED PHENOMENA VS. SYSTEM EFFECTS QUALIFICATION DATA BASE BWR - PIRT (Phenomena idenWicaton and.Rankinn Table) PHENOMENA 1 FACLifY -* TLTA TLTA TLTA TLTA FIST FIST FIST GIST GIST GIST GIST GIST SSTF FIX FIX GIRAFFE ARITOMI PSTF BO LB LBNE LSPP yLE SB LPCl* TT* BDB SB SB LL GB NB LB
- SB* contain.
N3 X X X X X X X N4 X X X X X NS X N6 X X X X X X X N7 X X X X X X N10
* 'These tests were used for the qualification of previous versions of1RACG, the qualification was not repeated with TRACG02 x =
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Yb t t a 3 IL g nS A o _ n U w m o Q n t r e i A a V t c T R A d e D n a diiv X X X x X X X X X X X S T oS l ( N n o HM n A dn L T o P o P _p i i t a c S df V ds A i B 2 f N a M X X X X X X X X X - i l E na K T" a M m KT .
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T R . A ( e I P t y dm X X X X X X X X R p -- ni eh s T A R R bat ieS c Y L B W L ._ 3- H 5 G I H 1 F ey X X X x X X X X X X X l e O nl eo shn b X I ai _ a R ts T T A -i M - _ 4' se 'p X X X x X X X X X X X eu m er wt _ ot a a cs Y _ T I
. L I
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. A F amT
. im sn o A N p e a E _ M O _. r N E 1 ) X X H P 3 E 5 E 6 E 1 F 2 F 3 F I G ( m e 3 4 S 7 2 G G G H 1 1 E D 1 L 2 L 1 L 1 M - 2 N 3 N pc g &5eai,I, . 3$p& .3lkk!: g
.-- Table 1.5-3c TRACG Qualification (Section 3) (Continued) m k
- MAIHix OF HIGHLY RANKED PHENOMENA VS. PLANT DATA QUALIFICATION DATA BASE BWR - PlHT iPhenomena jdenaRcahon and Bankng lgdt)
PHENOMENA 1 FACILITY -+ Peach Bot- Coerentes 1.mSeBe Leibetadt KKM Hatch Vermont Yankee Cofrontes Oscarshamn Forsmark tom TT Start Up
- 8es6Elly Stabety TT** 2PT MSIV Stability " Stabihty PT" Stability N4 N5 N6 N7 N10
- These tests were used for the qualification of previous versions of TRACG, the qualification was not repeated with TRACG02 4
** 'these tests were not used for the qualification of'IRACG. Significant phenomena are covered by other tests.
k p :: e g- iP = s E b if r a e I e a 1 g- = I I E o - 4 O - O _ _ O
(m (3 g Table 1.5-3c TRACG Q* ali:' cation (Section 3) (Continued) g h MATRIX OF HKMt.Y--RANKED P1.1,;dENA VS. QUALIFICATION DATA BASE g BWR - PIRT (PHENOMENA IDENTIFICATION and RANKING TABLE)
$ PHENOMENA g & DATATYPE -+
@ SEPARATE EFFECTS COMPONENT SYSTEM. EFFECTS PLANT DATA
;* TEST 8 PERFORMANCE DATA TEST 8 E
' At X X X e
& M X X X fF M A4 X
X X X X X o g A5 X X X
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l M X X X $ E B4 B5 X X X X f s M X X X 87 X X Be X B10 X C1AX X C1BX X X CICX X X C2AX X X X g C2BX X X X CSAX X X CSCX X GESTR DATA SASE k u X X fE C6 X X { m X X
" X C7 X X
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- m -_ h --_-- - - - - - - - _ - _ - _ - _ _ _ _ _ . -
r Table 1.5-3C TRACG Qualification (Section 3) (Continued! m
$ ID MATRIX OF HIGHLY-RANKED PHENOMENA VS. QUALIFICATION DATA BASE BWR - PIRT (PHENOMENA IDENTIFICATION and RANKING TABLE)
SEPARATE EFFECTS COMPONENT SYSTEM. EFFECTS PLANT DATA TESTS PERFORMANCE DATA TESTS Ce X X X C8X X X X ClO X X X C11 X X X C12 X X X C1s x x c14 X X C15 X X C16 X X C17 X X C1s DATA BASE FOR GE 3 PERFORATION MODEL g j- Cte x x y lE m X X { C21 X X C24 X X X X P X X ANS STANDARD C25
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im, . . $ Dry Mark i Mark 11 Mark 111 ABWR SBWR Pressure Suppression No Yes Yes Yes Yes Yes , s { Number of Barriers l Containment 1 2 2 3 2 2 l; i Fission 2 4 4 4 4 4 [ Volume, million ft3 2.5 0.4 0.5 1.6 0.5 0.3 s' p) Heat Capacity, BTU 109 0.3 1.7 1.3 1.3 1.3 1.3 E 5 l Design Pressure, psig 50 62 45 l. i 15 45 55 {t i LOCA Pressure, psig 50 44 42 9 39 42 I a 5 Figure 1.5-3 Comparison of BWR Containments g 9 O O
25A5113Rev. A SBWR senatardsafetyAutrsissepar 1.6 Material Incorporated by Reference l Table 1.6-1 is a list of all GE topical reports which are incorporated in whole or in part l I by reference in the SBWR SSAR. i 1.6.1 COL License Information None. 1.6.2 References None. , l l : D f n 1 l l l l Materialincorporated by Reference - Amendment 1 131 (
2SA5113R2v. A SBWR standardsarety Analysis neport l O Table 1.6-1 Referenced Reports SBWR SSAR Report No. Title Section No. 22A7007 GESSAR 11,238 Nuclear Island, BWR/6 Std Plant, General 3.7,19.2,19.3, Electric Company,3/82, & Amendments 1-21. 19D.3,19 D.7, 19E.2,19E.3 APED-5750 Design and Performance of GE-BWR Main Steam Line 5.4 isolation Valves, General Electric Company, Atomic Power Dept., 3/69 APED-5756 N. R. Horton, W. A. Williams and K. W. Holtzclaw, .115,15.7 Analytical Methods for Evaluating the Radiological Aspects of General Electric BWRs,346 XSH NEDO-10029 An Analytic Study on Brittle Fracture of GE-BWR Vessel 5.3 Subject to the Design Basis Accident NEDO-10871 J.M.Skarpelos and R. S. Gilbert, Technical Derivaticn of 12.2. BWR 1971 Design Basis Radioactive Materials Source Terms,3/85 NEDE-10958-PA, GE BWR Thermal Analysis Basis (GETAB): Data 48.16 NEDO-10958-PA Correlation and Design Application,1/17 NEDO-11209-04A Nuclear Energy Business Operations Quality Assurance 17.1 Program Description, Rev. 8,3/89 NEDO-20159 BWR Radioactive Waste Treatment System,806 15.7 NEDO-20206 D. R. Rogers, BWR Turbine Equipment N-16 Radiation 12.2. Shielding Studies,1203 NEDO-20533 W. J. Bilanin, The GE Mark lli Pressure Suppression 6.2.7 Containment Analytical Model,6/74 NEDO-20533-1 W. J. Bilanin, The GE Mark Ill Pressure Suppression 6.2.7 Containment Analytical Model, Supplement 1,995 NEDE-20566 Analytical Model of Loss of Coolant Accident in 3.9.7 Accordance with 10CFR50, Appendix K, Proprietary Document,1195 NEDE-20566-P-A Analytical Model of Loss of Coolant Accident in 6.3.7 Accordance with 10CFR50, Appendix K,9/86 NEDM-20609-01 R R Stancavage and D. G. Abbott, Liquid Discharge Doses 12.2.
- LIDSR Code,8R6 NEDO-21052 F. J. Moody, Maximum Discharge Rate of Liquid Vapor 6.2.7 Mixtures from Vessels, General Electric Company,9#5 NEDO-21143-1 H. Careway, V. Nguyen, and P. Stancavage, Radiological 15.7 Accident - The CONACO3 Code,12/81 NEDO-21159, Airborne Releases from BWRs for Environmental impact 12.2.3 NEDO 21159-2 Evaluations,12/84 NEDE-21175-P BWR/6 Fuel Assembly Evaluation of Combined SSE and 3.9.7 Loss of Coolant Accident Loadings,11/16 NEDE-21354-P BWR Fuel Channel Design and Deflection,996 3.9.7 1.6-2 Materialincorporated by Reference - Amendment 1
2SA5113R:v. A SBWR standardsarery Analysis neport / s Table 1.6-1 Referenced Reports (Continued) SBWR SSAR Report No. Title Section No. NEDE-21514 BWR Scram System Reliability Analysis,12/81, General 19D.6.6 ! Electric Cr'mpany NEDO-21778-A Transient Pressure Rises Affecting Fracture Toughness 5.3.4 Requirements for BWRs,1/79 NEDO-21985 Functional Capability Criteria for Essential Mark 11 3.9.7 Piping,9/78, Prepared by Battelle Columbus Labs for General Electric Company NEDE-22056 Failure Rate Data Manual for General Electric BWR 19.3.5,19D.3.6 Cornponents, Rev. 2,1/86, Class ill, General Electric Company NEDE-22056 Reliability Analysis Data Manual, General Electric 19E.2.5 Company,1/86 NEDO-22155 GE Report, Generation and Mitigation of Combustible Gas 6.2.7 Mixtures in inerted BWR Mark l Containments,6/82 NEDO-23909A H. A. Careway, V. Nguyen, and D. G. Weiss, Control Room 15.6 Accident Exposure Evaluation - CRDOS Program 2/81 s
\ NEDE-24011-P-A-US General Electric Standard Application for Reactor Fuel- 15.4 (y ,/ United States Supplement (Latest Applicable Revision)
NEDO-24057 Assessment of Reactor Internals Vibration in BWR/4 and 3.9.7 BWR/5 Plants,11/77 NEDO-24057-P Assessment of Reactor Internals Vibration in BWR/4 and 3.9.7 BWR/5 Plants, Amendment 1,12/78, Also NEDE-2-P24057, Amendment 2,6/79 NEDE-24222 Assessment of BWR Mitigation of ATWS,9/19 15.8 NEDE-24326-1-P GE Environmental Qualification Program, Proprietary 3.9.7,3.11.7 l Document,1/83 NEDO-25132A E. W. Bradley, Gamma and Beta Dose to Man from Noble 12.2. Gas Release to the Atmosphere - GEMAN Code,4/B0 NEDO-25250 Generic Criteria for High Frequency Cutoff of BWR 3.9.7 Equipment, Proprietary Document,1/80 , NEDO-25257 E. W. Bradley and V. D. Nguyen, Radiation Exposure from 12.2. f Airborne Effluents - the REFAE Code,7/80 NEDE-30130A J.A. Woolley, Three-Dimensional BWR Core Simulator, 15.1 1/77 NEDE-30637 B. M. Gordon, Corrosion and Corrosion Control in BWRs 5.2 12/84 NEDE-31152-P GE Bundle Designs,12/88 4.2.3 NEDC-31858-P BWROG for increasing Main Steam isolation Valve 15.8 Leakage Rate Limits and Elimination of Leakage Control (/ ,) Systems,2/91 MaterialIncorporated by Reference- Amendment 1 1.64'.1
l l 25AS113Rsv. A SBWR sisuutardsateerAnsirsisnepar f~ \ t 1.7 Engineering Drawings Engineering drawings for the SBWR are provided in Chapter 21. The drawings are , numbered with a prefix of 21.X.Y-Z where: l X is the chapter number that refers to the drawing; Yis the section number that refers to the drawing; and Z is the sequence number within the section. , The chapter volume is Gsize (17 x 22 inches) and includes the following: l m Piping and Instrumentation Diagrams (P& ids) and Process Flow Diagrams (PFDs) listed in Table 1.7-1. m Instrument Electrical Diagrams (IEDs) and Logic Diagrams (LDs) listed in Table 1.7-2. , a Building Arrangement Drawings listed in Table 1.7-3. m Radiation Zone Drawings listed in Table 1.7-4. m Miscellaneous drawings listed in Table 1.7-5. m Structural Desivn Drawings listed in Table 1.74 m Fire Protection Drawings listed in Table 1.7-7. , 1.7.1 COL License information None. ., 1.7.2 References l None. i
/O, U
1.7-1 Engineering Drawings - Amendment 1
$ Table 1.7-1 Piping and Instrumentation, Air Flow, and Process Flow Diagrams g l Rev. GE Drawing Rev. %
SSAR Figure No. Title Source Drawing No. No. No. No. MPL No. 23 Fig. 21.4.6-1 sh 1-3 Control Rod Drive System P&lD NA 107E5013 A C12-1010 l ! Fig. 21.4.6-2 sh 1-2 Control Rod Drive System PFD NA 107E6050 A C12-1020 Fig. 21.5.1-1 sh 1-Z 6 Nuclear Baller System P&lD SBW5100DNJXN001 00 107E6291 BA B21-1010 Fig. 21.5.4-1 sh 1-2 Isolation Condenser System P&lD NA 107E5154 1B B32-1010 Fig. 21.5.4-2 sh 1-4 Reactor Water Cleanup / Shutdown Cooling P&lD SBW5230DNJXN001 2 107E6277 A G31-1010 Fig. 21.5.4-3 sh 1-3 Reactor Water Cleanup / Shutdown Cooling PFD SBW5230DNIXN003 1 107E6294 A G31-1020 Fig. 21.6.2-1 Passive Containment Cooling System P&lD NA '07E5160 A T15-1010 Fig. 21.6.3-1 sh 1-2 Gravity Driven Cooling System PFD SBW5240DNIXN003 4 112D6027 HA E50-1020 s E Fig. 21.6.3-2 Gravity Driven Cooling System P&lD SBW5240DNJXN001 0 107E6228 RA E50-1010 3 Fig. 21.6.4-1 Emergency Breathing Air P&lD SBW2729DNJXN001 00 107E6284 A U40-1010 p Fig. 21.9.1-1 sh 1-5 Fuel & Auxiliary Pools Cooling System P&lD 60828-02-023-1G2 A 107E6299 A G21-1010 Fig. 21.9.1-2 sh 1-3 Fuel & Auxiliary Pools Cooling System PFD 60828-02-023-1G1 A 107E6305 A G21-1020 a g g Fig. 21.9.2-1 Plant Service Water System P&lD LS-M6-001 C 107E6253 A P41-1010 $. D j Fig. 21.9.2-2 sh 1-2 Reactor Component Cooling Water System - LP-M6-001 C 107E6256 A P21-1010 Ei. g g Trains A & B P&lD g i Fig. 21.9.2-3 sh 1-2 Makeup Water System P&lD BG-MS-001 C 107E6257 A P10-1010 g l Fig. 21.9.2-4 Fig. 21.9.2-5 sh 1-2 Condensate Storage and Transfer System P&lD Reactor Building Chilled Water System P&lD AP-M6-001 VW-MS-001 C D 107E6252 107E6255 A A P30-1010 P25-1010 {
,i?
- a
~
9 9 9
p J v v T Table 1.7-1 Piping and instrumentation, Air Flow, and Process Flow Diagrams (Continued) m W i g Rev. GE Drawing Rev. g g SSAR Figure No. Title Source Draw!ng No. No. No. No. MPL No. g Turbine Component Cooling Water P&lD E-M-1000 A 107E6301 A P22-1010 @ Fig. 21.9.2-6 sh 1-4 3 Equipment & Floor Drainage System P&lD RW-M6-001 D 107E6261 A U50-1010 g Fig. 21.9.3-1 sh 1-5 A C41-1010 [ 3 Fig. 21.9.3-2 Standby Liquid Control System P&lD NA 103E1588 A C41-1020 8 Fig. 21.9.3-3 Standby Liquid Control System PFD NA 103E1771 9 Fig. 21.9.3-4 Instrument Air System P&lD PJ-M6-001 D 107E6264 A PS2-1010 g Fig. 21.9.3-5 Service Air System P&lD PJ-M6-002 D 107E6263 A P51-1010 Fig. 21.9.3-6 High Pressure Nitrogen Supply System P&lD PG-M6-001 E 107E6262 A P54-1010 Fig. 21.9.4-1 Reactor Building Ctrl Rm Envelope HVAC Air Flow VN-MS-002 E 107E6270 A U40-1020 Diagram g Fig. 21.9.4-2 Reactor Building Ctrl Rm Envelope HVAC P&lD VN-M6-002 D 107E6269 A U40-1010 $ C Fig. 21.9.4-3 Reactor Building Refueling & Pool Area Vent VN-MS-005 B 107E6270 A U40-1020 P System HVAC Air Flow Diagram ( Fig. 21.9.4-4 Reactor Building Refuel & Pool Area Vent System VN-M6-005 B 107E6269 A U40-1010 P&lD Fig. 21.9.4-5 Turbine Area Ventilation System P&lD E-M-1015 A 107E6316 A U39-1010 Fig. 21.9.4.6 Reactor Building Clean Area Ventilation System VN-MS-001 E 107E6270 A U40-1020 HVAC Air Flow Diagram g Fig. 21.9.4-7 Reactor Building Clean Area Ventilation System VN-M6-001 C 107EG269 A U40-1010 P&lD g Fig. 21.9.4-8 sh 1-2 Reactor Building Controlled Area HVAC Air Flow VN-MS-003 D 107E6270 A U40-1020 D Diagram k Fig. 21.9.4-9 sh 1-2 Reactor Building Controlled Area Vent System VN-M6-003 C 107E6269 A U40-1010 h P&lD Fig. 21.9.4-10 sh 1-2 Drywell Cooling System P&lD NA 103E1593 A T41-1010 {h C k 6
r Table 1.7-1 Piping and instrumentation, Air Flow, and Process Flow Diagrams (Continued) to 2 t0 Rev. GE Drawing Rev. SSAR Figure No. Title Source Drawing No. No. No. No. MPL No. Fig. 21.9.4-11 sh 1-2 Drywell Cooling System PFD NA 107E6244 A T41-1020 Fig. 21.9.4-12 sh 1-2 Containment Atmospheric Control System P&lD NA 103E1775 A T31-1010 Fig. 21.9.4-13 Containment Atmospheric Control System PFD NA 103E1776 A T31-1020 Fia. 21.9.4-14 sh 1-3 Radwaste Buildina HVAC P&lDs 10109-0001 A M U38-1010 Fig. 21.9.5-1 sh 1-2 Fire Protection Reactor Building P&lD PF-M6-M001 D 107EG258 A U43-1010 Fig. 21.10.2-1 Hydrogen & Carbon Dioxide Bulk Storage System E-M-1013 A 107E6296 A N45-1010 P&lD Fig. 21.10.3-1 Main Steam System P&lD E-M-1001 A 107E6303 A N11-1010 % 5 Fig. 21.10.4-1 Circulating Water System P&lD E-M-1007 A 107E6295 A N71-1010 Fig. 21.10.4-2 Condensate & Feedwater P&lD E-M-1004 A 107E6297 A N21-1010 g A N61-1010
- Fig. 21.10.4-3 Main Condenser Evacuation System P&lD E-M-1008 A 107E6300 Fig. 21.10.4-4 Condensate Purification P&lD E-M-1014 A 107E6302 A N25-1010 Fig. 21.10.4-5 Low Pressure Extraction System P&lD E-M-1002 A 107E6304 A N36-1010 Fig. 21.10.4-6 High Pressure Extraction System P&lD E-M-1003 A 107E6304 A N36-1010
$s-3 Fig. 21.10.4-7 Feedwater Heater Vent System P&lD E-M-1005 A 107E6298 A N22-1010 $
=
s-Fig. 21.10.4-8 Feedwater Heater Drain System P&lD E-M-1006 A 107E6298 A N22-1010
- 3. Fig. 21.11.2-1 Liquid Waste Management System BFD 02-DBM-0001-2 CG 107E6265 A K10-1020 @
$ .L M5 001 k k Fig. 21.11.2-2 sh 1-4 Equipment Drain PFD RL-MS-002 ED 107E6266 A K10-1020 k a 5 Fia. 21.11.2-3 sh 1-19 Liould Radwaste P&lDs 02-DFM-0001-2 H M K10-1010 x;;;- 3
=
8 Fig. 21.11.3-1 sh 1-2 Offgas System PFD NA 107E5031 A K30-1020 4 2 3 O O O
T Table 1.7-1 Piping and instrumentation, Air Flow and Process Flow Diagrams (Continued) e W ig Rev. GE Drawing Rev. g g SSAR Figure No. Title Source Drawing No. No. No. No. MPL No. g y Fig. 21.11.3-2 sh 1-2 Offgas System P&lD NA 10735030 A K30-1010 Solid Waste Management System BFD RS-MS-001 GF 107E6267 A K20-1020 {3 Fig. 21.11.4-1 A K20-1020 [ Fig. 21.11.4-2 sh 1-3 Solid Waste PFD RS-MS-002 C 107E6268 K20-1010 Fia. 21.11.4-3 sh 1-4 Solid Radwaste P&lDs RS-MS-001 C NA t 4 i:: e s b P 5 a 2 4 e a-
; R
j Table 1.7-2 Instrument Electrical Diagrams and Logic Diagrams g Source Rev. GE Drawing Rev. % Drawing No. No. No. No. MPL No. 23 SSAR Figure No. Title NA 137C9712 A C12-1030 Fia. 21.4.6-3 sh.1-15 Control Rod Drive System LD Reactor Protection System IED NA 107E6031 A C71-1010 Fig. 21.7.2-1 sh 1-11 NA 137C9725 A C71-1030 Fig. 21.7.2-2 sh.1-62 Reactor Protection System LD Neutron Monitoring System LED NA 107E5056 A C51-1010 Fig. 21.7.2-3 sh 1-4 NA 137C9579 A C51-1030 Fig. 21.7.2-4 sh.1.14 Neutron Monitoring System LD Suppression Pool Temperature Monitoring System IED NA 107E6195 A T53-1010 Fig. 21.7.2-5 sh 1-2 NA 137C9634 A B21-1030 Fig. 21.7.3-1 sh.1-56 Nuclear Boiler System LD DNEXN002000 137C9639 A E50-1030 Fig. 21.7.3-2 sh.1-18 Gravity Driven Cooling System LD Leak Detection & lsolation System IED NA 107E5045 A C21-1010 y Fig. 21.7.3-3 sh 1-9 Fig. 21.7.3-4 sh.1-61 Leak Detection & Isolation System LD NA 137C9663 A C21-1030 f Fig. 21.7.3-5 sh 1-5 Safety System Logic & Control System IED NA 107E6039 A C74-1010 $ Fig. 21.7.3-6 Essential Multiplexing System IED NA 107E6250 A C62-1010
"?;;. 21 '.3 ' r!crm:S!!:ty Certre! Sy:ter !ED NA 107ESOS-' A T491010
"?;;. 21 '.3 8 ch ' 4 c!cr ab?!:ty Certre! Sy :cm LD NA 137C9515 A "O1030 Standby Liquid Control System LD NA 137C9164 A C41-1030 Fig. 21.7.4-1 sh.1-8 m A C61-1010 Fig. 21.7.4-2 Remote Shutdown System IED NA 107E6254
{ k Fig. 21.7.4-3 sh.1-18 Remote Shutdown System LD NA 137C9721 A C61-1030 $ s- G31-1030 g Fig. 21.7.4-4 sh.1-11 Reactor Water Cleanup / Shutdown Cooling LD NA 137C9640 A Fig. 21.7.4-5 sh.1-13 Isolation Condenser System LD NA 137C9292 IB B32-1030 & 1 Fig. 21.7.5-1 sh 1-3 Containment Atmospheric Monitoring System IED NA 107E5017 A D23-1010 k I N Containment Atmospheric Monitoring System LD NA 137C9703 A D23-1030 e, g Fig. 21.7.5-2 sh.16 - s Fig. 21.7.5-3 sh 1-9 Process Radiation Monitoring System IED NA 107E5015 A D11-1010 5 E = 5 D11-1030 S 9 Fig. 21.7.5-4 sh.1-16 Process Radiation Monitoring System LD NA 137C9680 A E O O . _ _ _ - _ _ _ _ _ - _ - - O _ _ - _ _ _ _ _ _ _
D O O T Table 1.7 2 Instrument Electrical Diagrams and Logic Diagrams (Continued) (4
- ID
[1 SSAR Figure No. Title Source Drawing No. Rev. GE Drawing No. No. Rev. No. MPL No. = y Fig. 21.7.5-5 Area Radiation Monitoring System IED NA 107E5016 A D21-1010 Fig. 21.7.5-6 sh.1-2 Area Radiation Monitoring System LD NA 137C9702 A D21-1030 i Fig. 21.7.7-1 sh 1-6 Rod Control & Information System LED NA 107E5041 A C11-1010 j Fig. 21.7.7-2 sh.1-87 Rod Control & Information System LD NA 137C9720 A C11-1030 Fig. 21.7.7-3 sh 1-3 Feedwater Control System IED NA 107E5008 A C31-1010 .I E Fig. 21.7.7-4 sh.1-14 Feedwater Control System LD NA 137C9594 A C31-1030 Fig. 21.7.7-5 Automatic Power Regulator System IED NP-1005865 A 107E6245 A C82-1010 Fig. 21.7.7-6 sh.1-4 Steam Bvoass & Pressure Control .^ utemctic Pc=cr NA 137C9611 A C85-1030 "cgu!:ter System LD Fig. 21.7.7-7 sh 1-2 Steam Bypass & Pressure Control System IED NA 103E1817 A C85-1010 Fig. 21.7.7-8 sh.1-9 Containment Atmosoheric Steam Bype:: ".: Preocurc NA 137C9238 A T31-1030 {g Control System LD p Fig. 21.7.7-9 sh 1-2 Process Computer IED NA 105E1133 A C91-1010 ( Fig. 21.7.7-10 (Number not used! Prec00: Ccmputer LD Fig. 21.7.7-11 Non-Essential Multiplexing System IED NA 107E6251 A CS2-1010 rig. 2 ? ' ' 12 Centcirment ^tmcepheric Centrc! Sy tem LD Fio. 21.9.1-4 sh.1-40 Fuel & Auxiliarv Pools Coolino System LD 60828-02-023- C NA G21-1030 1G3 !?
=
Fio. 21.9.4-15 sh.1-3 Drvwell Coolino Svstem LD NA 137C6932 A T41-1030 h 188C8940 K30-1030 a Fio. 21.11.3-3 sh.1-3 Offons System LD NA A %. e 4-I'as
.- 3 % R
j Table 1.7-3 Building Arrangement Drawings g Source Rev. GE Drawing Rev. $ Drawing No. No. No. No. MPL No. 33 SSAR Figure No. Title NA 107E6278 A A12-1010 Fig. 21.1.2-1 Site Plan NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 1 Reactor Building, Section A-A 0/180 degrees NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 2 Reactor Building, Section B-B 270/90 degrees NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 3 Reactor Building, Floor Plan at El. 32100 NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 4 Reactor Building, Floor Plan at El. 25300 Reactor Building, Plan at El. 22500 NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 5 NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 6 Reactor Building, Plan at El.19800/20550 Reactor Building, Plan at El.18500 NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 7 Reactor Building, Floor Plan at El.17200 NA 107E6281 A A12-2010 y Fig. 21.1.2-2 sh 8 Reactor Building, Plan at El.16000 NA 107E6281 A A12-2010 h Fig. 21.1.2-2 sh 9 Fig. 21.1.2-2 sh 10 Reactor Building, Plan at El.15500 NA 107E6281 A A12-2010 $ Fig. 21.1.2-2 sh 11 Reactor Building, Floor Plan at El.13200 NA 107E6281 A A12-2010 ( NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 12 Reactor Building, Plan at El.12500 Reactor Building, Ground Grade Plan at El.10000 NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 13 Reactor Building, Plan at El. 9000 NA 107E6281 A A12-2010 Fig. 21.1.2-2 sh 14 Reactor Building, Floor Plan at El. 6600 NA 107E6281 A A12-2010 of Fig. 21.1.2-2 sh 15 Fig. 21.1.2-2 sh 16 Reactor Building, Floor Plan at El. 3050 NA 107E6281 A A12-2010 $ g, s-Reactor Building, Plant at El.1000 NA 107E6281 A A12-2010 g Fig. 21.1.2-2 sh 17 Reactor Building, Floor Plan at El. -1000 NA 107E6281 A A12-2010 &
- g. Fig. 21.1.2-2 sh 18 -
= A A12-2010 Fig. 21.1.2-2 sh 19 Reactor Building, Plan at El. -2000 NA 107E6281 f-Reactor Building, Floor Plan at El. -6400 NA 107E6281 A A12-2010 $
g Fig. 21.1.2-2 sh 20 Fig. 21.1.2-3 sh 1 Turbine Building, Base Slab Plan at El.10000 E-M-0001 0 107E6309 A A12-3010 hno k 9 Turbine Building, Intermediate Floor Plan at El.17320 E-M-0002 0 107E6309 A A12-3010 4 5 Fig. 21.1.2-3 sh 2 3
~
i 1 I O ____________ - __ __--___ _ O O _ _ _ _ _ -
O O O T Table 1.7-3 Building Arrangement Drawings (Continued) m g 9. l Source Drawing No. Rev. GE Drawing Rev. No. No. MPL No. g g SSAR Figure No. Tetle No. g ; y Fig. 21.1.2-3 sh 3 Turbine Building, Intermediate Floor Plan at El. 21580 & 24020 E-M-0003 0 107E6309 A A12-3010 Fig. 21.1.2-3 sh 4 Turbine Building, Operator Floor Plan at El. 31340 E-M-0004 0 107E6309 A A12-3010 l Fig. 21.1.2-3 sh 5 Turbine Building, Laydown Areas at El. 31340 E-M-0005 0 107E6309 A A12-3010
$ Fig. 21.1.2-3 sh 6 Turbine Building, Section A-A E-M-0006 0 107E6309 A A12-3010 h Fig. 21.1.2-3 sh 7 Turbine Building, Section B-B E-M-0007 0 107E6309 A A12-3010 5 ; Fig. 21.1.2-3 sh 8 Turbine Building, Section C-C E-M-0008 0 107E6309 A A12-3010 Fig. 21.1.2-3 sh 9 Turbine Building, Section D-D E-M-0009 0 107E6309 A A12-3v iu Fig. 21.1.2-3 sh 10 Turbine Building, Section E-E E-M-0010 0 107E6309 A A12-3010 Fig. 21.1.2-4 Radwaste Building General Arrangement 5000-G2-001 CB 107E6273 A A12-4010 Fia. 21.1.2-5 Electrical Buildina Gen'eral Arranaement DG-D-0001 R NA A12-2012 h
V 5 a. L* e f a i
.i
. _ _ _ . _____ . ._ -_.-__~- . _.~ _ -~. . . - _ _ _ _ _ _ _ _
Table 1.7-4 Radiation Zone Drawings Source Rev. GE Drawing Rev. E SSAR Figure No. Title Drawing No. No. No. No. MPL No. ll13 Fig. 21.12.3-1 sh 1 Reactor Building Radiation Zone Full Power / Shutdown, NA 107E6321 A A12-2040 Section A-A Fig. 21.12.3-1 sh 2 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 El. 32100 Fig. 21.12.3-1 sh 3 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 El. 25300 Fig. 21.12.3-1 sh 4 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 El. 22500 Fig. 21.12.3-1 sh 5 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 El.17200 Fig. 21.12.3-1 sh 6 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 4 El.16000 h Fig. 21.12.3-1 sh 7 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 $ El.13200 E Fig. 21.12.3-1 sh 8 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 El.10000 Fig. 21.12.3-1 sh 9 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 El. 6600 NA A A12-2040 k Fig. 21.12.3-1 sh 10 Reactor Building Radiation Zone Full Power / Shutdown at 107E6321 j El. 3050 g k Fig. 21.12.3-1 sh 11 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 $ y El. -1000 a. s5 Fig. 21.12.3-1 sh 12 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 "% % El. -2000 I I D g E Fig. 21.12.3-1 sh 13 Reactor Building Radiation Zone Full Power / Shutdown at NA 107E6321 A A12-2040 !m El. -6400 j. j Fig. 21.12.3-2 sh 1 Reactor Building Post LOCA Radiation Zone, Section A-A NA 107E6322 A A12-2040 g
- a O O - -- --
O
,\ ,
!v) (s.s) (vl
!? Table 1.7-4 Radiation Zone Drawings (Continued) to i CD g Source Rev. GE Drawing Rev. g g SSAR Figure No. Title Drawing No. No. No. No. MPL No. g y Fig. 21.12.3-2 sh 2 Reactor Building Post LOCA Radiation Zone at El. 32100 NA 107E6322 A A12-2040 fi Fig. 21.12.3-2 sh 3 Fig. 21.12.3-2 sh 4 Reactor Building Post LOCA Radiation Zone at El 25300 Reactor Building Post LOCA Radiation Zone at El. 22500 NA NA 107E6322 107E6322 A A A12-2040 A12-2040 Fig. 21.12.3-2 sh 5 Reactor Building Post LOCA Radiation Zone at El.17200 NA 107E6322 A A12-2040 I Fig. 21.12.3-2 sh 6 Reactor Building Post LOCA Radiation Zone at El.16000 NA 107E6322 A A12-2040 8
- Fig. 21.12.3-2 sh 7 Reactor Building Post LOCA Radiation Zone at El.13200 NA 107E6322 A A12-2040 Fig. 21.12.3-2 sh 8 Reactor Building Post LOCA Radiation Zone at El.10000 NA 107E6322 A A12-2040 Fig. 21.12.3-2 sh 9 Reactor Building Post LOCA Radiation Zone at El. 6600 NA 107E6322 A A12-2040 Fig. 21.12.3-2 sh 10 Reactor Building Post LOCA Radiation Zone at El. 3050 NA 107E6322 A A12-2040 Fig. 21.12.3-2 sh 11 Reactor Building Post LOCA Radiation Zone at El. -1000 NA 107E6322 A A12-2040 Fig. 21.12.3-2 sh 12 Reactor Building Post LOCA Radiation Zone at El. -2000 NA 107E6322 A A12-2040 3 NA A Fig. 21.12.3-2 sh 13 Reactor Building Post LOCA Radiation Zone at El. -6400 A
107E6322 A A12-2040 A12-3020 { Fig. 21.12.3-3 sh 1 Turbine Building Radiation Zone Full Power / Shutdown, E-M-0020 107E6325 Base slab Fig. 21.12.3-3 sh 2 Turbine Building Radiation Zone Full Power / Shutdown, E-M-0021 A 107E6325 A A12-3020 El 17920 Fig. 21.12.3-3 sh 3 Turbine Building Radiation Zon6 Full Power / Shutdown, E-M-0022 A 107E6325 A A12-3020 El 21580 g Fig. 21.12.3-3 sh 4 Turbine Building Radiation Zone Full Power / Shutdown, E-M-0023 A 107E6325 A A12-3020 $. El 31940 El. Fig. 21.12.3-3 sh 5 Turbine Building Radiation Zone Full Power / Shutdown, E-M-0024 A 107E6325 A A12-3020 Section B-B k D Fig. 21.12.3-4 Radwaste Building Radiation Zone NA 107E6320 A A12-4030 E' 1 Z - a
{ Table 1.7-5 Miscellaneous Drawings g Source Rev. GE Drawing Rev. No. MPL No. SSAR Figure No Title Drawing No. No. No. 33 Fig. 21.8.3-1 Electrical Power Distribution System Single Line Diagram E3-001 E 107E5846 A R10-1010 Fig. 21.8.3-2 sh 1-2 125V DC Distribution System (Class 1E) Single Line Diagram E3-002 C 107E6287 A R42-1010 Fig. 21.8.3 ~; 125V DC & 250V DC Distrib Systems (non-1E) Single Line E3-003 C 107E6286 A R42-1010 Diagram Fig. 21.8.3-4 sh 1-4 480 Volt Power Centers Single Line Diagram E3-004 B 107E6285 A R23-1010 Fig. 21.8.3-5 120V AC UPS (Class 1E) Single Line Diagram E3-005 B 107E6288 A R46-1010 Fig. 21.8.3-6 120V AC UPS (non-1E) Single Line Diagram E3-006 B 107E6289 A R46-1010 Fig. 21.8.3-7 Instrument & Control (non-1E) Power Supply Single Line E3-007 B 107E6283 A R47-1010 Diagram k Fig. 21.9.1-3 Plant Refueling & Servicing Sequence NA 107E6027 HA F15-2020 s Fig. 21.9.5-2 Standby DG Fuel Oil & Intake & Exhaust Systems NA 107E6246 A R40-Z001 Fig. 21.9.5-3 Standby DG Jacket Cooling Water System NA 107E6247 A R40-Z001 Fig. 21.9.5-4 Standby DG Starting Air System NA 107E6248 A R40-Z001 Fig. 21.9.5-5 Standby DG Lubricating Oil System NA 107E6249 A R40-Z001 7 Y a P d Fig. 21.1BD-1 Arrangement of Equipment on Main Control Console NA 107E6306 A H10-2020 $ D E. NA 107E6323 A H10-2030 g { Fig. 21.18D-2 sh 1-2 Fixed-Position Displays A A32-1040 k '$ Fig. 21.18E-1 sh 1-3 SBWR Man-Machine interface Systems Design & NA 107E6159 g [ Implementation Process a o s. a 2 i= 0 0 0
O O O Table 1.7-6 Structural Desian Drawinas g { Source Rev. GE Drawing Rev. a [ SSAR Figure No Title Drawing No. No. No. No. MPL No. W[ k Fig. 21.3.8-1 RB & RCCV Section 0* - 180* 1000-CC-001 0 NA- A40-1090 1 0 NA A40-1090 E Fig. 21.3.8-2 RB & RCCV Section 270* - 90* 1000-CC-002 [ Fig. 21.3.8-3 RB Fir. El. -6400 1010-CC-001 0 NA NA A40-1090 A40-1090 l Fig. 21.3.8-4 RB Partial Floor Plan El. -2800 1010-CC-002 0 Fig. 21.3.8-5 RB & RCCV Basemat Sections & Details 1010-CR-001 0 NA A40-1090 k it RB & RCCV Basemat Reinforcing Steel 1010-CR-002 0 NA A40-1090
- Fig. 21.3.8-6 Fig. 21.3.8-7 RB Partial Fir. Plan at El.17200 Col Lines C-l 1070-CR-001 0- NA A40-1090 t Fig. 21.3.8-8 RB Partial Fir. Plan at EL 25300 Col. Lines 3-10 1080-CR-001 0 NA A40-1090 Fig. 21.3.8-9 Fuel & IC Pools Liner Plate Location Plan at El. 25300 1080-ML-001 0 NA A40-1090 y Fig. 21.3.8-10 Spent Fuel Pool Liner Plate Plan at El. 25300 Sections & 1080-ML-002 0 NA A40-1090 k Details 6 Fig. 21.3.8-11 Spent Fuel Pool Liner Plate Sections & Details 1080-ML-003 0 NA A40-1090 f Fig. 21.3.8-12 RB Partial Fir. Plan at El. 32100 Col. Lines 3-10 1090 iCR-001 0 NA A40-1090 Fig. 21.3.8-13 RB Firs at El. 32100 & 25300 Col Lines 3-10 Sections 1090-CR-002 0 NA A40-1090 Fig. 21.3.8-14 RB IC Pool Girders, Sections & Details 1090-CR-003 0 NA A40-1090 Fig. 21.3.8-15 RB Partial Ftr. Plan at EL 32100 & 25300 Col. Lines 1-3 (Spent 1090-CR-004 0 NA A40-1090 Fuel)
Fig. 21.3.8-16 RB Fuel Pool & Steam Tunnel Area Col. Lines C-l Sections & 1090-CR-005 0 NA A40-1090 Details
- Fig. 21.3.8-17 RB Fuel Pool & Steam Tunnel Area Col. Lines 1-3 Sections & 1090-CR-006 0 NA A40-1090 ft Details {
{ Fig. 21.3.8-18 RB Cask Loading Area Col. Lines 1-3 Sections & Details 1090-CR-007 O NA A40-1090 Fig. 21.3.8-19 RPV Pedestal Reinforcing Steel Sections & Details ' 1100-CR-001 0 NA A40-1090 f
.; Fig. 21.3.8-20 RPV Pedestal Reinf. Steel Developed Elevation 1100-CR-002 0 NA A40-1090 g e a
.- - . _ . _ - - _ _ - - - _ _ _ _ _ - - _ _ _ _ _ . _ _ _ _ _ _ _ - . _ - _ _ _ - - ,. . ~ - _ , - , - ., . _ _ , , , - . _ . --- _ _ _ _ _ _ _ _
" Table 1.7-6 Structural Desian Drawinas (Continued) M Z m ' Source Rev. GE Drawing Rev. Drawing No. No. No. No. MPL No. SSAR Figure No Title RCCV Wall Reinforcing Steel Section & Details 1100-CR-003 0 NA A40-1090 Fig. 21.3.8-21 RCCV Wall Developed Elevation 1100-CR-004 0 NA A40-1090 Fig. 21.3.8-22 1100-ML-001 0 NA A40-1090 Fig. 21.3.8-23 RCCV Liner Plate Sections & Details 1100-ML-002 0 NA A40-1090 Fig. 21.3.8-24 RCCV Liner Plate Plans RCCV Liner Plate Developed Elevations 1100-ML-003 0 NA A40-1090 Fig. 21.3.8-25 Vent Wall & RPV Support Bracket Plan & Sections 1100-SS-001 0 NA A40-1090 Fig. 21.3.8-26 Reactor Shield Wall Developed Elevation 1100-SS-002 0 NA A40-1090 Fig. 21.3.8-27 RCCV S/P Bottom Slab at El. 4650 Reinforcing Steel 1130-CR-001 0 NA A40-1090 Fig. 21.3.8-28 Diaphragm Floor at El.17200 Plan & Sections 1170-SS-001 0 NA A40-1090 Fig. 21.3.8-29 Diaphragm Floor Anchor Details 1170-SS-002 0 NA A40-1090 Fig. 21.3.8-30 1170-SS-003 0 NA A40-1090 6 Fig. 21.3.8-31 GDCS Pools at El.17200 RCCV Top Stab at El. 25300 Reinforcing Steet 1180-CR-001 0 NA A40-1090 E Fig. 21.3.8-32 RB Fir. at El. 25300 Concrete Reinforcing Details 1180-CR-002 0 NA A40-1090 Fig. 21.3.8-33 RB Fir. at El. 25300 Concrete Reinforcing Details 1180-CR-003 0 NA A40-1090 Fig. 21.3.8-34 RB Inner Box (Safety Envelope) Walls Elevation & Sections 1200-CR-001 0 NA A40-1090 Fig. 21.3.8-35 RB Outer Box Wall Elevation & Sections 1200-CR-002 0 NA A40-1090 ,S' Fig. 21.3.8-36 1200-SS-001 0 NA A40-1090 !? ( Fig. 21.3.8-37 RB Intermediate Box Wall El. at Col. Row J E g-k Fig. 21.3.8-38 RB Intermediate Box Wall EI. at Col. Row 2 1200-SS-002 0 NA A40-1090 O ii. 1220-CC-001 0 NA A40-1090 g [ Fig. 21.3.8-39 Fig. 21.3.8-40 RB Fir. at El. -1000 RB Ftr. at El. 3050 1230-CC-001 0 NA A40-1090 k [ Fig. 21.3.8-41 RB Fir. at El.3050 Concrete Reinforcing Details 1230-CR-001 0 NA A40-1090 k
.:r 3 0 NA A40-1090 g.
Fig. 21.3.8-42 RB Fir. at El. 6600 1240-CC-001 s ct 1250-CC-001 0 NA A40-1090 l Fig. 21.3.8-43 RB Fir. at El.10000 a O O O
N T Table 1.7-6 Structural Desian Drawinas (Continued) m g
- j. Source Rev. GE Drawing Rev. g g
Title Drawing No. No. No. No. MPL No. g g SSAR Figure No 0 NA A40-1090 y Fig. 21.3.8-44 RB Fir. at El.13200 1260-CC-001 f Fig. 21.3.8-45 Fig. 21.3.8-46 RB Fir. at El.17200 RB Fir. at El.19800 1270-CC-001 1270-CC-002 0 0 NA NA A40-1090 A40-1090 Fig. 21.3.8-47 RB Fir. at El. 22000 1270-CC-003 0 NA A40-1090 RB Partial Fir. Plan at El.17200 Reinforcing Details 1270-CR-001 0 NA A40-1090 h Fig. 21.3.8-48 3 1270-CR-002 0 NA A40-1090
; Fig. 21.3.8-49 RB Fir. at El.17200 Concrete Reinforcing Details Fig. 21.3.8-50 RB Fir. at El. 25300 1280-CC-001 0 NA A40-1090 Fig. 21.3.8-51 RB Operating Fir. at El. 32100 1290-CC-001 0 NA A40-1090 i3 2
P 5 a
!?
e. a-5 r S ut R
Table 1.7-7 Fire Protection Drawinas g Source Rev. GE Drawing Rev. E SSAR Figure No Title Drawing No. No. No. No. MPL No. 2 PF-M1-001 C NA A12-2045 Fig. 21.9A-1 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-002 C NA A12-2045 Fig. 21.9A-2 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-003 B NA A12-2045 Fig. 21.9A-3 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-004 C NA A12-2045 Fig. 21.9A-4 Fire Prot. Sys. Reactor Building Barrier Drawing FF-M1-005 C NA A12-2045 Fig. 21.9A-5 Fire Prot. Sys. Reactor Building Barrier Drawing Fire Prot. Sys. Reactor Building Barrier Drawing PF-M 1-006 C NA A12-2045 Fig. 21.9A-6 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-007 C NA A12-2045 Fig. 21.9A-7 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-008 8 NA A12-2045 Fig. 21.9A-8 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-009 B NA A12-2045 ,, Fig. 21.9A-9 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-010 C NA A12-2045 h Fig. 21.9A-10 Fig. 21.9A-11 Fire Prot. Sys. Reactor Building Barrier Drawing FF-M1-011 C NA A12-2045 $ Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-012 C NA A12-2045 Fig. 21.9A-12 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-013 C NA A12-2045 Fig. 21.9A-13 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-014 B NA A12-2045 Fig. 21.9A-14 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-015 C NA A12-2045 Fig. 21.9A-15 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-016 B NA A12-2045 f Fig. 21.9A-16 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-017 B NA A12-2045 @ f Fig. 21.9A-17 Fig. 21.9A-1B Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-018 C NA A12-2045 h fA Fig. 21.9A-19 Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-019 C NA A12-2045 tl? 1 a T Fire Prot. Sys. Reactor Building Barrier Drawing PF-M1-020 C NA A12-2045 4-Fig. 21.9A-20 I D
=
Fire Prot. Sys. Electrical Building Barrier Drawing 47809-0011 A NA A12-6045 b Fig. 21.9A-21 T
! Fig. 21.9A-22 Fire Prot. Sys. Turbine Building Barrier Drawing 47809-0001 A NA A12-3045 W
& 'a 3
Fig. 21.9A-23 Fire Prot. Sys. Turbine Building Barrier Drawing 47809-0002 A NA A12-3045 S 9
' E 9 _
9 e -
O O O T Table 1.7-7 Fire Protection Drawinas (Continuedl M l-g ID Source Rev. GE Drawing Rev. Drawing No. No. No. MPL No. g g SSAR Figure No Title No. g R Fig. 21.9A-24 Fire Prot. Sys. Turbine Building Barrier Drawing 47809-0003 A NA A12-3045 h Fig. 21.9A-25 Fire Prot. Sys. Turbine Bui ding Barrier Drawing 47809-0004 A NA A12-3045 i Fig. 21.9A-26 Fire Prot. Sys. Turbine Building Barrier Drawing 47809-0005 A NA A12-3045 x l Fig. 21.9A-27 Fire Prot. Sys. Turbine Building Barrier Drawing 47809-0006 A NA A12-3045 A NA ( Fig. 21.9A-28 Fire Prot. Sys. Turbine Building Barrier Drawing 47809-0007 A12-3045 $ Fig. 21.9A-29 Fire Prot. Sys. Turbine Building Barrier Drawing 47809-0008 A NA A12-3045 Fig. 21.9A-30 Fire Prot. Sys. Turbine Building Barrier Drawing 47809-0009 A NA A12-3045 Fig. 21.9A-31 Fire Prot. Sys. Radweste Building Barrier Drawing 47809-0010 A NA A12-4045 N 5 C
'O P
5
$. )
t 1 r t a 3
i t ' 25A5113Rev. A SBWR studsntsaretyAntysis nepon (" t L 1.9 Conformance with Standard Review Plan and Applicability of Codes and Standards 1.9.1 Conformance with Standard Review Plan This subsection provides the information required by 10CFR50.34(g) showing conformance with the Standard Review Plan (SRP). A summary of exceptions from the SRP acceptance criteria is presented by SRP section below. See Subsection 1.9.4 for COL license information requirements. 1.9.1.1 SRP Section 2 SRP Acceptance Criteria SRP acceptance criteria for Sections 2.2.1-2.2.2,2.2.3,2.3.1,2.3.4,2.4.1,2.4.4,2.4.5, 2.4.6,2.4.8,2.4.11.6 and 2.4.12 are provided in Subsection 11 of each SRP section. These criteria impose limits on the site characteristics for the acceptability of a plant site, with respect to such factors as public exposure, severe weather, hydrology, heat sink dependability and groundwater. m Summary Description of Exception Since the SBWR is not designed for a specific site location, the limits imposed on these SRP acceptance criteria are defined by: a the envelope of the SBWR Standard Plant site parameters a assumptions used in the evaluations of site characteristics Subsections 2.1 through 2.5 of this SSAR provide the results of the site characteristics evaluations, and Section 2.6 provides the requirements for the determination of SBWR site acceptability from the standpoint of both design basis events and severe accidents. j 1 1.9.1.2 SRP Subsection 2.5.2.7 j SRP Acceptance Criteria l SRP Subsection 2.5.2.7 specifies that the minimum value of the acceleration level for l the Operating Basis Earthquake (OBE) used in analyses should equal one-half of the reference acceleration for seismic design corresponding to the Safe Shutdown Earthquake (SSE). Summary Description of Exception OBE is not an SBWR design requirement. In the design analyses for SBWR, the value of OBE acceleration is chosen to be equal to one-third SSE acceleration. Consistent with , p the NRC's draft Appendix S to 10CFR50, the design requirements associated with the Q OBE, when the level of the OBE ground motion is chosen to be one-third of the SSE ground motion, are satisfied without pedorming explicit response or design analyses.
]
Conformance with Standard Review Plan and Applicability of Codes and Standards - Arnendment 1 1.9-1
25A5113 R1v. A SBWR standardsarety Analysis neport O A discussion of the vibratory ground motion analysis can be found in Subsection 2.5.2 of this SSAR. 1.9.1.3 SRP Sections 3.6.1 and 3.6.2 SRP Acceptance Criteria SRP Sections 3.6.1 and 3.6.2 provide acceptance criteria for postulated pipe mpture and guidelines for the piping analysis to meet the requirements of General Design Criterion 4. Summary Description of Exception For the SBWR, piping analyses for high energy fluid systems utilize the leak before break l l (LBB) option as provided in General Design Criterion 4 (GDC 4), November 27,1987,
" Modification of General Design Criterion 4."The mechanistic LBB approach, justified by appropriate fracture mechanics techniques,is recognized as an acceptable )
procedure under certain conditions to exclude design against the dynamic effects from l l postulation of breaks in high energy piping. A discussion of the LBB evaluation procedures can be found in Subsection S.6.3 of this
)
SSAR. 1 1.9.1.4 SRP Section 3.7.3 SRP Acceptance Criteria Paragraph 2.b of SRP Section 3.7.3 provides acceptance criteria for the determination of the number of earthquake cycles in the seismic analysis of the plant design. According to paragraph 2.b, the number of earthquake cycles should be obtained from the synthetic time history used for the system analysis, or a minimum of 10 maximum stress cycles per earthquake may be assumed. Summary Description of Exception in the seismic fatigue analysis for SBWR, two low-level earthquake events (lesser magnitude than the SSE) are considered, each with 10 peak stress cycles for a total of 20 cycles. Further discussion of the determination of number of earthquake cycles can be found in Subsection 3.7.3.2 of this SSAR. l 1.9.1.5 SRP Section 3.8.2 l SRP Acceptance Criteria l SRP Section 3.8.2 provides acceptance criteria for the design of steel containments. I l O 1.9-2 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1
2SAS113 Rey, A SBWR standardsuretyanalysis neport Summary Description of Exception SRP Section 3.8.2 applies only to the drywell head. In the SBWR Standard Plant design there is no steel containment used. 1.9.1.6 SRP Section 5.2.3 SRP Acceptanca Criteria Criterion 3.b.(S) of Subsection 11 of SRP Section 5.2.3 for the control of ferritic steel welding is based upon conformance to Regulatory Guide 1.71, " Welding Qualification for Areas of Limited Accessibility." Summary Description of Exception The SBWR design meets the intent of Regulatory Guide 1.71 by utilizing an alternate approach, as described in detail in Subsection 5.2.3.4.2 of this SSAR. 1.9.1.7 SRP Section 5.2.4 SRP Acceptance Criteria SRP Section 5.2.4 provides acceptance criteria for inspection of Class 1 pressure-containing components. According to Criterion 2, the design and arrangement of f system components are acceptable if adequate clearance is provided. Dj Summary Description of Exception For the SBWR, all items within the Class 1 boundary are designed, to the extent practicable, to proside access for examination. Items such as nozzle-to-vessel welds, however, often have inherent access restrictions when vessel internals are installed. Therefore presenice examination shall be performed on these items prior to installation ofinternals which would interfere with examination. Further details on the accessibility for inspection of Class I components can be found ; in Subsection 5.2.4.2 of this SSAR. 1.9.1.8 SRP Section 6.5.1 j SRP Acceptance Criteria SRP Section 6.5.1 provides acceptance criteria for the design of engineered safety feature (ESF) atmosphere cleanup systems. Summary Description of Exception In the SBWR Standard Plant design, no safety-related filter systems are used following a ; design basis accident (DBA), as discied in Subsection 6.5.1 of this SSAR. The control ! room is provided with self contained bottled air to maintain a safe control room atmosphere following a DBA. A) (
%d Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 133 I
l
1 2SAS113 Riv. A SBWR standardsafety Analysis aeport O1 l 1.9.1.9 SRP Seetion 0 5.2 l SRP Acceptance Criteria SRP Section 6.5.2 provides acceptance criteria for the design of containment spray systems. Summary Description of Exception In the SBWR Standard Plant design, neither the suppression chamber or upper drywell containment sprays are safety-related, and no credit is taken for removal of fission products under DBA evaluations (as discussed in Subsection 6.5.2 of this SSAR). 1.9.1.10 SRP Section 7.1 SRP Acceptance Criteria l SRP Section 7.1 specifies criteria establishing the necessary design, fabrication, construction, testing, and performance requirements for structures, systems, and components important to safety. The following Regulatory Guides (RG) should be used to provide guidance to applicants: a RG 1.22 m RG 1.75 m RG l.ll8 Summary Description of Exception m In the SBWR Standard Plant design, some actuators and digital filters, because of their locations, cannot be fully tested during actual reactor operation. Provisions for meeting the requirements are discussed in the Safety Evaluation subsections within Sections 7.2 through 7.7 of this SSAR. m Some alternate approaches to the independence and separation criteria for redundant systems specified in RG 1.75 are implemented in the SBWR Standard Plant design, as discussed in Subsection 7.1.2.2 of this SSAR. m The instrumentation and control systems are consistent with RG 1.118 requirements, with cladfications of those requirements prosided in Subsection 7.1.2.2 of this SSAR. l 1.9.1.11 SRP Section 8.1 SRP Acceptance Criteria I SRP Section 8.1 specifies criteria and guidelines for electric power systems. Conformance to the following Regulatory Guides (RG) provides assurance that systems will perform their design safety functions when required: 1.9-4 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
25A5113Rev. A 1 SBWR standantsuretyAnsorsis neport b C l m RG 1.6 ' l s RG 1.9 m RG 1.108 Summary Description of Exception In the SBWR Standard Plant design, no safety-related standby ac power sources and no - safety-related diesel generators are used, as discussed in Subsection 8.1.5.2.3 of this SSAR. Therefore, these RGs are not applicable. 1.9.1.12 SRP Section 9.3.2 SRP Acceptance Criteria Criterion K.5 of Subsection II of SRP Section 9.3.2 specifies conformance to item II.B.3 in NUREG-0737,' Clarifications of TMI Action Plan Requirements." Summary Description of Exception l For the SBWR, the liquid sampling capability is 1 Ci/g, instead of 10 Ci/cm3. O V A discussion of this exception can be found in Subsections I A.2.7 and 9.3.2.2.2 of this SSAR. 1.9.1.13 SRP Section 9.4.5 l SRP Acceptance Criteria SRP Section 9.4.5 provides acceptance criteria for the design of an engineered safety feature (ESF) ventilation system. Summary Description of Exception i In the SBWR Standard Plant design, an ESF ventilation system is not necessary, as discussed in Subsection 9.4.5 of this SSAR. Therefore, this SRP section is not applicable to the SBWR design. 1.9.1.14 SRP Section 11.1 SRP Acceptance Criteria SRP Section 11.1, Criterion 9 specifies the use of an acceptable method for evaluating source terms, such as those programmed in the BWR GALE computer code. Summary Description of Exception For the SBWR source term evaluation, a method similar to the BWR GALE computer code is used. The methods for evaluating source terms are discussed in Subsection 12.2.1 of this SSAR. Conformance with Standard Review Plan and Apnlicability of Codes and Standards - Amendment 1 1.95 l
2545113R1v. A SBWR StandardSafetyAnalysis Report O 1.9.1.15 SRP Sections 15.1.1 - 15.1.4 SRP Acceptance Criteria SRP Sections 15.1.1 through 15.1.4 specify the use of an acceptable analytical model for ATWS analyses. Models not listed in these SRP sections must be evaluated by the NRC staff for acceptability. Summary Description of Exception The models used to analyze the core and system performance have been approved by the USNRC or meet the criteria in Appendix 4B of this SSAR. For the SBWR A'1MS analyses, the TRACG code is used, as discussed in Subsection 15.0.3.4 of this SSAR. Licensing Topical Reports for the TRACG code are provided in References 1.9-1,1.9-2 and 1.9-3. 1.9.1.16 SRP Section 15.4.2 \ SRP Acceptance Criteria SRP Section 15.4.2 provides acceptance criteria for the analysis of uncontrolled control rod withdrawal at power. Summary Description of Exception For the SBWR, no quantitative analysis is provided because prevention of this transient from occurring is incorporated in the design of the Automated Thermal Limit Monitor (ATLM) system. The features of the ATLM are discussed in Subsection 15.4.2.1 of this SSAR. 1.9.1.17 SRP Section 15.4.9 SRP Acceptance Criteria SRP Section 15.4.9 provides acceptance criteria for the analysis of rod drop accidents. Summary Description of Exception For the SBWR, no quantitative analysis is provided because prevention of this accident from occurring is incorporated in the design of the Fine Motion Control Rod Drive I (FMCRD) system. 1 l The features of the FMCRD are discussed in Subsection 15.4.9.1 of this SSAR. j l 1.9.1.18 SRP Section 15.6.5 SRP Acceptance Criteria SRP Section 15.6.5 specifies the use of an approved ECCS evaluation model. O 1.% Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
l 2SAS113Rev. A l SBWR samrentseierr An*Irsis neport O V ; Summary Description of Exception For the SBWR ECCS mathematical model, analytical methods and associated i assumptions used in evaluation of LOCA (resulting from a spectrum of pipe breals inside containment) are considered to proside conservative assessment of the expected consequences of this event. Further details are discussed in Subsection 15.6.5.3 of this SSAR. i
- 1.9.1.19 SRP Section 17.1 ,
SRP Acceptance Criteria e Criterion 1 of Subsection 11 of SRP Section 17.1 specifies applicant responsibility for i the overall quality assurance (QA) program. m Critesia 2,3,4,7,13,17 and 18 of Subsecdon 11 of SRP Section 17.1 specify the commitment to comply with quality related Regulatory Guides. Summary Description of Exception C a For the SBWR Project, GE Nuclear Energy (GE-NE) and its Team Members conduct ; lk and perform their QA proginms as discussed in Subsection 17.1.2 of this SSAR. m For the SBWR, NRC-accepted alternate positions are employed to meet I conformance with Regulatory Guides 1.37, Rev. O and 1.38 Rev. 2. Quality related Regulatory Guide conformance is summarized in Table 17.0-1 of this SSAR. 1.9.2 Applicability of Codes and Standards Standard Review Plans, Branch Technical Positions, Regulatory Guides and Industrial ' Codes and Standards which are applicable to the SBWR design are provided in Tables 1.9-1,1.9-2 and 1.9-3. Applicable revisions are also shown. See Subsection 1.9.4 for COL license information requirements. 1.9.3 Applicability of Experience information , Experience information is routinely made available and distributed to design personnel in the design process. Nuclear field experience is maintained in functional component and library files and it, v ' GE world-wide computer retrieval system. Generic Letters and IE Bulletins, Information Notices and Circulars covering the period from 1980 through the current issues (early 1992) were reviewed for i
.s applicability to the SBWR design. The review was enhanced by associating related l experiences and tracing referenced occurrences. This was accomplished starting with L the current issues of the Generic Letters and proceeding back to 1980. The Circulars, Bulletins and Notices were reviewed in that order. Interfacing experience was included Conformance with Standard Review Plan and Applicabihty of Codes and Standards - Amendment 1 1.9-7 j l
J l
25A5113Rev. A SBWR statwardsaferyAntysisneport O l in the review. The selection of SBWR information was based on the significance to future design and operation guidance. Included is a list of NUREGs related to the l closing of current safety issues. Experience that resulted in applicable rules, codes and standards was not repeated. Table 1.94 lists the experience information that has been included in the SBWR Standard Plant design or impacts the COL applicant. See Subsection 1.9.4. for COL license information requirements. 19 A COL License Information Any COL license information requirements are identified in Tables 1.9-1,1.9-2 and 1.9-4. These COL license information requirements can be categorized into one of the following two types: m COL Applicant Action item (COL) -These items must be fulfilled by the COL applicant. m Bounding Site Parameter (BSP) -These requirements must be met by the plant site location chosen by the COL applicant. 1.9.5 References 1.9-1 Licensing Topical Report, TRACG Model Description, NEDE-32176P, Class 111 (Proprietary), February 1993. 1.9-2 Licensing Topical Report, TRACG Qualification, NEDE-32177P, Class 111 (Proprietary), February 1993. 1.9-3 Licensing Topical Report, Application of TRACG Model to SBWR Licensing Safety Analysis, NEDE-32178P, Class 111 (Proprietary), February 1993. l { O 1.9-8 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amyndment 1
I \ 2545113 R1v. A SBWR standardseteryAnarr sisneport v l Table 1.9-1 NRC Standard Review Plans I and Branch Technical Positions Applicability to SBWR SBWR Appl. Issued Appli- ; SRP No. SRP Title or BTP Rev. Date cable? Comments Chapter 1 Introduction and General Description of Plant j 1.8 Interfaces for Standard Design 1 7/81 Yes l Chapter 2 Site Characteristics 2.1.1 Site Location and Description 2 7/81 - BSP j 2.1.2 Exclusion Area Authority and Control 2 7/81 - BSP 2.1.3 Population Distribution 2 7/81 - BSP 2.2.1- Identification of Potential Hazards in Site 2 7/81 - BSP 2.2.2 Vicinity 2.2.3 Evaluation of Potential Accidents 2 7/81 - BSP 2.3.1 Regional Climatology 2 7/81 - BSP [,N 2.3.2 Local Meteorology 2 7/81 - BSP 2.3.3 Onsite Meteorological Measurements 2 7/81 - BSP Programs Appendix A 2 7/81 - BSP 2.3.4 Short-Term Diffusion Estimates for 1 7/81 - BSP Accidental Atmospheric Releases 2.3.5 Long-Term Diffusion Estimates 2 7/81 - BSP 2.4.1 Hydrologic Description 2 7/81 - BSP Appendix A 2 7/81 - BSP 2.4.2 Floods 2 7/81 - BSP 2.4.3 Probable Maximum Flood (PMF) on Streams 2 7/81 - BSP and Rivers 2.4.4 Potential Dam Failures 2 7/81 - BSP 2.4.5 Probable Maximum Surge and Seiche 2 7/81 - BSP Flooding 2.4.6 Probable Maximum Tsunami Flooding 2 7/81 - BSP 2.4.7 Ice Effects 2 7/81 - BSP 2.4.8 Cooling Water Canals and Reservoirs 2 7/81 - BSP [g 2.4.9 Channel diversions 2 7/81 - BSP k.l 2.4.10 Flood Protection Requirements 2 7/81 - BSP Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-9
25A5113Rw. A SBWR standedsuretyAurrsisneport O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 2.4.11 Cooling Water Supply ~ 2 7/81 - BSP 2.4.12 Groundwater 2 7/81 - BSP BTP HGEB1 2 7/81 - BSP Accidental Released of Liquid Effluents in 2 7/81 - BSP 2.4.13 Ground and Surface Waters 2.4.14 Technical Specifications and Emergancy 2 7/81 - BSP Operation Requirements Basic Geologic and Seismic Information 2 7/81 - BSP 2.5.1 Vibratory Ground Motion 2 8/89 - BSP 2.5.2 Surface Faulting 2 7/81 - BSP 2.5.3 2.5.4 Stability of Subsurface Materials and 2 7/81 - BSP Foundations 2.5.5 Stability of Slopes 2 7/81 - BSP Chapter 3 Design of Structures, Components, Equipment, and Systems 3.2.1 Seismic Classification 1 7/81 Yes 3.2.2 System Quality Group Classification 1 7/81 Yes Appendix A (Formerly BTP RSB 3-1) 1 7/81 Yes Appendix B (Formerly BTP RSB 3-2) 1 7/81 Yes 3.3.1 Wind Loadings 2 7/81 Yes 3.3.2 Tornado Loadings 2 7/81 Yes 3.4.1 Flood Protection 2 7/81 Yes 3.4.2 Analysis Procedures 2 7/81 Yes 3.5.1.1 Intarnally Generated Missiles (Outside 2 7/81 Yes Containment) 3.5.1.2 Internally Generated Missiles (Inside 2 7/81 Yes Containment) Turbine Missiles 2 7/81 Yes 3.5.1.3 Missiles Generated by Natural Phenomena 2 7/81 Yes 3.5.1.4 Site Proximity Missiles (Except Aircraft) 1 7/81 Yes 3.5.1.5 Aircraft Hazards 2 7/81 Yes 3.5.1.6 l 1.9 10 Conformance with Standard Revoer v Plan and Applicabilsty of Codes and Standards - Amendment 1 1
I i 25A5113 Rev. A SBWR standsedsafetyAns& sis Report O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) i SBWR ! I Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 3.5.2 Structures, Systems, and Components to be 2 7/81 Yes Protected from Externally Generated Missiles 3.5.3 Barrier Design Procedures 1 7/81 Yes 3.6.1 Plant Design for Protection Against 1 7/81 Yes Postulated Piping Failures in Fluid Systems , Outside Coctainment BTP ASB-3-1 1 7/81 Yes 3.6.2 Determination of Rupture Locations and 1 7/81 Yes Dynamic Effects Associated with the Postulated Rupture of Piping , BTP MEB-3-1 2 6/87 Yes 3.7.1 Seismic Design Parameters 2 8/89 Yes ( 3.7.2 3.7.3 Seismic System Analysis Seismic Subsystem Analysis 2 2 8/89 8/89 Yes Yes 3.7.4 Seismic Instrumentation 1 7/81 Yes 3.8.1 Concrete Containment 1 7/81 Yes Append:r. 0 7/81 Yes 3.8.2 Steels cement 1 7/81 Yes Applies for Drywell Head only 3.8.3 Concrete and Steel Internal Structures of 1 7/81 Yes Steel or Concrete Containments 3.8.4 Other Seismic Category I Structures 1 7/81 Yes Appendix A 0 7/81 Yes Appendix B 0 7/81 Yes Appendix C 0 7/81 Yes Appendix D 0 7/81 Yes 3.8.5 Foundations 1 7/81 Yes 3.9.1 Special Topics for Mechanical Components 2 7/81 Yes 3.9.2 Dynamic Testing and Analysis of Systems, 2 7/81 Yes
- h. Components, and Equipment G
Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-11
25AS113R:v. A SBWR standardsatory Ana!Ysis Report O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 3.9.3 ASME Code Class 1,2, and 3 Components, 1 7/81 Yes Component Supports, and Core Support Structures Appendix A 1 4/84 Yes 3.9.4 Control Rod Drive Systems 2 4/84 Yes 3.9.5 Reactor Pressure Vessel Internals 2 7/81 Yes 3.9.6 Inservice Testing of Pumps and Valves 2 7/81 Yes 3.10 Seismic Qualification of Category 1 2 7/81 Yes Instrumentation and Electrical Equipment 3.11 Environmental Design of Mechanical and 2 7/81 Yes Electrical Equipment Chapter 4 Reactor 4.2 Fuel System Design 2 7/81 Yes Appendix A 0 7/81 Yes 4.3 Nuclear Design 2 7/81 Yes BTP CPB 4.3-1 2 7/81 Yes 4.4 Thermal and Hydraulic Design 1 7/81 Yes 4.5.1 Control Rod Drive Structural Materials 2 7/81 Yes 4.5.2 Reactor Internal and Core Support Materials 2 7/81 Yes 4.6 Functional Design of Control Rod Drive 1 7/81 Yes System Chapter 5 Reactor Coolant System and Connected Systems 5.2.1.1 Compliance with the codes and Standard 2 7/81 Yes Rule,10 CFR 50.55a 5.2.1.2 Applicable Code Cases 2 7/81 Yes 5.2.2 Overpressure Protection 2 7/81 Yes BTP RSB 5-2 0 7/81 No PWR only 5.2.3 Reactor Coolant Pressure Boundary 2 7/81 Yes Materials BTP MTEB 5-7 (Superseded by NUREG 0313) 1.9-12 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
l 25A5113Rev. A SBWR standardsurety Analysis neport 1
- ~ V
1 I Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) 1 SBWR l1 Appl. Issued Appli-
! SRP No. SRP Title or BTP Rev. Date cable? Comments 5.2.4 Reactor Coolant Pressure Boundary 1 7/81 Yes Inservice inspection and Testing 5.2.5 Reactor Coolavt Pressure Boundary Leakage 1 7/81 Yes Detection a
5.3.1 Reactor Vessel! Aaterials 1 7/81 Yes 5.3.2 Pressuti. Temp 9rature Limits 1 7/81 Yes BTP MTEB 5-2 1 7/81 Yes 5.3.3 Reactor Vessel integrity 1 7/81 Yes 5.4 Preface (Deletedi j 5.4.1.1 Pump Flywheel integrity (PWR) 1 7/81 No PWR only 5.4.2.1 Steam Generator Materials 2 7/81 No PWR only i BTP MTEB 5-3 2 7/81 No PWR only No PWR only Q 5.4.2.2 Steam Generator Tube Inservice Inspection 1 7/81 5.4.6 Reactor Core Isolation Cooling System 3 4/84 Yes isolation (BWR) Condensers 5.4.7 Residual Heat Removal (RHR) System 3 4/84 Yes RWCU/SDC ,
&ICS BTP RSB 5-1 2 7/81 Yes RWCU/SDC
&ICS 5.4.8 Reactor Water Cleanup System (BWR) 2 7/81 Yes 5.4.11 Pressurizer Relief Tank 2 7/81 No PWR only 5.4.12 Reactor Coolant System High Point Vents 0 7/81 Yes Chapter 6 Engineered Safety Features 6.1.1 Engineered Safety Features Materials 2 7/81 Yes BTP MTEB 6-1 2 7/81 No PWR only 6.1.2 Protective Coating Systems (Paints)- 2 7/81 Yes Organic Materials 6.2.1 Containment Functional Design 2 7/81 Yes Og 6.2.1.1.A PWR Dry Containments, including 2 7/81 No PWR only Q 6.2.1.1.B Subatmospheric Containments Ice Condenser Containments 2 7/81 No PWR only Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-13
25A5113Rev. A SBWR standardsafety Analysis neport O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 6.2.1.1.C Pressure-Suppression Type BWR 6 8/84 Yes Containments Appendix A 2 1/83 No Appendix B 0 1/83 No 6.2.1.2 Subcompartment Analysis 2 7/81 Yes 6.2.1.3 Mass and Energy Release Analysis for 1 7/81 Yes Postulated Loss-of-Coolant Accidents 6.2.1.4 Mass and Energy Release Analysis for 1 7/81 No PWR only Postulated Secondary System Pipe Ruptures 6.2.1.5 Minimum Containment Pressure Analysis 2 7/81 No PWR only for Emergency Core Cooling System Performance Capability Studies BTP CSB 6-1 2 7/81 No PWR only 6.2.2 Containment Heat Removal Systems 4 10/85 Yes 6.2.3 Secondary Containment Functional Design 2 7/81 Part Safety Envelope BTP CSB 6-3 2 7/81 Yes 6.2.4 Containment isolation System 2 7/81 Yes BTP CSB 6-4 2 7/81 Yes 6.2.5 Combustible Gas Controlin Containment 2 7/81 Yes Appendix A 2 7/81 Yer, BTP CSB 6-2 (Superseded by Reg. Guide 1.7) , 6.2.6 Containment Leakage Testing 2 7/81 Yes 6.2.7 Fracture Prevention of Containment 0 7/81 Yes Pressure Boundary 6.3 Emergency Core Cooling System 2 4/84 Yes BTP RSB 6-1 1 7/81 No PWR only 6.4 Control Room Habitability Systems 2 7/81 Yes Appendix A 2 7/81 Yes 6.5.1 ESF Atmosphere Cleanup Systems 2 7/81 No No safety- l related filter systems O 1.9-14 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
2SA5113Rsv. A SBWR standardsatetr Annarsis aeport O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 6.5.2 Containment spray as a Fission Product 1 7/81 No Spray Cleanup System systems not safety-related 6.5.3 Fission Product Control Systems and 2 7/81 Yes Structures 6.5.4 Ice Condenser as a Fission Product Cleanup 2 7/81 No PWR only System 6.5.5 Pressure Suppression Pool as a Fission 0 12/88 Yes Product Cleanup System 6.6 Inservice inspection of Class 2 and 3 1 7/81 Yes Components 6.7 Main Steam Isolation Valve Leakage Control 2 7/81 Yes System (BWR)
/
I D) Chapter 7 Instrumentation and Controls 7.1 Instrumentation and Controls Introduction 3 2/84 Yes Table 7-1 Acceptance Criteria and Guidelines 3 2/84 Yes for instrumentation and Controls Systems important to Safety Table 7-2 TMI Action Plan Requirements for 0 7/81 Yes Instrumentation and Controls Systems important to Safety Appendix A 1 2/84 Yes Appendix B 0 7/81 Yes 7.2 Reactor Trip System 2 7/81 Yes Appendix A (Superseded by SRP 7.1 App. B) 7.3 Engineered Safety Features Systems 2 7/81 Yes Appendix A (Superseded by SRP 7.1 App. B) 7.4 Safe Shutdown Systems 2 7/81 Yes 7.5 Information Systems important to Safety 3 2/84 Yes 7.6 Interlock Systems important to Safety 2 7/81 Yes 7.7 Control Systems 3 2/84 Yes Appendix 7-A Branch Technical Positions 2 7/81 Yes l (ICSB) Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-15
25A5113Rw. A SBWR standan! Safety Analysis Report i O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP 11the or BTP Rev. Date cable? Comments BTP ICSB 1 (DOR) (Deleted) BTP ICSB 3 2 7/81 Yes BTP ICSB 4 (PSB) 2 7/81 No PWR only BTP ICSB 5 (Superseded by Std. Tech Specs) BTP ICSB 9 (Superseded by Std. Tech Specs) BTP ICSB 12 2 7/81 Yes BTP ICSB 13 2 7/81 No PWR only BTP ICSB 14 2 7/81 No PWR only BTP ICSB 16 (Deleted) BTP ICSB 19 (Deleted) BTP ICSB 20 2 7/81 No BTP ICSB 21 2 7/81 Yes BTP ICSB 22 2 7/81 Yes BTP ICSB 25 (Superseded by Std. Tech Specs) BTP ICSB 26 2 7/81 Yes Appendix 7-B General Agenda, Station Site 1 7/81 - COL Visits Chapter 8 Electric Power 8.1 Electric Power-Interaction 2 7/81 Yes Table 8-1 Acceptance Criteria and Guidelines 2 7/81 Yes for Electric Power Systems 8.2 Offsite Power System 3 7/83 Yes Interface Appendix A 0 7/83 Yes interface 8.3.1 AC Power Systems (Onsite) 2 7/81 Yes Appendix (Superseded by BTP PSB-2) 8.3.2 DC Power Systems (Onsite) 2 7/81 Yes Appendix 8-A Branch Technical Positions 2 7/81 Yes ( (PSB) i BTP ICSB 2 (PSB) (Superseded by IEEE-387) 1.9-16 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 l
2SA5113Rsv. A SBWR standardssiety Annirsis neport ! o l
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Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 2 7/81 No PWR only BTP ICSB 4 (PSB) BTP ICSB 8 (PSB) 2 7/81 No No safety. related diesel generators BTP ICSB 11 (PSB) 2 7/81 Yes BTP ICSB 15 (PSB) (Deleted) BTP ICSB 17 (PSB)(Superseded by Reg. Guide 1.9) BTP ICSB 18 (PSB) 2 7/81 Yes BTP ICSB 21 (PSB) 2 7/81 Yes BTP PSB 1 0 7/81 Yes BTP PSB 2 0 7/81 Yes s Appendix 8-B General Agenda, Station Site 0 7/81 - COL i
\ Visits Chapter 9 Auxiliary Systems 9.1.1 New Fuel Storage 2 7/81 Yes 9.1.2 Spent Fuel Storage 3 7/81- Yes 9.1.3 Spent Fuel Pool Cooling and Cleanup 1 7/81 Yes System 9.1.4 Light Load Handling System (Related to 2 7/81 ,Yes Refueling)
BTP ASB (9-1 (Superseded by NUREG 0554) 9.1.5 Overhead Heavy Load Handling Systems 0 7/81 Yes 9.2.1 Station Service Water System 4 6/85 Yes 9.2.2 Reactor Auxiliary Cooling Water Systems 3 6/86 Yes 9.2.3 Demineralized Water Makeup System 2 7/81 Yes 9.2.4 Potable and SanitaryWater Systems 2 7/81 - Interface 9.2.5 Ultimate Heat Sink 2 7/81 - Interface BTP ASB 9-2 2 7/81 Yes 9.2.6 Condensate Storage Facilities 2 7/81 Yes 9.3.1 Compressed Air System 1 7/81 Yes Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-17
25A5113Rrv. A SBWR standardsafety Analysis Repod O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 9.3.2 Process and Post-Accident Sampling 2 7/81 Yes Systems 9.3.3 Equipment and Floor Drainage System 2 7/81 Yes 9.3.4 Chemical and Volume Control System 2 7/81 No PWR only (PWR)(including Boron Recovery System) 9.3.5 Standby Liquid Control System (BWR) 2 7/81 Yes 9.4.1 Control Room Area Ventilation System 2 7/81 Yes 9.4.2 Spent Fuel Pool Area Ventilation System 2 7/81 Yes 9.4.3 Auxiliary and Radwaste Area Ventilation 2 7/81 Yes System 9.4.4 Turbine Area Ventilation System 2 7/81 Yes 9.4.5 Engineered Safety Feature Ventilation 2 7/81 No System 9.5.1 Fire Protection Program 3 7/81 Yes BTP CMEB 9.5-1 2 7/81 Yes Appendix A(Deleted) 9.5.2 Communication Systems 2 7/81 Yes 9.5.3 Lighting Systems 2 7/81 Yes 9.5.4 Emergency Diesel Engine Fuel Oil Storage 2 7/81 Yes and Transfer System 9.5.5 Emergency Diesel Engine Cooling Water 2 7/81 Yes System 9.5.6 Emergency Diesel Engine Starting System 2 7/81 Yes 9.5.7 Emergency Diesel Engine Lubrication 2 7/81 Yes System , 9.5.8 Emergency Diesel Engine Combustion Air 2 7/81 Yes intake and Exhaust System Chapter 10 Steam and Power Conversion System 10.2 Turbine Generator 2 7/81 Yes 10.2.3 Turbine Disk Integrity 1 7/81 Yes 10.3 Main Steam Supply System 3 4/B4 Yes 10.3.6 Steam and Feedwater System Materials 2 7/81 Yes 1.9-18 Confortnance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
l \ 25AS113Rw. A SBWR studentsafety Anlysis Report ( L Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. - Issued Appli. SRP No. SRP Title or BTP Rev. Date cable"/ Comments 10.4.1 Main Condensers 2 7/81 Yes 10.4.2 Main Condenser Evacuation System 2 7/81 Yes 10.4.3 Turbine Gland Sealing System 2 7/81 Yes 10.4.4 Turbine Bypass System 2 7/81 Yes 10.4.5 Circulating Water System 2 7/81 Yes 10.4.6 Condensate Cleanup System 2 7/81 Yes 10.4.7 Condensate and Feedwater System 3 4/84 Yes BTP ASB 10-2 3 4/84 No PWR only 10.4.8 Steam Generator Blowdown System (PWR) 2 7/81 No PWR only 10.4.9 Auxiliary Feedwater System (PWR) 2 7/81 No PWR only BTP ASB 10-1 2 7/81 No PWR only N,_) ' Chapter 11 Radioactive Waste Management 11.1 Source Terms 2 7/81 Yes Subsection 1.9.1.14 11.2 Liquid Waste Management Systems 2 7/81 Yes ( 11.3 Gaseous Waste Management Systems 2 7/81 Yes BTP ETSB 11-5 0 7/81 Yes 11.4 Solid Waste Management Systems 2 7/81 Yes BTP ETSB 11-3 2 7/b1 Yes Appendix 11.4-A 0 7/81 Ves ( 11.5 Process and Effluent Radiological 3 7/81 Yes l Monitoring Instrumentation and Sampling l l Systems Appendix 11.5-A 1 7/81 Yes Chapter 12 Radiation Protection 12.1 Assuring That Occupational Radiation 2 7/81 Yes Exposures are As Low As is Relatively Achievable i
/ 12.2 Radiation Sources 2 7/81 Yes Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-19 l
I 25A5113Rev. A SBWR stadardsarety Anarrsisneport O Table 1.9-1 NRC Standard Review Pians and Branch Technical Positions Applicability to SBWR (Continued) l SBWR Appl. Issued Appil-SRP No. SRP Title or BTP Rev. Date cable? Comments 12.3- Radiation Protection Design Features 2 7/81 Yes 12.4 12.5 Operational Radiation Protection Program 2 7/81 - COL Chapter 13 Conduct of Operations 13.1.1 Management and Technical Support 2 7/81 - COL Organization 13.1.2- Operating Organization 2 7/81 - COL
; 1.3 13.2 Training (Replaced by SRP Sections 13.2.1 and 13.2.2) 13.2.1 Reactor Operator Training 0 7/81 -
COL 13.2.2 Training For Non-Licensed Plant Staff 0 7/81 - COL 13.3 Emergency Planning 2 7/81 - COL 13.4 Operational Review 2 7/81 - COL 13.5 Plant Procedures (Replaced by SRP Sections 13.5.1 and 13.5.2) 13.5.1 Administration Procedures 0 7/81 - COL 13.5.2 Operating and Maintenance Procedures 1 7/85 -- COL Appendix A 0 7/85 - COL 13.6 Physical Security 2 7/81 Yes SBWR and COL; Safeguards Information provided i ! Chapter 14 initial Test Program ( 14.1 initial Plant Test Programs - PSAR (Deleted) 14.2 Initial Plant Test Programs - FSAR 2 7/81 Yes i 14.3 Standard Plant Design, initial Test Program-Final Design Approval (FDA)(Deleted) l l Chapter 15 Accident Analysis 1.9-20 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
l 2SAS113Rsv. A SBWR studerdsovery Antrsis Report i( O
'O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued)
SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 15.0 Introduction 2 7/81 Yes j 15.1.1- Decresse in Feedwater Temperature, 1 7/81 Yes 15.1.4 Increase in Feedwater Flow, increase in Steam Flow, and inadvertent Opening of a Steam Generator Relief or Safety Valve 15.1.5 Steam System Piping Failures inside and 2 7/81 No PWR only Outside of Contamination (PWR) Appendix A 2 7/81 No PWR only 15.2.1- Loss of External load, Turbine Trip, Loss of 1 7/81 Yes 15.2.5 Condenser Vacuum, Closure of Main Steam isolation Valve (BWR), and Steam Pressure Regulator Failure (Closed) 15.2.6 Loss of Nonemergency AC Power to the 1 7/81 Yes Station Auxiliaries y/ 15.2.7 Loss of Normal Feedwater Flow 1 7/81 Yes 15.2.8 Feedwater System Pipe Breaks inside and 1 7/81 No PWR only Outside Containment (PWR) 15.3.1- Loss of Forced Reactor Coolant Flow 1 7/81 No No forced 15.3.2 Including Trip of Pump and Flow Controller recirculatiot. Malfunctions systems in I SBWR
- 15.3.3- Reactor Coolant Pump Rotor Seizure and 2 7/81 No No forced 15.3.4 Reactor Coolant Pump Shaft Break recirculation systems in l SBWR 15.4.1 Uncontrolled Control Rod Assembly 2 7/81 Yes Withdrawal from a Subcritical of Low Power Startup Condition 15.4.2 Uncontrolled Control Rod Assembly 2 7/81 Yes Withdrawal at Power 15.4.3 Control Rod Misoperation (System 2 7/81 Yes Malfunction or Operator Error) 15.4.4- Startup of an inactive Loop or Recirculation 1 7/81 No No forced 15.4.5 Loop at an incorrect Temperature, and Flow recirculation Controller Malfunction Causing an increase systems in in BWR Core Flow Rate SBWR N) i Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-21 r
l 25AS113Rev. A SBWR standardsafety Analysis neport O Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SRWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments i 15.4.6 Chemical and Volume Control System 1 7/81 No PWR only Malfunction That Results in a Decrease in the Boron Concentration in the Reactor Coolant (PWR) 15.4.7 inadvertent Loading and Operations of a 1 7/81 Yes l Fuel Assembly in an Improper Position 15.4.8 Spectrum of Rod Ejection Accidents (PWR) 2 7/81 No PWR only Appendix A 1 7/81 No PWR only 15.4.9 Spectrum of Rod Drop Accidents (BWR) 2 7/81 Yes Appendix A 2 7/81 Yes l 15.5.1- Inadvertent Operation of ECCS and 1 7/81 Yes 15.5.2 Chemical and Volume Control System l Malfunction That increases Reactor Coolant i inventory 15.6.1 Inadvertent Opening of a PWR Pressurizer 1 7/81 Yes Relief Valve or a BWR Relief Valve 15.6.2 Radiological consequences of the Failure of 2 7/81 Yes l l Small Lines Carrying Primary Coolant l Outside Containment 15.6.3 Radiological Consequences of Steam 2 7/81 No PWR only l Generator Tube Failure (PWR) 15.6.4 Radiological Consequences of Main Steam 2 7/81 Yes Line Failure Outside Containment (BWR) 15.6.5 Loss-of-Coolant Accidents Resulting from 2 7/81 Yes Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary Appendix A 1 7/81 Yes Appendix B 1 7/81 Yes Appendix C (Deleted) Appendix D 1 7/81 No 15.7.1 Waste Gas System Failure (Deleted) 15.7.2 Radioactive Liquid Waste System Leak or Failure (Released to Atmosphere) (Deleted) 15.7.3 Postulated Radioactive Release Due to 2 7/81 Yes Liquid-Containing Tank Failures 1.9-22 Conformance with Standard Review Plan and Apphcability of Codes and Standards - Amendrnent 1
25A5113Rev. A SBWR standardsafety Ansirsis neport CN Table 1.9-1 NRC Standard Review Plans and Branch Technical Positions Applicability to SBWR (Continued) SBWR Appl. Issued Appli-SRP No. SRP Title or BTP Rev. Date cable? Comments 15.7.4 Radiological Consequences of Fuel Handling 1 7/81 Yes Accidents 15.7.5 Spent Fuel Cask Drop Accidents 2 7/81 Yes 15.8 Anticipated Transients Without Scram 1 7/81 Yes Appendix (Deleted) Chapter 16 Technical Specifications 16.0 Technical Specifications 1 7/81 Yes Chapter 17 Quality Assurance 17.1 Quality Assurance During the Design and 2 7/81 Yes j'3 Construction Phases t I U/ 17.2 Quality Assurance During the Operations 2 7/81 - COL Phase l Chapter 18 Human Factors Engineering 18.0 Human Factors Engineering / Standard 1 9/84 Yes Review Plan Development 18.1 Control Room 0 9/84 Yes Appendix A 0 9/84 Yes 18.2 Safety Parameter Display System 0 11/B4 Yes Appendix A 0 11/84 Yes I i 1 .t ) ! \d Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-23 i l
25AS113 Riv. A SBWR standardsafety Analysis Reput O Table 1.9 2 NRC Regulatory Guides Applicability to SBWR SBWR Appl. Issued Appli-RG No. Regulatory Guide Tatie Rev. Date cable? Comments Net Positive Suction Head for Emergency 0 11R0 No No safety 1.1 Core Cooling and Containment Heat related Removal System Pumps. pumps for SBWR 1.2 Thermal Shock to Reactor Pressure Vessels. 0 1100 No Withdrawn 7/31/91 1.3 Assumptions Used for Evaluating the 2 694 No SBWR will Potential Radiological Consequences of a comply with Loss-of-Coolant Accident for Boiling Water intent of EPRI URD, Reactors. NU REG-1465 assumptions, and proposed modifications to 10CFR50 and 10CFR100 1.4 Assumptions Used for Evaluating the 2 694 No PWR only Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors. 1.5 Assumptions Used for Evaluating the 0 3R1 Yes Potential Radiological Consequences of a Steam Line Break Accident for Boiling Water Reactors. 1.6 Independence Between Redundant Standby 0 391 Yes No safety (On-site) Power Sources and Between Their related Distribution Systems. Diesel Generators for SBWR 1.7 Control of Combustible Gas Concentrations 2 1198 Yes in Containment Following a Loss-of-Coolant Accident. 1.8 Qualification and Training of Personnel for 2 4/87 - See Table Nuclear Power Plants. 17.0-1 1.9 Selection, Design, and Qualification of 2 1299 No No safety l Diesel-Generator Units Used as Standby related (on-site) Electric Power Systems at Nuclear Diesel Plants. Generators for SBWR 1.9-24 Conforrnance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
l 25AS113Rev. A SBWR standardsarety Analysis neport l Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) SBWR Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cable? Comments l 1.11 Instrument Lines Penetrating Primary 0 292 Yes Reactor Containment. 1.12 instrumentation for Earthquakes. 1 494 Yes 1.13 Spent Fuel Storage Facility Design Basis. 1 1295 Yes 1.14 Reactor Coolant Pump Flywheel Integrity. 1 895 No PWR only 1.16 Reporting of Operating information- 4 8/75 - COL Appendix A Technical Specifications. , 1.17 Protection of Nuclear Power Plants Against 1 693 No Withdrawn l l Industrial Sabotage. 7/5/91 l 1.20 Comprehensive Vibration Assessment 2 596 Yes Performed l Program for Reactor internals During During Preoperational and Initial Startup Testing. Power Ascension Testing O 1.21 Measuring, Evaluating and Reporting 1 6R4 Yes (V) Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light Water Nuclear Power Plants. 1.22 Periodic Testing of Protection System 0 292 Yes Actuation - Functions. 1.23 On-site Meteorological Programs. 0 292 - BSP 1.24 Assumptions Used for Evaluating the 0 392 No PWR only Potential Radiological Consequences of a Pressurized Water Reactor Gas Storage Tank Failure. 1.25 Assumptions Used for Evaluating the 0 392 Yes Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors. 1.26 Quality Group Classifications and 3 2B6 - See Tables Standards for Water , Steam , and 3.2-1 and Radioactive-Waste-Containing Components 17.0-1 of Nuclear Power Plants. 1.27 Ultimate Heat Sink for Nuclear Power Plants. 2 106 Yes Interface A) 8/85 See Table (d Quality Assurance Program Requirements 3 - 1.28 (Design and Construction). 17.0-1 Conformance with Standard Review Plan and Applicabihty of Codes and Standards - Amendment 1 1.9-25 l
l l 25A5113Riv. A I F SBWR standardsurety Analysis nepois l 9 j Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) SBWR l Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cable? Comments 1.29 Seismic Design Classification. 3 998 - See Tables 3.2-1 and 17.0-1 1.30 Quality Assurance Requirements for the 0 892 - See Table Installation, inspection, and Testing of 17.0-1 instrumentation and Electric Equipment. 1.31 Control of Ferrite Content in Stainless Steel 3 498 Yes Weld Metal. 1.32 Criteria for Safety-Related Electric Power 2 2R7 Yes Systems for Nuclear Power Plants. 1.33 Quality Assurance Program Requirements 2 2R8 - COL (Operations). 1.34 Control of Electroslag Weld Properties. 0 12n2 Yes 1.35 In-Service Inspection of Ungrouted Tendons 3 8/90 No in Prestressed Concrete Containment Structures. 1.35.1 Determining Prestressing Forces for G 8/90 No inspection of Prestressed Concrete Containments. 1.36 Non-Metallic insulation for Austenitic 0 2/73 Yes Stainless Steel. 1.37 Quality Assurance Requirements for 0 3/73 - See Table Cleaning of Fluid Systems and Associated 17.0-1 Components of Water-Cooled Nuclear Power Plants. 1.38 Quality Assurance Requirements for 2 5/77 - See Table Packaging, Shipping, Receiving, Storage, 17.0-1 and Handling of items for Water Cooled Nuclear Power Plants. 1.39 Housekeeping Requirements for Water- 2 9/77 - See Table Cooled Nuclear Power Plants. 17.0-1 1.40 Qualification Tests of Continuous-Duty 0 393 Yes Motors installed inside the Containment of Water-Cooled Nuclear Power Plants. O 1.9 26 Conformance with Standard Review Plan and Apphcability of Codes and Standards - Amendment 1
25AS113Rsv. A SBWR stamtsidsafetyAnalysis Report (3 V Table 1.9 2 NRC Regulatory Guides Applicability to SBWR (Continued) I l SBWR Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cable? Comments l 1.41 PreoperationalTesting of Redundant On-site 0 3/73 Part No safety i Electric Power Systems to Verify Proper related , Load Group Assignments. Diesel l Generators for SBWR, therefore, l - only DC l portions are applicable 1.43 Control of Stainless Steel Weld Cladding of 0 5/73 No Potential of Low Alloy Steel Components. underclad I cracking nota concern 1.44 Control of Use of Sensitized Steel. 0 5/73 Yes 1.45 Reactor Coolant Pressure Boundary Leakage 0 5/73 Yes Detection Systems. l
/ 1.47 Bypassed and Inoperable Status Indication 0 5/73 Yes l
l for Nuclear Power Plant Safety Systems. 1.49 Power Levels of Nuclear Power Plants. 1 1293 Yes 1.50 Control of Preheat Temperature Welding of 0 5/73 Yes Low-Alloy Steel. 1.52 Design, Testing, Maintenance Criteria for 2 3/78 No Post Accident Engineered Safety-Feature Atmosphere Cleanup System Air Filtration l and Absorption Units of Light-Water-Cooled Nuclear Power Plants. 1.53 Application of the Single-Failure Criterion to 0 6/73 Yes ' Nuclear Power Plant Protection Systems. l 1.54 Quality Assurance Requirements for 0 6/73 Yes l Protective Coatings Applied to Water-Cooled Nuclear Power Plants. 1.56 Maintenance of Water Purityin Boiling 1 7/78 Yes Water Reactors. 1.57 Design Limits and Loading Combinations for 0 6/73 Yes l Metal Primary Reactor Containment System l Components. 1.58 Qualification of Nuclear Power Plant Superseded See Table gg Inspection, Examination, and Testing 17.0-1 ( V) Personnel. ; Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-27 1
25AS113Riv. A SBWR standardsarery Analysis Report O Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) I SBWR Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cable? Comments
]
1.59 Design Basis Floods for Nuclear Power 2 897 Yes Plants. 1.60 Design Response Spectra for Seismic 1 1293 Yes l Design of Nuclear Power Plants. 1.61 Damping Values for Seismic Design of 0 1093 Yes Nuclear Power Plants. 1.62 ManualInitiation of Protective Actions. 0 1093 Yes 1.63 Electric Penetration Assemblies in 3 2/87 Yes Containment Structures of Nuclear Power Plants. 1.64 Quality Assurance Requirements for the Superseded See Table i Design of Nuclear Power Plants. 17.0-1 l 1.65 Materials and Inspections for Reactor Vessel 0 10/73 Yes I Closure Studs. j l 1.68 initial Test Programs for Water-Cooled 2 8/78 Yes ' Reactor Power Plants. 1.68.1 Preoperational and initial Startup Testing of 1 107 Yes Feedwater and Condensate Systems for Boiling Water Reactor Power Plants. 1.68.2 Initial Startup Test Program to Demonstrate 1 7/78 Yes Remote Shutdown Capability for Water-Cooled Nuclear Power Plants. 1.68.3 Preoperational Testing of instrument and 0 4/82 Yes ! Control Air Systems. l 1.69 Concrete Radiation Shields for Nuclear 0 12/73 Yes Power Plants. 1.70 Standard Format and Content of Safety 3 11/78 Yes Analysis Reports for Nuclear Power Plants. 1.71 We! der Qualifications for Areas of Limited 0 12/73 - COL I Accessibility. 1.72 Spray Pond Piping Made from Fiberglass- 2 1198 No Reinforced Thermosetting Resin. 1.73 Qualification Tests of Electric Valve 0 1/74 Yes Operators installed inside the Containment of Nuclear Power Plants. 1.74 Quality Assurance Terms and Definitions. Superseded See Table 17.0-1 1.9-28 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
l 25A5113Rsv. A j SBWR standardsafety Anstrsis Reput , I m lU l Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) SBWR Appl. Issued Appli- l RG No. Regulatory Guide Title Rev. Date cable 7 Comments 1.75 Physical Independence of Electric Systems. 2 998 Yes 1.76 Design Basis Tornado for Nuclear Power 0 4R4 Yes Plants. ; 1.77 Assumptions Used for Evaluating a Control 0 594 No PWR only Rod Ejection Accident for Pressurized Water ; Reactors. l 1.78 Assumptions for Evaluating the Habitability 0 694 Yes l of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release. PreoperationalTesting of Emergency Core 1 995 No PWR only 1.79 Cooling Systems for Pressurized Water Reactors. 1.81 Shared Emergency and Shutdown Electric 1 105 .No SBWR is a Systems for Multi-Unit Power Plants. single unit plant (' No ECC 1.82 Water Sources for Long-Term Recirculation 1 11/85 Part Cooling Following Loss-of-Coolant Accident. pumps in I l SBWR 1.83 In-Service inspection of Pressurized Water 1 795 No PWR only Reactor Steam Generator Tubes. 1.84 Design and Fabrication Code Case 27 11/90 Yes . Acceptability, ASME Section lit, Division 1. 1.85 Materials Code Case Acceptability, ASME 27 11/90 Yes Section ill, Division 1. 1.86 Termination of Operating Licenses for 0 604 - COL I Nuclear Reactors. 1.87 Guidance for Construction Class 1 1 695 No Components in Elevated Temperature Reactor (Supplement to ASME Section 111 Code Cases 1592,1593,1594,1595, and 1596). 1.88 Collection, Storage, and Maintenance of Superseded See Table Nuclear Power Plant Quality Assurance 17.0-1 Records. 1.89 Environmental Qualification of Certain 1 6/84 Yes f ,O Electric Equipment important to Safety for Nuclear Power Plants.
'Q Conformance with Standard Review Plan and Applicability of Codes and Standards - Arnendment 1 1.9-29 i
25AS113Riv. A \ SBWR standardsafetyAnarrsisReport O Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) SBWR Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cable? Comments 1.90 In-Service inspection of Prestressed 1 897 No Concrete Containment Structures with Grouted Tendons. 1.91 Evaluations of Explosions Postulated to 1 298 - BSP Occur on Transportation Routes Near Nuclear Power Plants. 1.92 Combining Model Responses and Spatial 1 206 Yes Components in Seismic Response Analysis. Availability of Electric Power Sources. 0 1294 Part No safety-1.93 related diesel generators. Only DC portion is applicable 1.94 Quality Assurance Requirements for 1 496 - See Table Installation, inspection, and Testing of 17.0-1 Structural Steel During the Construction Phase of Nuclear Power Plants. 1.95 Protection of Nuclear Power Plant Control 1 197 Yes Room Operators Against an Accidental Chlorine Release. 1.96 Design of Main Steam isolation Valve 1 696 No No MSIV Leakage Control Systems for Boiling Water Leakage Reactor Nuclear Power Plants. Control System 1.97 instrumentation for Light-Water-Cooled 3 5/83 Yes Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident. 1.98 Assumptions for Evaluating the Potential 0 396 Yes Radiological Consequences of a Radioactive Offgas System Failure in a Boiling Water Reactor. 1.99 Effects of Residual Elements on Predicted 2 5/88 Yes Radiation Damage to Reactor Vessel Materials. 1.100 Seismic Qualification of Electric Equipment 2 6/88 Yes for Nuclear Power Plants. 2 10/81 Yes ! 1.101 Emergency Planning and Preparedness for Nuclear Power Reactors. l i I 1.9-30 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 l
2SAS113Rev. A SBWR stamtsrdsafety Anstrsis separ p i
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Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) SBWR Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cabla? Comments 1.102 Flood Protection for Nuclear Power Plants. 1 9#6 Yes 1.105 Instrument Setpoints for Safety-Related 2 2/86 Yes Systems. 1.106 Thermal Overload Protection for Electric 1 3n7 Yes Motors on Motor-Operated Valves. 1.107 Qualifications for Cement Grouting for 1 2n7 No Prestressing Tendons in Containment Structures. 1.108 Periodic Testing of Diesel Generator Units 1 8#7 No No safety-Used as Onsite Electric Power Systems at related Nuclear Power Plants. Diesel Generators for SBWR 1.109 Calculation of Annual Doses to Man from 1 1007 Yes Routine Releases of Reactor Effluents for the
]j Purpose of Evaluating Compliance with 10 v CFR Part 50, Appendix 1.
1.110 Cost Benefit Analysis for Radwaste Systems 0 396 Yes for Light-Water-Cooled Nuclear Power Plants. 1.111 Methods for Estimating Atmospheric 1 707 Yes Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors. 1.112 Calculation for Releases of Radioactive 0-R 5#7 Yes Materials in Gaseous and Liquid Effluents from Light-Water-Cooled Power Reactors. I 1.113 Estimating Aquatic Dispersion of Effluents 1 4#7 Yes l from Accidental and Routine Reactor Releases for the Purpose of implementing Appendix 1. 1.114 Guidance on Being Operating at the 2 5/89- - COL Controls of a Nuclear Power Plant. ) ! 1.115 Protection Against Low-TrajectoryTurbine 1 7/77 Yes I Missiles. 1.116 Quality Assurance Requirements for 0-R 5/77 - See Table Installation, inspection, and Testing of 17.0-1
) Mechanical Equipment and Systems.
1.117 Tornado Design Classification. 1 408 Yes l Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-31
25A5113R1v. A SBWR staniadsareryAnalysis neport Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) ! SBWR Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cable? Comments 1.118 Periodic Testing of Electric Power and 2 6#8 Yes l l Protection Systems. l 1.120 Fire Protection Guidelines for Nuclear Power 1 1197 Yes l Plants. l 1.121 Bases for Plugging Degraded PWR Steam 0 896 No PWR only ! Generator Tubes. \ l 1.122 Development of Floor Design Response 1 2R8 Yes l Spectra for Seismic Design of Floor-l Supported Equipment or Components. 1.123 Quality Assurance Requirements for Superseded See Table Control of Procurement of items and 17.0-1 Services for Nuclear Power Plants. 1.124 Service Limits and Loading Combinations 1 198 Yes for Class 1 Linear-Type Component Supports. 1.125 Physical Models for Design and Operation of 1 1098 Yes Hydraulic Structures and Systems for Nuclear Power Plants. 1.126 An Acceptable Model and Related Statistical 1 398 Yes Methods for the Analysis for Fuel Densification. 1.127 Inspection of Water-Control Structures 1 3n8 - COL Associated with Nuclear Power Plants. 1.128 Installation Design and Installation of Large 1 10/78 Yes Lead Storage Batteries for Nuclear Power Plants. 1.129 Maintenance, Testing, and Replacement of 1 298 Yes Large Lead Storage Batteries for Nuclear Power Plants. 1.130 Service Limits and Loading Combination for 1 1098 Yes Class 1 Plate-and-Shell-Type Comr.anect Supports. l l 1.131 Qualification Tests of Electric Cable, Field 0 897 Yes l Splices, and Connections for Light-Water-Cooled Nuclear Power Plants. 1.132 Site Investigations for Foundations of 1 3/79 Yes Nuclear Power Plants. 1.9-32 Conformance with Standard Review Plan a ud Applicability of Codes and Standards - Amendment 1
1 25AS113Rev. A SBWR senadantsaretyanssysis neport O 'V Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) SBWR Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cable? Comments 1.133 Loose-Part Detection Program for the 1 5/81 Yes Primary Systems of Light-Water-Cooled Reactors. 1.134 Medical Evaluation of Licensed Personnel 2 4/87 -
-COL for Nuclear Power Plants.
I 1.135 Normal Water Level and Discharge et 0 9R7 Yes l Nuclear Power Plants 1.136 Materials, Construction, and Testing of 2 6/81 Yes Concrete Containments (Articles CC-1000,
-2000, and -4000 through -6000 of the " Code i for Concrete Reactor Vessels and l Containment".
1.137 Fuel-Oil Systems for Standby Diesel 1 1099 No No safety-Generators. related Diesel l, Generators for SBWR l( 1.138 Laboratory Investigations of Soils for 0 4R8 Yes l Engineering Analysis and Design of Nuclear Power Plants. 1.139 Guidance for Residual Heat Removal. 0 598 Yes 1.140 Design, Testing, and Maintenance Criteria 1 1099 Yes for Normal Ventilation Exhaust System Air Filtration and Absorption Units of Light-Water-Cooled Nuclear Power Plants. 1.141 Containment Isolation Provisions for Fluid 0 4n8 Yes Systems. l 1.142 Safety-Related Concrete Structures for 1 10/81 Yes Nuclear Power Plants (Other than Reactor l Vessels and Containments). 1.143 Guidance for Radioactive Waste 1 10n9 Yes Management Systems, Structures, and Components installed in Light-Water-Cooled Nuclear Power Plants. 1.144 Auditing of Quality Assurance Programs Superseded See Table Nuclear Power Plants. 17.0-1 1.145 Atmospheric Dispersion Models for 1 11/82 Yes i ( n\ Potential Accident Consequences (,/ Assessment at Nuclear Power Plants. Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-33 l l
25A5113 R1v. A SBWR standardsaretyAnarr sisnepoa O l Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) SBWR Appl. Issued Appli- ! RG No. Regulatory Guide Title Rev. Date cable 7 Comments
=
1.146 Qualification of Quality Assurance Program Superseded See Table Audit Personnel for Nuclear Power Plants. 17.0-1 1.147 Inservice Inspection Code Case Acceptabilty 8 11/90 Yes
- ASME Section XI, Division 1.
1.148 Functional Specification for Active Valve 0 3/81 Yes Assemblies in Systems important to Safety in Nuclear Power Plants. 1.149 Nuclear Power Plant Simulation Facilities for 1 4/87 - COL Use in Operator License Examinations. 1.150 Ultrasonic Testing of Reactor Vessel Welds 1 2/83 Yes During Preservice and inservice Examinations. 1.151 Instrument Sensing Lines. 0 7/83 Yes 1.152 Criteria for Programmable Digital Computer 0 11/85 Yes System Software in Safety-Related Systems of Nuclear Power Plants. 1.153 Criteria for Power instrumentation, and 0 12/85 Yes Control Portions of Safety Systems. 1.154 Format and Contents of Plant Specific 0 1/87 No PWR only Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors. 1.155 Station Blackout. 0 8/88 Part No emergency power required for SBWR. Only . coping i analysis l applicable. l 1.156 Environmental Qualification of Connection 0 11/87 Yes Assemblies for Nuclear Power Plants. 1.157 Best Estimate Calculations of Emergency 0 5/89 Yes Core Cooling System Performance. 1.158 Qualification of Safety-Related Lead Storage 0 2/89 Yes Batteries for Nuclear Power Plants. 5.1 Serial Numbering of Fuel Assemblies for 0 12/72 Yes Light-Water-Cooled Nuclear Power Plants. 5.7 Entry / Exit Control for Protected Areas, Vital 1 5/80 No Areas, and Material Access. 1.9-34 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
2SA5113Rev. A SBWR staudardsaveurAnsirsisneport i J
)
Table 1.9-2 NRC Regulatory Guides Applicability to SBWR (Continued) SBWR Appl. Issued Appli-RG No. Regulatory Guide Title Rev. Date cable? Comments j 5.12 General Use of Locks in the Protection and 0 1193 Yes ) Control of Facilities and Special Nuclear Materials. l l 5.44 Perimeter Intrusion Alarm Systems. 2 5/80 Yes 5.61 Intent and Scope of the Physical Protection 0 6/80 Yes Safeguards Upgrade Rule Requirements for Fixed Sites. Information Provided 5.65 Vital Area Access Controls, Protection of 0 9/86 Yes ; l Physical Security Equipment, and Key and Lock Controls. 8.2 Guide for Administrative Practices in 0 2B3 - COL Radiation Monitoring. 8.5 Criticality and Other interior Evacuation 1 3/81- Yes Signals. 8.8 information Relevant to Ensuring inat 3 6/78 Yes j Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As is Reasonably Achicvable. 8.10 Operating Philosophy for Maintaining 1-R 5/77 - COL Occupational Radiation Exposures as Low as is Reasonably Achievable ; 8.12 Criticality Accident Alarm Systems. 2 10/88 Yes 8.19 Occupational Radiation Dose Assessmentin 1 6/79 - COL Light-Water Reactor Power Plants Design j Stage Man-Rem Estimates. BJS Radiation Dose to the Embrvo/ Fetus Q 282 yfa BJS Control of Access to Hiah and Very Hiah Q S$3 Ygs Radiation Areas in Nuclear Power Plants 7 i i l i I I r Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-35 l
, - . ~
i 25A5113Rev. A SBWR standardsafetyanalysis neport l 9 Table 1.9-3 Industrial Codes and Standards Applicable to SBWR 1 Code or Standard Number Year Title American Concrete institute (ACl) l 211.1 1981 Practice for Selecting Proportions for Normal, Heavy Weight, and Mass Concrete l 212 1981 Guide for Admixtures in Concrete 214 1977 Recommended Practice for Evaluation of Strength Test Results of Concrete 301 1984 Specifications for Structural Concrete for Buildings 304 1973 Practice for Measuring, Mixing, Transporting, and Placing of j Concrete l 305 1977 Recommended Practice for Hot Weather Concreting 306 1978 Recommended Practice for Cold Weather Concreting 307 1979 Specification for the Design and Construction of Reinforced j Concrete Chimneys 308 1981 Practice for Curing Concrete l 309 1972 Practice for Consolidation of Concrete 311.1R 1981 ACI Manual of Concrete inspection 311.4R 1981 Guide for Concrete Inspection 315 1980 Details and Detailing of Concrete Reinforcement l 318 1989 Building Code Requirements for Reinforced Concrete l I 349 1990 Code Requirements for Nuclear Safety-Related Concrete Structures 1 359 (See ASME BPVC Section Ill) l l American Institute of Steel Construction (AISC) ! i i N690 1984 Specifications for the Design, Fabrication, and Erection of Steel i Safety-Related Structres for Nuclear Facilities j ! SG-673 1986 Specification for the Design of Cold-Formed Steel Structural l Members
- - Load and Resistance Factor Design (LRFD), Specification for Structural Steel Buildings - - Manual of Steel Construction - - Specification for Structural Steel Buildings- Allowable Stress Design (ASD) and Plastic Design 1.9-36 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 l
2SA5113Rev. A SBWR standardsakry Analysis norart Table 1,9-3 Industrial Codes and Standards Applicable to SBWR (Continued) Code or Standard Number Year Title American Nuclear Society (ANS) 2.3 1983 Standard for Estimating Tornado and Other Extreme Wind Characteristics at Nuclear Power Sites 2.8 1981 Determining Design Basis Flooding at Power Reactor Sites 4.5 1980 Criteria for Accident Monitoring Functions in Light-Watur-Cooled Reactors 5.1 1979 Decay Heat Power in LWRs 6.4 1985 Guidelines on the Nuclear Analysis and Design of Concrete Radiation Shielding for Nuclear Power Plants 18.1 1984 Nuclear Power Plants-Source Term Specification 52.1 1983 Nuclear Safety Design Criteria for the Design of Stationary Boiling Water Reactor Plants 55.4 1979 Gaseous Radioactive Waste Processing Systems for Light Water Reactors i 56.8 1987 Containment System Leakage Testing Requirements 56.11 1988 Design Criteria for Protection Against the Effects of Compartment Flooding in Light Water Reactor Plants 57.1 1980 Design Requirements for LWR Fuel Handling Systems 57.2 1983 Design Requirements for LWR Spent Fuel Storage Facilities at NPP 57.5 1981 Light Water Reactor Fuel Assembly Mechanical Design and Evaluation 58.2 1988 Design Basis for Protection of Light Water NPP Against Effects of Postulated Pipe Rupture 58.14 Safety and Pressure Integrity Classification Criteria for Light Water Reactors 59.51 1976 Fuel Oil Systems for Standby Diesel-Generators American National Standards institute (ANSI) A58.1 1982 Minimum Design Loads for Buildings and Other Structures B3.5 1960 American Standard Tolerance for Ball and Roller Bearings C1 1985 Specifications of General Requirements for a Quality Program C37.06 1987 AC High-Voltage Circuit Breakers Rated on a Symmetrical Current Basis - Preferred Ratings l C37.11 Power Circuit Breaker Control Requirements i Conformance with Standard Review Plan and Applicabihty of Codes and Standards - Amendment 1 1.9-37
2SAS193Rev. A SBWR standardsarety Analysis neport O Table 1.9-3 Industrial Codes and Standards Applicable to SBWR (Continued) Code or Standard Number Year Title C37.16 1988 Preferred Ratings and Related Requirements for Low Voltage AC Power Circuit Breakers and AC Power Service Protectors C37.17 1979 Trip Devices for AC and General-Purpose DC Low Voltage Power Circuit Breakers C37.20 Switchgear Assemblies and Metal-Enclosed Bus C37.50 1989 Test Procedures for Low Voltage AC Power Circuit Breakers Used in Enclosures MC11.1 1976 Quality Standard for instrument Air N5.12 1972 Protective Coatings (Paint) for Nuclear Industry N13.1 1969 Guide to Sampling Airborne Radioactive Materials in Nuclear Facilities N14.6 1986 Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More for Nuclear Materials N45.2.1 1973 Cleaning of Fluid Systems and Associated Components During Construction of Nuclear Power Plar.ts N45.2.2 1972 Packaging, Shipping, Receiving, Storage, and Handling of items for Nuclear Power Plants (During Construction Phase) N45.4 1972 Leakage-Rate Testing of Containment Structures for Nuclear Reactors N101.2 1972 Protective Coatings (Paints) for Light Water Nuclear Containment Facilities N101.4 1972 QA for Protective Coatings Applied to Nuclear Facilities N195 (See ANS 59.51) N237 (See ANS 18.1) N270 (See ANS 57.2) American Petroleum Institute (API) 620 1986 Rules for Design and Construction of Large, Welded, Low-Pressure Storage Tanks 650 1988 Welded Steel Tanks for Oil Storage I American Society of Civil Engineers (ASCE) 4-86 1986 ASCE Standard for Seismic Analysis of Safety-Related Nuclear Structures 1 S 38 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
25A5113Rsv. A SBW3 standardsareryAnsirsis neport Table 1.9-3 Industrial Codes and Standards Applicable to SBWR (Continued) : Code or Standard Number Year Title 7-88 1988 Minimum Design Loads for Buildings and Other Structures American Society of Heating, Refrigerating and Air Conditioning Engineers, Inc. (ASHRAE) 30 1978 Methods of Testing Liquid Chilling Packages 33 1978 Methods of Testing Forced Circulation Air Cooling and Air Heating Coils l l American Society of Mechanical Engineers (ASME) B30.2 1990 Overhead and Gantry Cranes (Top Running Bridge, Single or Multiple Girder, Top Running Trolley Holst) B30.9 1990 Slings j B30.10 1987 Hooks B30.11 1980 Monorail and Underhung Cranes l B30.16 1987 Overhead Hoists (Underhung) Power Piping ~ B31.1 1989 B31.3 1990 Chemical Plant and Petroleum Refinery Piping B96.1 1989 Welded Aluminum-Alloy Storage Tanks N509 1989 Nuclear Power Plant Air Cleaning Units and Components N510 1989 Testing of Nuclear Air-Cleaning Systems NOG-1 1989 Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder) l N OA-1 1989 Quality Assurance Program Requirements for Nuclear Facilities N OA-la 1989 Addenda to ANSI /ASME NOA-1 NOA-2 1989 Quality Assurance Requirements for Nuclear Power Plants OM3 1987 Requirements for Preoperational and initial Startup Vibration Test i Programs for Water-Cooled Power Plants l OM7 1986 Requirements for Thermal Expansion Testing of Nuclear Plant l Piping Systems [ September 1986 (Draft-revision 7)] Sec 11 1989 BPVC Section 11, Material Specifications Sec til 1989 BPVC Section lil, Rules for Construction of Nuclear Power Plant Components Sec V 1989 BPVC Section V, Non-Destructive Examination l ( Sec Vill 1989 BPVC Section Vill, Rules for Construction of Pressure Vessel ) 1 Conformance with Standard Review Plan and Applicabiliry of Codes and Standards - Amendment 1 1.9-39 J l
25A5113Rsv. A SBWR standantsafety Analysis neport O Table 1.9-3 industrial Codes and Standards Applicable to SBWR (Continued) Code or Standard Number Year Title SecIX 1989 BPVC Section IX, Qualification Standard for Welding and Brazing Procedures Welder, Brazers and Welding and Brazing Operators SecXI 1989 BPVC Section XI, Rules for Inservice inspection of Nuclear Power Plant Components - 1967 Thermodynamic and Transport Properties of Steam American Society for Testing and Materials (ASTM) A614 Specification for Special Requirements for Botting Material for Nuclear and Other Special Applications E185 1982 Practice for Conducting Surveillance Tests for Light-Water-Cooled Nuclear Power Reactor Vessels E380-89a 1989 Standard Practice for Use of the International System of Units (SI) (The Modemized Metric System) E399 Plane-Strain Fracture Toughness of Metallic Materials E813 Test Mr. hod for J,c, a Measure of Fracture Toughness D975 1989 Classif cation of Diesel Fuel Oils Additional (See ASME BPVC Section lil) I American Welding Society, Inc. (AWS) ; A4.2 1986 Procedures for Calibrating Magnetic Instruments to Measure the Delta Ferrite Content of Austenitic Stainless Steel Weld Metal D1.1 1990 Structural Welding Code - Steel D1.4 1979 Structural Welding Code - Reinforcing Steel D14.1 1985 Welding of industrial and Mill Cranes and other Material Handling Equipment American Water Works Association (AWWA) CMAA70 1983 SpeciScation for Electric Overht ad Traveling Cranes D100 1984 Welded Steel Tanks for Water Storage insulated Cable. Engineer Association (ICEA) P-46-426/iEEF 1982 Ampacities including Effect of Shield Losses for Angle Conductor S-135 Solid-Dielectric Power Cable 15 kV through 69 kV 1.940 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
l 25A5113Rev. A SBWR studsntsareryAnso rsesneport r k Table 1.9-3 Industrial Codes and Standards Applicable to SBWR (Continued) Code or Standard Number Year Title P-54-440/ 1987 Ampacities of Cables in Open-Top Cable Trays NEMA WC-51 S-66-524/ 1982 Cross Linked Thermosetting Polyethylene Insulated Wire and Cable NEMA Transmission and Distributor of Electrical Energy W C-7 Institute of Electrical and Electronics Engineers, Inc. (IEEE) 279 1971 Criteria for Protection Systems for NPGS 308 1980 Criteria for Class 1E Power Systems for NPGS 317 1983 Electrical Penetration Assemblies in Containment Structures for NPGS 323 1984 Qualifying Class 1E Equipment for NPGS l 334 1974 Type Tests of Continuous Duty Class 1E Motors for NPGS 338 1987 Criteria for the Periodic Surveillance Testing of NPGS Safety Systems l 344 1987 Recommended Practices for Seismic Qualifications of Class 1E Equipment for NPGS 379 1988 Application of the Single Failure Criterion to NPGS Safety Systems 382 1985 Qualification of Actuators for Power Operated Valve Assemblies with Safety-Related Functions for NPP , l 383 1974 Type Test of Class 1E Cables, Field Splices and Connections for I NPGS l 384 1981 Criteria for independence of Class 1E Equipment and Circuits 387 1984 Criteria for Diesel-Generator Units Applied as Standby Power l Supplies for NPGS 450 1987 Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations 484 1987 Practice for the installation Design and Installation of Large Lead Storage Batteries for NPGS 485 1983 Recommended Practice for Sizing Large Lead Storage Batteries for NPGS
- 535 1986 Qualification of Class 1E Lead Storage Batteries for Nuclear Power Generating Stations 603 1980 Safety Systems for Nuclear Power Generating Stations i
Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment I 1.941 i.____.
l 25A5113 Rev. & l SBWR standardSafety Analysis Report O Table 1.9-3 Industrial Codes and Standards Applicable to SBWR (Continued) Code or Standard Number Year Title . 730.1 1989 Software Quality Assurance Plans 802.1d 1990 Local Area Network Mac (Media Access Control) Bridges 802.1e 1990 Local Area Network System Load Protocol 802.3h 1989 CSMA/CD Access Method and Physical Layer Specifications 802.5 1989 Local Area Networks: Token Ring Access Method and Physical Layer Specifications 828 1990 Software Configuration Management Plans 829 1983 Software Test Documentation 944 1986 Recommended Practice for the Application and Testing of Uninterruptible Power Supplies for Power Generating Station 946 1985 Design of Safety-Related DC Auxiliary Power Systems for Nuclear Power Generating Stations 1012 1987 Software Verification and Validation Plans 1023 1988 IEEE Guide for the Application of Human Factors Engineering to Systems, Equipment and Facilities of Nuclear Generating Stations C37.04 1979 Rating Structure for AC High-Voltage Circuit Breakers Rated on a Symmetrical Current Basis C37.09 1979 Test Procedure for AC High-Voltage Circuit Breakers Rated on a Symmetrical Current Basis C37.010 1979 Application Guide for AC High-Voltage Circuit Breakers Rated on a Symmetrical Current Basis C37.013 1990 AC High-Voltage Generator Circuit Breakers Rated on a Symmetrical Current Basis C37.13 1991 Low Voltage AC Power Circuit Breakers Used in Enclosures C37.100 1981 Definitions for Power Switchgear C57.12.00 1987 General Requirements for Distribution, Power, and Regulating Transformers, Liquid-Immersed C57.12.11 1980 Guide for installation of Oil-immersed Transformers (10MVA and Larger,69-287 kV rating) j C57.12.80 1978 Terminology for Power and Distribution Transformers C57.12.90 1987 Test Code for Distribution, Power, and Regulating Transformers, Liquid-immersed O' 1.942 Conformance with Standard Review Plan and Applicabiliry of Codes and Standards - Amendment 1
I 25A5113Rev. A SBWR standardsareryAnalysis neport [
\
Table 1.9-3 Industrial Codes and Standards Applicable to SBWR (Continued) Code or Standard Number Year Title j instrument Society of America (ISA) j S7.3 1981 Quality Standard for Instrument Air S67.02 1980 Nuclear Safety-Related instrument Sensing Line Piping and Tubing Standards S 67.04 1988 Setpoints for Nuclear Safety-Related Instrumentation international Standards of Organization (ISO) 8802.3 1988 Local Area Networks - Carrier Sense Multiple Access with Collision Detecting (CSMA/CD) Access Method and Physical Layer Specifications National Electrical Manufacturers Association (NEMA) p AB-1 Molded Case Circuit Breakers k ICS 1 1983 General Standards for Industrial Control MG1 1987 Motors and Generators ICS 2 1988 Industrial Control Devices, Controllers and Assemblies National Fire Protection Association (NFPA) 10 1990 Portable Fire Extinguishers 10A 1973 Portable Fire Extinguishers - Maintenance and Use 11 1988 Low Expansion Foam and Combined Agent Systems 12 1989 Carbon Dioxide Extinguishing Systems 13 1989 Installation of Sprinkler Systems l 14 1990 Standpipe and Hose Systems 15 1990 Water Spray Fixed Systems for Fire Protection I l 20 1990 Centrifugal Fire Pumps ; 22 1987 Water Tanks for Private Fire Protection 24 1987 Private Service Mains and Their Appurtenances 26 1988 Supervision of Valves Controlling Water Supplies for Fire Protection 30 1990 Flammable and Combustible Liquids Code n 37 1990 Stationary Combustion Engines and Gas Turbines l 70 1990 National Electrical Code Conforrnance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.943 i
i 2SAS113Rev. A SBWR standardsafety Analysis Report i Table 1.9-3 Industrial Codes and Standards Applicable to SBWR (Continued) Code or Standard I Number Year Title 72 1990 Installation, Maintenance and Use of Protective Signaling Systems 72D 1986 Proprietary Protective Signaling Systems l 72E 1990 Automatic Fire Detectors
, 78 1989 Lightning Protection Code 80 1990 Fire Doors and Windows 80A 1987 Protection of Buildings from Exterior Fire Exposures 1
90A 1989 Air Conditioning and Ventilating Systems 91 1990 Blower and Exhaust Systems 101 1991 Life Safety Code 204M 1991 Smoke and Heat Venting 251 1990 Fire Tests of Building Construction and Materials 252 1990 Fire Tests of Door Assemblies 255 1990 Surface Burning Characteristics of Building Materials 321 1991 Basic Classification of Flammable and Combustible Liquids 801 1991 Facilities Handling Radioactive Materials 802 1988 Nuclear Research Reactors 803 1988 Fire Protection for Light Water Nuclear Power Plants 1961 1987 Fire Hose 1963 1985 Screw Threads and Gaskets for Fire Hose Connections Steel Structures Painting Council (SSPC) PA-1 1972 Shop, Field and Maintenance Painting PA-2 1973 Measurements of Paint Film Thickness with Magnetic Gages SP-1 1982 Solvent Cleaning SP-5 1985 White Metal Blast Cleaning S P-6 1986 Commercial Blast Cleaning SP-10 1985 Near-White Blast Cleaning Underwriters Laboratories, Inc. (UL) l UL-44 1983 Rubber-Insulated Wires and Cables UL-489 1985 Molded-Case Circuit Breakers and Circuit Breaker Enclosures 1.944 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
25A5113Rev. A SBWR stemtantsunnyAnsorsis neport Table 1.9-3 Industrial Codes and Standards Applicable to SBWR (Continued) Code or Standard Number Year Title UL-845 1987 Motor Control Contracts Others TEMA C 1978 Standards of Tubular Exchanger Manufacturers Association
- - Aluminum Construction Manual by Aluminum Association
- - Crane Manufacturers Association of America, Specification No. 70
- - Factory Mutual Approval Guide
- - Gear Classification Manual by AGMA
- - Stainless Steel Cold-Formed Structural Design Manual by AISI 1970 Standards for Steam Surface Condensers by Heat Exchange institute O
U l l %Y l Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.945
25AS113Rev. A SBWR StandardSafety Analysis Report i Table 1.9-4 Experience Information Applicable to SBWR No. Issue Date Title Comment Generic Letters 80-06 4/25/80 Clarification of NRC Requirement for Emergency Response Facilities at Each Site 80-30 12/15/80 Periodic Updating of Final Safety Analysis Reports (FSARs) COL 80-31 12/22/80 Control of Heavy Loads 81-03 2/26/81 Implementation of NUREG-0313, Rev.1 81-04 2/25/81 Emergency Procedures and Training for Station Blackout COL Events 81-07 2/3/81 Control of Heavy Loads 81-10 2/18/81 Post-TMI Requirements for the Emergency Operations Facility 81-11 2/22/81 Error in NUREG-0619 Subsection 19H.2.6 81-20 4/1/81 Safety Concerns Associated with Pipe Breaks in the BWR Scram System 81-37 12/29/81 ODYN Code Reanalysis Requirements 81-38 11/10/81 Storage of Low-Level Radioactive Wastes at Power Reactor COL Sites 82-09 4/20/82 Environmental Qualification of Safety-Related Electrical Equipment 82-21 10/6/82 Technical Specifications for Fire Protection Audits COL 82-22 10/30/82 Inconsistency between Requirements of 10 CFR 73.40(d) and Standard Technical Specifications for Performing Audits of Safeguard Contingency Plans 82-27 11/15/82 Transmittal of NUREG-0763, " Guidelines for Confirmatory in-Plant Tests of Safety-Relief Valve Discharges for BWR Plants," and NUREG-0783, " Suppression Pool Temperature Limits for BWR Containments." 82-33 12/17/82 Supplement 1 to NUREG-0737 82-39 12/22/82 Problems with the Submittals of 10CFR73.21 Safeguards COL Information Licensing Review 83-05 2/83 Safety Evaluation of " Emergency Procedure Guidelines, COL Revision 2," NEDO-24934, June 1982 83-07 2/16/83 The Nuclear Waste Policy Act of 1982 COL 83-13 3/2/83 Clarification of Surveillance Requirernents for HEPA Filters and Charcoal Absorber Units in Standard Technical Specifications on ESF Cleanup Systems i 1.946 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
e 2SAS113Rev. A l SBWR standardsafety Analysis Report
~
'(Q
\
Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issi.e Date Titie Comment 83-28 7/8/83 Required Actions Based on Generic Implications of Salem Subsection ATWS Events 19H.2.37 83-33 10/19/83 NRC Positions on Certain Requirements of Appendix R to COL 10CFR50 84-11 4/19/84 Inspections of BWR Stainless Steel Piping Subsection 19H.2.40 84-15 7/2/84 Proposed Staff Actions to improve and Maintain Diesel Generator Reliability 84>23 10/26/84 Reactor Vessel / Water Level Instrumentation in BWRs 85-01 1/9/85 Fire Protection Policy Steering Committee Report 85-06 4/16/85 QA Guidance for ATWS Equipment That is Not Safety-Related 86-02 1/23/86 Technical Resolution of Generic issue B-19 Thermal Hydraulic Stability 86-10 4/24/86 Implementation of Fire Protection Requirements 87-06 3/13/87 Periodic Verification of Leak Tight integrity of Pressure COL isolation Valves 87-09 6/4/87 Sections 3.0 and 4.0 of the Standard Technical Specifications (STS) on the Applicability of Limiting Conditions for Operations and Surveillance Requirements 88-01 1/25/88 NRC Position on IGSCC in BWR Austenitic Stainless Steel Subsections Piping 19H.2.15 & 19H.2.40 88-02 1/20/88 Integrated Safety Assessment Program 11(ISAP 11) 88-14 8/8/88 Instrument Air Supply System Problems Affecting Safety-Related Equipment Past Related Correspondence: IE Notice 87-28, Supp.1 NUREG-1275, Volume 2 88-15 9/12/88 Electric Power Systems-inadequate Control Over Design Process Past Related Correspondence: IE Notice 88-45 88-16 10/4/88 Removal of Cycle-Specific Parameter Limits from Technical Specifications 88-18 10/20/88 Plant Record Storage on Optical Disks COL Past Related Correspondence: NUREG-0800 Reg. Guide 1.28, Rev. 3 ( 1 88-20 11/23/88 Individual Plant Examination for Severe Accident Vulnerabilities-10CFR50.54(f) Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-47
25A5113Rsv. A SBWR standardsarery Analysis Report O Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment 88-20 8/29/89 Individual Plant Examination for Severe Accident Supp.1 Vulnerabilities-10CFR50.54(f) 88-20 4/4/90 Individual Plant Examination for Severe Accident Supp. 2 Vulnerabilities-10CFR50.54(f) 88-20 7/6/90 Individual Plant Examination for Severe Accident Supp. 3 Vulnerabilities-10CFR50.54(f) 88-20 6/28/91 !ndividual Plant Examination for Severe Accident S upp. 4 Vulnerabilities-10CFR50.54(f) 89-01 1/31/89 Implementation of Programmatic Controls for Radiological COL Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Off-site Dose Calculation Manual or to the Process Control Program 89-02 3/21/89 Actions to improve the Detection of Counterfeit and COL Fraudulently Marketed Products Past Related Correspondence: EPRI-NP-5622, " Guideline for the Utilization of Commercial-Grade items in Nuclear Safety-Related Applications". Bulletins 87-02 and Supplements 1 and 2,88-05 and Supplements 1 and 2,88-10 IE Notices 87-66,88-19,88-35, 88-46 and Supplements 1 and 2,88-48 and Supplement 1, 88-97 89-04 4/3/89 Guidance on Developing Acceptable Inservice Testing Program 89-06 4/12/89 Task Action Plan item 1.D.2-Safety Parameter Display System,10CFR50.54(f) 89-07 4/28/89 Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs 89-07 /21/90 Power Reactor Safeguards Contingency Planning for Supp.1 Surface Vehicle Bombs 89-08 5/2/89 Erosion / Corrosion-induced Pipe Wall Thinning 89-10 6/28/89 Safety-Related Motor-Operated Valve Testing and COL j Surveillance l 89-10 6/13/90 Results of the Public Workshop COL Supp.1 89-10 8/3/90 Availability of Program Descriptions COL Supp. 2 89-10 10/25/90 Consideration of the Results of NRC-Sponsored Tests of COL Supp. 3 MOVs 1.9-48 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 l
2SA5113 Rsv. A SBWR standardsurety Analysis neport m Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment I 89-11 6/30/89 Resolution of Generic issue 101, " Boiling Water Reactor l Water Level Redundancy" ! 89-13 7/18/89 Service Water System Problems Affecting Safety-Related Subsection l Equipment 19H.2.55 89-13 4/4/90 Service Water System Problems Affecting Safety-Related Subsection Supp.1 Equipment 19H.2.55 89-14 8/21/89 Line item improvements in Technical Specifications-Removal of the 3.25 Limit on Extending Surveillance Intervals 89-15 8/21/89 Emergency Response Data Fystem COL 89-18 9/6/89 Resolution of USI A-17, Systems Interactions 89-19 9/20/89 Request for Action Related to Resolution of Unresolved Subsection Safety issue A-47, " Safety implication of Control Systems in 19H.2.18 LWR Nuclear Power Plants," Pursuant to 10CFR50.54(f) l 89-22 10/19/89 Potential for increased Roof Loads and Plant Area Flood Subsection Runoff Depth at Licensed Nuclear Power Plants Due To 19H.2.43 f^,
'v Recent Change in Probable Maximum Precipitation Criteria Developed By The National Weather Service l
90-09 12/11/90 Alternative Requirements for Snubber Visual Inspection l Intervals and Corrective Actions 91-03 03/06/91 Reporting of Safeguards Events COL 91-04 04/02/91 Changes in Technical Specification Surveillance Intervals to l Accommodate a 24-Month Fuel Cycle 91-05 04/04/91 Licensee Commercial Grade Procurement and Dedication Programs 91-06 04/29/91 Resolution of Generic issue A-30," Adequacy of Safety-Related DC Power Supplies," Pursuant to 10CFR50.54(f) 91-10 07/08/91 Explosive Searches at Protected Area Portals COL ) 91-11 07/19/91 Resolution of Generic issues 48, "LCOs for Class 1E Tie Breakers" and 49, " interlocks and LCOs for Class 1E Tie Breakers," Pursuant to 10CFR50.54(f) 91-14 09/23/91 Emergency Telecommunications 91-16 10/03/91 Licensed Operators' and Other Nuclear Facility Personnel COL Fitness for Duty 91-17 10/17/91 Generic Safety issue 29, " Bolting Degradation or Failure in Nuclear Power Plants" !/ 92-04 8/19/92 Resolution of the issues Related to Reactor Vessel Level COL i (_/ Instrumentation in BWRs Pursuant to 10CFR50.54(f) l l~ Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-49 l 1
25A5113 Rett. A SBWR standardsarety Analysis neport l O Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment IE Bulletins l 79-02 6/21/79 Pipe Support Base Plate Designs Using Concrete Expansion Rev.1 Anchor Bolts 79-08 4/14/79 Events Relevant to BWR Identified During TMl incident 79-24 9/27/79 Frozen Lines Subsection 19H.2.33 80-01 1/11/80 ADS Valve Pneumatic Supply 80-03 2/6/80 Loss of Charcoalfrom Absorber Cells 80-06 3/13/80 ESF Reset Controls 80-08 4/7/80 Containment Lines Penetration Weids COL 80-10 5/6/80 Non-Radioactive System-Potential for Unmonitored COL Release 80-12 5/9/80 Decay Heat Removal System Operability COL 80-15 6/18/80 Possible loss of Emergency Notification System with Loss of Offsite Power 80-20 7/31/80 Westinghouse Type W-2 Switch Failures 80-21 11/6/80 Valve Yokes Supplied by Mole COL 80-22 9/11/80 Automatic Industries, Model 200-500-008 Sealed Source COL Con. 80-24 11/21/80 Prevention of Damage due to H 2O Leakage inside NUREG/ Containment CR-4524 80-25 12/19/80 Operating Problems with Target Rock SRVs at BWRs 81-01 1/27/81 Surveillance of Mechanical Snubbers 81-02 4/9/81 Failure of Gate Type Valves to Close COL , 81-02 8/19/81 Failure of Gate Type Valves to Close Against Differential COL Supp.1 Pressure 81-03 4/10/81 Flow Blockage of Cooling Water to Safety System COL 82-03 10/14/82 Stress Corrosion Cracking in Thick-Wall Large Diameter, Subsection Stainless Steel Recirculation System Piping at BWR Plants 19H.2.40 l 82-04 12/3/82 Deficiencies in Primary Containment Electrical Penetration COL Assemblies 83-02 3/4/83 Stress Corrosion Cracking in Large Diameter Stainless Steel Subsection Recirculation System Piping at BWR Plants 19H.2.40 83-06 7/22/83 Non-Conforming Materials Supplied by Tube-Line Corp. COL 84-01 2/3/84 Cracks in Boiling Water Reactor Mark l Containment Vent Header 1.9-50 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendmen: 1
25AS113Rev. A SBWR standardsafety Analysis Report 1
\.) i l
Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment 84-03 8/24/84 Refueling Cavity Water Seal 85-03 11/15/85 Motor-Operated Valve Common Mode Failures During Plant COL Transients Due to improper Switch Settings 85-03 4/27/88 Motor-Operated Valve Common Mode Failure During Plant COL Supp.1 Transients Due to improper Switch Settings l Past Related Correspondence: j IE Bulletin 85-03,IE Notice 86-29, and 87-01 87-01 7/9/87 Thinning of Pipe Walls in Nuclear Power Plants 87-02 11/6/87 Fastener Testing to Determine Conformance with COL J Applicable Material Specifications 87-02 4/22/88 Fastener Testing to Determine Conformance with COL , Supp.1 Applicable Material Specifications l Past Related Correspondence: i IE Notice 88-17 87-02 6/10/88 Fastener Testing to Determine Conformance with COL Supp.2 Applicable Material Specifications (O l l j 88-04 5/5/88 Potential Safety-Related Pump Loss v Past Related Correspondence: I lE Notice 87-59 88-07 6/15/88 Power Oscillations in Boiling Water Reactors (BWRs) Past Related Correspondence: . IE Notice 88-39 88-07 12/30/88 Power Oscillations in Boiling Water Reactors (BWRs) Supp.1 88-10 11/22/88 Nonconforming Molded-Case Circuit Breakers COL 88-10 8/3/89 Nonconforming Molded Case Circuit Breakers COL Supp.1 89-02 7/19/89 Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steet internal Preloaded Botting in Anchor Darling Model S350W Swing Check Valves or Valves of Similar Design 90-01 03/09/90 Loss of Fill-Oil in Transmitters Manufactured by Rosemount
- 90-02 03/20/90 Loss of Thermal Margin Caused by Channel Box Bow 91-01 10/18/91 Reporting Loss of Criticality Safety Controls )
l IE Information Notices ) 79-22 9/14/79 Qualifications of Control Systems COL 80-12 3/31/80 instrumentation Failure Causes PORV Opening Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-51 l
2SAS113Rev. A SBWR standardsa/ety Analysis neport Ol Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 80-21 5/16/80 Anchorage & Support of Safety-Related Electrical Equipment 80-22 5/28/80 Breakdowns in Contamination Control Programs COL 80-40 11/7/B0 Excesu,ve N2 Supply Pressure 80-42 11/24/80 Effect of Radiation on Hydraulic Snubber Fluid 81-05 3/13/81 Degraded DC Systems at Palisades COL : 81-n7 1/16/81 Potential Problem with Water Soluble Purge Dam Materials COL Used During Inert Gas Welding j 81-10 3/25/81 Inadvertent Containment Spray COL 81-20 7/13/81 Test Failures of Electrical Penetrations , 81-21 7/21/81 Potential Loss c f Direct Access to Ultimate Heat Sink COL 81-31 10/8/81 Failure of Safety injection Valves COL 81-38 12/17/81 Potential Significant Equipment Failure Resulting frorn COL Contamination of Air-Operated Systems 82-03 3/22/82 Environmental Tests of Electrical Terminal Block 82-10 3/3/82 Following Up Symptomatic Repairs COL 82-12 4/21/82 Surveillance of Hydraulic Snubbers 82-22 7/9/82 Failures in Turbine Exhaust Lines 82-23 7/16/82 Main Steam isolation Valve Leakage 82-25 7/20/82 Failures of Hiller Actuators Upon Gradual Loss of Air Pressure 82-32 8/19/82 Contamination of Reactor Coolant System by Organics COL 82-39 9/21/82 Service Degradation of Thick-Walled Stainless Steel Subsection Recircu!ation Systems at BWR Plants 19H.2.40 82-40 9/22/82 Deficiencies in Primary Containment Electrical Penetration Assemblies 82-43 11/16/82 Deficiencies in LWR Air Filtration / Vent System 82-49 12/16/82 Correction for Sample Conditions for Air & Gas Monitor COL 1 83-03 1/28/83 Calibration of Liquid LevelInstruments COL ' 83-07 3/7/83 Nonconformities with Materials Supplied by Tube Line COL Corp. 83-08 3/9/83 Component Failures Caused by Elevated DC Control Voltage l 83-17 3/31/83 Electrical Control Logic Problem Resulting in inoperable Auto Start of Emergency Diesel Generator 83-30 5/11/83 Misapplication of Generic EOP Guidelines COL 1.9-52 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
i 2SA5113Rsv. A SRWR standardsuretr^nalysis neport ( Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment 83-35 5/31/83 Fuel Movement with Control Rods Withdrawn at BWRs COL 83-44 7/1/83 Potential Damage to Redundant Safety Equipment from Backflow Through the Equipment and Floor Drain System 83-44 8/30/90 Potential Damage to Redundant Safety Equipment from Supp.1 Backflow Through the Equipment and Floor Drain System 83-50 8/1/83 Failure of Class 1E Circuit Breakers ta Close 83-51 8/5/83 Diesel Generator Events 83-62 9/26/B3 Failure of Toxic Gas Detectors 83-64 9/29/83 . Lead Shielding Attached to Safety-Related Systems COL : 83-70 10/25/83- Vibration Induced Valve Failures 83-70 3/4/85 Vibration-Induced Valve Failures j Supp.1 1 83-72 10/28/83 Environmental Qualification Testing Experience 83-75 11/3/83 Improper Control Rod Manipulation COL ! l ! l t 83-80 11/23/83 Use of Specialized Stiff" Pipe Clamps N 84-09 2/13/84 Lessons Learned from NRC Inspections of Fire Protection I Safe Shutdown Systems (10CFR50, App. R) 84-09 3/7/B4 Lessons Learned from NRC Inspections of Fire Protection Rev.1 Safe Shutdown Systems (10CFR50, App R) 84-10 2/24/84 Motor Operating Valve Torque Switches Set Below the COL Manufacturers Recommended Value ) 84-17 3/5/84 Problems with Liquid Nitrogen Cooling Components Below the Nil-Ductility Temperature 84-22 3/29/84 Deficiency in Comsip, Inc. Standard Bed Catalyst 84-23 4/5/84 Results of the NRC-Sponsored Qualification Methodology l on ASCO Solenoid Valves 84-32 4/18/84 Auxiliary Feedwater Sparger and Pipe Hanger Damage 84-35 4/23/84 BWR Post-Scram Drywell Pressurization 84-38 5/17/84 Problems With Design, Maintenance, and Operation of Offsite Power Systems ! 84-47 6/15/84 Environmental Qualification Tests of Electrical Terminal Blocks ! 84-67 8/17/84 Recent Snubber Inservice Testing With High Failure Rates COL ~ Q 84-69 8/29/84 Operation of Emergency Diesel Generators COL 84-69 2/24/86 Operation of Emergency Diesel Generators COL Supp.1 l Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-53
25A5113Rev. A SBWR standardsareer Ana!Ysis Report l Ol1 Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 84-70 9/4/84 Reliance on Water LevelInstrumentation with a Common COL Reference Leg 84-70 8/26/85 Reliance on Water LevelInstrumentation with a Common COL Supp.1 Leg 84-76 10/19/84 Loss of All AC Power 84-87 12/3/84 Piping Thermal Deflection Induced by Stratified Flow 84-93 12/17/84 Potential for Loss of Water from the Refueling Cavity 85-08 1/30/85 industry Experience on Certain Materials Used in Safety-Related Equipment 85-13 2/21/85 Consequences of Using Soluble Dams COL 85-17 4/1/85 Possible Sticking of ASCO Solenoid Valves 85-17 10/1/85 Possible Sticking of ASCO Solenoid Valves Supp.1 85-24 3/26/85 Failures of Protective Coatings in Pipes and Heat COL Exchangers 85-25 4/2/85 Consideration of Thermal Conditions in the Design and installation of Supports for Diesel Generator Exhaust Silencers 85-28 4/9/85 Partial Loss of AC Power and Diesel Generator Degradation 85-30 4/19/85 Microbiologically Induced Corrosion of Containment Service Water System 85-32 4/22/85 Recent Engine Failures of Emergency Diesel Generators 85-33 4/22/85 Undersized Nozzle-to-Shell Welded Joints in Tanks and Heat Exchangers Constructed Under the Rules of the ASME Boiler and Pressure Vessel Code 85-34 4/30/85 Heat Tracing Contributes to Corrosion Failure of Stainless COL Steel Piping 85-35 4/30/85 Failure of Air Check Valves to Seat 85-35 5/17/88 Failure of Air Check Valves to Seat Supp.1 85-47 6/18/85 Potential Effect of Line-Induced Vibration on Certain Target Rock Solenoid-Operated Valves 85-51 7/10/85 Inadvertent Loss of improper Actuation of Safety-Related COL l Equipment 85-59 7/17/85 Valve Stem Corrosion Failures 85-66 8/7/85 Discrepancies Between As Built Construction Drawings and COL Equipment Installations l 1 1.9-54 Conformance with Standsid Review Plan and Applicability of Codes and Standards - Amendment 1 \
2SAS113Rsv. A SBWR stamrantsataryAnnoysisnepart O Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment 85-76 9/19/85 Recent Water Hammer Events 85-77 9/20/85 Possible Loss of Emergency Notification System Due to COL Loss of AC Power 85-81 10/17/85 Problems Resulting in Erroneously High Reading With COL Thermoluminescent Dosimeters 85-84 10/30/85 Inadequate Inservice Testing of Main Steam isolation Valves 8b-85 10/31/B5 Systems Interaction Event Resulting in Reactor System Safety Relief Valve Opening Following a Fire-Protection Deluge System Malfunction 85-86 11/5/85 Lightning Strikes at Nuclear Power Generating Stations 85-87 11/18/85 Hazards of inerting Atmospheres COL 85-89 11/19/85 Potential Loss of Solid-State Instrumentation Following Failure or Control Room Cooling 85-90 11/19/85 Use of Sealing Compounds in an Operating Plant COL i 85-91 11/27/85 Load Sequencers for Emergency Diesel Generators COL 85-92 12/2/85 Surveys of Wastes Before Disposal From Nuclear Reactor COL ! Facilities - 85-96 12/23/85 Temporary Strainers Left installed in Pump Suction Piping COL 86-01 1/6/86 Failure of Main Feedwater Check Valves Causes Loss of Feedwater System Integrity and Water-Hammer Damage 86-03 1/14/86 Potential Deficiencies in Environmental Qualification of Limitorque Motor Valve Operator Wiring 86-09 2/3/86 Failure of Check and Stop Valves Subjected to Low Flow Conditions 86-10 2/13/86 Safety Parameter Display System Malfunctions 86-29 4/25/86 Effects of Changing Valve Motor-Operator Switch Settings COL Past Related Correspondence: IE Bulletin 85-03 86-30 4/29/86 Design Limitations of Gaseous Effluent Monitoring Systems 86-43 6/10/86 Problems with Silver Zeolite Sampling of Airborne COL Radioiodine 86-48 6/13/86 Inadequate Testing of Boron Solution Concentration in the Standby Liquid Control System 86-50 6/18/B6 Inadequate Testing to Detect Failures of Safety-Related Pneumatic Components or Systems Past Related Correspondence: IE Notices 82-25,85-35,85-84,85-94 Conformance with Standard Review Plan and Applicabihty of Codes and Standards - Amendment 1 1.9-55
25A5193Rev. A SBWR standantsafety Analysis neport O Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 86-51 6/18/86 Excessive Pneumatic Leakage in the Automatic ! Depressurization System l Past Related Correspondence: l IE Bulletins 80-01,80-25;IE Notice 85-35; ! IE Inspection Report 50-458/84-18 (8/16/84) I 86-53 6/26/86 Improper installation of Heat Shrinkable Tubing COL 86-57 7/11/86 Operating Problems With Solenoid Operated Valves at ' Nuclear Power Plants l 86-60 7/28/86 Unanalyzed Post-LOCA Release Paths l Past Related Correspondence: NUREG-0737 l 86-68 8/15/86 Stuck Control Rod 86-70 8/18/86 Potential Failure of All Emergency Diesel Generators 86-71 B/19/86 Recent Identified Problems With Limitorque Motor Operators Past Related Correspondence: IE Notice 86-03 86-76 8/20/86 Problems Noted in Control Room Emergency Ventilation Systems Past Related Correspondence: Item lli D.3.4 of NUREG-0737, Generic issue 83,IE Notice 85-89 86-83 9/16/86 Underground Pathwaysinto Protected Areas, Vital Areas, COL Material Access Areas, and Controlled Access Areas Past Related Correspondence: NUREG-0908, ANSI 3.3 86-87 10/10/86 Loss of Offsite Power Upon An Automatic Bus Transfer 86-89 10/16/86 Uncontrolled Rod Withdrawal Because of A Single Failure 86-96 11/20/86 Heat Exchanger Fouling Can Cause inadequate Operability COL of Service Water Systems Past Related Correspondence: IE Bulletin 81-03,IE Notice 81-21 86-100 12/12/86 Loss of Offsite Power to Vital Buses at Salem 2 86-104 12/16/86 Unqualified Butt Splice Connectors identified in Qualified Penetrations 86 106 12/16/86 Feedwater Line Break 86-106 2/13/87 Feedwater Line Break Supp.1 Past Related Correspondence: IE Notice 82-22, EPRI Report NP-3944,4/85 1 1.9-S6 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
i l 25A5113 Rev. S SBWR standardsafety Analysis Report O (V Table 1.9-4 Experience information Applicable to SBWR (Continued) No, issue Date Title Comment 86-106 3/18/87 Feedwater Line Break Supp.2 86-106 10/10/88 Feedwater Line Break Supp.3 86-109 12/29/86 Diaphragm Failure in Scram Outlet Valve Causing Rod COL Insertion Past Related Correspondence: IE Notice 86-08 87-06 1/30/87 Loss of Suction to Low-Pressure Service Water System COL Pumps Resulting From Loss of Siphon 87-08 2/4/87 Degraded Motor Leads in Limitorque DC Motor Operators Past Related Correspondence: (Unrelated problems involving wiring stalled in Limitorque motor actuators) IE Notices 83-72,86-03 and 86-71 87-09 2/5/87 Emergency Diesel Generator Room Cooling Deficiency Past Related Correspondence:
! IE Notice 86-50,86-51 and 86-89 87-13 2/24/87 Potential For High Radiation Fields Following Loss of Water
, From Fuel Pool l Past Related Correspondence: : ! IE Notice 84-93,IE Bulletin 84-03 87-14 3/23/87 Actuation of Fire Suppression System Causing inoperability l of Safety-Related Ventilation Equipment l Past Related Correspondence: l IE Notice 83-41,85-85,86-106 Supp. 2 87-28 6/22/87 Air Systems Problems at U.S. Light Water Reactors Past Related Correspondence: AEOD-C701 87-28 12/28/88 Air Systems Problems at U.S. Light Water Reactors Supp.1 Past Related Correspondence:
- AEOD-C701, NUREG-1275 Vol. 2
! 87 36 8/4/87 Significant Unexpected Erosion of Feedwater Lines ,
Past Related Correspondence: l IE Notice 82-22,86-106 plus Supp.1&2 IE Bulletin 87-01 l 87-43 9/8/87 Gaps in Neutron-Absorbing Material in High-Density Spent Fuel Storage Racks Past Related Correspondence: r EPRI NP-4724 I s ! ( ! 87-49 10/9/87 Deficiencies in Outside Containment Flooding Protection Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-57 l
25AS113Riv. A SBWR StamfardSafety Analysis Report O Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 87-50 10/9/87 Potential LOCA at High- and Low-Pressure Interfaces from Fire Damage 88-01 1/27/88 Safety Injection Pipe Failure 88-04 2/5/88 Inadequate Qualification and Documentation of Fire Barrier Penetration Seals Past Related Correspondence: 10CFR50 Appendix R, Appendix Ato BTP APCSB 9.5-1, NUREG-0800, ASTM E-119, BTP CMEB 9.5-1, Generic Letter 86-10 88-04 8/9/88 Inadequate Qualification and Documentation of Fire Barrier Supp.1 Penetration Seals 88-05 2/12/88 Fire in Annunciator Control Cabinets 88-12 4/12/88 Overgreasing of Electrical Motor Bearings COL Past Related Correspondence: LER 387/84-036 88-13 4/18/88 Water Hammer and Possible Piping Damage Caused by Misapplication of Kerotest Packless Metal Diaphragm Globe Valves 88-17 4/22/88 Summary of Responses to NRC Bulletin 87-01, " Thinning of Pipe Walls in Nuclear Power Plants" Past Related Correspondence: IE Bulletin 87-01 IE Notice 82-22,86-106,87-36 88-21 5/9/88 .tnadvertent Criticality Events at Oskarshamn and at U.S. COL Nuclear Power Plants 88-24 5/13/88 Failures of Air-Operated Valves Affecting Safety-Related Systems Past Related Correspondence: IE Notice 87-28 & Supp.1, NUREG-1275 88-27 5/18/88 Deficient Glectrical Terminations identified in Safety- COL Related Components 88-35 6/3/88 Inadequate Licensee Performed Vendor Audits COL Past Related Correspondence: IE Bulletin 88-05 88-37 6/14/B8 Flow Blockege of Cooling Water to Safety System COL Components Past Related Correspondence: IE Notice 81-21,86-96 IE Bulletin 81-03 0 1.9-58 Conformance with Standard Review Plan and Applicebility of Codes and Standards - Amendment 1
25A5113Rsv. A SBWR standedsafetyAnlysisReport Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 88-43 6/23/88 Solenoid Valve Problems Past Related Correspondence: lE Notices 85-17 & Supp.1,86-57 IE Circular 81-14 88-46 7/8/B8 Licensee Report of Defective Refurbished Circuit Breakers COL 88-46 7/21/88 Licensee Report of Defective Refurbished Circuit Breakers COL Supp.1 88-46 12/20/88 Licensee Report of Defective Refurbished Circuit Breakers COL Supp. 2 88-46 6/8/89 Licensee Report of Defective Refurbished Circuit Breakers COL Supp.3 88-46 9/11/89 bcensee Report of Defective Refurbished Circuit Breakers COL Supp. 4 88-51 7/21/88 Failures of Main Steam Isolation Valves 88-61 8/11/88 Control Room Habitability-Recent Reviews of Operating Experience V 88-63 8/15/88 High Radiation Hazards from Irradiated Incore Detectors and Cables COL I 88-63 6/25/91 High Radiation Hazards from Irradiated Incore Detectors COL Supp. 2 and Cables 88-65 8/18/88 Inadvertent Drainings of Spent Fuel Pools 88-70 8/29/88 Check Valve Inservice Testing Program Deficiencies Past Related Correspondence: IE Notice 86-01 Generic Letter 87-06 88-72 9/2/88 Inadequacies in the Design of DC Motor-Operated Valves 88-76 9/19/88 Recent Discovery of a Phenomenon Not Previously Considered in the Design of Secondary Containment Pressure Control Past Related Correspondence: NUREG-0800 88-77 9/22/88 Inadvertent Reactor Vessel Overfill ! 88-81 10/7/88 Failure of AMP Window indent Kynar Splices and Thomas and Betts Nylon Wire Caps During Environmental I Qualification Testing ] 88-85 10/14/88 Broken Retaining Block Studs on Anchor Darling Check Valves 88-86 10/21/88 Operating with Multiple Grounds in Direct Current ( Distribution Systems and Supplement 1 ; Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-69
l 25AS113Rsv. A SBWR StandardSafety Analysis Report O Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment 88-89 11/21/88 Degradation of Kapton ElectricalInsulation Past Related Correspondence: IE Notices 87-08,87-16 88-92 11/22/88 Potential for Spent Fuel Pool Draindown 88-92 11/29/91 Potential for Spent Fuel Pool Draindown Supp.1 88-95 12/8/88 Inadequate Procurement Requirements imposed by COL Licensees on Vendors 89-01 1/4/89 Valve Body Erosion Past Related Correspondence: IE Notice 88-17 89-04 1/17/89 Potential Problems frorn the Use of Space Heaters COL 89-07 1/25/89 Failures of Small-Diameter Tubing in Control Air, Fuel Oil, and Lube Oil Systems Which Render Emergency Diesel Generators Inoperable 89-10 1/27/89 Undetected Installation Errors in Main Steam Line Pipe COL Tunnel Differential Temperature-Sensing Elements at Boiling Water Reactors 89-11 2/2/89 Failure of DC Motor-Operated Valves to Develop Rated Torque Because of improper Cabling Sizing 89-14 2/16/89 Inadequate Dedication Process for Commercial Grade Components Which Could Lead to Common Mode Failure of a Safety System 89-16 2/16/89 Excessive Voltage Drop in DC Systems Past Related Correspondence: Generic Letter 88-15 89-17 2/22/89 Contamination and Degradation of Safety-Related Battery Cells i 1 89-20 2/24/B9 Weld Failures in a Pump of Byron-Jackson Design j 89-21 2/27/B9 Changes in Performance Characteristics of Molded-Case Circuit Breakers 89-26 3/7/89 Instrument Air Supply to Safety-Related Equipment Past Related Correspondence: Generic Letter 88-14 89-30 3/15/89 High Temperature Environments at Nuclear Power Plants 89-30 11/1/90 High Temperature Environments at Nuclear Power Plants Supp.1 89-36 4/4/89 Excessive Temperatures in Emergency Core Cooling System Piping Located Outside Containment 1.9-60 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
_ _ _ __ _ ~ _._ - _ _ . . _ _ _ _ _ _ _ _ _ _ _ __ 25A5113Rev. A ; SBWR standedsdety Analysis Reput L Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment 89-37 4/4/89 Proposed Amendments to 40CFR61, Air Emission Standards for Radionuclides 89-39 4/5/89 List of Parties Excluded from Federal Procurement of Non- COL Procurement Programs j 89-52 6/8/89 Potential Fire Damper Operational Problems ) 89-61 8/30/89 Failure of Borg-Warner Gate Valves to Close Against Differential Pressure 89-63 9/5/89 Possible Submergence of Electrical Circuits Located Above COL the Flood Level Because of Water Intrusion and Lack of 4 Drainage 89-64 9/7/89 Electrical Bus 8ar Failures COL 89-66 9/11/89 Qualification Life of Solenoid Valves 89-68 9/25/89 Evaluation of Instrument Setpoints During COL 89-69 9/29/89 Loss of Thermal Margin Caused by Channel Box Bow COL < Modifications Possible Indications of Misrepresented Vendor Products COL i 89-70 10/11/89 89-70 4/26/90 Possible Indications of Misrepresented Vendor Products COL Supp.1 l 89-72 10/24/89 Failure of Licensed Senior Operators to Classify COL Emergency Events Properly 89-73 11/1/89 Potential Overpressurization of Low Pressure Systems COL i 89-76 11/21/89 Biofouling Agent: Zebra Mussel 89-77 11/21/89 Debris in Containment Emergency Sumps and incorrect i Screen Configurations 89-80 12/1/89 Potential for Water Hammer, Thermal Stratification, and . Steam Binding in High-Pressure Coolant Injection Piping { 89-81 12/6/89 Inadequate Control of Temporary Modifications to Safety- COL Related Systems 89-83 12/11/89 Sustained Degraded Voltage on the Offsite Electrical Grid COL , and Loss of Other Generating Stations as a Result of a Plant Trip , ! 89-87 12/19/89 Disabling of Emergency Diesel Generators by Their Neutral Ground-Fault Protection Circuitry 89-88 12/16/89 Recent NRC-Sponsored Testing of Motor-Operated Valves 90-02 01/22/90 Potential Degradation of Secondary Containment
\, 90-08 02/01/90 KR-85 Hazards From Decayed Fuel Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.941 I
2545113Rsv. A SBWR standardSafety Analysis Report O Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 90-13 03/05/90 Importance of Review and Analysis of Safeguards Event COL Logs 90-20 03/22/90 Personnel Injuries Resulting Form Improper Operation of COL Radwaste incinerators 90-21 03/22/90 Potential Failure of Motor-Operated Butterfly Valves to COL Operate Because Valve Seat Friction was Underestimated 90-22 03/23/90 Unanticipated Equipment Actuation Following Restoration COL ni Power to Rosemount Transmitter Trip Units j 90-25 04/16/90 Loss of Vital AC Power With Subsequent Reactor Coolant COL ; System Heatup 90-25 03/11/90 Loss of Vital AC Power With Subsequent Reactor Coolant COL Supp.1 System Heatup 90-26 04/24/90 Inadequate Flow of Essential Service Water to Room COL Coolers and Heat Exchangers for Engineered Safety-Feature Systems 90-30 05/01/90 Ultrasonic Inspection Techniques for Dissimilar Metal Welds 90-33 05/09/90 Sources of Unexpected Occupational Radiation Exposure COL at Spent Fuel Storage Pools 90-39 06/01/90 Recent Problems with Service Water Systems COL 90-40 06/05/90 Results of NRC-Sponsored Testing of Motor-Operated COL Valves 90-42 06/19/90 Failure of Electrical Power Equipment Due to Solar Magnetic Disturbances 90-47 07/27/90 Unplanned Radiation Exposures to Personnel Extremities COL Due to improper Handling of Potentially Highly Radioactive Sources 90-50 08/08/90 Minimization of Methane Gas in Plant Systems and COL Radwaste Shipping Containers 90-53 08/16/90 Potential Failures of Auxiliary Steam Piping and the Possible Effects on the Operabirrty of Vital Equipment 90-54 08/28/90 Summary of Requalification Program Deficiencies COL 90-63 10/03/90 Management Attention To The Establishment And COL Maintenance of a Nuclear Criticality Safety Program j 90-67 10/29/90 Potential Security Equipment Weaknesses 90-69 10/31/90 Adequacy of Emergency and Essential Lighting 90-72 11/28/90 Testing of Parallel Disc Gate Valves in Europe 90-74 12/04/90 Information on Precursors to Severe Accidents { 1.9-62 Conforrnance with Standard Review Plan and Applicability of Codes and Standards - Arnendment 1
l 2SAS113Rsv. A - \ SBWR staatsrdseery Aamorsis neport l(% i Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 90-78 12/18/90 Previously Unidentified Release Path From Boiling Water Reactor Control Re>d Hydraulic Units ! 90-81 12/24/90 Fitness For Duty COL l 90-82 12/31/90 Requirements For Use of Nuclear Regulatory Commission- COL (NRC) Approved Transport Packages For Shipment of Type A Quantities of Radioactive Material 91-04 01/28/91 Reactor Scram Following Control Rod Withdrawal ! Associated With Low Power Turbine Testing 91-06 01/31/91 Lock-up of Emergency Diesel Generator and Load ' ! Sequencer Control Circuits Preventing Restart of Tripped ! Emergency Diesel Generator 91-12 02/15/91 Potential Loss of Net Positive Suction Head (NPSH) of l Standby Liquid Control System Pumps 91-13 03/04/91 Inadequate Testing of Emergency Diesel Generators ; (EDGs) 91-14 03/05/91 Recent Safety-Related incidents at Large Irradiators ! l > (. - 91-17 03/11/91 Fire Safety of Temporary Installation of Services COL 91-18 03/12/91 High-Energy Piping Failures Caused by Wall Thinning 91-18 12/18/91 High-Energy Piping Failures Caused by Wall Thinning Supp.1 1 91-22 03/19/91 Four Plant Outage Events involving Loss of AC Power or Coolant Spills j 91-23 03/26/91 Accident Radiation Overexposuret. y Personnel Due to COL ; Industrial Radiography Accessory Equipment Malfunctions 91-29 04/15/91 Deficiencies identified During Electrical Distribution , System Functionalinspections l 91-33 05/31/91 Reactor Safety Information for States During Exercises and COL Emergencies 91-34 06/03/91 Potential Problems in identifying Causes of Emergency Diesel Generator Malfunctions 91-37 06/10/91 Compressed Gas Cylinder Missile HeaMe COL 91-38 06/13/91 Thermal Stratification in Feedwater System Piping 91-40 06/19/91 Contamination of Nonradioactive System and Resulting COL ) Possibility for Unrnonitored, Uncontrolled Release to the Environment C 91-41 06/27/91 Potential Problems with the Use of Freeze Seals COL
~
Y l l Conformance with Standard Review Plan and Applicability of Codes .- Standards- Amendment 1 1.943
25A5913 Rett. A SBWR stamtardsaretYAnalYsisReport O' Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 91-42 07/27/91 Plant Outage Events involving Poor Coordination Between COL Operations and Maintenance Personnel During Valve Testing and Manipulations 91-46 07/18/91 Degradation of Emergency Diesel Generator Fuel Oil COL Delivery Systems 91-47 08/06/91 Failure of Thermo-Lag Fire Barrier Material to Pass Fire Endurance Test 91-49 08/15/91 Enforcement of Safety Requirements for Radiographers COL 91-50 08/20/91 A Review of Water Hammer Events After 1985 91-51 08/20/91 Inadequate Fuse Control Programs COL 91-57 09/19/91 Operational Experience on Bus Transfers 91-58 09/20/91 Dependency of Offset Disc Butterfly Valve's Operation of Orientation With Respect to Flow 91-59 09/23/91 Froblems With Access Authorization Programs COL 91-60 *1/01/91 Reissuance of information Notice 91-60: False Alarms of COL Alarm Ratemeters Because of Radio Frequency Interference 91-61 09/30/91 Preliminary Results of Validation Testing of Motor-Operated Valve Diagnostic Equipment 91-63 10/03/91 Natural Gas Hazards at Fort St. Vrain Nuclear Generating COL Station 91-64 10/09/91 Site Area Emergency Resulting From a Loss of Non-Class 1E Uninterruptable Power Supplies 91-65 10/17/91 Emergency Access to Low-Level Radioactive Waste COL Disposal Facilities 91-66 10/18/91 (1) Erroneous Data in " Nuclear Safety Guide, TID-7016, Revision 2," (NUREG/CR-0095, ORNL/NUREG/ CSD-6 (1978) and (2) Thermal Scattering Data Limitation in the Cross-Section Sets Provided With the Keno and Scale Codes COL 91-68 10/28/91 Careful Planning Significantly Reduces the Potential Adverse impacts of Loss of Offsite Power Events During j Shutdown l 91-72 11/19/91 Issuance of a Revision to the EPA Manual of Protective ! l Action Guides and Protective .htions for Nuclear Incidents IE Circulars i l 80-03 3/6/80 Protection from Toxic Gas Hazards COL l 80-05 4/1/80 Emergency D/G Lube Oil COL l 1.9-64 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
25A5113Rev. A SBWR standardsareryAnrysisneport O ; Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment 80-08 4/18/80 RPS Response Time 1 1 80-09 4/28/80 Problems with Plant internal Communications Systems COL j 80-10 4/29/80 Failure to Maintain Environmental Qualification of COL Equipment 80-11 5/13/80 Emergency Diesel Generator Lube Oil Cooler Failures COL 80-14 6/24/80 Radioactive Contamination of Demin Water System COL 80-18 8/22/80 10CFR50.59, Safety Evaluation for Changes to Radioactive COL Waste Treatment Systems 81-03 3/2/81 Inoperable Seismic Monitoring Instrument COL 81-05 3/31/81 Self-Aligning Rod End Bushing for Pipe Supports COL 81-07 5/14/81 Control of Radioactivity Contaminated Material COL 81-08 5/29/81 Foundation Materials COL 81-09 7/10/81 Containment Effluent Water 81-11 7/24/81 Inadequate Decay Heat Removal COL 81-13 9/25/81 Torque Switch Electrical Bypass Circuit COL 81-14 11/5/81 Main Steam isolation Valve Failures to Close COL i NUREGs : i 75/014 1095 Reactor Safety Study, An Assessment of Accident Risks in US Commercial Nuclear Power Plants 0138 11/76 Staff Discussion of Fifteen Technical issues Listed in Subsection Attachment to November 3,1976 Memo from Director, 19H.2.11 NRR to NRR Staff 0193 10/77 FRANTIC-A Computer Code for Time Dependent Subsection Unavailability Analysis 19H.2.24 0313 7/80 Technical Report on Material Selection and Processing Subsection Rev1 Guidelines for BWR Coolant Pressure Boundary Piping 19H.2.15 0313 1/88 Technical Report on Material Selection and Processing Subsection Rev. 2 Guidelines for BWR Coolant Pressure Boundary Piping 19H.2.40
,0371 10/78 Task Action Plans for Generic Activities Category A Appendix 19H l 0460 3/80 ATWS for LWRs Subsection 19H.2.5 0471 6/78 Generic Task Problem
Description:
Category B, C & D Tasks Subsection 19H.2.24 \ 0484 9/78 Methodology for Combining Dynarnic Responses Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.9-65
1 1 25A5113Rxv. A SBWR standardsaretyAnarrsisReport O Table 1.9-4 Experience information Applicable to SBWR (Continued) No. Issue Date Title Comment 0562 6/79 Fuel Rod Failure as a Consequence of Departure from Nucleate Boiling or Dryout 0578 9/80 Performance Testing of BWR and PWR Relief and Safety Valves 0588 12/79 Interim Staff Position On Environmental Qualification of Subsections Safety-Related Electrical Equipment 19H.2.8 & 19H.2.28 0606 Latest Unresolved Safety issues Summary Subsection 19H.2.8 0612 7/80 Control of Heavy Loads at Nuclear Power Plants; Subsection Resolution of Generic Technical Activity A-36 19H.2.12 0619 4/80 BWR Feedwater Nozzle and Control Rod Drive Return Line Subsection Nozzle Cracking 19H.2.6 0626 1/80 Generic Evaluation of Feedwater Transients and Small Break LOCA in GE-Designed Operating Plants and Near-Term Operating License Applications 0654 1/80 Criteria for Preparation and Evaluation of Radiological Emergency Response 0660 5/80 NRC Action Plan Developed as a Result of the TMI-2 Accident 0661 8/82 Safety Evaluation Report-Mark l Containment Long-Term Subsection Supp.1 Program-Resolution of Generic Technical Activity A-7 19H.2.3 0710 6/81 Licensing Requirements for Pending Applications for Rev.1 Construction Permits and Manufacturing License. 0737 12/82 Clarification of TMi Action Plan Requirements S upp.1 0744 10/82 Resolution of the Task A-11 Reactor Vessel Materials Rev.1 Toughness Safety issue 0763 5/81 Guidelines for Confirmatory Inplant Tests of SRV Subsection Discharges for BWR Plants 19H.2.13 0783 11/81 Suppression Pool Temperature Limits for BWR Subsection Containments 19H.2.13 0802 10/82 SRV Quencher Loads: Evaluation for BWR Mark ll and til Subsection ; Containments 19H.2.13 0803 8/81 Generic Safety Evaluation Report Regarding integrity of Subsection BWR Scram System Piping 19H.2.32 l I 0808 8/81 Mark 11 Containrr.ent Program Load Evaluation and Sub-n . ion Acceptance Criteria 19P 4 1.9-66 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
. ~ _ - _ ,
25AS113Rev. A SBWR staminedsatetr Anstrsis Report
%)
Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 0813 9/81 Draft Environmental Statement Related to the Operation of Calloway Plant, Unit No.1 . 0927, 3/84 Evaluation of Water Hammer Occurrences in Nuclear Subsection Rev.1 Power Plants 19H.2.2 0933, 3/88 A Prioritization of Generic Safety issues Appendix Supp. 7 19H 0977 3/83 NRC Fact-Finding Task Force Report on th' e ABVS Events at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25,1983 1000 4/83 Generic Implications of ATWS Events at the Salem Nuclear Subsection Power Plant 19H.2.37 1048 7/86 Safety Evaluation Report Relating to the Operation of Hope , Supp. 6 Creek Generating Station l l 1061, 8/84 Report of the USNRC Piping Review Committee Subsection Vol.1 19H.2.40 1061 11/84 Evaluation of Potential for Pipe Breaks, Report of the l Vol. 3 USNRC Piping Review Committee l 1150 6/89 Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, Vol.1 and 2 1174 5/89 Evaluation of Systems interactions in Nuclear Power Plants App.19H 1212 6/86 Status of Maintenance in the U.S. Nuclear Power Industry 1985, Vol.1 and 2 1216 8/86 Safety Evaluation PP2 Related to Operability and Reliability of Emergency Diesel Generators i 1217 4/88 Evaluation of Safety implications of Control Systems in Subsection LWR Nuclear Power Plants-Technical Findings Related to 19H.2.18 , USI A-47
- 1218 4/88 Regulatory Analysis for Resolution of USl A-47 Subsection l 19H.2.18 1229 8/89 Regulatory Analysis for Resolution of USI A-17 1233 9/89 Regulatory Analysis for USl A-40 Subsection 19H.2.14 1273 4/88 -
Containment integrity Check-Technical Finds Regulatory Analysis 1275 2/91 Volume 6, Operating Experience Feedback Report-Solenoid Operated Valve Problems 1289 11/88 Regulatory and Backfit Analysis: Unresolved Safety issue Subsection A-45, Shutdown Decay Heat Removal Requirements 19H.2.17 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.947 l
25A5113Rzv. A SBWR standantsarety Anatrsis Report Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment 1296 2/88 Peer Review of High Level Nuclear Waste 1339 6/90 Resolution of Generic Safety issue 29: Bolting Degradation or Failure in Nuclear Power Plants 1341 5/89 Regulatory Analysis for Resolution of Generic issue 115, Enhancement 1353 4/89 Regulatory Analysis for the Resolution of Generic Issue 82, "Beyond Design Basis Accidents in Spent Fuel Pools" 1370 9/89 Resolution of USl A-48 Subsection 19H.2.19 1465 6/92 Accident Source Terms for Light-Water Nuclear Power Plants (Draft Report for Comment) CR-1161 5/80 Recommended Revisions to USNRC-Seismic Design Subsection Criteria 19H.2.14 CR-1278 8/83 Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications CR-1677 8/85 Piping Benchmark Problems Dynamic Analysis Independent Support Motion Response Spectrum Method CR-1740 7/84 Data Summary of LER of Selected Instrument and Control Rev.1 Components of US Commercial Nuclear Power Plants CR-1924 4/81 FRANTIC ll-A Computer Code for Time Dependent Subsection Unavailability Analysis 19H.2.24 CR-2728 1/83 Interim Reliability Evaluation Program (IREP) Procedure Guide CR-2815 1/84 Probabilistic Safety Analysis Procedures Guide CR-3922 1/85 Survey and Evaluation of Systern Interaction Events and Sources, Vol.1 and 2 CR-4261 3/86 Assessment of Systems interactions in Nuclear Power Plants CR-4262 5/85 Effects of Control System Failures on Transients, Accidents at a GE BWR, Vol.1 and 2 CR-4387 12/85 Effects of Control System Failures on Transient and Accidents and Core-Melt Frequencies at a GE BWR CR-4470 5/86 Survey and Evaluation of Vital Instrumentation and Control Power Supply Events CR-5055 5/BB Atmospheric Diffusion for Control Room Habitability Assessment CR-5088 1/89 Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire Risk, including Previously Unaddressed Issues 1.9-68 Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1
E i 25A5113 Rev. A SBWR sesmtantsarsty Analysis napart O Table 1.9-4 Experience Information Applicable to SBWR (Continued) No. Issue Date Title Comment CR-5112 3/89 Evolution of Boiling Water Reactor Water-Level Sensing Line Break and Single Failure CR-5210 8/88 Technical Findings Document for GI 51: Improving the Subsection Reliability of Open-Cycle Service-Water Systems 19H.2.34 CR-5230 4/B9 Shutdown Decay Heat Removal Analysis: Plant Case Studies and Special issues 1 1 CR-5347 6/89 Recommendations for Resolution of Public Comments on Subsection USl A-40 19H.2.14 CR-5458 12/89 Value-Impact Assess for Candidate Operating Procedure l Upgrade Program ! CR-4674 84/89 Precursors to Potential Severe Core Damage Accidents: Series I O o o Conformance with Standard Review Plan and Applicability of Codes and Standards - Amendment 1 1.945V70
25A5113 Rev. A SBWR smkrdsdnyAndysis Repet Chapter 2 ( Table of Contents Table of Contents.... ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........................i List of Tables........ ... ... .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ... ...... iii 2.0 Site Characteristics ... .. . . ...... . ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... .... .... 2.0-1 2.1 Geography and Demograph'y ... ....... ...... ....... ....... ... ... . . . . ..... ...... 2.0 1 2.1.1 Site and location Description . .... ... ... .. ...... ... ... ...... .. .. ... .... ... .. ... ... 2.0-1 2.1.2 Exclusion Area Authority and Control..... ... ..... .... .. .. ...-....................2.0-1 2.1.3 Population Distribution . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . . ..... . . .. . 2.0-1 t 2.2 Nearby Industrial, Transportation, and Military Facilities .... ..... ... . . .. 2.0-1 2.2.1-2.2.2 Identification of Potential Hazards in Site Vicinity.... ... .. . ....... ... 2.0-1 2.2.3 Evaluation of Potential Accidents......... .... ........ .... ........... ...... 2.0 1 2.3 M e teo rol ogy..... .. .. .... . . . ... ... . .... ...... . .. . .. . . ...- ..... .. . . ....... . . . . .... . ... ..... .. .. . .. ...... .. 2.0-1 2.3.1 Regional Climatology . . ................ . ...... .. ... ... ............... ... .. . .... . .. .. ... .. . 2.0-1 2.3.2 Iocal Meteorol ogy ...... ... ... . ..... ....... ................ .. ....... ..................... ............. 2.0-2 2.3.3 On. site Meteorological Measurements Program.. .. .. .. ..... .... ... .... . ... .. ..... 2.0-2 2.3.4 Short-Term Diffusion Estimates .. . ...... ... . ....... ... .. .......................2.0-2 2.3.5 Ieng-Term Diffusion Estimates ............ ........... . ..... .... .. . ...... ................... 2.0-2
/m \
2.4 Hydrologic Engineering .... .... .. = .............................2.0-3 ( 2.4.1 Hydrologic D escription .. . ....................... ...... ........... . . . ... ........... .. .... .... 2.0-3 2.4.2 Floods .. ..... ............ ...... . ..................................................................2.0-3 2.4.3 Probable Maximum Flood on Strcams and Riven .............. . . ....... .. .. .... ...... 2.0-3 2.4.4 Poten tial Dam Fail ures ...... .. . . ........................ ................ ......... .... ... ....... .... ... 2.0-3 2.4.5 Probable Mavimum Surge and Seiche Flooding........ .......... .. ..... ........ ......... 2.0-3 2.4.6 Probable Marimum Tsunami Floodmg ............ .............. ................ ..... ......... 2.0 3 2.4.7 IceEffects...............................................................................................2.0-3 2.4.8 Cooling Water Canals and Reservoirs . ............ .... ......... .. ...... .. ...... .. ........... 2.0-3 2.4.9 Channel D ivers i ons .. .. .. .. .... . .............. ... ............. . ... ... .. ... ....... . . .. . ..... . ... 2.0-3 2.4.10 Flooding Protection Requiremen ts .. ... ... ............ ............. . .. .... . ... ........... .... 2.0-3 2.4.11 Cooling Water Supply .. . . ................. ...... .. . .............. ...... ................ 2.0 4 2.4.12 Accidental Releases of Liquid Effluents in Surface Waters... .. . ....... ........ ...... 2.0 4 2.4.13 Groun dwater .. ... .. .. . ... . . ... . . ... . .. . ... . .... ..... . . . .. ... . . . ... .... .. .... - - .. ...... . 2.0-4 2.4.14 Technical Specification and Emergency Operation Requirements........ ...... 2.0-4 3 Geology, Seismology, and Geotechnical Engineering.... .... ... ....... . . . . . .. ... . .. ... ...... 2.0-4 2.5.1 Basic Geologic and Seismic Information ..... . .. . ... .... .. . ... .......... 2.0-4 2.5.2 Vibratory Ground Motion..... . . ...................... ... .. . ............. 2.0-4 2.5.3 Surface Faulting.. . ....... . .. . ..... . . . . . . . . . . . . . . . - . . . ....... ... 2.0 4 2.5.4 Stability of Subsurface Materials and Foundations . . .. . ... .. . ... .............. 2.0-4 2.5.5 Stability of Slopes... ..... .. ........ .................. ....... ......... ........... .. 2.0-5 2.5.6 Embankments and Dams .. . .. . ... . . . . . . . . . . . . . . . . .... . ...... ... . . ... 2.0-5 ( ,/ 2.6 Requirements for Determination of Site Acceptability . ......... .. .. .. ... .. ... ............. 2.0-5 2.6.1 Design Basis Events... . ... .... ......... .......... .................. . . ... .. .. 2.0-5 2.6.2 Severe Accidents... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ... .. ..... ..... 2.0-5 Chapter 2 Table of Contents 2.0.i 4
4 i 2SA5113 Rxv. A SBWR StandantSafety Analysis Repon
- Table of Contents (Continued) f 2.7 COL License Information .... . . . . . . . . . . . . . . . . . . . . .... . . .. 2.0-6 2.7.1 Site Characteristics . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 2.0-6 2.7.2 Diffusion Estimates. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . 2.M 2.7.3 Dam Failures..... . . .. .. . . . . . . . . . . . . . . . . . . . . .. . . .. 2.0-6
. 2.7.4 Seismic Design Parameters. .. . . . . . . . . . . ... . . . . . . . . . . . .. .. 2.0-6 2.7.5 MACCS Computer Code Calculations . . . . .. .. .... .. . . . . .. 2.0-7 j 2.7.6 Site Acceptability - Design Basis Events.... .. . . . . . . . . . .. . . . . . 2.0-7 3 2.7.7 Site Acceptability - Severe Accidents ..... .... .. ..... . . . . . . . .. . .2.07 ] 2A Input to the MACCS Computer Code ... .. . . . . . . . . . . . . . . - . . 2A-1 b O O 2.0-il Chapter 2 Table of Contents
i l i 2fA5113 Rav. A r ( SBWR sarsw a s w w naport i I I l List of Tables Table 2.61 Dos e-Rela ted Goals .. . ...... . .... .... .... ... . .... . ..... .... .. .. .. .. ........ .. ....... ... . ... 2.0 G Table 2.6-2 MACCS Data Input Listing . ... ... ...... .. ... ...........................................2.08 Table 2A-1 Atm os Mod ule Input .... .. ........ .......... .. .. . .... ...................................2A-2 Table 2A-2 Early Module Input ......... .. ............. ............. ... ....................................2A-13 Table 2A-3 Chronic Module Inpu t ... ..... ... .... .. . .. . . . .. . ...... . .. ........ .. . ......... .... ... 2A-24 Table 2A-4 Site D ata I n p u t.. . .. . .. .... . . ..... . .... . .. ... . . . . . . . . . .. ... . .. .... .. . . ... . . . . . .. .. .... ..... .. . ... . .. 2A-3 8 ; i l l v i I E l i I L I h I i r a l lI h F I r
'l 6
f List of Tables y,g.;;;py
a.a . -s - w - & s A. .a___ l .I O I 4 l I J l O i l l l t O
25A5113 Rzv. A SBWR saadard sakry Analysis neport l /^\ V 2.0 Site Characteristics The site characteristics information will be provided in the combined operating license j (COL) applicant's Safety Analysis Report (SAR) in accordance with 10CFR52.79. (See l Subsection 2.7.1 for COL applicant license information requirements.) Sections 2.1 through 2.5 of this chapter, which has the same format as Chapter 2 of NUREG4800 Standard Review Plan (SRP), define the limits imposed on the SRP Section II acceptance criteria by (1) the envelope of site-related parameters that th: Simplified Boiling Water Reactor (SBWR) plant is designed to accommodate, and (2) the assumptions, both implicit and explicit, related to site characteristics employed in the evaluation of the SBWR design. 2.1 Geography and Demography 2.1.1 Site and Location Description None. 2.1.2 Exclusion Area Authority and Control (n %]
\ None.
2.1.3 Population Distribution None. 2.2 Nearby industrial, Transportation, and Military Facilities 2.2.1-2.2.2 Identification of Potential Hazards in Site Vicinity None. 2.2.3 Evaluation of Potential Accidents None. 2.3 Meteorology 2.3.1 Regional Climatology The basic speed of extreme winds used for design of structures is 49.2m/s (110 mph),a_L an elevation of 10m (33 ft) above crade, and it has a recurrence interval of 50 years. The j following importance factors, as defined in ANSI A58.1, are used for scaling wind forces for types of stnictures: g (1) Safety-Related Structures 1.11 (2) Non-Safety-Related Structures 1.00 ! l Geography and Der:wgraphy- Amendment 1 2.0-1 l l
2SA5113 Rev. A SBWR saadantsafety Analysis Report l The maximum design ambient temperature corresponding to a one percent I exceedance value is 37.8 C (100 F) dry bulb with a coincident wet bulb temperature of ) 25 C (77 F) and 26.7 C (80 F) for non-coincident wet bulb. The minimum design I temperature corresponding to a one percent exceedance value is -23.3 C (-10 F). The zero percent exceedance dry bulb temperature is 46.1 C (115 F) with a coincident wet bulb temperature of 26.7 C (80 F) and 27.2 C (81*F) for non<oincident wet bulb. The minimum temperature for this exceedance value is -40 C (-40*F). The maximum rainfall rate for roof design is 0.49m/h (19.4 in./h), which is based on the probable maximum precipitation (PMP) for one hour over one square mile with a ratio of 5 minutes to one hour PMP of 0.32, as found in National Weather Source Publication HMR No. 52. The maximum snow load for roof design is 2394 Pa (50 lb/ sq ft). The 10,000,000-year tornado has a mnimum wind speed of IS4m/s (300 mi/h), a translational velocity of 26.8m/s (60 mph), and a radius of 45.7m (150 ft). The maximum atmospheric pressure differentialis 13.8 kPa (2 psi) and the rate of pressure change is 4.85 kPa/s (1.2 psi /s). The missile spectra is per Spectra I of Standard Review Plan 3.5.1.4. Missile velocity is 35% of the maximum horizontal wind speed with an altitude of 9.1m (30 ft) above grade for large soft and rigid missiles. Small rigid missiles are postulated at all elevations. 2.3.2 Local Meteorology COL applicants will provide local meteoroloev for NRC review. l 2.3.3 On-site Meteorological Measurements Program COL applicants will provide the oneite meteorolocical measurements procram. 2.3.4 Short-Term Diffusion Estimates Short-term chffusion estimates are given in Chapter 15. The COL applicant will provide short-term diffusion estimates in accordance with Regulatory Guide 1.145 for comparison to dose values given in Chapter 15. They must be shown to result in doses less than stipulated in 10CFR100 and in the applicable portions of SRP Sections 11 and
- 15. (See Subsection 2.7.2 for COL license information requirements.)
2.3.5 Long-Term Diffusion Estimates Ieng-term diffusion estimates are given in Chapter 12. The COL applicant will provide long-term diffusion estimates in accordance with Regulatory Guide 1.112 for comparison to Chapter 12 values. (See Subsection 2.7.2 for COLlicense information requirements.) 2.62 Meteorology- Amendment 1
i j 25A5113 P.ov. A l SBWR saadant sarry Andysis neport 2.4 Hydrologic Engineering 2.4.1 Hydrologic Description The maximum ground water level is 0.61m (2 ft) below grade. I 2.4.2 Floods None. 2.4.3 Probable Maximum Flood on Streams and Rivers The probable manmum flood (PMF), as defined in ANSI /ANS 2.8,is 0.3m (1 ft) below grade or less. 2.4.4 Potential Dam Failures Itwill be demonstrated by the COL applicant that seismically induced failure ofexisting and potential upstream or downstream water control structures will not exceed flooding 0.3m (1 ft) below grade. (See Subsection 2.7.3 for COL license information requirements.) 2.4.5 Probable Maximum Surge and Seiche Flooding The probable maximum surge and flooding level is 0.3m (1 ft) below grade. 2.4.6 Probable Maximum Tsuna'mi Flooding The PMF presented in Subsection 2.4.3 applies to probable maximum tsunami l flooding. 2.4.7 Ice Effects , None. , 2.4.8 Cooling Water Canals and Reservoirs - None. 2.4.9 Channel Diversions ; None. l l 2.4.10 Flooding Protection Requirements The plant is located above the flood and groundwater levels described in ( Subsections 2.4.3,2.4.6, and 2.4.13; hence, there are no floodmg protection requirements. Hydrologic Engineering - Amendment 1 2.0-3
25AS113 Rev. A SBWR standantsareryAnalysis Report O 2.4.11 Cooling Water Supply None. 2.4.12 Accidental Releases of Liquid Effluents in Surface Waters None. 2.4.13 Groundwater The maximum ground water levelis 0.61m (2 ft) below grade. 2.4.14 Technical Specification and Emergency Operation Requirements None. 2.5 Geology, Seismology, and Geotechnical Engineering 2.5.1 Basic Geologic and Seismic Information None. 2.5.2 Vibratory Ground Motion The safe shutdown earthquake (SSE) peak ground acceleration (PGA) is 0.30g in the free field at plant grade elevation. This value envelopes all present U.S. nuclear sites, except those on the California coastline. The SSE design response spectra is per Regulatory Guide 1.60 and it is enveloped by the SSE time history. The operating basis earthquake (OBE) is not an SBWR design requirement. Consistent with the NRC's draft Appendix S to 10CFR50, the design requirements associated with the OBE, when the level of the OBE ground motion is chosen to be one-third of the SSE ground motion, are satisfied without performing explicit response or design analyses. (See Subsection 2.7.4 for COL applicant license information requirements.) 2.5.3 Surface Faulting None. 2.5.4 Stability of Subsurface Materials and Foundations The minimum static bearing capacity of the soil is 718 kPa (15 ksf), and the minimum shear wave velocity is 300m/s (1000 fps). The minimum shear wave velocity is at low strams after the soil property uncertainties have been applied. These values of bearing capacity and shear wave velocity are used to assure wide application of a standard mat-type foundation design. The design must be evaluated parametrically against ranges of possible soil properties to verify wide application. There is no liquefaction potential resulting from an SSE. See Appendix 3A for additional information. (See Subsection 2.7.4 for COL applicant license information requirements.) 2.04 Geology, Seismology. and Gootechnical Engineering - Amendment 1
I I I 25A5113 RN. A l SBWR standanisatirtyAnalysis neport l l i - 0 lV l 1 l 2.5.5 Stability of Slopes l l Noner ! l 1 l l Stability of slooes is site characteristic information and will be prosided by the COL applicant (see Subsection 2.7.1h 2.5.6 Embankments and Dams None, Embankments and dams are site characteristic information and will be prosided by the COL applicant (see Subsection 2.7.1). 2.6 Requirements for Determination of Site Acceptability
- This section provides the requirements for the determination of SBWR site i acceptability. Acceptabilityis required from the standpoint of both design bases events l and severe accidents.
. 2.6.1 Design Basis Events k For design bases events, the site is acceptable if all of the site characteristics fall within the envelope of the SBWR Standard Plant site design parameters prosided in Sections 2.3,2.4, and 2.5. For cases where a characteristic exceeds its envelope, it will be necessary for the COL applicant to submit analyses to demonstrate that the overall set ofsite characteristics does not exceed the capability of the design. (See Subsection 2.7.6 l for COL applicant license information requirements.)
2.6.2 Severe Accidents The SBWR probabilistic risk assessment (PRA) results are calculated for an average or typical site, as outlined in Appendix 19E. Although these results form a good basis for assessing the general SBWR capability to satisfy off site dose-related goals, they do not i form a basis for concluding that the SBWR would meet dose-related goals at a specific I site whose characteristics cannot be defined at the point of SBWR certification. Consistent with the certification concept that all key technical issues be resolved before certification, itis appropriate to define the process for determining future site l acceptability. This process is defined below in terms of (1) acceptance, (2) data input, ! and (3) analysis. l Acceptance Criteria Site acceptability for severe accidents will be based upon a calculation using the MACCS ,()m\ computer code for determination of SBWR site acceptability. The results of such a
\ calculation will be compared to the goals ofAppendix 19E as shown in Table 2.6-L The site will be deemed acceptable if the results fall within the given goals.
Requirements for Determination of Site Acceptability- Amendment 1 2.0-5 l
I 2SAS113 R:v, A SBWR standant safety Analysis neport O Data input The input to MACCS computer code will be a combination of SBWR and site l parameters. The MACCS code input is divided into specific areas. The items defined as l SBWRin Table 2.6-2 will be used as input with the specific data listed in Appendix 2A, j Input to MACCS Computer Code for Determination of SBWR Site Acceptability. The l items defined as GENERAL are also provided in Appendix 2A. The items defined as l UTILTIY are to be supplied by the licensing utility as specified in the MACCS manual and are site specific. l The basis reference case assumes no evacuation or radiation shielding (Subgroup ! Evacuation) for risk and dose calculations. However,if the results of such an evaluation l for a specific site are unacceptable, site 4pecific evacuation and shielding parameters ! may be substituted in lieu of the reference values in the Subgroup Evacuanon. I Analysis l The analysis for evaluation of a specific site will be accomplished with the MACCS computer code. Basic input and code characteristics are described in Appendix 2A. See Subsection 2.7.7 for COL applicant license information requirements. 2.7 COL License Information 2.7.1 Site Characteristics The COL applicant will provide the site characteristics information in their SAR in accordance with 10CFR52.79. 2.7.2 Diffusion Estimates The COL applicant will provide short-term and long-term ddfusion estimates in accordance with Subsections 2.3.4 and 2.3.5, respectively. 2.7.3 Dam Failures The COL applicant will demonstrate that seismically induced failure of existing and j potential upstream or downstream water control structures will not exceed flooding ! 0.3m (1 ft) below grade (see Subsection 2.4.4). l 2.7.4 Seismic Design Parameters To confirm the seismic design adequacy, the COL applicant will demonstrate that the sitempecific conditions specified in Appendix SA are satisfied. In meeting these conditions, compliance with the SBWR site envelope conditions in Section 2.5 for soil properties and seismology is also established. 2.0-6 COL License information - Amendment 1
25AS113 Rev. A SBWR sarndardsarety Analysis neport 2.7.5 MACCS Computer Code Calculations Compliance with acceptance criteria, data input, and analysis of Subsection 2.6.2 for determining SBWR site-specific acceptability for severe accidents will be demonstrated. 2.7.6 Site Acceptability-Design Basis Events i The COL applicant will submit analyses to demonstrate that the overall set of site characteristics do not exceed the capability of the design (see Subsection 2.6.1). 2.7.7 Site Acceptability-Severe Accidents The COL applicant will apply the acceptance criteria, data input, and analysis outlined in Subsection 2.6.2 to demonstrate severe accident site acceptability. i l I l
/%
<t l l l l
)
1 l I n ! COL License information - Amendment 1 2.07 l
25AS113 Rev. A SBWR standantSafety Analysis Report O Table 2.6-1 Dose-Related Goals Individual Risk <3.9 x 10-7 (0.1% of normal risk) Societal Risk <1.7 x 104 (0.1% of normal risk) Probability of 25 Rem Whole Body Dose at 0.8 km (0.5 mi) <10-6 per year Table 2.6-2 MACCS Data input Listing MACCS Parameter Group Defined By Purpose l 1. Spatial SBWR Site radial mesh
- 2. Site SBWR Meteorological selection
- 3. Economic General Not used; required to run code
- 4. Population Utility Population description
- 5. Topography Utility Topography description
- 6. Isotope SBWR Reactor core inventory
- 7. Leakage SBWR Release parameters
- 8. Dispersion SBWR Building parameters
- 9. Evacuation SBWR Evacuation modeling
- 10. Acute General Health physics
- 11. Latent General Health physics
- 12. Chronic General Health physics
- 13. File 20 Not used Same data an 4 and 5
- 14. File 21 General Health pnydcs
- 15. cite 27 Utility Meteorology data O
2.S8 COL License information - Amendment 1
4 < 1 l 1 > 25A5113 Rev. A l l > l l SBWR sondardsatireyAmarr sisneport i , l l Chapter 3 l 'V Table of Contents l Table of Con tents .. ... .... ........ .. . ......... ....... ... ... . ....... ... .. .. .. ...... .... .... ........ ... .. 3. 0-i Lis t of Tabl es .. . . . .. .... .. . . . . .... ... .. . . . . . .... . ... ... ... ... ... ... . . .. . .... . . . .. . . .. .... . .... . ...... ... . .. 3.0-v - List of Figures ... .... .. . ..... . ..........................................................~.3.0-xv 3.0 Design of Structures, Components, Equipment, and Systems ..... . ........... .. ................ .. 3.1-1 3.1 Conformance with NRC General Design Criteria.. .. ....... ..... . ... .... . .......... 3.1-1 3.1.1 Group I - Overall Requirements . .......... .......... .. .. ... .. .. ... ... .... . . .... . .. 3.1-1 3.1.2 Group H-Protection by Multiple Fission Product Barriers....... .... . .. 3.1-6 3.1.3 Group III- Protection and Reactivity Control Systems . ..... .... ........ ..... . 3.1-20 3.1.4 Group IV - Fluid Systems .. .. .. .... ................... ... . .. ....... .. ................ .. . . 3.1-34 3.1.5 Group V - Reactor Con tainment................ . ......... .... .... .. .... .......... ....... .... 3.1-50 3.1.6 Group VI- Fuel and Radioactivity Control . . ........ . .... ... .. ....... .......... 3.1-56 l 3.1.7 COL License Information .......... .............. .. .. ... ... .. . . .......... ...... ... ... 3.1-61 l 3.1.8 References .. . ............................................................................3.1-62 l 3.2 hification of Structures, Systems, and Components ... .... . .. .... ..... . ......... .......... .. 3.2-1 3.2.1 Safety Classification ..... ............. ..... ...... ..........................................3.2-1 3.2.2 System Quality Group Classification ..... .......... .......-...... ... ..... ........ ... . .. ..-.. 3.2-1 3.2.3 Seismic Ciassification... . . ... ..... .. . .... ........... .... .... .. .............. . . ..... . .......... . 3.2-3 C 3.2.4 3.2.5 COL License Information ........ ..................... .... -.. ........... ...... . . . .. ........ .. 3.2-4 Refere nc es ..... ... ... .... . ...... ... .... ..... .. .. .. . ...... . . .... .. . . .. .... ... ... .. .... .. . . .... ... . .... 3.2-4 l 3.3 Wind and Tornado Loadin gs .... ............... . ... .......... .............. . ......... ..... ..... ....... ..... .. .. 3.3-1 3.3.1 Win d Lo a din gs .. . ... ... ..... ... ... .. . . ... ..... .. . ... ... .. .. . ... .. .... . . . .. . ... .... .... . ..... . .. .. 3.3-1 i ! 3.3.2 Tornado Ioadings ........ ...... ... ..... ............................. ....... . ........ . ................ . 3.3-2 3.3.3 COL License Information ... ......... ............ ...... ...... .... ................. ............... . 3.3-2 3.3.4 Refer e n ces .... .... ... ..... . ........ . .... . .... . .. . ... . .. .... .. . ............ .. . ..... . ... ... .. . . ... . 3. 3 3 3.4 Water Level (Flood) Design ... ... ... .. ........ .. ..... ................... ..... ...... .. ........ . . ... . . .. . 3.4-1 3.4.1 Flood Pro tection ..... .. ........ ..... .............. ....... .. .......... . .......... . ........ ........ 3. 4-1 l 3.4.2 Analytical and Test Procedures ... ... ...... .. ..... ... .......... ............ ............. ... . . 3.4-3 l 3.4.3 COL License Information ........ ..... . ........ . .. .... . . . . . . . . . . . . . . . . ..... .. . 3.4-4 3.4.4 References...... ..............................................................................3.4-4 ,
- 3.5 Missile Protection... . ... ..............................................................................3.5-1 j l
3.5.1 Missile Selection and Description .. . . ... ......... . ...... ...... ....... ....... ........ ...... . 3.5-1 l 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles.. .. .. . ......... .....................................3.5-10 l 3.5.3 Barrier Design Procedures ....... . ... ..... ............ .... . .. ...... .... ... ........... 3.5-10 l l 3.5.4 COL License Information.. ... .... ............ .............................................3.5-11 ! 3.5.5 References .... .... .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . ... .. ... ... 3.5-12 O 3.6 Protection Agamst Dynamic Effects Associated with the Postulated Rupture of Piping.. 3.6-1 3.6.1 Postulated Piping Failures in Fluid Systems Inside and Outside of Containment.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-2 l i Table of Contents- Amendment 1 3.0-i
25AS113 Rev. A SBWR standardsafetyAnalysis neport Table of Contents (Continued) 3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping.... ..... . ......... . ... ... . .... ... . .. 3.6 8 3.6.3 Irak.Before-Break Enluation Procedures .... .. .. .. . . . . . . . . . . . . . . . . . .. 3.6-26 3.6.4 COL License Information.. . . . . . . . . . . . . . . . . . . . 3.6-29 3.6.5 References.. . .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.6-29 3.7 Seismi c D esign .. .... . . .. ..... .. . .. . . . . ... .. ... .. ..... .. . .. .. .. .. . . . . . . . . . ... 3.7-1 3.7.1 Seismic Design Parameters .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .... 3.7-2 3.7.2 Seismic System Analysis... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 3.7-3 3.7.3 Seismic Subsystem Analysis.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-18 3.7.4 Seismic Instrumen tation .. . . .... . .......... ....... . .. .. .. . .. .. . . . . ... .. . .... . 3.7-24 3.7.5 COL License Information.. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-27 3.7.6 References .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. 3.7-27 3.8 Design of Seismic Category I Structures ..... . ............... ......... . ..... . ... ... ...... 3.8-1 3.8.1 Reinforced Concrete Contamment.. .. . ...... ........ ...... .. .......... . ....... 3.8-1 3.8.2 Steel Con tainm ent.. ... . . . ... . ......... . . . .... ... .. .. .. . .. . ... .. ... . . ... . ..... ... . 3.8-9 3.8.3 Containment Internal Structures .......... .............. .... ....... .. . . . .. 3.8-9 3.8.4 Other Seismic Category I Structures .. .... ....... ...... . . . . . . . . . . 3.8-15 3.8.5 Foundation Design = . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . . .. 3.8-2 5 3.8.6 COL License Information . .... ............. ... .. . ... .... .......... ..... .. . . ........... . ... 3.8-27 3.8.7 Referen ces .. .. .... . . . .. . .. ... . ..... . . ...... .... .. .. .. .. . .. ... ... ... - . . . . . . . .3.8-28 3.9 Mechanical Systems and Components ... .. ......... .... ......... .... ............. .. .. .... ........... 3.9 1 3.9.1 Special Topics for Mechanical Components .. . ... . . .... . . . . . . . . . . . ... 3.9-1 3.9.2 Dynamic Testing and Analysis of Systems, Components, and Equipment. .. ... 3.9-5 3.9.3 ASME Code Class 1,2, and 3 Components, Component Supports, and Core Support Structures .. .. .. ....... ........ ..... .. ..... . . .. ..... ........... ... .... .. .... .. 3.9-21 3.9.4 Control Rod Drive System (CRDS) ............ ..... .... .... . ...... .. . . ....... .. . .. 3.942 3.9.5 Reactor Pressure Vessel Internals.. .... ............ . .. ... .. ... .... ....... ..... 3.9-43 3.9.6 Inservice Testing of Pumps and Valves .. . ........ .......... ....... . . .. 3.9-51 3.9.7 COL License Information....... . .. ......... ...............................................3.9-54 3.9.8 Referen ces .. . .. . .. ... . .. . . ... . . . . . .. ..... ... .... . ... . . . . ... . . . .. .. . . . .... .... ... ..... . . 3.9-5 5 3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment . . . 3.10-1 3.10.1 Seismic and Dynamic Qualification Criteria..... . . . . . . . . . . . .... 3.101 3.10.2 Methods and Procedures for Qualifying Electrical Equipment.. ... . . 3.10-2 3.10.3 Analysis or Testing of Electrical Equipment Supports........ .. .. . ............ . ... 3.10 8 3.10.4 COL License Information . . .. . . . . ..... .. .. .. .......... ....... ..... .. .... . ... .. .. 3.10-11 3.10.5 Referen ces .. . . .. .. . . . . ... . . .. . .. . . . .. . . . ... . .. .. . . .. . . . .. . . .. .. . . . . .. . ... .. 3.10-12 3.11 Environmental Quali5 cation of Safety-Related Mechanical and Electrical Equipment. 3.11-1 3.11.1 Equipment Identification . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3.11-1 3.11.2 Emironmental Conditions . . . . . . . . . . . . . .. . . . .. .. .. . 3.11-2 3.11.3 Qualification Program, Methods and Documentation . ... ... . ... .. .. .... .. . . 3.11-4 3.11.4 COL License Information.- . . . . . . . .. . ..... . . . . . .. . . . . .. . .. 3.1 1 -5 3.11.5 References ....... . .. . ... . . . . . . . . ..... . . . . . . . . . . . . . . . . . . . .. . . . . .. .. . 3.1 1-5 3.04i Table of Contents - Amendment 1
25A5113 Rev. A SBWR sunknesakryAnarrsisnepet i e t
\ Table of Contents (Continued) 3A Seismic Soil-Structure Interaction Analysis .. .. ....... ... .. ... ....... .. ... . .. ........ .... ... . . . . 3A-1 3A.1 In trod uc tion . . . .. .. .. . ..... . .. . . .. . . . . . .. . .. . . . ... .. .. ... . . ... . .. . . . . . . . .. .... ... 3A-1 SA.2 SBWR Standard Plan t Site Plan .. ........................... ........... .. ... .. .. . ............. . 3A-1 SA.3 Generic Site Conditions ... .... ...... ..... . ... .. ..... .. . ... ...... .. ..... . .. . . .. .. ... 3A-2 ,
3A.4 Input Motion and Damping Values.. .. .... ................ .......... ........... ....... . ... . . 3A-3 3A.5 Soil-Structure Interaction Analysis Method................. .. ..... ........ . .... . 3A.4 SA.6 Free-Field Site Responses Analysis ........ . ........... .... .... . . .. .... . . .. ...... .. .. 3A-10 SA.7 Soil-Structure Interaction Analysis Cases ...... . ... ........... ...... . ....... ..... .. ......... 3A-11 SA.8 Analysis Models........... .... .......................................................3A-11 SA.9 A na lysis Resul ts ......... .... ... .. ..... .. ...... ..... .................. .. .. .... .... .. ... .. .. ... 3A-16 3A.10 Site Envelope Seismic Responses.... ......... .. . ....... .. ........ .. ....... ... .. . .. ......... 3A-19 3A.11 Refere n ces . ... . ... ....... .. . . . . .. . ... .. ... . . . .. . . ... . .......... ... . ... .. .. .. .. ... . ..... ... .... 3A-20 3B Computer Programs Used . .. .. ..... .. ........................................................3B-1 3B.1 In trod uc tion ......... . . ... ... . .. . . . ....... ... ... .. . . .... ..... .... . . ... ... .. . . . .. ......... .. . . 3B-1 3B.2 Fine Motion Con trol Rod Drive . ..... ..-........... ............ ........ . ...... ............ .. . 3B-1 3B.3 Piping...........................................................................................................3B-1 SB.4 Pumps an d M otors ..... .. .. . .. .. ..... ...... .. . ..... . .. . .. . ....... ..... . ............ ..... ... .. . ... .. .. .... . 3B-4 3B.5 Heat Exchangers. ..... .. ....... .................................................3B.4 3B.6 Soil. Structure In teraction .. ........ .. ..................................... ....................... ....... 3B-5 (A SC Guidelines For LBB Application ................ .. . ... ...... .... ..........._......... ... ...... ... ..... .. ... 3C.1 3C.1 In trod uc tion . .. .. . ..... .... ..... . ... ... .. . ... .. . . ... .... .. .... ...... ...... ... . .... .. ... ... ....... . .. 3G 1 3C.2 Material Fracture Toughness Characterization ...................... ......... .. . . . .... . 3C-4 3C.3 Fracture Mechanics Methods...................... .............................. ...... .. ..... .. . .... 3C-11 3C.4 Leak Rate Calculation Methods....... ................................ ....... ........ ...... ......... 3C-17 3C.5 Leak Detection Capabilities ............................ . ............ .... .......... ............. .. 3C-22 SC.6 Guidelines for Preparation of an LBB Report................. .................... ............ 3C-23 3C.7 References............ ... ....................................................................3C-28 3D Equipment Qn2Mcation Design Environmental Conditions........ ... ...... ... ................. 3D-1 3D.1 In trod uc tion . .. .. ..... ..... . .... .. ...... . .. .. . . .... . . . . ..... . .. ...... . ...... ... ... .. ...... . . . . SD-1 3D.2 Plan t Zon es ............ .. . . .... ... . ... .. ... . .. .. . . .. . . . .............. ... ... .. . . .. . ...... . .. . .. . ... .. 3D-1 3D.3 Environm ental Con ditions............ ......... . .... ............ .... ................. .. ... ... ...... 3D-2 3E Evaluation of Results of Seismic Category I Structures.................. ... ........ ..... ............ 3E.1-1 SE.1 Obj ective And Scope . ......... ... .. .. ....... ....... . ............... .. .. ... . .................. .. .. SE.1 -1 3E.2 Con cl usions .. .. .. .. . ... .... ..., ... .. . .. . .. . .. .. .. . . . ..... . .... .. . .... . ... . ........ .... .. . .. . .. SE. 2-1 SE.3 Structural Design Criteria ... ... .... ....................................................SE.3-1 SE.4 Structural Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . .... ........... .. 3E.4-1 3E.5 Structural Design of RCCV . . . ..... . ....... ...... . .. ....... ....... .... . ......... ........... 3E.5-1 SE.6 Containment Internal Structures Design.. .... ............. ..... .... . .. ....... . ..... ... 3E.6-1 SE.7 Reactor Building Design Outside Containment.. ..... ........ .... ... .. ............3E.7-1 0
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Table of Contents - Amendment 1 3.0-liihv
. _<. ._ ~ m .. tm _ . . . _ . . .. _-.m _- 2 -
e d t N e a 1 I f i 4 i O . i I O
2SA5113 Rev. A SBWR senadant sarety Analysis neport (m N Chapter 3 List of Tables l Table 3.2-1 nawification Summary. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3.2-5 I l Table 3.2-2 Minimum Safety Designation Requirements.. ... . . .. . . . . ... . . . 3.2-34 i 1 Table 3.2-3 Quality Group Designations - Codes and Industry Standards . ... . .. . . .. 3.2-35 Table 3.3-1 Importance Factor (I) for Wind Loads. .. .. . .. ... . .... .... . . . . .. ... 3.3-4 Table 3.3-2a Design Pressure Distdbution at Vadous Heights for Safety-Related Structures (Importance Factor = 1.11)... . .. . . . . . . . .3.3-5 Table 3.3-2b Design Pressure Distribution at Various Heights for Non-Safety-Related Structures (Importance Factor = 1.00) ........ . .. . .. . 3.34
- Table 3.3-3 Factor (qzn/qzl30) to Adjust Table 2 Loads for Building Heights l Other Than 39.5 Meters (130 ft)...... . ...... . .. ......... . . . . . . . . . . .. .. . 3.3-8 Table 3.4-1 Stmctures, Penetrations and Access Openings Designed for Flood Protection.... . .. ... ... ... . . . . . . . . . . . . . . . . 3.4-5 Table 3.5-1 Requirement for the Probability of Missile Generation for SBWR r'~g Standard Plant... ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.5-13 l!
> V) Table 3.61 Safety-Related Systems, Components, and Equipment for Postulated Pipe Failures Inside Containment .... .....-. .... ... .... .. . .. . 3.430 Table 3.62 Safety-Related Systems, Components, and Equipment for Postulated Pipe Failures Outside Containment .... ..... .... ... .. ......... . . 3.6-30 Table 3.63 High Energy Piping Inside Containment .... ......... ...... ... .... . . . .. ... ... .... 3.431 Table 3.6-4 High Energy Piping Outside Containment. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . 3.431 Table 3.7-1 Damping Values for SSE Dynamic Analysis .... .. .. ............. .... . .. . ..... 3.7-28 l Table 3.8-1 Load Combinations, Load Factors and Acceptance Criteria for the Design of Reinforced Concrete Contamment.... . . . . . . . . . . . . . 3.8-29 Table 3.9-1 Plan t Events ... ....... .... .. ...... . . . . . . . . . . . . _ . . . . . . . . . . . . . ... ... 3.9-56 Table 3.9-2 Load Combinations and Acceptance Criteria for Safety-Related, ASME Code Ona 1,2 and 3 Components, Component Supports, l and Class CS Structures.... . . . . . . .. .. . . . . . . 3.9-58 Table 3.9-3 Deformation Limit for Safety Class Reactor Internal Structures Only.. .3.941 g3 Table 3.9-4 Primary Stress Limit for Safety Class Reactor Internal Structures Only . . 3.942 t l ('v) Table 3.9-5 Buciding Stability Limit for Safety Class Reactor Intemal Structures Only.. 3.9 64 Table 3.94 Fatigue Limit for Safety Class Reactor Internal Structures Only... . .. .. .3.944 LW of Tables - Amendment 1 10-v
25A5113 Rev. A SBWR standantsaretyAnarrsisneport List of Tables (Continued) Table 3.9-7 Pressure Differentials Across Reactor Vessel Internals ....... . . . . . . . . . . .3.945 Table 3.9-8 inservice Testing.... ................................................ . 3.9-66 Table SA-1 Soil Properties for VP5 Profile... ... . . . . . . . . . . .. . . . . . . . . . . . . . . . .
. ., . . . .. 3A-22 Table SA-2 Average Shear Wave Velocities in Layers . ... .... . . . . .. .. ... ... . .. . . . . . . . . . 3A-2 2 Table SA-3 Strain-Dependent Shear Modulus . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3A-23 Table SA-4 Strain-Dependent Soil Damping .. . ......... . .... . .. . .. ........ . .. . .. .... .. 3A-23 Table 3A-5 Case ID's for Site Conditions Considered ... ......... . ... . . ....... . .... ..... . 3A-24 Table SA-6 SSE Free-Field Site Response Results for All Soil Profiles (Average Properties) .......... .. ... ..... . . .. ... .... . ...... ... .... ..... .. .. .. . 3A-24 Table SA-7 Seismic SSI Cases (Reactor Building) ......... .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3A-25 Table SA-8 Effect of Soil Stiffness on Maximum Force-. .......................................3A.26 Table SA-9 Effect of Separation Between the Side Soil and Foundation Walls . .... . . .. 3A-26 Table SA-10 Effect of Con cre te Crackin g ., .. ..... . .. ... . . . ..... .. . ......-..... ... ... . ....... . 3A-27 Table SA-11 Effect of LOCA ..... ... ...... .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... 3A-27 Table SA-12 SBWR Reactor Building Summary of Enveloping Seismic Loads:
TBS Stick (Reactor Building Wall and Safety Erwelope) ............. ............. .. 3A-28 Table SA-13 SBWR Reactor Building Summary of Enveloping Seismic Loads: RCCV S ti ck . . .. . . . . . . .. .. . .. .. .. . ... .. . . . .... . . ... . . . . .. .. . . . . . .. ... .. ..... .... . ... . ... . . . . ... 3A-2 9 Table SA-14 SBWR Reactor Building Summary of Enveloping Seismic leads: VW/ PED Stick .. ..... ....... .... . .. . ... .... .. . ..... ... ..... ..... . . . .. ..... . . .. 3A-30 Table SA-15 SBWR Reactor Building Summary of Enveloping Seismic Loads: RSW Stick ... .. .... ..... ...................................................3A-31 Table SA-16 SBWR Reactor Building Summary of Enveloping Seismic Loads: l RPV Stick...... ..... . . . . . . . . - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3A-31 l l l Table SA-17 SBWR Reactor Building Summary of Enveloping Maximum l Vertical Acceleration: TBS (Reactor Building Walls and Safety Envelope.... 3A-32 Table SA-18 SBWR Reactor Building Summary of Enveloping Manmum Vertical Acceleration: RCCV.. ... . . . . . . . . . . . . . . . . . . . . . . . .. .. .. .. . . . .. 3A-Si' l Table SA-19 SBWR Reactor Building Summary of Enveloping Manmum Vertical Acceleration: VW/ Pedestal.. . .... . . ..... .. . . . .. .. 3A-33 l 3.0-vi List of Tables - Amendment 1 l l
] 1 1 i l 2SA5113 Rev. A l SBWR smaarsanyAnrysisneport l 4 s -! J List of Tables (Continued) 1 l Table SA-20 SBWR Reactor Building Summary of Enveloping Maximum i j Vertical Acceleration: RSW....... ........ .. ... .............. ......... ..... ......... .... ...... . 3A-33 l l 2 Table 3A-21 SBWR Reactor Building Summary of Enveloping Maximum : Acceleration: TBS (Reactor Building Walls and Safety Envelope) ............. . 3A-34 , s ! Table SA-22 SBWR Reactor Building Summary of Enveloping Maximum i Acceleration: RCCV...... ............................ ... . ..............................3A-34 - 1 3 Table SA-23 SBWR Reactor Building Summary of Enveloping Maximum ,
; Acceleration: VW/ Pedestal .......... m.................................. ... 3A-35 l 1
l Table SA-24 SBWR Reactor Building Summary of Enveloping Maximum J 4 Accel erati on : RSW.. ................. ................ . . . ........ .... .. ......... .. .............. 3A-35 , i l Table 3A-25 SBWR Reactor Building Summary of Enveloping Maximum , Accelera ti on : RPV. .... ...... ................. ..... ........ . . .. . ........... .. ... . ....... ... 3A-36 , Table SA-26a Description of Figures SA-28a through SA-311 . ...... ..... .. .. ........ .. ... .. 3A-37 Table SA-26b Description of Figures SA-32a through 3A-46j . .......... ..... .................. ........ 3A-39 Table SC.1 - Electrodes and Eller Metal Requirements for Carbon Steel Welds .............. 30-31 v Table 3G2 Supplier Provided Chemical Composition and Mechanical Properties Info rmati o n ... ..... ...... ....... . . ... .. . . ..... .. ...... . . ... ... ... ....... .. .. .... . ..... . ... ... ...... .. 3C-31 Table SC-3 Standard Tension Test Data at Temperature....... . .. ........... . ........ . ........ 3C-32 Table SG4 Summary of Carbon SteelJ.R Curve Tests........ ................ ...... ..... ............ 3C-32 Table SG5 Mass Flow Rate for Several fl/Dh Values ... .. .... ........... ............. . ....... ..... 30-33 Table SC 6 Stresses in the Main Steam Lines (Assumed for Example) ........................... 3C-33 Table SC-7 Critical Crack 12ngth and Instability lead Margin Evaluations for Main Steam Lines (Example) ....... ... .......................................3C-33 Table 3C.8 Data for Feedwater System Piping (Erample).. . ... .. .. .. ..... .... . . ............. 3C-34 Table 3G9 Stresses in Feedwater Lines (Assumed for Example)............... ..... ... .......... 3C-34 Table 3C.10 Critical Crack length and Instability Load Margin Evaluations for Feedwa ter Lines (Example) .. ...... . .... . ....... .. ..... ... ......... . . .... .. ..... 3C-34 Table 3D-1 Cross Reference of Plant Environmental Data and Location........ .... .. . ..... 3D-4 Table 3D-2 Thermodynamic Eravironment Conditions Inside Containment
\ Vessel for Normal Operating Conditions .. . . . . . . . . .. . ... .. .. . . . .. 3D-5 List of Tables- Amendment 1 a0-vii
I 25AS:13 Kev. A l SBWR studant sarety Anatnis aeron List of Tables (Continued) Table 3D-3 Thermodynamic Environment Conditions Inside Safety Envelope for Normal Operating Conditions .. .. . . . . . . . . . . . . . . . . . . . 3D-6 Table 3D-4 Thermodynamic Environment Conditions Inside Reactor Building j for Normal Operating Conditions .. ...............................3D-8 Table 3D-5 Thermodynamic Environment Conditions Inside Control Room Envelope for Normal Operating Conditions... .... . . . . . . . . . . . . 3D-8 Table 3D-6 Radiation Environment Conditions Inside ContainmentVessel for Normal Operating Conditions . . .. . . . . . . . . . . . . . . . .. .. . . . . .. 3D-9 l Table 3D-7 Radiation Environment Conditions Inside Safety Envelope l for Normal Operating Conditions .. ..................................3D-10 Table 3D-8 Radiation Environment Conditions Inside Reactor Building for Normal Operating Conditions .. . .. .. ............ . ......... ............ ... ... . 3D-12 Table 3D.9 Radiation Environment Conditions Inside Control Envelope l for Normal Operating Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3D-12 l Table 3D-10 Thermodynamic Environment Conditions Inside Contamment l Vessel for Acciden t Conditions..... . .. .. ..... ... . .... ..... ... .... ........ ... .. ... .. 3D-13 Table 3D-11a Thermodynamic Environment Conditions Inside Safety Envelope for Accident Conditions... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3D-14 Table 3D-11b Thermodynamic Environment Conditions Inside Safety Envelope for Acciden t Conditions... . ... . ......... ... . ..... ... ...... . ......... ... ... . ..... .. . . 3D-15 Table 3D-12 Thennodynamic Environment Conditions Inside Reactor Building for Accident Conditions... ..... .....................................................SD-16 Table 3D-13 Thermodynamic Environment Conditions Inside Control Room Envelope for Accident Conditions........ .. . ......................................3D-17 Table 3D-14 Radiation Environment Conditions Inside Containment Vessel for Accident Conditions .... ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 3D-18 Table 3D-15 Radiation Environment Conditions Inside Safety Envelope for Accident Conditions... . ... .. . . . . . . . . .. .. ... . .. .. . . .. . . . .. . 3 D-19 i Table 3D-16 Radiation Emironment Conditions Inside Reactor Building for Accident Conditions..... . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 3D-21 Table 3D-17 Radiation Environment Conditions Inside Control Room Envelope for Accident Conditions... .. ..... ... . . . . . . . . . . . . . . ... ... .. .... .. . . . 3D-21 Table 3E.3-1 Temperature loads for DBA Conditions .. ... ... .. ..... . .. ........ ........ ..... . 3E.3-9 10-viii List of Tables - Amendment 1
i l 2SAS113 RDv. A SBWR sonderdsatirty Analysis neport O List of Tables (Continued) Table 3E.3-2 SBWR - Containment Pressure Loads . . .......... .... . . . . . . . . . . . . . . . . ..3E.S11 Table 3E.3-3 Dynamic Load Factors (DLF) for Suppression Pool Hydrodynamic Loads . .. . .. . . ...... .... .. . ..... ... .. . ..... ....................... . . .. SE.312 Table 3E.3-4 Maximum Vertical Acceleration... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3E.3-13 ! Table SE.3-5 SBWR-Selected Governing Load Combinations and Acceptance Criteria for the Design of Reinforced Concrete Containment and RPV Pedestal... ... ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. ... SE.3-15 Table 3E.4-1 Results of STARDYNE Analysis for Unit (1 t/m2) Dgwell Pressure ( PDw) . ... . ....... . .. ... .. . . ..... ........ . ..... ..... . . ... .. ... .. . . . ... ... .. 3E.4-9 Table SE.4-2 Results ofSTARDYNE Analysis for Unit (1 t/m2) Suppression Cham ber Pressure (Pww) ........ ...................... . ...................... ... ... . .. ... 3E.4-22 Table 3E.4-3 Results of STARDYNE Analysis for SSE Load (SRSS) ... .. .. ..... ...... ... . . . 3E.4-35 l Table 3E.4-4 Results of STARDYNE Analysis for Long Term Thermal (Ta-111).. .... . . . 3E.4-48 Table 3E.4-5 load Combination 1-1: SIT-1. .. ....... ........... ........... ............... .... ........ . SE.4-61 i' ! Table 3E.4-6 Ioad Combination 8-3: DBA (Period-11I) .......................... ...... .... . .. . . 3E.4-127 Table 3E.4-7 Load Combination 15-3: DBA + SSE (Period-III) .......... .. . ... ........... . ... . 3E.4-191 l l Table 3E.4-8 Load Combination 15-2B: DBA + SSE (Period - H) ............ ....... .. .. . . 3E.4-256 I Table 3E.5-1 Rebar Ratios Used in the Design .. ......... . ........... ....... . ..... ........... ..... ....... SE.5-3 Table 3E.5-2a Rebar and Concrete Stresses Due to Load Combination 1-1...... ..... ... ....... 3E.5-6 l Table 3E.5-2b Rebar and Concrete Stresses Due to Load Combination 1-J ....... ..... .... ... 3E.5-7 Table 3E.5-2c Rebar and Concrete Stresses Due to Load Combination 1-1........ . . ... .... 3E.5-8 l l Table 3E.5-2d Rebar and Concrete Stresses Due to Load Combination 1-1. .......... ... .. .. 3E.5-9 l l Table SE.5-2e Rebar and Concrete Stresses Due to Load Combination 1-1... ..... .... ....... 3E.5-10 l l 4 l Table 3E.5-2f Rebar and Concrete Stresses Due to Load Combination 1-1... ... . . . . ... 3E.5-11 l l l Table SE.5-Sa Rebar and Concrete Stresses Due to Load Combination 8-3...... . . . .. 3E.5-12 l Table SE.5-Sb Rebar and Concrete Stresses Due to Load Combination 8-3.. . . . 3E.5-13
- gp) Table SE.5-Sc Rebar and Concrete Stresses Due to Load Combination 8-3.. . . ... .... . .. 3E.5-14 O Table 3E.5-3d Rebar and Concrete Stresses Due to Load Combination 8-3. .... . .. . .. 3E.5-15 l
l List of Tables- Amendment 1 3.0-ix
2SAS193 Rev. A l SBWR saadant sark ty Analysis neport l 91 i List of Tables (Continued) ; Table 3E.5-Se Rebar and Concrete Stresses Due to Load Combination 8-3.. . . . .. .. .. 3E.5-16 Table 3E.5-3f Rebar and Concrete Stresses Due to Load Combination 8-3. . ... . 3E.5-17 Table 3E.5-4a Rebar and Concrete Stresses Due to Load Combination 15-3...... . - SE.5-18 i 1 Table SE.54b Rebar and Concrete Stresses Due to Load Combination 15-3.. .......3E.519 Table SE.5-4c Rebar and Concrete Stresses Due to Load Combination 15-3 ... .. 3E.5-20 Table 3E.5-4d Rebar and Concrete Stresses Due to Load Combination 15-3.. . . . .. .... . 3E.5-21 n.t>1e 3E.5-4e Rebar and Concrete Stresses Due to Load Combination 15-3. .. .. .... ..... 3E.5-22 Table 3E.5-4f Rebar and Concrete Stresses Due to Load Combination 15-3. . .... .. 3E.5-23 Table SE.5-5a Rebar and Concrete Stresses Due to Load Combination 15-2B ... .. .. .... 3E.5-24 Table SE.5-5b Rebar and Concrete Stresses Due to Ind Combination 15-2B.. . ....... ... . 3E.5-25 Table SE.5-5c Rebar and Concrete Stresses Due to Load Combination 15-2B.. . .. ... .... 3E.5-26 Table 3E.5-5d Rebar and Concrete Stresses Due to Load Combination .... .. .. . .... .. ..... 3E.5-27 Table 3E.5-5e Rebar and Concrete Stresses Due to Load Combination 15-2B.. . ......... . 3E.5-28 Table 3E.5-5f Rebar and Concrete Stresses Due to Load Combination 15-2B.. . . ... . .. .. 3E.5-29 Table 3E.5-6a Containment Liner Plate Strains (Max) in/in ...... ... .. ... .. ............. .. .. 3E.5-30 Table 3E.5-6b Containment Liner Plate Strains (Max) in/in .... . . ...... ..... .... .... .... .... 3E.5-31 Table SE.6-1 Diaphragm Floor (D/F) Slab Structural Elements Stress Summary... .. .. ... 3E.6-3 Table 3E.6-2 Diaphragm Floor (D/F) Slab Anchorage Structural Capacity... ... ....... .... 3E.6-3 Table 3E.6-3 Vent Wall Structural Elements Stress Summary........... ........ ....... ......... . . 3E.6-4 Table 3E.64 Reactor Shield Wall (RS'W) Structural Element Stress Summary .... ...... .... 3E.6-4 Table 3E.6 5 Gravity-Driven Cooling System (GDCS) Pool Structural Elements - Stress Summarv.... .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. ... ... ... .. SE. 6-4 Table 3E.6-6 Gravity-Driven Cooling System (GDCS) Pool Anchorage Structural Capacity.. . ..... ... . . . . . . . . . . . .............................3E.6-5 Table 3E.6-7 RPV Support Bracket Structural Elements Stress Summary... . .. ... . .... . . 3E.6-5 Table 3E.71 IC Pool Main Girder Section 1 Forces and Moments for Basic loads .. .... . .. .. . . . . . . . . . . . . . . .. .... .. . . . . .. .. . . .. .. . . 3 E. 7-6 ) l 3.0-x List of Tablen - Amendment 1
d 2M5113 Rev. A SBWR senadant sannyAn@sisRepen e
/
List of Tables (Continued) Table SE.7-2 IC Pool Main Girder Section 2 Forces and Moments for Basic Loads.......... 3E.7-8 Table SE.7-3 IC Pool Main Girder Secdon 3 Forces and Moments for Basic loads .... . 3E.7-10 Table 3E.7-4 IC Pool Main Girder Section 4 Forces and Moments for Basic loads.... .. 3E.7-12 Table SF 7. i IC Pool Main Girder Section 5 Forces and Moments for Basic Loads........ 3E.7-14 , l Table SE./-6 IC Pool Main Girder Section 6 Forces and Moments for Basic loads........ 3E.7-16 Table 3E.7-7 IC Pool Secondary Girder Section 7 Forces and Moments for Basic Loa ds .. . .. .... . . .. .. . . .. .. . ... . .. ... .. .. .. . . . . . .. .. ...... . . . . .. .. .. .. . . ... . .. . . .. . .. 3E.7-18 Table 3E.7 8 IC Pool Secondary Girder Section 8 Forces and Moments for Bc uc Loads ............ .... . . ... . ........ . ..... ... . ..... . . . . . . . . . . . . . . . . . . . 'tE.7-20 Table SE.7-9 IC Pool Cross Girder Section 9 Forces and Moments for Basi c Ioads ..... . ....... ... .......... ............. ........... .......... .. ...... .... . 3E.7-22 Table 3E.7-10 IC Pool Cross Girder Section 10 Forces and Moments for Basi c Loads.... ......... . ... ... ... ....... ......... ........... ... .. .................. ..... . . 3E.7-24 Table SE.7-11 IC Pool Main Girder Section 1(1) Forces and Moments v for Load Combinations.. . .................................................................3E.7-26 Table SE.7-12 IC Pool Mam Girder, Section 2(1) Forces and Moments for load Combina tions . . ...... ..... ...... ... ....... ........ .. ........... . . ... ..... ..... ... 3E.7-27 Table SE.7-13 IC Pool Main Girder, Section 3(1) Forces and Moments for Load Combina tio ns .. ...... ........... ............. ... ....... ... _ ......... ........ . SE.7-28 Table SE.7-14 IC Pool Main Girder, Section 4(1) Forces and Moments for Ioad Com binations .. . . ... . .... ....... ............ ....... ... .... ...... ......... . ......... 3E.7-29 Table 3E.7-15 IC Pool Main Girder, Section 5(1) Forces and Moments for load Combinations ... ... ... ........ .......... .. .... .. ........ . . .. . ...... . 3E.7-30 Table SE.7-16 IC Pool Main Girder, Section 6(1) Forces and Moments for load Com bina tions .. . ... ..... . .. ...... ........... ............. .... . . ...... . .... ..... ...... 3E.7-31 Table SE.7-17 IC Pool Secondary Girder, Section 7(1) Forces and Moments for Ioad Com bin a tions . ...... . . ......... ............. . .... .......... .... ...... ...... . . .... . 3E.7-3 2 Table SE.7-18 IC Pool Secondary Girder, Section 8(1) Forces and Moments for load Combinations = . . . . . . . . . . . . . . . . . . . . . . . . . . ..... .... ... ..... . . 3E.7-33 g Table SE.7-19 IC Pool Main Girder, Section 9(1) Forces and Moments for load Combinations... . ... .......................................................3E.7-34 Ust of Tables - Amendment 1 3.0-xl
25AS193 Rev. A SBWR standans saretyAnarysis neport List of Tables (Continued) Table 3E.7-20 IC Pool Cross Girder, Section 10(1) Forces and Moments for Load Combinations ... . . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 3E.7-35 Table 3E.7-21 Rebar Stresses at Sections for Cdtical Load Combinations... . ... . SE.7-36 Table SE.7-22 Rebar and Concrete Stresses Due to Load Combination 1-1. . . . SE.7-37 Table 3E.7-23 Rebar and Concrete Stresses Due to Load Combination 8-3. . .. ... . . . .. 3E.7-38 Table 3E.7-24 Rebar and Concrete Stresses Due to Load Combination 15-3. . . ... . . 3E.7-40 Table 3E.7-25 Rebar and Concrete Stresses Due to Load Combination 15-2B... . ...... ... 3E.7-42 Table 3E.7-26 Facton of Safety for Foundation Stability....... .... ....... . .. . .. .....3E.7-14 O l l l I i O. i 3.0-xii List of Tables- Amendment 1
25AS113 Rev. A SBWR saudantsurrryAaarrsisneport Chapter 3 List of Figures Figure 3.5-1 Missile Velocity and Displacement Characteristics Resulting from Saturated Steam and Water Blowdowns (1050 psia Stagnation Pressure) .. 3.5-14 l Figure 3.6-1 ,Je t Ch aracteristics .. .. ........ ........ ... ............. ... ... . .. .......... .......... .. . . . .... 3.6-3 2 Figure 3.6-2 Typical Pipe Whip Restraint Configuration ... ... .. . . .. . . .. ... . . ... ... ... 3.6-33 Figure 3.7-1 Horizontal SSE Design Spectra ............ ....... .. . .. ... . . .. . . . .... . . 3.7-2 9 Figure 3.7-2 Vertical SSE Design Spectra....... . .. ...... ........ ...... .. .... ... . . . . ... . ... .... . . 3.7-30 Figure 3.7-3 Horizontal, HI Component Time History ..... . .... ............ ......... ..... .......... 3.7-31 Figure 3.7-4 Horizon tal, H2 Component Time History ................ .... .. ............ .. . .... .. 3.7-32 Figure 3.7-5 Vertical, Component Time History . . ........ .. .... ... .. ...... . ...... .... ... ...... 3.7-33 Mgure 3.7-6 2% Damped Response Spectra, H1 Component ......... ... . ......... .... .. .... ... 3.7-34 Mgure 3.7-7 3% Damped Response Spectra, H1 Component ................................3.7-35 lp 'I Figure 3.7-8 4% Damped Response Spectra, H1 Component .... ...... .. .. .......... ......... 3.7-36
- ,\ Figure 3.7-9 7% Damped Response Spectra, H1 Component . ... ...... . ....... . . ... .... ..... 3.7-37 ,
l Mgure 3.7-10 2% Damped Response Spectra, H2 Component ..... ... .. ............. ..... .... . ... 3.7-38 t Figure 3.7-11 3% Damped Response Spectra, H2 Component ........... .. ............ .. . .......... 3.7-39 Figure 3.7-12 4% Damped Response Spectra, H2 Component ........... .............. ........ ... ... 3.7 40 Figure 3.7-13 7% Damped Response Spectra, H2 Component ..... ............. ..... .. ..... . ... . 3.7-41 Mgure 3.7-14 2% Damped Response Spectra, Vt Component .... .. .. . . ....... . .. .... 3.7-42 l Figure 3.7-15 3% Damped Response Spectra, Vt Component .... ... ............. ... ........ ..... 3.7-43 i Figure 3.7-16 4% Damped Response Spectra, Vt Component ..... . .. .. .. .... ............ ..... 3.7-44 Figure 3.7-17 7% Damped Response Spectra, Vt Component.. .. ...... ... .. ..... ... ... .......... 3.7-45 ! Figure 3.7-18 Power Spectral Density Function, H1 Component... . . . ... .... .... .... .. . . . .. 3.7-4 6 ! Figure 3.7-19 Power Spectral Density Function, H2 Component.... . . .. .. .. . . .... .. . .. .3.7-47 l Mgure 3.8-1 Stress Strain Relations for Concrete (Lower Bound) at Elevated Temperatures.. . ......................................................................3.8-31 lD Figure 3.8-2 Stress Strain Relations for Reinforcing Steel at Elevated Temperatures ... 3.8-32 l Figure 3.9-1 Stress. Strain Curve for Blowout Restraints... . ... .. .. . .. . . . . . . . .. ... 3.9-8 7 i List of Figures - Amendment 1 3.0-xiii
25A5113 Rw. A SBWR sandantsaretyAnalysis aeport i List of Figures (Continued) O1l Figure 3.9-2 Minimum Boodable Volume . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 3.9-88 l l Figure 3.9-3 SBWR Recirculation Flow Path.. .... .. . .. . . . . . . .. .. . ... . . . ..... . . ... 3.9-8 9 i l Mgure 3.9-4 Control Rod Guide Tube.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3.9-90 l l l Figure 3.9-5 Pressure Nodes for Depressurization Analysis... ... . . .. . . . . . . . . .... . 3.9-91 ! 1 i Egure SA-1 Shear Wave Velocity Profiles Considered for SSI Analyses.. . ... . ......... ... .... 3A-42 ; 1 Figure SA-2 Range of Shear Wave Velocities for Nuclear Power Plant Sites in High l Seismic Areas .. ... ...... ... .. ... .. ... .. .... .. . . . . . . . . . . . . . . . . . . . .... 3A.43 1 i Figure SA-3 Strain Dependent Soil Properties... .... ... . . ... ... . . . . . . . . . . . ...... ..... . 3A-44 Mgure SA-4 Strain Dependent Rock Properties .... ...... ... ...... ..... . . .. .. .. 3A-45 Figure SA-5 Substructuring ofInteraction Model...... . ..... .. .. . . .. .... ... . .. 3A-4 6 Figure SA-6 TBS FEM Segmen t - El. 4.4m to 17.2m .... .. .... ......... ...... . ..... .. .. .. 3A-47 Figure SA-7 A Typical Uniform Section FEM Cantilever Model ....... ..... ... .. ...... .... .... 3A-48 i l Figure SA-8 TBS Stick - XZ Plane... ... .. .. ... . ... ... .... . ..... ... . .. . . . .. . . ... . .... . . 3A-4 9 Hgure SA-9 TBS Stick-YZ Plane .. ...................................................................3A-50 Egure SA-10 RCCV Stick - XZ Plane ..... . ...... ....... .. . .. . .... .... .. .. . ... . . . .. . ... ...... . 3A-51 Egure SA-11 RCCV Stick - YZ Plan e .. . . ............... . .. .. . . . ... ....... . .. ...... .... . ....... .... ..... 3A-5 2 Figure 3A-12 Pedestal /Ventwall Stick - XZ Plane ...... . .... . . . . . .. . . .. . ... ..... .... ...... . 3A-5 3 i Figure SA-13 Pedestal /Ventwall Stick -YZ Plane... .. ......... .... ....... .... ..... ..... 3A-54 l 1 Figure 3A-14 Reactor Shield Wall Stick- XZ and YZ Plane ... . ..... . ... ... .......... ....... .. 3A-55 l l Figure SA-15 SBWR Reactor Builomg Seismic Model.. ... . . .. ... . . .... ... ............. ..... 3A-56 Figure SA-16 RPV M odel .. . . . . ... .. ... . . .. .. . . .. . . . . .. . .. .. . .. . . . . .. . . ... .. .. . .. .. .. . ...... . .. . SA-5 7 l Figure SA-17 Stick Basement Model.. .. . . . . . . . . . . . . . . . . . . .. .. . . .. . .. . . .. .. . . 3A-5 8 I Mgure SA-18 Basemat Plate Elements . .... . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . . ... . 3A-5 9 Mgure 3A-19 Side Wall Plate Elements (Y = 32.45m).. . . . . . . . . . . . . . . . .. . .. ... . . ... 3A-60 Egure SA-20 Side Wall Plate Elements (X = 32.45m) .. ... . .. .... . . .. .. ..... .. ... ........... ... 3A-61 Figure SA-21 Side Wall Plate Elements (X = -32.45m) .. . . ... . . . . . . . . . . . . .. . 3A-62 3.0-xiv List of Figures- Amendment 1
r f 25A5113 Rev. A SBWR saadard saouryAnalysis neport C List of Figures (Continued) Mgure 3A-22 Excavated Soil Volume... ... ............ ............... ... .. ........ . . . . . . . . . . . .. ... ... 3A-63 Mgure 3A-23 Excavated Soil Elements Between El. -8.45m to -4.7m.... .. . ...... .. ... .. . ...... 3A-64 Figure 3A-24 Excavated Soil Elements Between El. -4.7m to -1.0m... .................. .. ..... 3A-65 Hgure SA-25 Excavated Soil Elements Between El. -1.0m to 3.0m... ... . .. ..... . . ... . .. ... . 3A-66 l l Figure 3A-26 Excavated Soil Elements Between El. 3.0m to 6.5m....... .. ... .. ... . .... .. . .. 3A-67 Figure 3A-27 Excavated Soil Elements Between El. 6.5m to 10.0m ...... .. .. . ... .. 3A4 Figure SA-28a Soil Stiffness Effect - Node 1101 X ............ .... ... ...... .. .. .... . ... . .... . . . .. . 34-69 l Egure SA-28b Soil Stiffness Effect - Node 1821 X .......... . .......... .... .... . . .. . .. . ... ... 3A-69 Figure SA-28c Soil StifYn ess Effect - Node 2703 X .......... ........... ..... ..... .............. ... ...... 3A-70 i Figure SA-28d Soil Stiffness Effect - Node 5021 X ....... ................................3A-70 ; t Figure SA-28e Soil Stiffness Effect - Node 1101 Y............ . ............. ...... .. . . .. ....... .. . ..... 3A-71 O
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Figure SA-28f Soil Stiffness Effect - Node 1821 Y... ..... ............... ... .. ... .. .. - 3A-71 ! Mgure 3A-28g Soil Stiffness Effeet - Node 2703 Y........ ...... .................................3A-72 ! Hgure SA-28h Soil Stiffness Effect - Node 5021 Y............ ..... .. .... ........ .. ...... . .. .. .. SA-72 7 Mgure 3A-28i Soil Stiffness Effect - Node 1101 Z............ ......... .. ........... ....................... . 3A-73 Mgure SA-28j Soil Stiffness Effect - Node 1821 Z........ . .. .... .. ........ .... ............ . . .. . ...... 3A-73 Figure SA-28k Soil Stiffness Effect - Node 2703 Z...... .... ........... ........... ..... .. ... 3A-74 i Figure SA-281 Soil Stiffness Effect - Node 5021 Z.. . ...... .. . .. ......... ...... ..... ... .. . ........ 3A-74 I Egure SA-28m Soil Stiffness Effect - Node 101 XZ..... ... ... .. ............. .. ..... ... . .. .. .. SA-75 Figure 3A-28n Soil Stiffness Effect - Node 161 YZ........... . . . .. .. ........ . .. .... .... ... ........ 3A-75 Hgure 3A-29a Effect of Side Soil Separation - Node 1101 X .... ............. ........ ... . .. ...... 3A-76 Hgure SA-29b Effect of Side Soil Separation - Node 1821 X . . . . . . . . .. . .. ........ . .... 3A-76 Hgure SA-29c Effect of Side Soil Separation - Node 2703 X . . .. ..... .... . . .. . . ... ..... .. 3A-77 Figure 3A-29d Effect of Side Soil Separation - Node 5021 X ... . . . .. ... .... . . . ... ... ... 3A-77
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Egure 3A-29e Efrect of Side Soil Separation - Node 1101 Y..... ........ ... . . .... .. 3A-78 b Mgure SA-29f Effect of Side Soil Separation - Node 1821 Y ... .. ........ . . .... ... . . . . . .. 3A-78 List of Figures - Amendment 1 3.0-xv
25A5113 Rzv. A SBWR standardsarety Analysis neport List of Figures (Continued) Figure 3A-29g Mect of Side Soil Separation - Node 2703 Y. ... ... . . . . . . . . . . .. ... .. 3A-79 , i I Mgure 3A-29h Mect of Side Soil Separation - Node 5021 Y... . . . . . . . . . . . . . . .. . . . . 3A-79 l Egure SA-29i Mect of Side Soil Separation - Node 1101 Z.. . .. .. ... ... . . .. .. ... ... 3A-80 Mgure SA-29j Mect of Side Soil Separation - Node 1821 Z.. .. . .. .. . . . . . . . . . . . . . . . 3A-80 Egure SA-29k Mect of Side Soi1 Separation - Node 2703 Z... . . . . . . . . . .. 3A-81 Egure SA-291 Mect of Side Soil Separation - Node 5021 Z.. . . . . . . . . . . . . . . .... 3A-81 Egure SA-30a Mect of Concrete Cracking - Node 1101 X .. ... ...... . . . . . . . .. . 3A-82 Egure SA-30b Mect of Concrete Cracking- Node 1821 X . .. . . . . . .. .......... . 3A-82 Egure SA-30c Mect of Concrete Cracking- Node 2703 X. .. .... . . . ... .. . . ... . .. . .. . . . . . . 3A-8 3 Figure SA-30d Mect of Concrete Cracking- Node 5021 X . .. . ..... .. . . . . .... . . .. . .. . 3A-8 3 Figure 3A-30e Mect of Concrete Cracking- Node 1101 Y.. .. . .. . .. ... . . . . .. . .. . . . . . 3A-8 4 Egure SA-30f Mect of Concrete Cracking - Node 1821 Y .... . ... . ... ... ..... . . .. .. ....... . 3A-84 i Figure SA-30g Mect of Concrete Cracking- Node 2703 Y. . . . .. .... . . .... .. .. .. . 3A-85 Figure SA-30h Mect of Concrete Cracking - Node 5021 Y.. . ..... ... .. .... . . . . .... .. . . 3A.85 Figure SA-31a Mect of LOCA - Node 1101 X . . .. . ..... .. ... . ... . ............ .... .. ....... 3A-86 Mgure SA-31b Mect of LOCA - Node 1821 X ... . .. .. ... ..... .. ..... ......... .. . -.. .. . ..... ... 3A4 6 Hgure 3A-31c Mect of LOCA- Node 2703 X.. .. . ......................................3A-87 Figure SA-31d Mect of LOCA - Node 5021 X ........ ... ...... .. ... ... . ... .. .. . .. .. .. ... .. .. . 3A-8 7 Figure 3A-31e Mect of LOCA - Node 1101 Y.. ... . .... . .... .. ... . . ... ..... . . . .. .. . . . 3A-8 8 Figure SA-31f Mect of LOCA- Node 1821 Y .. . . . . . . . . . . . . . . . . . . . . . . . . . ... .. . .... . ..... . 3A48 Mgure 3A-31g Mect of LOCA - Node 2703 Y..... . . .. ... . . . ........ .. . ..... . . .. . . ... . 3A-89 Figure SA-31h Mect of LOCA - Node 5021 Y........ . ... .. .. . . . ... .... .. . . . .. ..... . 3A-89 Egure 3A.31i Mect of LOCA- Node 1101 Z.. . . . .. . . . . . . . . . . . . . . . ...... . 3A-90 Egure 3A-31j Mect of LOCA- Node 1821 Z.. .. . . . . . . . . ... . .. . . 3A-90 Hgure 3A-31k Mect of LOCA - Node 2703 Z -....... . . .. . .. ...... .... . . . . .. ........ . . . 3A-91 Egure SA-311 Mect of LOCA- Node 5021 Z ... . . . .. . . . . . . . . . . . . . . . ... ..... ... 3A-91 10-xvi List of Figures - Amendment 1
l l l 25A5113 Rsv. A SBWR sondent sakryAutrsis nepon C
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List of Figures (Continued) Mgure SA-32a Env. of Nodes 1101,101, I l l,161 and 2101...... ... ..... .......... ..... . . .. . . ... . 3A-92 Mgure SA-32b Env. of Nodes 1201, 301, 311 and 361. ................ . . ...... ................ . .. . ....... 3A-92 Mgure 3A-32c Env. of Nodes 1301, 471, 472 and 4 73 . . ............. .... ................. ... ... 3A-93 Mgure SA-32d Env. of Nodes 1401, 671, 672 and 673 ..... ......... ...... .................... .. .......... . .. 3A-93 I Mgure SA-32e Env. of Nodes 1501,1502,1503,1505,1507,1508 and 1509 ... ... . .... ...... 3A-94 Mgure 3A-32f Env. of Nodes 1601,1602,1603,1605,1606,1607,1608 and 1609 - = 3A-94 Mgure SA-32g Env. of Nodes 1701,1703,1705,1706,1707,1708 and 1709...... ....... ... . ... 3A-95 Mgure SA-32h Env. of Nodes 1801,1807,1808 and 1809 ....... ................. ..................... . . 3A-95 Mgure SA-32i Env. of Nodes 1811,1817,1818 and 1819 .. ......... ........... ....... .... ........ .. .. 3A-96 l l Mgure 3A-32j Env. of Nodes 1821,1827,1828 and 1829 .... ... ... ....... ....... ......-............ ... 3A-96 l l Mgure 3A-33a Env. of Nodes 2201, 2207, 2208 and 2209 ............ ........... .......... ........ .. ... 3A-97 Mgure 3A-33b Env. of Nodes 2301, 2307, 2308 an d 2309 ............ ....................... . ... ..... . 3A-97 Mgure 3A-33c Env. of Nodes 2401, 2402, 2403, 2407, 2408 and 2409....... ................. ..... .. 3A-98 Mgure SA-33d Env. of Nodes 2501,2502,2503,2505,2506,2507,2508 and 2509...... .. . 3A-98 Mgure SA-33e Env. of Nodes 2601,2602,2603,2605,2606,2607,2608 and 2609 .......... . 3A-99 Mgure SA-33f Env. of Nodes 2701, 2702, 2703, 2707, 2708 and 2709........ ........... .............. 3A-99 I Mgure SA-34a Node 3177 Gnly ....... .......... ...... ... ... ....... ...... .. ... . .. ........... ......... . .. .. 3A-100 Mgure 3A-34b Env. of Nodes 3201, 3205, 3206 an d 3207 .......... .................... .. ... 3A-100 Mgure SA-34c Env. of Nodes 3301, 3307, 3308 an d 3309 . ... .... ............ ....... .................. 3A-101 Mgure SA-34d Env. of Nodes 3401, 3407, 3408 an d 3409 .. ... .... .............. ......... . .... ... . 3A-101 Mgure SA-34e Env. of Nodes 3501, 3507, 3508 and 3509. .. .. .. . .... .. . ............ . 3A-102 ! Mgure SA-35a Env. of Nodes 4301, 4307, 4308 an d 4309 ...... ........... ..... . ............ ........ .. 3A-102 i i
- Mgure SA-35b Env. of Nodes 4401, 4407, 4408 and 4409 ...... .... ... ..... ..... .......... ...... ... 3A-103 i
Figure SA-35c Env. of Nodes 4501, 4507, 4508 and 4509 ............ .. ........ .. ... ...... .......... .. 3A-103 , Mgure 3A-35d Env. of Nodes 4511, 4517, 4518 and 4519 .. . .... . . . .... . .. . .... ...... ....... 3A-104 Mgure SA-36a Node 5001 Only..... ... .. . ..... . .. . . . . . ... ....... .. .. ..... .. .. . . . .......... ... .... .. 3A-104 Ust of Figures - Amendment 1 3.0-xvii
25AS993 %v A SBWR standard safetyAnalysis Report List of Figures (Continued) l l Figure SA-36b Env. of Nodes 5013,5101,5102 and 5103. . . . . . . . . . .... . .. . 3A-105 Figure SA-36c Env. of Nodes 5015,5106,5107 and 5108. .. . . . . . . . . . . . . . . . . . . 3A-105 l Figure 3A-36d Env. of Nodes 5017,5111,5112 and 5113. . .. . . . ... . .. ... . . . . 3A-106 l Mgure SA-36e Env. of Nodes 5021,5116,5117 and 5118. .. .. ... . .. .. . ... 3A-106 Mgure SA-36f Node 5023 Only....... . . ........... .. ....... . . . . . . . . . . . . . . . . . . . . .... 3A-107 I Figure SA-36g Env. of Nodes 5026,5121,5122 and 5123.. ... . . . . . . . . . . . . . 3A-107 Figure 3A-36h Env. of Nodes 5032,5126,5127 and 5128 . . . ..... . . . . . . . . . . . . . . . . . ...... 3A-108 Mgure 3A-36i Node 5057 Only...... ..... ......... .... .. . . . . . . . . . ... . 3A-108 Figure 3A-36j Env. of Nodes 5060, 5131, 5132 and 5133 . .... . .. . . . .. .... . . .. ... .. ....... 3A-109 Mgure SA-36k Env. of Nodes 5064, 5136, 5137 and 5138 .. ... . . . . . . . . . . . . . . . . . . . . . . . . ... 3A-109 Figure 3A-361 Env. of Nodes 5607, 5141, 5142 and 5143 . .... . .. .... .. . .......... .. .... . . 3A-110 Egure 3A-37a Env. of Nodes 1101,101, Ill,161 and 2101. .... ... .. .... .. ... 3A-110 Egure SA-37b Env. of Nodes 1201,301,311 and 361 .... .. . . . . . . . . . . . . . 3A-111 Mgure SA-37c Env. of Nodes 1301, 471, 472 and 473 .. .. ... . . . ..... .. ... ... ... . 3A-111 Mgure 3A-37d Env. of Nodes 1401, 671, 672 and 673 ...... . . ... ... ............ . .... . .......... .. 3A-112 i Mgure SA-37e Env. of Nodes 1501,1502,1503,1505,1507,1508 and 1509. .... .... . .. 3A-112 1 l Figure SA-37f Env. of Nodes 1601,1602,1603,1605,1606,1607,1608 and 1909..... ... . 3A-113 Mgure SA-37g Env. of Nodes 1701,1703,1705,1706,1707,1708 and 1709 = ... .... . 3A-113 Mgure SA-37h Env. of Nodes 1801,1807,1808 and 1809 ... .....................................3A-114 Mgure SA-37i Env. of Nodes 1811,1817,1818 and 1819.. . . ..... .. ..... . 3A-114 Figure SA-37j Env. of Nodes 1821,1827,1828 and 1829.. . . . . . ...... .. ..... . .. . ... ... 3A-1 15 Figure 3A-38a Env. of Nodes 2201,2207,2208 and 2209. .... .. . . . . . .. .. ... ... . . . .. . . .. 3A-1 15 Figure 3A-38b Env. of Nodes 2301,2307,2308 and 2309. . . . . .. .. . ... ... 3A-116 Figure SA-38c Env. of Nodes 2401,2402,2403,2407,2408 and 2409. .. .. . ... . . .. .... . 3A-116 Figure 3A-38d Env. of Nodes 2501,2502,2503,2505,2506,2507,2508 and 2509. . . . 3A-117 Mgure SA-38e Env. of Nodes 2601,2602,2603,2305,2606,2607,2608 and 2609 . .... 3A-117 3.0-xviii List of Figures- Amendment 1
I 25A5113 Rcv. A SBWR standardsakryAnarrsisnapart n v List of Figures (Continued) Egure SA-38f Env. of Nodes 2701, 2702, 2703, 2707, 2708 and 2709.. . ...... ... ...... ... ...... 3A-118 Mgure SA-39a Nod e 317 7 Only.. . .. ..... ..... ... . . .. . ., . . . . . .. .... . . . .. . . .. .. . .. . . .. . . .. . .. ... .. .. ... .. 3A-118 Hgure 3A-39b Env. of Nodes 3201, 3205, 3206 and 3207 ........ ... ... . .. .. ..... ...... ... .. .... 3A-119 Figure 3A-39c Env. of Nodes 3301, 3307, 3308 and 3309 .. . . .. . . .. . . . ...._...... .......... . 3A-119 , f Egure SA-39d Env. of Nodes 3401,3407,3408 and 3409........ . . . . . . .. ..... .... .. 3A-120 Hgure SA-39e Env. of Nodes 3501, 3507, 3508 and 3509 ... . ..... ... ... ..... ... ......... .. . ..... 3A-120 Mgure 3A40a Env. of Nodes 4301,4307,4308 and 4309..... .. .... ..... . .. ......... . 3A-121 Egure 3A40b Env. of Nodes 4401, 4407, 4408 and 4409 .. . .. ................. ... ....... .... ....... 3A-121 i Egure SA40c Env. of Nodes 4501, 4507, 4508 and 4509 .. . .... ..... ..... ........... .... ............. 3A-122 l l l Mgure SA-40d Env. of Nodes 4511, 4517, 4518 and 4519 ..... ... . ...... .... .. .. ............ . ....... 3A-122 l Hgure 3A41a Node 5001 O nly.. ..... ............ . . .. . ...... ...... ... .... ...... .......... . . ....... .. ... 3A-123 , /m '/ Egure 3A.41b Env. of Nodes 5013, 5101, 5102 and 5103 .... .... . ..... ... .. .. ........ ... . . ..... 3A-123
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Mgure SA-41c 5015, 5106, 5107 and 5108... . . ... ....... ....... ...... ......... ........... . .......... 3A-124 Egure 3A.41d 5017, 5111, 5112 an d 5113 ...... .. ................ .......... . ....... ....... ... . .......... .. 3A-124 Figure SA-41e Env. of Nodes 5021, 5116, 5117 and 5118 ... ................ .... ........... .. . .. .... 3A-125 Egure 3A41f Node 5023 Only .. .. ... . ...... . .... . . .. . ......... ............. .......... . ....... .. ... .. ... 3A-125 Figure SA-41g Env. of Nodes 5026, 5121, 5122 and 5123 . . ... . ... .. ... .. ............. ........ .... 3A-126 Egure SA-41h Env. of Nodes 5032, 5126, 512 7 and 5128 . . ........ ..... ... ..... ..... ... .......... 3A-126 Hgure SA41i Node 5057 Only ... ....... ................................................................3A-127 Egure SA-41j Env. of Nodes 5060, 5131, 5132 and 5133 ........ . .. . . ..... .... .. .. ... .. ........ 3A-127 Figure SA-41k Env. of Nodes 5064, 5136, 5137 and 5138 .. . . . ........... ...... ... ............... ..... 3A-128 l l Hgure 3A.411 Env. of Nodes 5607, 5141, 5142 and 5143 .. . . ....... .. . ..... ... .. . ... ...... .... 3A-128 Figure 3A42a Env. of Nodes 1101,101, Ill,161 and 2101. .. ... . . . ... . .. .. ... . . . .. ... . . . . .. . 3A-129 Egure SA42b Env. of Nodes 1201,301,311 and 361... .. .. . ... ... ... . .. .. . . . .. . . 3A-129 \
. Figure 3A42c Env. of Nodes 1301,471,472 and 473.. ... .... . .. .. . ......... .. .. . .... . ... 3A-130
( Mgure 3A42d Env. of Nodes 1401, 671, 672 and 673 .... . ... . .. . . ... .. . . ....... ...... ........ 3A-130 i List of Figures -Amendment 1 10-xix
l l 25A5193 Rett. A l SBWR standantsakty Anasysis nepon 1 List of Figures (Continued) Figure 3A42e Env. of Nodes 1501,1502,1503,1505,1507,1508 and 1509. . .. . . ... . . 3A-131 Mgure SA-42f Env. of Nodes 1601,1602,1603,1605,1606,1607,1608 and 1609.. ... . . 3A-131 l Figure 3A42g Env. of Nodes 1701,1703,1705,1706,1707,1708 and 1709.. . . . ..... ..... 3A-132 ) Egure 3A42h Env. of Nodes 1801,1807,1808 and 1809 ..... .... . . . . . . . . . . . ... 3A-132 Figure 3A42i Env. of Nodes 1811,1817,1818 and 1819.. . .. .. . . . . . . . . . . .. ...... 3A-133 Figure 3A-42j Env. of Nodes 1821,1827,1828 and 1829.. . ... . . ... . .. ... . . .. 3A-13 3 Figure 3A43a Env. of Nodes 2201, 2207,2205 cnd 2209. . .... .. ... 3A-134 Figure SA-43b Env. of Nodes 2301,2307, f 308 and 2309. ... . . . . . . .. . . . .. .. . .. ... . 3A-134 Figure 3A43c Env. of Nodes 2401, 2402, 2403, 2407, 2408 and 2409.. .. . .... ...... .. ...... 3A-135 Mgure 3A43d Env. of Nodes 2501,2502,2503,2505,2506,2507,2508 and 2509........ .... 3A-135 Figure 3A43e Env. of Nodes 2601, 2602,2603,2605,2006,2607,2608 and 2609. ..... . 3A-136 Figure 3A43f Env. of Nodes 2701,2702,2703,2707,2708 and 2709. .. . . ... ...... ... .. .. . 3A-136 Figure SA44a Env. of Nodes 3201,3205,3206 and 3207........ . . . . . . . .. .... ... . .. ... 3A-137 Mgure 3A44b Env. of Nodes 3301, 3307,3308 and 3309... .... . . .. . .. .. .... ....... 3A-137 Figure 3A44c Env. of Nodes 3401,3407,3408 and 3409.... ..... ... . . ... ..... . .... 3A-138 Mgure 3A44d Env. of Nodes 3501,3507,3508 and 3509...... . . .. . . . . . . ... . . 3A-138 Mgure SA45a Env. of Nodes 4301,4307,4308 and 4309 ... .. . . .. .. ... .. ... . . ....... . .... .. SA-139 Figure SA-45b Env. of Nodes 4401,4407,4408 and 4409.. .... . . .. .... . . . . .. .... .... 3A-139 Figure 3A45c Env. of Nodes 4 501, 4507, 4508 and 4509 .... . . . . .... . . ......... . . .... ... 3A-140 Mgure 3A45d Env. of Nodes 4511, 4517, 4518 and 4519 ...... ...... .... ...... . .... .. .. 3A-140 Mgure SA-46a Node 5001 Only.. . . .._.... . . . . . . . . . . . . . . . . . . . . . . . . .... . .. .. . .. .. .. .. .. 3A-141 Mgure SA-46b Env. of Nodes 5013,5101,5102 and 5103. .... . . . . . . . . . . .. .. 3A-141 Figure 3A46c 5015, 5106, 5107 and 5108 ... . ....... . ... . ... ... .. .. ... . .... ..... . . . .... . 3A-142 Mgure 3A-46d 5017,5111,5112 and 5113.... . . . . . . . . . . . . . . . . . . . . . . . .... .. 3A-142 Figure SA-46e Env. of Nodes 5021,5116,5117 and 5118.. . . ... . . .. ..... .. . 3A-143 Figure SA-46f Node 5023 Only... ... . . . . .. . ........... .. . .. . .. .. .. .. .. . . . . ..... 3A-14 3 3.0-xx List ofFigures- Amendrnent 1
t 25AS113 Rev. A
- SBWR sondentsonrryAnalysis neport 4
N j List of Figures (Continued) Mgure 3A46g Env. of Nodes 5026, 5121, 5122 and 5123 ............ ................ . ... ...... .. . ... 3A-144 i Mgure SA46h Env. of Nodes 5032, 5126, 5127 and 5128 ... .... ...... ............... . . ..... ....... 3A-144
- Mgure 3A46i Env. of Nodes 5060, 6131, 5132 and 5133 ........... .... ........... . ..... . ... ...... 3A-145 4
j Mgure 3A46j Env. of Nodes 5607, 5141, 5142 an d 5143 .... . . ... ... ........... ... .... .... .. .. ..... 3A-145 1 Figure SGI Schematic Representation of MaterialJ-Integal R andJ-T Curves. . ........ . 3C-35 1
- Figure 3C-2 Carbon Steel Test Specimen Orientation Code............-....... . . . . . . ........... 3G36 i
1 j Mgure SG3 Toughness Anisotropy of ASTM Pipe (6 in. Sch. 80) . ...... ........ . .. .. . . .. 3C37 j Figure 3G4 Charpy Energies for Pipe Test Material 3 as a Function of Orientation and Temperature ................. ......... .. .. .. .. . 3C-38 l q Egure 3G5 Charpy Energies for Plate Test Material i as a Function of Orientation and Temperature ... . . .. ... ... .... ...... . .......... 3C-39 i Figure SG6 Comparison of Base Metal, Weld and HAZ Charpy Energies for p SA 333 Gr. 6 ....... ... .. ...... . . . .. . ... . . . .. . . .. . ..... .. .. .. .. ... ... . . . .. ..... .. .. ... .... .. . . . 3 C.40 l (s Figure SG7 Plot of 550T True Stress-True Strain Cunes for SA 333 Gr.6 Carbon S teel ...... . ....... .... ... . ... ... . ... .... . . ........ .. .. .. .. .. ... ... . ... . . ..... . .. . . . .. . 3G41 i 1
- Mgure SG8 Plot of 550T True Stress-True Strain Curves for SA 516 Gr.70
- Carbon Steel ... ................. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3G42 i Figure 3C.9 Plot of 350 F True Stress-True Strain Cunes for SA 333 Gr.6 Carbon Steel .. ..................................................................................3C43 l
i Egure SG10 Plot of 350T True Stress-True Strain Curves for SA 516 Gr.70
- Carbon S teel .. ............. . . . .... ... . . ..... .. .. ... .. ... ... . .. ... . ... ... . ... ... .. .... . . .. ..... 3C44 i
j Hgure SG11 Plot of 550T TestJ-R Curve for Pipe Weld .... ... ... . ......... ........... . ............. 3C45 i l Mgure SG12 Plot of 550TJmod, Tmod Data from TestJ-R Curve . ..... . .. ... . ... .......... 3C46 Figure SC-13 Carbon SteelJ-T Curve for 420 F .. ... .... ......... ...... ... ....... .. .. ...... .. . ...... 3C47 j Mgure SC-14 Schematic Illustration of Tearing Stability Evaluation.. . . .. . . ... . . .. . .. . .. .. . ... . 3G4 8 Figure 3C-15 A Schematic Representation ofInstability Tension l { and Bending Stresses as a Function of Flaw Strength.. ... ... .. . ..... .. ....... 3C49 1
.r l j g Figure SC.16 SA 333 Gr.6 Stress Strain Data at 550 F in the Ramberg-Osgood Format... . 3G50 .
)
.-( 5 Mgure 3C.17 Carbon Steel Stress-Strain Data at 350 F in the Ramberg-Osgood Format... 3C-51 4 Mgure 3C-18 Comparison of PICEP Predictions with Measured Leak Rates... .. .. .... 3C-52 j List ofFigures - Amendment 1 10-xwi 4 J
25A5113 Rsv. A SBWR Standant Safety Analysis Report List of Figures (Continued) O1l Mgure 3C-19 Pipe Flow Model.... . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. .. .. 30-53 Figure 3C-20 Mass Mow Rates for Steam / Water Mixtures.. . . . . . . . . . . . . . . . . . . . . . . . . . .. 3C-54 Figure 3C-21 Friction Factors for Pipes... .. ... . . . . . . . . . . . . . . . . . . . . . . . .. 3 0-55 Figure 3C-22 Irak Rate as a Function of Crack Length in Main Steam Pipe (Example) . . 3C-56 Mgure 3D-1 Environmental Zones in the Containment Vessel.... .. ........ . . 3D-23 Figure 3E.3-1 Reactor Building Section 0 -180 . . . .. ... . .. . . . . . . . . .. . ..SE.S17 Figure SE.3-2 Reactor Building Section 90 -270 . . .. ... . .... . ... .. .......... .. ..... = 3E.S18 Mgure SE.&3 Locations for Thermal Distribution Analysis.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3E.3-19 Figure 3E.14 Distribution of CO Pressure in the Suppression Pool ..... ........ .. .. .3E.S20 Mgure 3E.&5 Distribution of Chugging Pressure in the Suppression Pool.. . . . .... .... . . 3E.3-21 Figure 3E.S6 Distribution of SRV Pressure in the Suppression Pool.. ... . ... . . . . ... 3E.3-22 Mgure 3E.&7 Design Seismic Shears and Moments for Three Box Structure of the Reactor Building...... .. .. . .. . ... ................................................3E.3-23 Mgure 3E.S8 Design Seismic Shears and Moments for RCCV... .. ..... ........ .... . . ..... ... 3E.S24 Mgure SE.&9 Design Seismic Shears and Moments for RPV Pedestal /Ventwall Structure ..... . ... ....... ... . ... . . . . . . . . . . . .... ... . . .. . . ... ... . .. . . . .. . ... . . 3E. S2 5 Figure SE.3-10 Design Seismic Shears and Moments for Reactor Shield Wall.. .. . .. ...... 3E.3-26 Mgure SE.&11 Design Lateral Soil Pressure for RB Outer Walls . . .. .. ... .. . . ... . . . ... 3E.S27 Mgure SE.4-1 Critical Sections for Design ... . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . ..... . 3E.4-320 Mgure SE.4-2 Sections Outside the RCCV.. . ..................................................3E.4-321 Figure 3E.4-3 Moor Slab Sections at El. 32100.. ... .... .... .. .... . ...... .. .......... ....... ... 3E.4-322 Figure SE.4-4 3D Finite Element Model 270 ... . ... .... . . . . . . . . . . . . . . . .. .. 3E.4-323 Figure 3E.4-5 Unit Pressure in Drywell .... . . . . . . . . . . . . . . . ... .. . .. ... .. . ., .. . . . . 3E.4-324 Figure 3E.4 6 Unit Pressure in the Suppression Chamber .. ... ........ ....... . ..... .. .. ... 3E.4-325 Figure 3E.4-7 Iong-Term Thermal Load.. . . . . . . . . . . ... . . . . 3E.4-326 Mgure SE.4-8 SSE E-W .. .... . ... .. . . .. . . . . . . . . . .. . . .. . . ... . .. ........ . .. SE.4-327 Figure SE.7-1 IC Pool Main Girder.. .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . .. .. 3E.7-45 10 xxii List of Figurse - Amendment 1
1 l I
- 25A5113 Rev. A l SBWR sondantsannyAnarysis aeron i
! List of Figures (Continued)
- Figure SE.7-2 IC Pool Secondary Girder Cross Girder.. ... ..... . ...... .. .... ....... ..... ....... 3E.7-46 e
i it I a 3 i , i 1 i ) 4 i 4 ' i 1 F 1 1 1 4 1 4 4 e ) 1 4 , l , i 1 : I d t ' 4 3 J > li f i 1 l i 4 List of Figures - Amendr: rent 1 10-alii /niv
I i 25A5113 Rsv. A l SBWR sMdaniSaktyAnalysis Report i i i j 3.0 Design of Structures, Components, Equipment, and
- Systems
! 3.1 Conformance with NRC General Design Criteria ] This section contains an evaluation of the principal design cdteria of the SBWR ! Standard Plant as measured against the NRC General Design Criteria for 10CFR50 Appendix A. The General Design Criteria are intended to establish minimum requirements for the principal design criteria for nuclear power plants. t 4 The NRC General Design Criteria were intended to guide the design ofall water-cooled l 3 nuclear power plants; separate BWR-specific criteria are not addressed. As a result, the ? criteria are subject to a variety ofinterpretations. For this reason,in some cases ] conformance to a particular critedon is not directly measurable. In these cases, the I conformance of the SBWR design to the interpretation of the criteria is discussed. For j each criterion, the SBWR design is specifically assessed and a complete list of references is included to identify where detailed design information pertinent to that criterion is l treated in this standard safety analysis report (SSAR). ) 3.1.1 Group I - Overall Requirements ;
- ( 3.1.1.1 Criterion 1 - Quality Standards and Records -
k Criterion 1 Statement j Structures, sptems, and components important to safety shall be designed, fabricated, ) erected, and tested to quality standards commensurate with the importance of the safety j functions to be perfortned. Where generally recognized codes and standards are used,
- they shall be identified and evaluated to determine their applicability, adequacy, and
{ sufficiency and shall be supplemented or modified as necessary to assure a quality ; j product in keeping with the required safety function. A quality assurance program shall j be established and implemented in ottier to provide adequate assurance that these j structures, systems, and components will satisfactorily perform their safety functions. j Appropriate records of the design, fabrication, erection and testing of structures, i systems, and components important to safety shall be maintained by or under the l control of nuclear power unit licensee throughout the life of the unit. 1 Evaluation Against Criterion 1
- Safety-related and non-safety-related structures, systems, and components are identified l in Table 3.2-1. The total quality assurance program is described in Chapter 17 and j applies to the safety-related items. The quality requirements for non-safety-related items
- are controlled by the quality assurance program described in Chapter 17 in accordance j with the functional importance of the item. The intent of the quality assurance program j is to assure sound engineering in all phases of design and construction through conformity to regulatory requirements and design bases described in the license 4 q
? Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1 1 I 4
2SA5113 Rov, A SBWR SumfantSafety Anlysis Report 9 application. In addition, the program assures adherence to specified standards of workmanship and implementation of recognized codes and standards in fabrication and construction. The quality assurance program also includes the observance of proper preoperational and operational testing and maintenance procedures as well as the appropriate documentation. The quality assurance program is responsive to and in conformance with the intent of the quality-related requirements of 10CFR50 Appendix B. Structures, systems, and components are identified in Section 3.2 with r:spect to their location, service, and their relationship to the safety-related or non-safety-related function to be perfonned. Applicable codes and standards are applied to the equipment commensurate with their safety-related function. Documents are maintained to demonstrate that the requirements of the quality assurance program are satisfied. This documentation shows that appropdate codes, standards, and regulatory requirements are identified, correct matedals are specified, correct procedures are utilized, qualified personnel are provided, and the finished parts and components meet the applicable specifications. These records are available so that any desired item ofinformation is retrievable for reference. These records will be maintained for the life of the operating licenses. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 3.2 Classification of Stnictures, Systems, and Components 17 QualityAssurance 3.1.1.2 Criterion 2- Design Bases for Protection Against Natural Phenomena Criterion 2 Statement Structures, systems, and components important to safety shall be designed to withstand the effect of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions. The design bases for these structures, systems and components shall reflect: (1) appropdate consideration of the most severe of the natural [ enomena that have been histoncally reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and pedod of time in which the historical data have been accumulated; (2) appropriate combination of the effects of normal and accident 3.1-2 Conformance ,vith NRC General Design Criteria - Amendment 1
i i i l 25A5113 Rsv. A SBWR smekntsonroy Analysis nerort !C i\ conditions with the effects of the natural phenomena; and (3) the importance of the safety functions to be performed. Evaluation Against Criterion 2 Since the SBWR design is designated as a standard plant, the design bases for safety-related structures, systems and components, cannot accurately reflect the most severe of the natural phenomena that have been historically reported for each possible site and their surrounding areas. However, the envelope of the site-related parameters which encompass the majority of the potential sites in the contiguous United States is defined in Chapter 2. The design bases for these structures, systems, and components reflect this envelope of natural phenomena including appropriate combinations of the effects of no: mal and accident conditions with this envelope. The design bases meet die requirements of Criterion 2. Detailed discussions of various phenomena considered and design criteria developed are presented in the following sections: 1 Chapter / Section Title lle '\ 2.0 Site Characteristics 3.2 Classification of Structures, Systems, and Components 3.3 Wind and Tornado Loadings 3.4 Water level (Flood) Design 3.5 Missile Protection l ! 3.7 Seismic Design l l 3.8 Design of Seismic Category I Structures 3.9 Mechanical Systems and Components i 3.10 Seismic and Dynamic Qualification of Mechanical and Electrical I Equipment 3.11 Environmental Quahfication of Mechanical and Electrical Equipment 3.1.1.3 Criterion 3-Fire Protection Criterion 3 Statement Structures, systems, and components important to safety shall be designed and located l to minimize, consistent with other safety requirements, the probability and effect of l Conformance with NRC General Design Criteria - Amendment 1 11-3 l l
2SA5113 Rev. A SBWR standard saktyAnalysis Report O fires and explosions. Noncombustible and heat-resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the 1 containment and control room. Fire detection and fighting systems of appropriate l capacity and capability shall be provided and designed to minimize the adverse effects ) of fires on structures, systems, and components important to safety. Firefighting systems j shall be designed to assure that their rupture or inadvertent operation does not < significantly impair the safety capability of the stnictures, systems, and components. Evaluation Against Criterion 3 Fires in the plant are prevented or mitigated by the use of noncombustible and heat-resistant materials such as metal cabinets, metal wireways, high melting pointinsulation, and flame resistant markers for identification wherever practicable. Cablingis suitably rated and cable tray loading is designed to avoid unacceptable internal heat buildup. Cable trays are suitably separated to avoid the loss of redundant channels ofprotective cabling if a fire occurs.The arrangement ofequipment in reactor protection channels provides physical separation to limit the effects of fire. Combustible supplies, such as logs, records, manuals, etc., are limited in such areas as the control room, thus limiting the potential a fire. The plant fire protection system includes the following provisions: m automatic fire detection equipment in those areas where fire danger is greatest; and a suppression services which include suppression systems with automatic actuation with manual overnde as well as manually +perated fire extinguishers. The design of the fire protection system meets the requirements of Criterion 3. For further discussion, see the following sections: Chapter / Section Title 7 Instnamentation and Control Systems 8 Electric Power 9.5.1 Fire Protection System Appendix 9A Fire Hazard Analysis 13 Conduct of Operations , I r l 3.14 Conformance with NRC General Design Criteria - Amendment 1 i l
2SA5113 Rev. A SBWR seausannyAnarysis soport l 3.1.1.4 Criterion 4- Environmental and Dynamic Effects Design Bases Criterion 4 Statement Structures, systems, and components important to safety shall be designed to l accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, , including lossof. coolant accidents. These structures, systems, and components shall be l app ;priately protected agamst dynamic effects, including the effects of missiles, pipe ! whipping, and discharging fluids that may result from equipment failures and from l events and conditions outside the nuclear power unit. However, dynamic effects . associated with postulated pipe ruptures in nuclear power plant units may be exduded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under
- conditions consistent with the design basis for the piping.
j Evaluation Against Criterion 4 l Safety-related structures, systems, and components (SSC) are designed to accommodate l I the dynamic effects of, and to be compatible with, environmental conditions associated with normal operation, maintenance, and postulated pipe failure accidents induding loss-of coolant accidents (LOCA). These safety-related structures, systems, and components are appropriately protected against dynamic effects in cluding the effects ofmissiles, pipe whipping, and discharging fluids that may result from equipmen t failure. The effects ofmissiles originating outside , the SBWR Standard Plant are also considered. Design requirements specify the ! I duration that safety-related SSC must survive the extreme environmental conditions l following a LOCA. The design of these structures, systems, and components meets the l requirements of Criterion 4. l Subsection 3.6.3 identifies the requirements for the piping that is to be excluded from postulation of pipe ruptures for design of the plant against dynamic effects from the associated pipe ruptures. l For further discussion, see the following sections: Chapter / Section Title ! 2.0 Site Characteristics 3.3 Wind and Tornado Loadings 3.4 Water level (Flood) Design l Conformance with NRC GeneralDesign Criteria - Amendment 1 3,1-5 l
25AS113 Rev. A SBWR Standsnf Safety Anlysis Repon O Chapter / Section Title 3.5 Missile Protection 3.6 Protection Agamst Dynamic Effects Associated with the Postulated Rupture of Piping 3.8 Design of Seismic Category I Structures 3.11 Environmental Qualification of Mechanical and Electrical Equipment 5.2 Integrity of Reactor Coolant Pressure Boundary 6 Engineered Safety Features 7 Instrumentation and Control Systems 8 Electric Power 3.1.1.5 Criterion 5- Sharing of Structures, Systems, and Components Criterion 5 Statement Structures, systems, and components important to safety shall not be shared among nuclear power units unless it can be shown that such sharing will not significantly impair their ability to perform their safety functions, including, in the event of an accident in one unit, an orderly shutdown and cooldown of the remaining units. Evaluation A, gainst Criterion 5 i Since the SBWR design is for a single-unit station, this criterion is not applicable, j 3.1.2 Group II- Protection by Multiple Fission Product Barriers 3.1.2.1 Criterion 10 - Reactor Design Criterion 10 Statement The reactor core and associated coolant, control, and protection systems shall be designed with appropriate margm to assur( that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. Evaluation Against Criterion 10 The reactor components consist of fuel assemblies, control rods, incore ion chambers, neutron sources, and related items. The mechanical design is base 3 on conservative application of stress limits, operating experience, and experimental test results. The fuel is designed to maintain integrity over a complete range of power levels, including 3.1-6 Conformance with NRC GeneralDesign Criteria - Amendment 1
__ _ _ . _ -_ _ _ ___ _ _ . - , __ __ __. _ . _ _ . ,.._ m _ t 25A5113 Rev. A SBWR sondentsannyAnafysis Reput _
- V i
transient conditions. The core is sized with sufficient heat transfer area and coolant flow to ensure that fuel design limits are not exceeded under normal conditions or anticipated opemtional occurrences. The Reactor Protection System (RPS) is designed to monitor certain reactor ! parameters, sense abnormalities, and to senm the reactor, thereby preventing fuel design limits from being exceeded when the trip points are exceeded. Scram setpoints are based on operating experience and by the safety design basis. There is no case for which the scram setpoints allow the reactor core to exceed the thennal-hydraulic safety limits. Power for the RPS is supplied by four independent, uninterruptible ac power supplies. An alternate battery power supply is available for each bus. The reactor will scmm on loss of power or CRD hydraulic pressure. An analysis has been made of the effects on core fuel following adverse plant openting conditions. The results of abnormal operational transients are presented in Chapter 15 i and show that the mirumum critical power ratio (MCPR) does not fall below the - l ' transient MCPR limit, thereby satisfymg the transient design basis. 8 l l j The reactor core and associated coolant, control, and protection systems are designed
\ to assure that the specified fuel design limits are not exceeded during conditions of normal or abnormal operation and, therefore, meet the requirements of Criterion 10.
For further discussion, see the following sections: ; l l l Chapter / ' j l Section Tide I 1.2 Genen! Plant Description 4.2 FuelSystem Design 4.3 Nuclear Design 4.4 Thermal and Hydraulic Design i 5.4.8 Reactor Water Cleanup / Shutdown Cooling System 7.2 Reactor Trip System j 15- AccidentAnalyses 1O l i Conformence with NRC GeneralDesign Criterie - Amendment 1 11 l
25A5193 Rev. A SBWR standardsafety Analysis Report O 3.1.2.2 Criterion 11 - Reactor inherent Protection Criterion 11 Statement The reactor core and associated coolant systems shall be designed so that in the power operating range the net effect of the prompt inherent nuclear feedback characteristics tend to compensate for a rapid increase in reactivity. Evaluation Against Criterion 11 ne reactor core is designed to have responses that regulate or damp changes in power level and spatial distribution of power production to a level consistent with safe and efficient operation. The inherent dynamic behavior of the core is characterized in terms of: a fuel temperature or Doppler reactivity coefficient; a moderator void reactivity coefficient; and a moderator temperature reactivity coefficient. The combined effect of these coefficients in the power range is termed the power coefficient. A negative Doppler reactivity coefficient is maintained for any operating condition. Doppler reactivity feedback occurs simultaneously with a change in fuel temperature and opposes the power change that caused it; it contributes to sprem stability. A negative core moderator void reactivity coefficient resulting from boiling in the active flow channels is maintained for any operating condition. The negative void reactivity coefficient provides an inherent negative feedback during power transients. Because of the large negative moderator void reactivity coefficient, the SBWR has a number of inherent advantages, such as: a the inherent self-flattening of the radial power distribution; a the ease of control; and a the spatial xenon stability. The reactor is designed so that the moderator temperature reactivity coefficient is negative above hot standby, and the overall power reactivity coefficient is negative. Typically, the power coefficient at full power is about -0.04Ak/k/AP/P at the beginning of life and about -0.03Ak/k/AP/P at 10,000 mwd /T (short). These values are well l within the range required for adequate damping of power and spatial xenon l disturbances. The reactor core and associated coolant system are designed so that in the 3.14 Conformance with NRC General Design Criteria - Amendment 1 l l
2SAS113 Rev. A SBWR standard sarety Anairsis neport (A) V power operating range, prompt inherent dynamic behavior tends to compensate for any rapid increase in reactivity in accord with Criterion 11. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 4.3 Nuclear Design 4.4 Thermal and Hydraulic Design 3.1.2.3 Criterion 12- Suppression of Reactor Power Oscillations Critorion 12 Statement The reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in conditions exceeding specified acceptable fuel design limits are not possible or can be reliably and readily
) detected and suppressed.
(w/ Evaluation Against Criterion 12 The reactor core is designed to ensure that no power oscillation will cause fuel design limits to be exceeded. The power reactivity coefficient is the composite simultaneous effect of the fuel temperature or Doppler reactivity coefficient, moderator reactivity void coefficient and moderator temperature reactivity coefficient to the change in power level. It is negative and well within the range required for adequate damping of power and spatial xenon disturbances. Analytical studies indicate that for large boiling water reactors, under<iamped, unacceptable power distribution behavior (i.e., xenon instability) could only be expected to occur with power coefficie'nts more positive than about -0.01 Ak/k/AP/P. Operating experience has shown large boiling water reactors to be inherently stable against renon induced power instability. The negath e reactivity coefficients provide: n good load following with well< lamped behavior and little undershoot or overshoot in the heat transfer response; and a strong damping of spatial power disturbances. The RPS design provides protection from excessive fuel cladding temperatures and protects the reactor coolant pressure boundag (RCPB) from excessive pressures which
' threaten the integrity of the system. Local abnormalities are sensed, and,if protection
( C systems are reached, corrective action is initiated through an automatic scram. High integrity of this protection system is achieved through the combination oflogic Conformance with NRC General Design Criteria -- Amendment 1 3.1-9
l 25AS113 Rev. A l l SBWR seemiantsaktyAnalysis Repon , arrangement, tdp channel redundancy, power supply redundancy, and physical l separauon. l l The reactor core and associated coolant, control, and protection systems are designed to suppress any power oscillations which could result in exceeding fuel design limits. These systems assure that Criterion 12 is met. For further discussions, see the following sections: ! Chapter / l Section Title l 1.2 General Plant Description l l 4.3 Nuclear Design l l 4.4 Thermal and Hydraulic Design 1 7.2 Reactor Trip System 7.7 ControlSystems 15 Accident Analyses 3.1.2.4 Criterion 13-instrumentation and Control Criterion 13 Statement Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those l variables and systems that can affect the fission process, the integrity of the reactor core, the RCPB, and the contamment and its associated systems. Appropriate controls shall l be provided to mamtain these variables and systems within prescribed operating ranges. 1 Evaluation Against Criterion 13 The neutron flux in the reactor core is monitored by four subsystems. The Startup l Range Neutron Monitor (SRNM) Subsystem measures the flux from startup through l 15% power (into the power range). The power range is monitored by many detecton l which makeup the Local Power Range Monitor (LPRM) Subsystem. 'Ite output of these detectors is used in many ways. The output of selected core-wide sets of detectors is averaged to provide a core-average neutron flux. This output is called the Average Power Range Monitor (APRM) Subsystem. The Automated Fixed Incore Probe (AFIP) Subsystem provides a means for calibrating the LPRM. Both the SRNM and APRM Subsystems generate scram trips to the Reactor Protection System. They also generate rod-block trips. l l 1110 Conformance with NRC GeneralDesign Criteria - Amendment 1 I
2SAS113 Rzv. A SBWR standant safety Anatysis Report /m\ V The Reactor Protection System protects the fuel barriers and the nuclear process barrier by monitoring plant parameters and causing a reactor scram when predetermined setpoints are exceeded. Separation of the scram and normal rod control function prevents failures in the reactor manual control circuitry from affecting the scram circuitry. To provide protection against the consequences of accidents invohing the release of radioactive materials from the fuel and RCPB, the Leak Detection and Isolation System (LD&lS) initiates automatic isolation of appropriate pipelines whenever monitored variables exceed preselected operationallimits. The LD&lS, described in Subsections 5.2.5 and 7.3.3, provides instrumentation and controls to detect, annunciate and,in some cases, isolate the RCPB to ensure its integrity. Also see the evaluation of GDC 30. The Process Radiation Monitoring System (PRMS) monitors radiation levels of various processes and provides trip signals to the RPS and LD&lS whenever pre-established limits are exceeded. Adequate instrumentation has been provided to monitor system variables in the reactor core, RCPB, and reactor containment. Appropriate controls have been prosided to maintain the variables in the operating range and to initiate the necessary corrective l Q action in the event of abnormal operational occurrence or accident. The design ofinstrumentation and control systems meets the requirements of Criterion 13. Additionalinformation on the instrumentation and controls is given in Chapter 7. 3.1.2.5 Criterion 14- Reactor Coolant Pressure Boundary Criterion 14 Statement The reactor coolant pressure boundary shall be designed, fabricate of, erected, and tested so as to have an extremely low probability of abnormal leakage, rapidly propagating failure, and of gross rupture. Evaluation Against Criterion 14 The piping and equipment pressure parts within the RCPB (as defmed by Section 50.2 l of 10CFR50) are designed, fabricated, crected, and tested in accordance with 10CFR50.55a to provide a high degree ofintegrity throughout the plant lifetime. Systems and components within the RCPB are classified as Quality Group A (Section 3.2). The design requirements and codes and standards applied to this quality group help ensure high integrity in keeping with the safety-related function. t To minimize the possibility of brittle fracture within the RCPB, the fracture toughness j (m") properties and the operating temperature of ferritic materials are controlled to ensure adequate toughness. Section 5.2 describes the methods utilized to control toughness properties of the RCPB materials. Materials are to be impact teswd in accordance with ; Confortnance with NRC GeneralDesign Criteria - Amendment 1 3.1-11
l l PSAS113 RDv. A SBWR standard saretyAnairsis aeport , l O' ASME Boiler and Pressure Vessel Code Section III, where applicable. Where RCPB piping penetrates the containment, the fracture toughness temperature requirements of the RCPB materials apply. Piping and equipment pressure parts of the reactor coolant pressure boundary are assembled and erected by welding unless applicable codes perndt flanged screwed joints. Welding procedures are employed which produce welds of complete fusion that are free of unacceptable defects. All welding procedures, welders, and welding machine operators used in producing pressure containing welds are quahfied in accordance with the requirements of the ASME Boiler and Pressure Vessel Code Section IX for the materials to be welded. Qualifications records, including the results of procedure and performance quahfication tests and identiEcation symbols assigned to each welder, are maintained. Section 5.2 contains the detailed material and exammation requirements for the piping and equipment of the RCPB prior to and after its assembly and erection. Leakage testing and smveillance is accomplished as described in the evaluation against Critedon 30 of the General Design Criteria. The design, fabrication, erection, and testing of the reactor coolant pressure boundary help assure an extremely low probability of failure of abnormal leakage, thus satisfying the requirements of Criterion 14. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 3 Design of Structures, Components, Equipment, and Systems 5.2 Integrity of Reactor Coolant Pressure Boundary : l 5.3 Reactor Vessel 15 Accident Analyses 17 Quality Assurance 3.1.2.6 Criterion 15- Reactor Coolant System Design Criterion 15 Statement The Reactor Coolant System (RCS) and associated auxiliary, control, and protection systems shall be designed with sufEcient margin to assure that the design conditions of 3.1-12 Conformance with NRC General Design Criteria - Amendment 1
.. . _ . _ . . _ _ . - - _ _ _. - - __. . - _ . _ _ =_ 2SA5113 Rw. A SBWR sondentsonnyAnarrsisserat lC
- t the reactor coolant pressure boundary are not exceeded dudng any condition of normal operation, including anticipated operational occurrences.
Evaluation Against Criterion 15 The Reactor Coolant System (RCS), as identified in Section 5.1, consists mainly of the j reactor vessel and appurtenances, and the nuclear boiler system including the main l steamlmes, feedwater lines and pressure-relief discharge system, the Isolation , l Condenser System, and portions of the Reactor Water Cleanup / Shutdown Cooling System and Control Rod Drive System. ! The auxiliary, control, and protection systems associated with the RCS act to provide sufBcient margin to assure that the design conditions of the RCPB are not exceeded i during any condition of normal operation, including anticipated operational
- occurrences. As described in the evaluation ofCriterion 13, instrumentation is provided l to monitor essential variables to verify that they are within prescribed operating limits. l If the monitored variables exceed their predetermined settings, the aimliary, control, !
l and protection systems automatically respond to maintain the variables and systems ; within allowable design limits. [ ' An example of the integrated protective action scheme which provides sufficient l margm to assure that the design conditions of the RCPB are not exceeded is the automatic initiation of the pressure relief system of the Nuclear Boiler System (NBS) upon receipt of an overpressure signal. To accomplish overpressure protection of the l reactor pressure vessel system and RCPB, a number of pressure-operated relief valves l are provided that can discharge steam from the main steamlines to the suppression l pool. The pressure reliefsystem also provides for automatic depressurization of the RCS in the event of an LOCA in which the vessel is not depressurized by the accident. The depressurization of the RCSin this situation allows operation of the Gravity-Driven Cooling System to supply enough cooling water to adequately cool the core. In a similar manner, other aurihary, control, and protection systems provide assurance that the design conditions of the RCPB are not exceeded during any conditions of normal I operat2on, mcluding anticipated operational occurrences. j For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 3 Design of Structure, Components, Equipment, and Systems 5.2.2 Overpressure Protection Conformance with NRC GeneralDeeign Criteria - Amendment 1 3.1-13 i
2SA5113 Rect. A SBWR saadant sehnyAnalysis nepon l O l l Chapter / Section Title (Continued) 5.2.5 Reactor Coolant Pressure Boundary (RCPB) Leakage Detection 5.3 Reactor Vessel 15 Accident Analpes 3.1.2.7 Criterion 16-Containment Design Criterion 16 Statement Reactor containment and associated systems shall be provided to establish an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment and to assure that the contamment design conditions important to safety are not exceeded for as long as postulated accident conditions require. Evaluation Against Criterion 16 The Primary Containment System consists of the following major structures and components: . a Aleaktight containment vessel (CV) encloses the reactor pressure vessel, the reactor I coolant pressure boundary, and other branch connections of the reactor primary coolant system. The CVis a reinforced concrete cylindrical structure with an internal leaktight steel liner providing the primary contamment boundary. The CV structure consists of the drywell top slab, cylindrical containment wall, suppression pool floor slab, RPV pedestal, and the basemat. The opening in the top of the CV for servicing and refueling the RPV is closed by a steel drywed head. The upper drywell encloses the upper portion of the RPV, the major piping systems (main steam, feedwater, GDCS, and IC lines, SRVs, DPVs), drywell cooling systems, GDCS pools, and other miscellaneous systems. The lower drywell encloses the lower portion of the RPV below the support skirt, cooling system, fine motion control drives (FMCRD), other miscellaneous systems, and provides maintenance space . below the RPV. a A suppression pool-the water volume in the wetwell--serves as a heat sink to condense the steam released during a LOCA or SRV discharge. The gas space-the gas volume in the wetwell-serves as the olowdown resenoir for the upper and lower drywell nitrogen during a LOC 4 after it passes through the horizontal vents and pool. m Associated containment penetrations and isolation devices. The drywell and wetwell zones condense the steam and contain fission product releases from the postulated design basis accident (i.e., the double ended rupture of the largest 3.1-14 Conformance with NRC General Design Criteria - Amendment 1
F 25A5113 Rev. A SBWR StandantSafetyAnalysisRepon l iO b) pipe in the Reactor Coolant System). The leaktight containment vessel prevents the l release of fission products to the emironment. l The safety envelope completely encloses and integrates structurally the CV. The safety l envelope contains, dilutes, and holds up leakage from the containment. Temperature and pressure in the CV are limited following an accident by using the l ( PCCS, an engineered safety feature system to condense steam in the containment atmosphere. Additionally, the isolation condensers and the RWCU/SDC System can assist in cooling reactor steam and reactor water coolants following an accident The Fuel and Auxiliaq Pools Cooling Systern can be used to cool the suppression pool water. The design of the contamment systems meets the requirements of Criterion 16. For further discussion, see the folloting sections: Chapter / Section Title
/N 1.2 Principal Plant Description k'
3.8.1 Reinforced Concrete Containment 6.2 Containment Systems 15 AccidentAnalyses 3.1.2.8 Criterion 17 - Electric Power Systems Criterion 17 Statement An on-site electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety. The safety function of each system (assuming the other system is not functioning) shall be to provide sufficient capacity and capability to assure that (1) specified acceptable fuel design limits and design conditie of the reactor coolant pressure boundary are not ; exceeded as a result of anticipated operatiom Luccurrences and (2) the core is cooled l and containment integrity and other vital functions are maintained in the event of postulated ace.idents. 1 The on-site electric power supplies, including the batteries and the on4ite electrical distribution system, shall have sufficient, independence, redundancy, and testability to l perform their safety functions assuming a single failure. (j Electric power from the transmission network to the on-site electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate Conformance with NRC GeneralDesign Criteria - Amandment 1 3.1 16
I l 25AS113 Recs. A SBWR stadadsafetyAnar rsisneport O rights of way) designed and located so as to minimize to the extent practical the likelihood of simultaneous failure under operating and postulated accident and environmental conditions. A switchyard common to bath circuits is acceptable. Each of these circuits shall be designed to be available in sufficient time following a loss of all on-site alternating current power supplies and the other off-site electric power circuit, to assure that specified acceptable fuel design limits and design conditicas of the reactor coolant pressure boundary are not exceeded. One of these circuits shall be designed to be available within a few seconds following a loss-of-coolant accident to assure that core cooling, contamment integrity, and other vital safety functions are mamtained. Provisions shall be included to minimize the probability oflosing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power
=
generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the on-site electric power supplies. Evaluation Against Criterion 17 On-site Electric Power System - The on-site power system is divided into two power trams at the 6.9 kVlevel for design and operational flexibility of the plant non-safety fluid and mechanical systems. Separate unit auxiliary transformers feed each of the power tnins. Each power train supplies power to a three-tier power distribution system arrangement: a The first tier supplies power to nc,n-safety-related loads required prunarily for unit operation and supplies power to the second tier. m The second tier supplies power to permanent non-safety-related load,which, due to their specific functions, are generally required to remain operational at all times or when the unit is shutdown, and gplies power to the uird tier. m The third tier supplies power to safety-related loads. Both trains of the permanent non<afety-related system (second tier) have,in addition to their normal preferred power supply (first tier), an alternate preferred power supply through the reserve transformer to an independent off-site source, and a standby power supply from separate on-site standby genemtors. Each division of the safety-related power distribution system (third tier) is provided with physically separated and electrically independent batteries sized to supply emergency power to the engineered safety systems in the event ofloss of all other power sources. The on-site dc power system includes the plant batteries and banery chargers and that portion of the system ou its load side, except for the dc/ac inverters and that portion of l the system on the load side of the inverters. 3,1-16 Confonnance with NRC GeneralDesign Cateria - Amendment 1 I
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The safety loads utilize various Class IE ac and/or dc sources for instrumentation and motive or ccentrol power or both for all systems required for safety. Combinations of power sources may be involved in performing a single safety-related function. The systems required for safety are: a Reactor Protection System (RPS); a Engineered Safety Features Systems;
- a Isolation Condenser System (ICS);
a Standby Liquid Control System (SLCS); and a information systems important to safety. The onsite electric power systems are designed to meet the requirements of Criterion 17. Off site Electric Power System-The off-site power system consists of the set of electrical circuits and associated equipment that is used to interconnect the off-site ( ) transmission system with the plant main generator and the on4ite electrical power
\._/ distribution system.
The system includes the plant switchyard, the high voltage tie lines, the main step-up transformers, the generator breaker, the isolated phase bus, and the 6.9 kV bus duct from the unit aunhary transformers to the unit aimbary switchgear and from the reserve transformer to the DG switchgear. The off4ite power system begins at the terminals on the transmission system side of the circuit breakers which connect the switching stations to the off-site transmission systems. It ends at the 6.9 kV switchgear main circuit breakers which are supplied power from the unit aimbary and reserve transformers, and at the terminals of the main generator. Power is supplied to the plant from two electrically independent and physically separate off-site power sources as follows: i a " Normal Preferred" source through the plant main and unit aimbary transformers; l and a "Altemate Preferred" source through the reserve transformer. During plant stanup, normal or emergency shutdown, or during plant outages, the off-
,q site power system serves to supply power from the off-site transmission system to the ]
/ plant aimbary and service loads !
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Conformarwe with NRC Senoral Design Criteria - Amendment 1 a l-17 l I
2SA5113 Rev. A SBWR saadantsareryAnarrsisneport l During normal operation, the off-site power system is used to transmit generated power to the off-site transmission system and to the plant auxiliary and service loads. The generator breaker allows the on-site distribution system to be powered ! continuously and unswitched throughout plant startup, norral operation, and normal or emergency shutdown. When the generator breaker is tripped, power to the plant is backfed from the off-site power sprem through the main step-up transformer. The design of the off-site power systems is outside the scope of the SBWR design. The SBWR Standard Plant interfaces are addressed in Subsection 8.2.3. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 3,10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment 3.11 Environmental Quahfication of Safety-Related Mechanical and Electrical Equipment 8.2 Off-site Power Systems 8.3 Onsite Power Systems 3.1.2.9 Criterion 18-Inspection and Testing of Electric Power Systems Criterion 18 Staternent
; Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing ofimportant areas and features, such as wiring, msulation, connections, and switchboards, to assess the continuity of the systems and 1 condition of their components. The systems shall be designed with a capability to test i periodically (1) the operability and functional performance of the component of the l systems such as on-site power sources, relays, switches, and buses, and (2) the operability of the systems as a whole and, under conditions as close to design as practical, the full operation sequence that brings the systems into ope ation, including operation of , applicable portions of the protection system, and the transfer of power among the nuclear power unit, the off-site power system, and the on-site power system.
t Evaluation Against Criterion 18 l All Class 1E loads are normally sapplied directly from the Class 1E dc power supply xstem (e.g., batteries, chargers) or from the Class IE dc system through dc-to-ac al-18 Conformance with NRC GeneralDesign Criteria -Amandment 1 I o
2SAS113 R:v. A SBWR standant sarety AnarysisReport
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() inverters. Capability is provided for testing each battery, batte y charger, and inverter without disrupting power to the Class IE loads. Design of the Class IE power system provides testability in accordance with the requirements of Criterion 18. For further discussion, see the following sections: Chapter / Section Title 8.3 On-site Power Systems 14 InitialTest Program 3.1.2.10 Criterion 19- Control Room Criterion 19 Statetr- nt A control room shall i e provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to mamtain it in a safe condition under accident conditions, including lossef-coolant accidents. Adequate radiation (U; protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem ) whole body, or its equivalent to any part of the body, for the duration of the accident. i Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrument action and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent co!d shutdown of the reactor through the use of suitable procedures. Evaluation Against Criterion 19 The control room contains the controls and nestsny surveillance equipment for operation of the plant systems, including the reactor and its auxiliary systems, engineered safety features, turbine generator, steam and power conversion systems, and station electrical distribution. The control room is located in the reactor building. Safe occupancy of the control room l during abnormal conditions is provided for in the design. Adequate shielding is l provided to maintain tolerable radiation levels in the control room in the event of a i desi;,;n basis accident for the duration of the accident. !n The control room ventilation system has redundant equipment and includes radiation, { J G toxic and smoke detectors with appropriate alarms and interlocks. The control room i intake air can be filtered through high efficiency particulate air / absolute (HEPA) and l Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1-19
25AS113 Rev. A SBWR sandard saretyAnalysis aeport \ O charcoal Elters. If any of the above hazards exist at the normal control room ventilation intake, habitability is assured by the Emergency Breathing Air System (EBAS) which, upon isolation of the control room envelope, provides a positive air purge. The control room is continuously occupied by qualified operating personnel under all l operating and accident conditions. In the unlikely event that the control room must be vacated and access is restricted, instrumentation and controls are provided outside the l control room which can be utilized to safely perform a hot shutdown and a subsequent cold shutdown of the reactor. The control room design meets the requirements of Criterion 19. l For further discussion, see the following sections: Chapter / Section Title l 1.2 General Plant Description 6.4 ControlRoom Habitability Systems l 7 Instrumentation and Control Systems l l 7.4.2 Remote Shutdown System 9.4.1 Control Room Area Ventilation System 9.5.1 Fire Protection System 12.3 Radiation Protection 12.3.3 Ventilation 18.2 Design Goals and Design Bases 3.1.3 Group til- Protection and Reactivity Control Systems 3.1.3.1 Criterion 20-Protection System Functions Criterion 20 Statement The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specHied acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety. 3.1 20 Conformance with NRC GeneralDesign Criteria - Amendment 1 t
2SAS113 R:v. A SBWR stamtantsannyAnarysis napon \ i
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Evaluation Against Criterion 20 The Reactor Protection System (RPS) is designed to provide timely protection against ; the onset and consequences of conditions that threaten the integrity of the fuel barrier and reactor coolant pressure boundary barrier. Fuel damage is prevented by initiation of an automatic reactor shutdown if monitored variables of nuclear steam supply systems (Section 7.2) exceed pre <stablished limits of anticipated operational occurrences. Scram trip settings are selected and verified to be far enough above or below operating levels to provide proper protection but not be subject to spurious j scrams. The RPS includes the uninterruptible power sources, sensors, transmitters, j bypass circuitry, and switches that signal the control rod system to scram and shutdown l the reactor. The scrams initiated by the Neutron Monitoring System signals, nuclear boiler high pressure, and reactor vessel low and high water levels prevent fuel damage following abnormal operational transients. Specifically, these process parameters initiate a scram in time to prevent the core from exceeding thermal hydraulic safety ; I limits during abnormal operational transients. Response by the Reactor Protection System is prompt and the total scram time is short. In addition to the Reactor Protection System, which provides for automatic shutdown ; O of the reactor to prevent fuel damage, protection systems are provided to sense accident i V conditions and to initiate automatically the operation of other systems and components ! important to safety. Other systems automatically isolate the reactor vessel or the containment to prevent the release of significant amounts of radioactive materials from l the fuel and the reactor coolant pressure boundary. The controls and instnunentation for the ECCS and the isolation systems are initiated automatically when monitored variables exceed pre 4 elected operational limits. The design of the protection system satisfies the functional requirements as specified in Criterion 20. For further discussion, see the following sections: Chapter / l Section Title 4 1 1.2 General Plant Description 4.6 Functional Design of Fine Motion Rod Drive Systems 5.2.2 Overpressure frotection ,p 5.4.5 Main Steamline Isolation System Y 6.3 Emergency Core Cooling Systems Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1-21
f 25A5113 Rev. A SBWR sundant SafetyAnalysis Report O Chapter / Section Title 7.2 Reactor Trip System 7.3.1 Emergency Core Cooling System 7.3.2 Passive Contamment Cooling System 7.3.3 Leak Detection and Isolation System 7.3.4 Safety System Logic and Control System 15 AccidentAnalyses 3.1.3.2 Criter;on 21 - Protection System Reliability and Testability Criterion 21 Statement The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and i independence designed into the protection system shall be sufficient to assure that l (1) no single failure results in loss of the protection function and (2) removal from senice of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designer' a permit periodic testing ofits functioning when the reactor is in operation,inclui y a capability to test channels independently to determine failures and losses of redundancy that may have occurred. Evaluation Against Criterion 21 Reactor Protection System design provides assurance that, through redundancy, each channel has sufficient reliability to fulfill the single-failure criterion. No single component failure, int:ntional bypass maintenance operation, calibration operation, or test to verify operational availability impairs the ability of the system to perform its intended safety function. Additionally, the system design assures that when a scram trip point is exceeded, there is a high scram probability. However, should a scram not occur from the Reactor Protection System, the Alternate Rod Insertion will actuate when the l trip points are exceeded (Subsection 7.4.5). There is sufficient electrical and physical separation between channels and between logics monitoring the same variable to prevent environmental factors, electrical transients, and physical events from impairing the ability of the system to respond correctly. 3.1-22 Conformance with NRC GeneralDesign Criteria - Amendment 1 l
2SA5113 Rev. A SBWR saadant safetyAnalysis Report in i ) LJ The Reactor Protection System includes design features that permit inserdce testing. This ensures the functional reliability of the system should the reactor variable exceed the corrective action setpoint. The Reactor Protection System initiates an automatic reactor shutdown if the monitored plant variables exceed preestablished limits. This system is arranged as four separately powered divisions. Each division has a logic which can produce an automatic trip signal. The logic scheme is a twoout-of-four arrangement. The Reactor Protectic n System can be tested dudng reactor operation. t ianual scram testing is performed by operating one of the four manual scram controh; this tests one division. The total tests vedfy the ability to de-energize the scram pilot vaive solenoids. Indicating lights verify that the actuators contacts h:.ve opened. This capability for a thorough testing program significantly increases reliability. Control rod drive operability can be tested during normal reactor operation. Rod position indicators and in< ore neutron detectors are used to verify control rod movement. Each control rod can be withdrawn one step and then reinserted to the original position without significantly perturbing the nuclear steam sup;,:y systemt at (g} v most powerlevels. One control rod is tested at a time. Control rod mechanism overdrive demonstrates rod-to<irive coi., Tg integrity. Hydraulic supply subsystem pressure can be observed on control room ins . mentation. More importantly, the hydraulic control unit scram accumulator level is continuously monitored. The high functional reliability, redundancy, and inservice testability of the protection system satisfy the requirements specified in Cdterion 21. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 4.6 Functional Design of Fine Motion Control Rod Drive Systems 7.2 ReactorTrip System 7.3.4 Safety System logic and Control { 7.4.5 Alternate Rod Insertion (y i ) N.) Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1-23
25A5113 Rev. A l SBWR standant sareny Am seport ! 3.1.3.3 Criterion 22-Protection System independence elj Criterion 22 Statement The protectiori system shall be designed to assure that the effects of natural phenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protection function, or shall be demonstrated to be acceptable on some other defined basis. Design techniques, such as functional divenity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function. Evaluation Against Criterion 22 Components of the protection system are designed so that the mechanical, therinal and radiological environmental conditions resulting from any accident situation in which the components are required to function do not interfere with the operation of that function. The redundant sensors are electrically and physically separated. Only circuits of the same didsion are run in the same raceway. Multiplexed signals are carried out by fiber optic medium to assure control signal isolation. The Reactor Protection System is designed to permit maintenance and diagnostic work while the reactor is operating without restricting the plant operation or hindering the output ofsafety functions.The flexibility in design afforded the protection system allows operational system testing by the use ofindependent input for each actuator logic. When an essential monitored variable exceeds its scram trip point, it is sensed by four independent senson each located a in separate instrumentation channel. A bypass of any single channel is permitted for maintenance operation, test, etc. Tids leaves three channels per monitored variable, each of which is capable ofinitiating a scram. Only two actuator logics must trip to initiate a scram. Thus, the twoout-of-four arrangement assures that a scram occurs as a monitored variable exceeds its scram setting The protection system meets the design requirements for functional and physical independence as specified in Criterion 22. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 3.11 Environmental Qualification of Safety-Related Mechanical and Electrical Equipment 4.6 Functional Design of Fm' e Motion Control Rod Drive Systems 3,1-24 Conformance with NRC GeneralDesign Criteria - Amendment 1
I I 2SA5113 Rsv. A SBWR saadant sanny Analysis aerort r k Chapter / Section Title 5A.5 Main Steamline Isolation System 1 6.3 Emergency Core Cooling Systems 7.2 Reactor Trip System , 7.3.1 Emergency Core Cooling System 7.3.3 Leak Detection and Isolation System 3.1.3.4 Criterion 23-Protection System Failure Modes Criterion 23 Statement The protection system shall be designed to fail into a safe state or into a state demonstnted to be acceptable on some other defined basis ifconditions such as disconnection of the system, loss of energy (e.g., electric power, instrument air), or (q postulated adverse environments (e.g., extreme heat or cold, fire, pressure, steam, water, and radiation) are experienced. Evaluation Against Criterion 23 The Reactor Protection (trip) System is designed to fail into a safe state. Use of an independent channel for each actuator logic allows the system to sustain any logic channel failure without preventing other sensors monitoring the same variable from initiating a scram. Any two-out-of four logic channel trips initiate a scram. Intentional bypass for maintenance or testing causes the scram logic to revert to two out-of-three. A failure of any one reactor protection input or subsystem component produces a trip in one channel. This condition is insufficient to produce a reactor scram, but the system will perform its protective function upon trip of another channel. The environmental conditions in which the instrumentation and equipment of the reactor protection must operate were considered in establishing the component specifications. Instrumentation specifications are based on the worst expected ambient conditions in which the instruments must operate. The failure mode of the Reactor Protection (trip) System are such thatit fails into a safe state as required by Criterion 23. O V 3.1-25 Conformance with NRC GeneralDesign Criteria - Amendment 1
t l 25A5113 Rev. A SBWR standardsaktyAnalysis aeport O For further discussion, see the following sections: Chapter / l Section Title l 1.2 General Plant Description 7.2 Reactor Trip System 3.1.3.5 Criterion 24-Separation of Protection and Control Systems Criterion 24 Statement The protection system shall be separated from control systems to the extent that failure j of any single control system component or charmel or failure or removal from service ! of any single protection system component or channel which is common to the control i and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the l protection and control systems shall be limited to assure that safety is not significantly I impaired. Evaluation Against Criterion 24 There is separation betwi-en the Reactor Protection System and the process control systems. Logic channel and acmator logics of the Reactor Protection System are not used directly for automatic control of process systems. Sensor outputs may be shared, but each signal is optically isolated before entering a redundant or not ufety-related channel interface. Therefore, failure in the controls and instnunentation of process systems cannot induce failure of any portion of the protective system. Scram reliability is designed into the Reactor Protection System and hydrarlic control unit for the control rod drive. 'Ihe scram signal and mode of operation override all other signals. ; The systems that isolate containment and the reactor pressure vessel are designed so that any one failure, maintenance operation, calibration operation, or test to verify operational availability does not impair the functional ability of the isolation systems to j respond to safety-related variables. l l The protection system is separated fiom control systems as required in Criterion 24. 9 3.1-26 Conformance with NRC GeneralDes:yn Criteria - Amendment 1
25AS113 Rsv. A SBWR sondantsenrryAnalysisneport f r
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j For further discussion, see the following sections: l Chapter / Section Tide 1.2 General Plant Description 4.6 Functional Design of Fine Motion Control Rod Drive Systems 7.2 Reactor Trip System 7.3.1 Emergency Core Cooling System 7.3.3 Leak Detection and Isolation System 7.3.4 Safety System Logic and Control 7.4.5 Alternate Rod Insertion 7.5.3 Process Radiation Monitnring System 7.7.2 Rod Control and Information System 3.1.3.6 Criterion 25- Protection System Requirements for Reactivity Control Malfunctions Criterion 25 Statement The protection system shall be designed to assure that specified acceptable fuel design limits are not exceeded for any single malfunction of the reactivity control systems, such as accidental withdrawal (not ejection or dropout) of control rods. Evaluation Against Criterion 25 The Reactor Protection System provides protection against the onset and consequences of conditions that threaten the integrity of the fuel barrier and the reactor coolant pressure boundary. Any monitored variable which exceeds the scram setpoint will initiate an automatic scram and not impair the remaining variables from bemg l monitored, and if one channel fails, the remaining portion shall function. { The Rod Control and Information System (RC&IS) is designed so that no single failure can negate the effectiveness ofa reactor scram. The circuitry of the RC&IS is completely independent of the circuitry controlling the scram valves. This separation of the scram and normal rod control functions prevents failures in the reactor normal circuitry from affecting the scram circuitry. Because one or two control rods are controlled by an individual hydraulic control unit, a failure that results in continued energizing of an y insert solenoid valve on a hydraulic control unit can affect, at most, two control rods. The effectiveness of a reactor scram is not impaired by the malfunctioning of any one hydraulic control unit or two control rods. Contormance with NRC GeneralDesign Criteria - Amendment 1 3.1-27 l
25A5113 R:v. A l SBWR sundaniSafety Analysis Report l 9 The design of the protection system assures that specified acceptable fuellimits are not exceeded for any single malfunction of the reactivity control systems as specified in i "ritedon 25. l l For further discussion, see the following sections: ! Chapter / Section Title ( 1.2 General Plant Descdption 4.3 Nuclear Design ! 4.4 Thermal and Hydraulic Design 4.6 Functional Design of Fine Motion Control Rod Drive Systems 7.2 Reactor Trip System 7.7.2 Rod Control andInformation System 15 AccidentAnalyses 3.1.3.7 Criterion 26-Reactivity Contro! System Redundancy and Capability Criterion 26 Statement Two independent reactivity control systems of different design principles shall be l provided. One of these systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions. Evaluation Against Criterion 26 Two independent reactivity control systems utilizing different design principles are provided. The normal method of reactivity control employs control rod assemblies which contain boron carbide (B 4C), hafnium or other approved material. Positive insertion of these control rods is provided by means of the control rod drive electrical and hydraulic system. The control rods are capable of reliably controlling reactivity changes during normal operation (e.g., power changes, power shaping, xenon burnout, normal startup and shutdown) via operator controlled insertions and withdrawals. The 3.1-28 Conformance with NRC Genc 11 Design Criteria - Amendment 1
2SAS113 Rev. A SBWR standant safety AnalysisReport p
\ l %s' control rods are also capable of maintaining the core within acceptable fuel design limits during anticipated operational occurrences via the automatic scram function.
The unlikely occurrence of a limited number of stuck rods during a scram will not adversely affect the capability to maintain the core within fuel design limits. The Control Rod Drive System is capable of maintaining the reactor core subcritical ! under cold conditions, even when the pair of the control rods of the highest worth controlled by a hydraulic control unitis assumed to stickin the fullywithdrawn position. 1 This shutdown capability of the Control Rod Drive System is made possible by designing , the fuel with burnable poison (Gd2 03) to control the high reactivity of fresh fuel. The circuitry for manual insertion or withdrawal of control rods is completely independent of the circuitry for reactor scram.This separation of the scram and normal rod control functions prevents failures in the reactor manual-control circuitry from j affecting the scram circuitry. Two sources of energy (accumulatw pressure and electrical power to the motors of fine motion cor.acl . od drives, FMCRDs) are available for control rod insertion over the entire range of reactor pressure (i.e., from operating conditions to cold shutdown). The design of the Cont:ol Rod Drive System includes rm appropriate margin for malfunctions such as stuck rods in the unlikely event that they (v) do occur. Control rod withdrawal sequences and patterns are selected prior to operation to achieve optimum core performance and, simultaneously, low individual 1 rod worths. The operating procedures to accomplish such patterns are supplemented by the Rod Control and Information System, which prevent rod withdrawals yielding a rod worth greater than permitted by the preselected rod withdrawal pattern. Because of , the carefully planned and regulated rod withdrawal sequence, prompt shutdown of the reactor can be achieved with the insertion of a small number of the many independent controf rods. A Standby Liquid Control System containing a neutron-absorbing sodium pentaborate solution is the independent backup system. This system has the capability to shut the reactor down from full power and maintain it in subcritical condition at any time during the core life. The reactivity control provided to reduce reactor power from rated power to cold shutdown conditions with the control rods withdrawn in the power pattern accounts for the reactivity effects of the xenon decay, elimmation of steam voids, change in water density due to the reduction in water temperature, Doppler effect in uranium, change in the neutron leakage from boiling to cold, and change in the rod worth as boron affects the neutron migration length. The redundancy and capabilities of the reactivity control systems for the SBWR satisfy the requirements of Criterion 26.
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, LJ Conformance with NRC General Design Criteria - Amendment 1 3.1-29
l 25AS113 Rev. A SBWR sandantsareryAnalysis neport O For further discussion, see the following sections: l Chapter / j Section Title 1.2 General Plant Description 4.6 Functional Design of Fine Motion Control Rod Drive Systems 7.2 Reactor Trip Systems 7.3 Engineered Safety Features Systems l 1 7.4.1 Standby Liquid Control System 7.4.5. Alternate Rod Insertion 7.7.2 Rod Control and Information System 3.1.3.8 Criterion 27 - Combined Reactivity Control Systems Capability Criterion 27 Statement The reactivity control systems shall be designed to have a cornbined capability,in conjunction with poison addition by the emergency core cooling system of reliably controlling reactivity changes to assure that under postulated accident conditione and with appropriate margm for stuck rods the capability to cool the core is maintained. Evaluation Against Criterion 27 There is no credible event applicable to the SBWR which requires combined capability of the Control Rod Drive System and the Standby Liquid Control System (SLCS). The SBWR design is capable of maintaining the reactor core subcritical, including allowance for a pair of stuck rods controlled by a hydraulic control unit (HCU), without addition of any poison to the reactor coolant. The primary reactivity control system for the SBWR during postulated accident conditions is the Control Rod Drive System. Abnormalities are sensed, and, if protection system limits are reached, ccerrective action is initiated through automatic insertion of control rods. High integrity of the protection system is achieved through the combination of a arrangement, actuator redundancy, power supply redundancy, and physical separation. High reliability of reactor scram is further achieved by separation ofindividual HCUs controlling a pair of control rods and by fail safe design features built into the Control Rod Drive System. Response by the Reactor Protection System is prompt and the total scram time is short. In the very unlikely event that more than one control rod fails to insert and the core cannot be maintained subcritical by control rods alone, the SLCS can be actuated to insert soluble boron into the reactor core. The SLCS has sufficient capacity to ensure 3.1 30 Conformance with NRC General Design Criteria - Amendment 1
l 25AS113 R v. A SBWR saadard saktyAnalysis Report p\ t O that the reactor can always be maintained subcritical; and, hence, only decay heat will be generated by the core which can be removed by the appropdate decay heat removal systems (e.g., Isolation Condenser System), thereby ensuring that the core will always be coolable. 2 The design of the reactivity control systems assure reliable control of reactisity under postulated accident conditions with appropriate margin for stuck rods. The capability to cool the core is maintained under all postulated accident conditions; thus, Criterion 27 is satisfied. For further discussion, see the following sections: 1 Chapter / Section Title 1.2 General Plant Description 4.3 Nuclear Design 4.4 Thermal and Hydraulic Design
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v) 4.6 Functional Design of Fine Motion Control Rod Drive Systems 7.2 Reactor Trip System 7.4.1 Standby Liquid ControlByg. m 15 AccidentAnalyses 1 3.1.3.9 Criterion 28-Reactivity Limits Criterion 28 Statement The reactivity control systems shall be designed with appropriate linits on the potential amount and rate of reactivity increase to assure that the effects of pestnlated reactisity accidents can neither (1) result in damage to the reactor coolant pressure loundag greater than limited local yielding nor (2) sufficiently disturb the core, its support stnictures, or other reactor pressure vessel intemals to impair significantly the capability to cool the core.These postulated reactivity accidents shall include consideration of rod ejection (unless prevented by positive means), rod dropout, steam line rupture, changes in reactor coolant temperature and pressure, and cold water addition. Evaluation Against Criterion 28 The combined features of the Control Rod Drive System and the Rod Control and j [r] V Information System (RC&IS) designs incorporate appropriate limits on the potential l amount and rate of reactivity increase. Control rod withdrawal sequences and patterns l Conformance with NRC General Design Criteria - Amendment 1 3.1-31
25A5113 ReJ. A SBWR standantSafety Analysis Report are selected to achieve optimum core performance and low individual rod uorths. The RC&lS prevents any withdrawal other than the preselected rod withdrawal pattere.. The RC&IS function assists the operator with an effective backup control rod monitor.ng routine that enforces adherence to established startup, shutdown and power opera dons control rod procedures. The control rod drive mechanical design incorporates a passive brake and hydraulic inlet check valve which, individually, prevent rapid rod ejection. The brake spring holds the rod in position if there is a break in the FMCRD primary pressure boundary. The check valve prevents rod ejection if there is a failure of the scram insert line. The FMCRD includes a separation switch that detects when withdrawal of a stuck control rod is being attempted and stops rod motion. Normal rod movement and the rod withdrawal rate is limited through the fine motion control motor. The accident analyses (Chapter 15) evaluates the postulated reactivity accidents, as well as abnormal operational transients, in detail. Analyses are included for rod drop, steam line break, changes in reactor coolant temperature and pressure, and cold water addition. The initial conditions, assumptions, calculational models, sequences of vents, and anticipated results of ear _h postulated occurrence are covered in detail. The results of these analyses indicate that none of the postulated reactivity transients or accidents results in damage to the reactor pressure vesselinternals, so that the capability to cool the core is notimpaired. The design features of the RC&lS which limit the potential amount and rate ofreactivity increase ensure that Criterion 28 is satisfied for all postulated reactivity accidents. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 3.9.4 Control Rod Drive System ; 1 3.9.5 Reactor Pressure VesselInternals 4.3 Nuclear Design l l 4.5.1 Control Rod Drive System Structural Materials I 1 4.6 Functional Design of Fine Motion Control Rod Drive Systerns j l 3.1-32 Conformance with t,,'.: General Design Criteria - Amendment 1 l
i 25M113 Rev. A SBWR samtantsarayAurysis neport b l l Chapter / j Section Title (Continued) 5.2.2 Overpressure Protection 5.3 Reactor Vessel ; I 5.4.4 Mean Steamline Flow Restrictors 5.4.5 Mean Steamline Isolation System i 7.7.2 Rod Control and Information System l 15 AccidentAnalpes 3.1.3.10 Criterion 29- Protection Against Anticipated Operational Occurrences Criterion 29 Statement The protection and reactivity control systems shall be designed to assure an extremely l high probability of accomplishing their safety functions in the event of anticipated c ( operational occurrences. Lb Evaluation Against Criterion 29 f The high functional reliability of the Reactor Protection (trip) System and reactivity con trol system is achieved through the combination ofIc g'.c arrangement, redundancy, i physical and electrical independence, functional separation, fail-safe design, and ; inservice testability.These design features are discussed in detailin Criteria 21,22,23, l 24, and 26. An extremely high reliability of timely response to anticipated operational occurrences is maintained by a thorough program ofinservice testing and surveillance. l Safety-related components, such as control rod drives, Reactor Protection System components, etc., are testable during normal reactor operation. Functional testing and calibration schedules are developed using available failure rate data, reliability analyses, and operating experience. These schedules represent an optimization of protection and reactivity control system reliability effects during individual component testing on the portion of the sptem not undergoing test. The capability for inservice testing i ensures the high functional reliability of protection and reactivity control systems if a reactor variable exceeds the corrective action setpoint. The capabilities of the protection and reactivity control systems to perform their safety I' functions in the event of anticipated operational occurrences satisfy the requirements ; i (~ of Criterion 29. I ! I i Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1-33 t i l
l 2SA5113 Rw. A l l SBWR standantsaretyAnalysis neport O' For further discussion, see the following sections: Chapter / l Section Title 1.2 General Plant Description 4.6 Functional Evaluation of Fine Motion Control Rod Drive Systems 7.2 Reactor Trip System 7.3 Engineered Safety Features Systems 15 AccidentAnalyses 3.1.4 Group IV- Fluid Systems 3.1.4.1 Criterion 30-Quality of Reactor Coolant Pressure Boundary Criterion 30 Statement Components which are part of the reactor coolant pressure boundary shall be designed, fabricated, erected, and tested to the highest quality standards practical. Means shall be provided for detecting and, to the extent practical, identifying the location of the source of reactor coolantleakage. Evaluation Against Criterion 30 By utilizing conservative design practices and detailed quality control procedures, the pressure retaining components of the reactor coolant pressure boundary (RCPB) are designed and fabricated to retain their integrity during normal and postulated accident conditions (Subsection 3.1.2.5). Accordingly, components which comprise the RCPB l are designed, fabricated, erected, and tested in accordance with recognized industry codes and standards listed in Chapter 5 and Table 3.2-1. Further product and process quality planning is provided as described in Chapter 17 to assure conformance with the applicable codes and standards, and to retain appropriate documented evidence venfymg compliance. Because the subject matter of this criterion deals with aspects of the RCPB, further discussion on this subject is treated in the response to Criterion 14. Means are provided for detecting leakage in the RCPB. The Izak Detection and Isolation System (LD&IS) consists of sensors and istruments to detect, annunciate, and, in some cases, isolate the RCPB from potentially hazardous leaks before predetermined limits are exceeded. Small leaks are detected by temperature and pressure changes, increased frequency of sump pump operation, and increased airborne radioactivity. In addition to these means of detection, large leaks are detected by changes in flow rates in process lines, and changes in reactor water level. The allowable leakage rates have been based on the predicted and experimentally 3.1-34 Conformance with NRC GeneralDesign Criteria - Amendment 1
25AS113 Rev. A SBWR senadardsonrtyAnalysis Report A U determined behavior of cracks in pipes, the ability to makeup the RCS, the normally expected background leakage due to equipment design, and the detection capability of the various sensors and 'actruments. The RCPB and the LD8dS are designed to meet requirements of Criterion 30. For further discussion, s(e the following sections: Chapter / Section Title 1.2 General Plant Description 3.2 Classification of Structures, Systems, and Components 5.2.2 Overpressure Protection 5.2.5 Reactor Coolant Pressure Boundary Leakage Detection 5.3 ReactorVessel - V 7.3.3 Leak Detection and Isolation System 7.7.1 Nuclear Boiler System 17 QualityAssurance 3.1.4.2 Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary Criterion 31 Statement The reactor coolant pressure boundag shall be designed with sufficient margin to assure that when stressed under operating, maintenance, testing; and postulated accident conditions (1) the boundary behaves in nonbrittle manner and (2) the probability of rapidly propagating fracture is minimized. The design shall reflect consideration of service temperatures and other conditions of the boundary material under operating, maintenance, testing, and postulated accident conditions and the uncertainties in determuung (1) anaterial properties, (2) the effect ofinadiation on material properties, (3) residual, steady-state, and transient stresses, and (4) size of flaws. Evaluation Against Criterion 31 gS Brittle fracture control of pressure. retaining ferritic materials is provided to ensure protection against nonductile fracture. To minimize the possibility of brittle fracture failure of the reactor pressure vessel, the reactor pressure vesselis designed to meet the requirements of ASME Code Section III. Conformance with NRC General Desigr. Criteria - Amendment 1 3.1-35
i l t 25A5113 Rav. A l \ SBWR standardSafetyAnalysis Report l l l Tne nil-ductility transition (NDT) temperature is defined as the temperature below i j which ferritic steel behaves in a brittle rather than ductile manner. The NDT i l temperature increases as a function of neutron exposure at integrated neutron j exposures greater than about 1x10" mt with neutron energies in excess of 1 MeV. ) l l The reactor assembly design provides an annular space from the outermost fuel l assemblies to the inner surface of the reactor vessel that serves to attenuate the fast l neutron flux incident upon the reactor vessel wall. This annular volume contains the l core shroud and reactor coolant. Assuming plant operation at rated power and j availability 100% of the plant lifetime, the cumulative neutron fluence at the inner surface of the vessel causes a slight shift in the transition temperature. Expected shifts in transition temperature during design life as a result of environmental conditions, such as neutron flux, are considered in the design. Operationallimitations assume that l NDT temperature shifts are accounted for in the reactor operation. l The reactor coolant pressure boundary is designed, maintained, and tested to provide adequate assurance that the boundary will behave in a non-brittle manner throughout l the life of the plant. Therefore, the reactor coolant pressure boundary is in conformance with Criterion 31. For further discussion, see the following sections: l Chapter / Section Title 3 Design of Structures, Components, Equipment, and Systems 5.2 Integrity of Reactor Coolant Pressure Boundary 3.1.4.3 Criterion 32-Inspection of Reactor Coolant Pressure Boundary Criterion 32 Statement Components which are part of the reactor coolant pressure boundary sh?H be designed to permit (1) periodic inspection and testing ofimportant areas and features to assess their structure and leaktight integrity, and (2) an appropriate material surveillance program for the reactor pressure vessel. Evaluation Against Criterion 32 The reactor pressure vessel design and engineering effort includes provisions for inservice inspection. Access tothe annulus between the shield wall and vessel plug removable panels in the insulation provide access for examination of the vessel and its appurtenances. Also, removable insulation is provided in the nuclear boiler system safety / relief valves, and on the main steam and feedwater systems extending out to and including the firstisolation valve outside containment. Inspection of the reactor coolant 3.1-36 Conformance with NRC GeneralDesign Criteria - Amendment 1
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4 t l l l 2SA5113 R:v. A SBWR senaderd sawy Aasinis neport i pressure boundary is in accordance with ASME Boiler and Pressure Vessel Code i Section XI. Section 5.2 defines the Insenice Inspection Plan, access provisions, and ! areas of restricted access. l Vessel matedal surveillance samples will be located within the reactor pressure vessel. The program will include specimens of the base metal, weld metal, and heat affected zone metal. The plant testing and inspection program ensure that the requirements of Critedon 32 will be met. For further discussion, see the following sections: Chapter / Section Title 3.9 Mechanical Systems and Components 5.2 - Integrity of Reactor Coolant Pressure Boundag (j 3.1.4.4 Criterion 33-Reactor Coolant Makeup l l Criterion 33 Statement l A system to supply reactor makeup for protection against small breaks in the reactor ! l coolant pressure boundary shall be provided. The system safety function shall be to i assure that specified acceptable fuel design limits are not exceeded as a result of reactor coolant loss due to leakage from the reactor coolant pressure boundag and rupture of small piping or other small components which are part of the boundag. The system l shall be designed to assure that for on-site electdc power system operation (assuming l l off site power is not available) and for off-site electric power system operation (assuming ! ! on-site power is not available) the system safety function can be accomplished using the piping, pumps, and valves used to maintain coolant inventory during normal reactor operation. Evaluation Against Criterion 33 with or without preferred power and with a loss of feedwater supply, makeup is provided by the Control Rod Drive (CRD) System, CRD System and Isolation Condenser System (for coolant inventory conservation), or ADS (blowdown) with GDCS operation. Safety-related makeup is provided for the complete range of makeup i i rates by the Gravity-Driven Cooling System (GDCS). For small breaks where depressurization of the reactor vesselis necessary to achieve GDCS flow, the Automatic l Depressurization Subsystem (ADS) of the Nuclear Boiler System operates to j (- depressunze the vessel. l I Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1-37 l
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25A5113 Rev. A SBWR standardsarety Analysis Report O The plant is designed with systems that provide ample reactor coolant makeup for protection against small leaks in the reactor coolant pressure boundary during anticipated operational occurrences and posta'ated accident conditions.The requirements of Criterion 33 are met with these systems. For further discussion, see the following sections: Chapter / Section Title 3.9.4 Control Rod Drive System 5.4.6 1 solation CondenserSystem 6.3 Emergency Core Cooling Systems 3.1.4.5 Criterion 34-Residual Heat Removal Critarion 34 Statement A system to remove residual heat shall be provided. The safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that, for on-site electric power system operation (assuming off-site power is not available) and for off-site electric power system operation (assuming on-site power is not available), the system safety function can be accomplished, assuming a single failure. Evaluation Against Criterion 34 The Isolation Condenser System (ICS) provides the means to remove decay heat and residual heat from the Nuclear Steam Supply Systems (NSSS) at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundag are not exceeded. The major equipment of the ICS consists of heat exchangers. The equipment is connected to the reactor by associated valves and piping, and controls and , instnzmentation are provided for proper system operation. ! Each ICS subloop is actuated simply by opening one of a pair of redundant, diverse dram line valves.Two of the three ICS subloops are adequate operating alone to remove i residual heat from the reactor core and to assure fuel and RCBP design limits are not
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exceeded following an NSSS isolation event.The ICS provides the capability to reliably ! remove decay heat and residual heat from the reactor as required by Criterion 34. l l 3,s.38 Conformance with NRC GeneralDesign Criteria - Amendment 1 j
2SA5113 Rev. A SBWR medardsaktyAnalysis Report L) b The design of the ICS meets the requirements of Criterion 34. For further discussion, see the following sections: Chapter / Section 'I"dle 5.4.6 Isolation Condenser System 7.4.4 Isolation Condenser System 15 AccidentAnalyses 3.1.4.6 Criterion 35-Emergency Core Cooling Criterion 36Staternent A system to proside abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts. L.) Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for l on4ite electric power system operation (assuming off-site power is not available) and for j off-site electric power system operation (assuming on4ite power is not available) the ! system safety funcdon can Oc accomplished, assuming a single failure. Evaluation Against Criten*on 35 The Emergency Core Cooling System consists of the followmg: a Gravity Driven Cooling System (GDCS); and I a Automatic Depressurization Subsystem (ADS).
'nie ECCS is designed to limit fuel cladding temperature over the complete spectnam of possible break sizes in the RCPB including the complete circumferential rupture of the largest pipe connected to the reactor pressure vessel. The SBWR ECCS does not rely on pumps, off4ite ac power, or standby diesel generators to accomplish its safety function.
The GDCS provides flow to the annulus region of the reactor through its own nozzles. It provides gravity-driven flowfrom three separate water pools located within the drywell q ,/ at an elevation above the active core region. It also provides water flow from the Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1 39
2SA5113 Rett. A / SBWR standard sak tyAutrsis Report O suppression pool to meet long-term post-LOCA core cooling requirements. Refer to Subsection 6.3.2 for a complete description of the GDCS.
'Ibe ADS provides reactor depressurization capability in the event of a pipe break which does not rapidly depressurize the reactor. The ADS is a function of the Nuclear Boiler System and is accomplished through the combined use ofsquib-type permanently-opening depressurization valves (DFVs) and nitrogen operated safety /reliefulves (SRVs).
The ADS operates as follows: when a confirmed low-low water level (Level 1) signr.1 is received and sealed-in to the ECCS iogic,8 safety / relief valves and 6 depressurization valves actuate in a sequence describe iin Subsection 6.3.3. This sequence of SRV and DPV openings ensures that the RPV is depressurized rapidly so as to allow GDCS initiation, prior to core uncovery. Results of the performance of the ECCS for the entire spectrum of reactor pressure boundary line breaks are discussed in Subsection 6.3.4, which provides an analysis to show that the ECCS conforms to 10CFR50 Appendix K. This analysis shas complete compliance with Criterion 35 with the following results: a Peak clad temperatures are well below the 1200 C (2200'F) NRC acceptable limit. m The amount of fuel cladding reacting with steam is well below the 1% acceptable limit. m The accident is terminated while the core is maintained in a coolable geometry. i a The core temperature is reduced and the decay heat can be removed for an l extended period of time. The SBWR ECCS is powered by the safety-related station batteries. The redundancy and capability of the on-site electrical power systems are presented in the evaluation against Criterion 17. The design of the ECCS, including the power supply, meets the requirements of Criterion 35. For further discussion, see the following subsections: Chapter,- Section Title 6.3 Emergency Core Cooling System 7.3 Engineered Safety Features Systems 1140 Conformance with NRC GeneralDesign Criteria - Amendment 1
25A5113 Rev. A SBWR standardsareryAnalysis neport ()
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v Chapter / Section Tide 8.3 On4ite Power Systems 15 Accident Analyses 3.1.4.7 Criterion 36-Inspection of Emergency Core Cooling System Criterion 36 Statement The Emergency Core Cooling System shall be designed to permit appropriate periodic inspection ofimportant coinponents, such as spray rings in the reactor pressure vessel, water injection nozzles, and piping, to assure the integrity and capability of the system. Evaluation Against Criterion 36 The ECCS discussed in Criterion 35 includes inservice inspection considerations. Removable plugs in the reactor shield wall and/or panels in the insulation are provided on the ECCS pipingin the drywell. ('s (w, ) During plant operations, the instrumentation valves, Mument piping, instrumentation, wiring, and other components that are outside the drywell can be visually inspected at any time. Components inside the drywell can be inspected when the drywellis open for access during outages. Portions of the ECCS which are part of the reactor pressure boundary are designed to specifications for inservice inspection to detect defects which might affect the cooling performance. Particular attention is given to the GDCS nozzles. Design of the reactor vessel and intemals for inservice inspection and the plant testing program ensure that the requirements of Criterion 36 will be met. For further discussion see the following subsections: Chapter / Section Title 4.1.2 Reactor Internal Components 5.2.4 Preservice and Inservice Inspection and Testing of Reactor Coolant Pressure Boundary 5.3 Reactor Vessel 6.3 Emergency Core Cooling Systems 3 c) Conformance with NRC GeneralDesign Criteria - Amendment 1 1141
2SAS113 Rev. A SBWR standant sareryAnalysis neport O 3.1.4.8 Criterion 37-Testing of Emergency Core Cooling System l Criterion 37 Statement The Emergency Core Cooling System shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity ofits components, (2) the operability and performance of the active components of the systera, and (3) the operability of the system as a whole and, under conditions as close to the design as practical, the performance of the full operational sequence that brings the system into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system. Evaluation Against Criterion 37 Each of the ECCS subsystems (ADS and GDCS) is designed to p _it periodic testing to assure operability and performance of active components of e :h system. q The ADS DPVs and the GDCS valves cannot be tested during power operation; selected actuators will be removed and test fired during refueling outages. The GDCS check valves can be functionally tested via dedicated test line connections every refueling outage. GDCS flow testing is conducted as part of preoperational testing. Provisions for flushing the GDCS injection lines and venturi within the GDCS injection nozzle are provided. The ECCS will be subject to periodic tests to verify the logic sequence thatinitiates ADS and the GDCS system. A periodic self-test of the logic circuitry is performed to verify operability. The requirements of Criterion 37 are met. For further discussions, see the following subsections: Chapter / Section Title l I 5.2.2 Overpressure Protection 6.3 Emergency Core Cooling Systems I f 7.3.1.1 Automatic Depressurization Subsystem i 7.3.1.2 Gravity-Driven Cooling System 16 Technical Specifications 3.1 42 Conforrnance with NRC General Design Criteria - Amendment 1
25A5113 Rev. A SBWR StandaniSanrty Analysis Report 3.1.4.9 Criterion 38-Containment Heat Removal Criterion 38 Staternent A system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated system, the containment pressure and temperature following any LOCA and maintain them at acceptably low levels. Suitable redundancy in components and features and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for on-site electric power system operation (assuming off4ite power is not available) ant for off-site electric power system operation (assummg on4ite power is not available) the system safety function can be accomplished, assuming a single failure. Evaluation Against Criterion 38 The contamment heat removal function is accomplished by the Passive Contamment Cooling System (PCCS). The PCCS provides sufficient decay heat removal post-LOCA, to assure that containment pressure never exceeds its design pressure. The PCCS consists of three, totally independent, closed-loop extensions of the containment. Each loop contains a heat exchanger (passive containment cooling condenser) that condenses steam on the tube-side and transfers heat to water in the ICS/PCCS pool which is vented to atmosphere. The ICS/PCCS pool is positioned above, and outside, the SBWR contamment (drywell). To assure availability, no valves are employed, thus precluding inadvertent isolation of the PCC heat exchangers. The PCCS loops receive a steamgas mixture supply directly from the drywell. PCCS flow is driven by the pressure difference created between the containment dryw.;it and the suppression pool during a LOCA. The PCCS does not require power supphes, sensors, control logic, power-actuated devices or operator actions to function. During normal plant operation, the PCCS loops are in " ready standby." The PCCS is designed to Quality Group B Requirements per RG 1.26. The system is designed as Seismic Category I per RG 1.29. The common pool that the PCC condensers share with the ICs of the Isolation Condenser System is an Engineered Safety Feature (ESF). This pool is designed such that no locally generated force (such as an IC tube rupture) can destroy its function. Protection requirements against mechanical damage, fire and flood apply to the common ICS/PCCS pool. Portions of the PCCS outside the con %1 ment are located in a subcornpartment of the safety-related ICS/PCCS pool to comp.'v with 10CFR50, Appendix A, Criteria 2 and 4. The FCC condenser will not fail in a manner that damages the safety-related ICS/PCCS pool because it is designed to withstand the induced dynamic loads, which are caused Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1-43
25A5113 Rev. A SBWR scandant sareryAnstysis aeport O l I by combined seismic, DPV/SRV or LOCA conditions in addition to PCC operating i loads. The PCCS provides the containment heat removal function required in Criterion 38. l For further discussion, see the following subsections: Chapter / Section Tide i 6.2.2 Passive Containment Cooling System 3.1.4.10 Criterion 39-Inspection of Containment Heat Removal System Criterion 39 Statement The Containment Heat Removal System shall be designed to permit appropriate periodic inspection ofimportant components, such as torus, sumps, spray nozzles, and I piping, to assure the integnty and capability of the system. Evaluation Against Criten~on 39 "Ibe PCC condenser is an extension of the containment (dgwell) pressure boundary and it is used to mitigate the consequences of an accident. This function classifies it as safety- related Engineered Safety Feature (ESF). The PCCS is designed to ASME Code Section III, Class II and Section XI requirements for design and accessibility ofwelds for inservice inspection apply to meet 10CFR50 Appendix A, Criterion 16. Ultrasonic testing tube-to-header welds and eddy current testing of tubes can be done with the PCC condenser in place. The Containment Heat Remon! System is designed to permit periodic inspection of major components. This design meets requirements of Criterion 39. For further discussion, see the following subsections: Chapter / Section Title 6.2.2 Passive Containment Cooling System 3.1.4.11 Criterion 40- Testing of Containment Heat Removal System Criterion 40 Statement The Containment Heat Removal System shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural and leaktight integrity ofits components, (2) the operability and performance of the active components of the system, and (3) the operability of the system as a whole, and, under 3.14 Conformance with NRC GeneralDesign Criteria - Amendment 1 l j
25A5113 Rev. A SBWR saadudsannyAnarrsisneport (s conditions as close to the design as practical, the performance of the full operational sequence that brings the system into operation including operation, of applicable portions of the protection system, the transfer between normal and emergency power sources, and the operation of the associated cooling water system. Evaluation Against Criterion 40 The containment heat removal function is accomplished by the Passive Contamment Cooling System. The PCCS is an extension of the containment system, and it will be periodically pressure tested as part of overall containment pressure testing (Subsection 6.2.6.2). Also, the PCCS loops can be isolated for individual pressure testing during maintenance. The intent of Criterion 40 is satisfied as follows: m The structural and leaktight integrity can be tested by periodic pressure testing. m Functional and operability testing is not needed because there are no active i components of the system. p a Performance testing during power operation is not feasible; however, the ) performance capability of the PCCS will be proven by full scale PCC condenser prototype tests at a test facility before their application to the plant contamment system design. Performance will be established for the range ofin<ontainment environmental conditions following a LOCA. Integrated containment cooling tests have been completed on a full height, reduced section test facility, and the results have been correlated with TRACG computer program analytical predictions; this computer program is used to show acceptable containment performance. The design of the testing of Containment Heat Removal System meets the requirements of Criterion 40. For er discussion, see the following subsections: Chapter / j Section Title 6.2.2 Passive Containment Cooling System 7.3.2 Passive Containment Cooling Systern 3.1.4.12 Criterion 41-Containment Atmosphere Cleanup Criterion 41 Statement Systems to control fission products, hydrogen, oxygen, and other substances which may be released into the reactor containment shall be provided as necessary to reduce, consistent with other associated systems, the concentration and quantity of fission products released to the emironment following postulated accidents, and to control the Conformance with NRC GeneralDesign Criteria - Amendment 1 11-45
2SAS113 Rev. A SBWR stanwsareryAnalysis neport O concentration of hydmgen or oxygen and other substances in the containment atmosphere following postulated accidents to assure that containment integrity is maintained. Each system shall have suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and contamment capabilities to assure that for on-site electric power system operation (assuming off-site poweris not available) and for off-site electric power system operation (assuming onsite power is not available) its safety function can be accomplished, assuming a single failure. Evaluation Against Criterion 41 1 Fission products, hydrogen, oxygen, and other substances released from the reactor are contained within the low-leakage containment. Except for bypass leakage, leakage from the containment after an accident enters the safety envelope, which is isolated on an accident signal and which contains, dilutes, and holds up leakage from the containment such that the dose guidelines of 10CFR100 are not exceeded. Contamment leakage that bypasses the safety envelope enters the reactor building.
- The containment is inerted with nitrogen during normal operation. A Flammability Control System controls post-accident hydrogen and oxygen levels with oassive autocatalvtic recombiners a re-ies of to prevent deflagration or detonation of hydrogen and oxygen, thus assuring that containment integrity is maintained.
These systems have sufficient redundancy to withstand a single failure. A - ope =ble frc. . either off eite er en cite pcrier sc=ces. Criterion 41 is satisfied. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 6.2.3 Safety Envelope Functional Design 6.2.5 Flammability Control System 6.5.3 Fission Product Control Systems and Structures 7 Instrumentation and Control Systems 8 Electric Power 15 AccidentAnalyses 1146 Conformance with NRC General Design Criteria - Amendment 1
25AS113 Rev. A SBWR samratsaktyAnalysisnerat i\ 3.1.4.13 Criterion 42-Inspection of Containment Atmosphere Cleanup Systems Criterion 42 Statement The Containment Atmosphere Cleanup Systems shall be designed to permit appropriate periodic inspection ofimportant components, such as filter frames, ducts, and piping to assure the integrity and capability of the systems. Evaluation Against Criterion 42 Except for components located in the containment, all components of the Flammability Control System can be inspected during norTnal plant operation at power. The components within the containment may be inspected during refueling and maintenance outages. The design of the system, therefore, meets the requirements of Critenon 42. i For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description l 6.2.3 Safety Envelope Functional Design l 6.2.5 Flammability Control System l 6.5.3 Fission Product Control Systems and Structures 6.6 Inservice Inspection of Class 2 and 3 Components l Instrumentation and Control Systems 7 8 Electdc Power 3.1.4.14 Criterion 43-Testing of Containment Atmosphere Cleanup Systems Criterion 43 Statement i The Containment Atmosphere Cleanup Systems shall be designed to permit appropriate pedodic pressure and functional testing to assure (1) the structural and leaktight integrity ofits components, (2) the operability and performance of the active components of the systems such as fans, filters, dampers, pumps, and valves and (3) the operability of the systems as a whole and, under conditions as close to the design as p practical, the performance of the full operational sequence that bdng the systems into operation, including operation of applicable portions of the protection system, the transfer between normal and emergency power sources, and th e operation of associated systems. Conformance with NRC GeneralDesign Criteria -Amendment 1 3.1-47
I I 25A5113 Rev. A SBWR standardsafety Anairsis Report O Evaluation Against Criterion 43 The design of the safety envelope is provided in Subsections 6.2.3 and 6.5.3. The isolation dampers can be periodically tested. The Flammability Control System can be periodically tested as described in Subsection 6.2.5. The design of the system, therefore, meets the requirements of Criterion 43. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 6.2.3 Safety Envelope Functional Design 6.2.5 Flammability Control System 6.5.3 Flssion Product Control System 7 Instrumentation and Control Systems 8 Electric Power 3.1.4.15 Criterion 44-Cooling Water Criterion 44 Statement A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided. The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for on-site electric ; power system operation (assuming off-site poweris not available) and for off4ite electric l power system operation (assummg on site power is not available) the system safety function can be accomplished, assuming a single failure. Evaluation Against Criterion 44 l The SBWR ultimate heat sink is the IC/PCC pool. In the event of a design basis accident, heat is transferred to the IC/PCC pool through the Passive Contamment i Cooling System (PCCS). The PCCS has no active components and requires no electrical motive power or control and instrtunentation functions to perfonn its safety-related function of transferring heat to the ultimate heat sink. Therefore, no credible single failure can prevent the PCCS from performing its safety-related function. 3,14 Contormance with NRC GeneralDesign Criteria - Amendment 1
l 25A5113 R;v. A SBWR StandardSaktyAnalysis Report
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vl The nonstfety-related Plant Service Water System, described in Subsection 9.2.1, and its supported plant systems provide cooling water dudng nonnal operation. 1 The requirements of Criterion 44 for heat transfer to the ultimate heat sink are met. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 6.2.2 Passive Containment Cooling System 9.2.1 Plant Service Water System 3.1.4.16 Criterion 45-Inspection of Cooling Water System Criterion 45 Statement The Cooling Water System shall be designed to permit appropriate periodic inspection i 7 ofimportant components, such as heat exchangers and piping, to assure the integrity
- (d% and capability of the system.
Evaluation Against Criterion 45 The IC/PCC pool is located outside containment and is accessible for periodic inspections. Dudng outages, the IC/PCC pool compartments can be dramed tc permit inspection of the condensers. PCCS piping inside containment can be inspected dudng outages (see the evaluation of Critenon 39). These features meet the requirements of Criterion 45. l For further discussion, see the following sections- ! l l l Chapter / Section Title 1.2 General Plant Descdption ! 6.2.2 Passive Containment Cooling System i 14 InitialTest Program o i I t. (- l Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1-49 l i
2SA5113 Rev. A SBWR standardsaretyAnalysis aeport O 3.1.4.17 Criterion 46-Testing of Cooling Water System Criterion 46 Statement The Cooling Water System shall be designed to permit appropriate periodic pressure and functional testing to assure (1) the structural leaktight integrity ofits components, (2) the operability and the performance of the active components of the system, and (3) the operability of the system as a whole and, under conditions as close to the design as practical, the perfonnance of the full operational sequence that brings the system into operation for reactor shutdown and for lossef-coolant accidents, including operation of applicable portions of the protcotion system and the transfer between normal and emergency power sources. Evaluation Against Criterion 46 Redundancy and isolation are provided to allow periodic pressure testing of the PCCS as a whole. As discussed in the evaluation of Criterion 44, the PCCS contains no active components; therefore, functional testing is not necessary. The pedodic inspections described in the response to Criterion 45 will verify system integrity (see the evaluation of Cdterion 40). The design of the system, therefore, meets the requirements of Criterion 46. For further discussion, see the following sections: Chapter / Section Title 1.2 General Plant Description 6.2.2 Passive Containment Cooling System 14 Initial Test Program 1 1 16 Technical Specifications 3.1.5 Group V- Reactor Containment 3.1.5.1 Criterion 50-Containment Design Basis l Criterion 50 Statement The reactor containment structure, including access openings, penetrations, and the Coctainment Heat Removal System, shall be designed so that the containment structee and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect consideration of (1) the effects of potential energy sources which have not been included in the determination of peak conditions, such as energy in steam generators 3.1-50 Conformance with NRC GeneralDesign Criteria - Amendment 1
r I 25AS113 Rev. A SBWR Standard Safety Analysis Report l/^N l\v) ( and, as required by Section 50.44, energy from metal-water and other chemical l reactions that may result from degradation but not total failure of emergency core l cooling functioning, (2) the limited experience and experimental data anilable for j defining accident phenomena and containment responses, and (3)the conservatism of ) the calculational model and input parameters. l l Evaluation Against Criterien 50 Design of the containment is hued on consideration of a full spectrum of postulated accidents which would result in the release of reactor coolant to the containment. These l accidents include liquid breaks, steam breaks, and partial breaks (both steam and liquid), as <iescribed in Subsection 6.2.1. The evaluation of the containment design will be based on enveloping the results of this range of analyses, plus provision for appropriate margins. The most limiting short-term and long-term pressure and temperature responses will be assessed to verify adequacy of the contairunent structure. The design of the contamment system thus meets the requirements of Criterion 50. For further discussion, see the following sections i; ) Chapter / Section Title 3.7 Seismic Design 3.8 Design of Seismic Category I Structures 6.2.1 Containment Functional Design 6.2.2 Passive Containment Cooling System 15 Accident Analyses 3.1.5.2 Criterion 51 - Fracture Prevention of Containment Pressure Boundary Criterion S1 Statement i The reactor containment boundary shall be designed with sufficient margm to assure that under operating, maintenance, testing, and postulated accident conditions (1) its ferritic materials behave in a nonbrittle manner and (2) the probability of rapidly i propagating fracture is mirumized. ! \ ! l The design shall reflect consideration of service temperatures and other conditions of l the containment boundary material during operation, maintenance, testing, and ) postulated accident conditions and the uncertainties in determining (1) material (v' ) properties, (2) residual, steady-state, and transient :; tresses, and (3) size of flaws. l Conformance with NRC GeneralDesign Criteria - Arnendment 1 1 1-51
l l 2SAS113 Rev. A i SBWR standardsareryAnalysis neport O' Evaluation Against Criterion 51 The containmentvessel (CV) is a reinforced concrete structure with ferritic parts, such as a liner and a removable head, widch are made of material that has a nil-ductility transition temperature of at least 16.7 C (30 F) below the minimum senice temperature. The CV is enclosed by and integrated with the reinforced concrete reactor building. The preoperational test program and the quality assurance program ensure the integrity of the contamment and its ability to meet all normal operating and accident requirements. The containment design thus meet the requirements of Criterion 51. For further discussion, see the following sections: Chapter / Section Title 3.8 Design of Seismic Category I Structures 17 QualityAssurance 3.1.5.3 Criterion 52- Capability for Containment Leakage Rate Testing Criterion 52 Statement The reactor containment and other equipment which may be subjected to containment test conditions shall be designed so that periodic integrated leakage rate testing can be conducted at contamment design pressure. Evaluation Against Criterion 52 The containment system is designed and constructed and the necessary equipment is provided to permit periodic integrated leak-rate tests d; ring the plant lifetime. The testing program is conducted in accordance with 10CFR50 Appendixj. The testing provisions provided and the test program meet the requirements of Criterion 52. I For further discussion, see the following subsection: l l Chapter / ) Section Title 6.2.6 Containment Irakage Testing 3.1-S2 Conformance with NRC General Design Criteria - Amendment 1
4 2SA5113 R2v. A SBWR Standant Safety Analysis Report (n
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3.1.5.4 Criterion 53 - Provisions for Containment Testing and inspection Criterion 53 Statement The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate suneillance program, and (3) periodic testing at containment design pressure of the leaktightness of penetrations which have resilient seals and expansion bellows. Evaluation Against Criterion 53 ) There are special provisions for conducting individual leakage rates tests on applicable penetrations. Penetrations are visually inspected and pressure tested for leaktightness at periodic intervals in accordance with 10CFR50 AppendixJ. The provistous made for protection testing meet the requirements of Criterion 53. For further discussion, see the following sections: Chapter / Section Title p) (v 3.8 Design of Seismic Category I Struc+ures 6.2.6 Containment Leakage Testing 3.1.5.5 Criterion 54-Piping Systems Penetrating Containment i l Criterion 54 Statement j Piping systems penetrating the primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety ofisolating these piping systems. Such piping systems shall be designed with a capability to test periodically the ! operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. Evaluation Against Criterion 54 Piping systems penetrating containment are designed to provide the required isolation and testing capabilities. These piping systems are provided with test connections to allow periodic leak detection tests as necessary to determine if valve leakage is within acceptable limits. The actuation test circuitry provides the means for testing isolation vahre operability as necessary to determine if operability is within acceptable limits. I V) The design and provisions made for piping systems penetrating containment meet the requirements of Criterion 54. Conformance with NRC GeneralDesign Criteria - Amendment 1 3.1-53
25A5113 Rev. A SBWR scamtanisaretyAnalysis aeport O 3.1.5.6 Criterion 55-Reactor Coolant Pressure Boundary Penetrating Containment Criterion 55 Statement Each line that is part of the reactor coolant pressure boundary and that penetrates the primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions for a specific class oflines, such as instrument lines, are acceptable on some other defined basis: (1) one locked closed isolation valve inside and one locked closed isolation valve outside containment; or (2) one automatic isolation valve inside and one locked closed isolation vah>e outside containment; or (3) one locked closed isolation valve inside and one automatic isolation uh*e outside containment (a simple check valve may not be used as the automatic isolation valve outside containment); or (4) one automatic isolation valve inside and one automatic isolation valve outside containment (a simple check valve may not be used as the automatic isolation valve outside containment). Isolation valves outside containment shall be located as close to the containment as practical and, upon loss of actuating power, automatic isolation vah es shall be designed ) to take the position that provides greater safety. l Other appropriate requirements to minimize the probability or consequences of an accidental rupture of these lines or oflines connected to them shall be provided as necessary to assure adequate safety. Determination of the appropriateness of these requirements, such as higher quality in design, fabrication, and testing, additional provisions for inservice inspection, protection against more severe natural phenomer.a. and additionalisolation valves and containment, shallinclude consideration of the population density, use characteristics, and physical characteristics of the site emirons. Evaluation Against Criterion 55 The reactor coolant pressure boundary, as defined in 10CFR50, Section 50.2, consists of the reactor pressure vessel, pressure-retaining appurtenances attached to the vessel, valves and pipes which extend from the reactor pressure vessel up to and including the outermost isolation valves. The lines of the reactor coolant pressure boundary which penetrate the containment have suitable isolation valves capable ofisolating the containment, thereby precluding any significant release of radioactivity. The specific design for the isolation of RCPB lines penetrating containment are described in Section 6.2A and the defined basis is provided for those arrangements that are different than described in Criterion 55. 3.1-54 Conformance with NRC General Design Criteria - Amendment 1
l 25AS113 Rev. A i ? l SBWR sondantsanyAndysis neport l I For further discussion, see the following sections: Chapter / Section Title 5.2 Integrity of Reactor Coolant Pressure Boundary 6.2A Containment Isolation System ( 7 Instrumentation and Control Systems 15 AccidentAnalyses 5 16 Technical Specifications 3.1.5.7 Criterion 56-Primary Containment isolation Qiterion 56 Statement Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with contamment isolation valves as follows, unless it can be demonstrated that the contamment isolation provisions for a specific d class oflines, such as instruments lines, are acceptable on some other defined basis: , (1) one locked closed isolation valve inside and one locked closed isolation valve outside contamment; or (2) one automatic isolation valve inside and one locked closed isolation valve # ! outside containment; or
. (3) one locked closed isolation valve inside and one automatic isolation valve outside containment (a simple check valve may not be used as the automatic l isolation valve outside containment); or ;
(4) one automatic isolation valve inside and one automatic isolation valve outside containment (a simple check valve may not be used as the automatic isolation , valve outside containment). l Isolation valves outside containment shall be located as close to the containment as i l practical and upon loss of actuating power, automatic isolation valves shall be designed l to take the position that provides greater safety. T Evaluation Against Criterion 56 O Lines penetrating containment and connecting directly to the containment I atmosphere are isolatable by one of the methods specified in Criterion 56 or are j exempted on a defined basis as described penetration-by-penetration in Section 6.2.4. :
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Conformance with NRC General Design Criteria - Amendment 1 1 1-55
2SAS113 Rett. A SBWR senadantsakw Analysis sepan O The manner in which the containment isolation system meets this requirement is discussed further completely in the following sections: Chapter / Section Title 6.2 4 ContainmentIsolation System 7 Instrumentation and Control Systems 15 Accident Analyses 16 Technical Specifications 3.1.5.8 Criterion 57-Closed System Isolation Valves Criterion 57 Statement Each line that penetrates prunary reactor containment and is neither part of the reactor coolant pressure boundary nor connected directly to the containment atmosphere shall have at least one contamment isolation valve which shall be either automatic, or locked ! closed, or capable of remote manual operation. This valve shall be outside the containmen t and located as close to the containment as practical. A simple check valve may not be used as the automatic isolation valve. Evaluation Against Criterion 57 Each line that penetrates the contamment and is not connected to the containment atmosphere and is not part of the reactor coolant pressure boundary has at least one isolation valve outside containment. Details demonstrating conformance with Criterion 57 are provided in the following subsection: Chapter / Section Title j 6.2.4 Containment Isolation Systems 3.1.6 Group VI- Fuel and Radioactivity Control l 3.1.6.1 Criterion 60 -Control of Releases of Radioactive Materials to the Environment Criterion 60 Statement i The nuclear power unit design shall include means to control suitably the release of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during normal reactor operation inclnding anticipated operational occurrences. Sufficient holdup capacity shall be provided for retention of gaseous and a l-56 Conformance with NRC General Design Criteria - Amendment 1
25AS113 Rev. A SBWR StandaniSafetyAnalysis Report / fm \ 'sv/ liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operational limitations upon the release of such efDuents to the environment. Evaluation Against Criterion 60 The SBWR is designed so that releases of radioactive materials, in their gaseous, liquid, and solid form are minimized. Gaseous releases come primarily from the turbine condenser offgas and the ventilation systems. Noble gas and iodine activity that enters the turbine offgas system is held by ambient charcoal beds. Ventilation releases are through the plant stack. The plant stack and the major streams feeding the plant stack are monitored by the process radiation monitoring system so that suitable action may be taken to avoid releases in excess of regulatory limits. Th e radwaste systems process liquid and solid wastes. Processes are provided to treat and package solid wastes, as required by applicable state and federal regulations. In addition, the SBWR liquid radwaste system can be operated in a mode where all non-detergent and non-chemical waste streams are treated to allow mnimum recycle to the condensate storage tank. This mode of operation would minimize releases of p radioactivity via the liquid or discharge pathway but would increase solid waste
) generated.
The radwaste system has significant hold-up capacity, both in waste collection tanks and in sample tanks containing processed water. This hold-up or surge capacity prosides the i plant operator flexibility in operations when deciding when and how to release efDuents ! to the emironment. l l For further discussion, see the followmg sections. , 1 Chapter / l Section Title 11.2 Liquid Waste Management System 11.3 Gaseous Woe Management System 11A Solid Waste Management System 11.5 Process and EfDuent Radiological Monitoring Instrumentation and Sampling Systems 12.2 Plant Sources
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% )1 Conformance with NRC GeneralDesign Crinria - Amendment 1 3.1-57
i 25A5113 RCv. A j SBWR standardsaretyAnalysis neport l l O 3.1.6.2 Criterion 61 - Fuel Storage and Handling and Radioactivity Control Criterion 61 Statement The fuel storage and handling, radioactive waste, and other sptems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropdate containment, confinement, and filtering systems, (4)with a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other r:sidual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventog under accident conditions. Evaluation Against Criterion 61 The spent fuel storage pool has adequate water shielding for stored spent fuel. Adequate shielding for transporting fuel is also provided in the intermediate pools between the vessel and spent fuel storage pool. Liquid level senson are installed to detectlow pool water level. The reactor building is designed to meet Regulatory Guide 1.13 criteria. The spent fuel storage pool is designed with no penetrations below the water level needed for adequate shielding at the operating floor. Anti. siphoning , provisions protect against dramng the spent fuel storage poolin the event of a line break. New fuel storage racks are located in the concrete fuel storage vault. No cooling or air filtering system is required. New fuel storage racks are also provided in the new fuel storage pool adjacent to the vessel cavity. These storage racks preclude accidental criticality (see evaluation agamst Criterion 62). The new fuel storage racks do not require any special insenice inspection and testing for nuclear safety purposes. l The non4afety-related Fuel and Aumhary Pool Cooling System (FAPCS) normally removes decay heat from fuel storage pools. In addition, cooling of the spent fuel pools can be backed up from one train of the non4afety-related RWCU/SDC System. Without the FAPCS or the RWCU/SDC System available, the safety-related method of cooling the spent fuel is to allow the spent fuel pools to boil. Sufficient pool water inventog is provided to permit boiling for at least 72 hours without makeup. After 72 hours, make-up water is provided using safety-related piping and portable water sources. l l l The fuel storage and handling system is designed to assure adequate safety under normal and postulated abnormal conditions. The design of these systems meets the requirements of Criterion 61. O 1 1-58 Conformance with NRC General Design Criteria - Amendment 1
l l 25AS113 Rev. A SBWR Standard Safety Analysis Report l V l For further discussion, see the following sections: l Chapter / Section Title l 9.1.3 Fuel and Auxiliary Pools Cooling System 11 Radioactive Waste Management System 12 Radiation Protection 3.1.6.3 Criterion 62 - Prevention of Criticality in Fuel Storage and Handling Criterion 62 Statement Cdticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. l Evaluation Against Criterion 62 l Appropriate plant fuel handling and storage facilities are provided to preclude i accidental criticality for new and spent fuel. Criticality in both the new fuel storage pool l (m) and dry new fuel storage vault is prevented by physical separation. Criticality in the spent fuel storage pool is prevented by presence of fixed neutron absorbing material. The new and spent fuel racks are Seismic Category I components. l l New fuel is placed in dry storage in the top-loaded new fuel storage vault. Elis vault contains a drain to prevent the accumulation of water. Spacing of fuel bundles in the l new fuel storage vault prevents an accidental critical array, even if the vault becomes flooded or subject to seismic loadings. After installation of the fuel channels, new fuel is stored in the wet new fuel storage pool. Spacing of the fuel in the new fuel storage pool prevents a cdtical array even in a seismic event. The center-to-center new fuel assembly spacing limits the ke g of the array to not more than 0.95 for new dry fuel. The spent fuel is stored under water in the spent fuel storage pool. A full array ofloaded spent fuel racks is designed to be subcritical, by at least 5%Ak. Neutron-absorbing material, as an integral part of the design,is employed to assure that the calculated ke g, including biases and uncertainties, will not exceed 0.95 under all normal and abnormal conditions.The abnormal conditions accounted for are an earthquake, accidental dropping of equipment, or impact caused by the hodzontal movement of the fuel handling equipment without first disengaging the fuel from the hoisting equipment. O ('~ Refueling interlocks include circuitry which senses conditions of the refueling equipment and the control rods. These interlocks reinforce the operational procedures that prohibit making the reactor critical. The fuel handling system is designed to Conformance with NRC General Design Criteria - Amendment 1 3.1-59
2SA5113 Rev. A SBWR standantsafery Analysis neport O provide a safe, effective means of transporting and handling and is designed to minimize the possibility of mishandling or maloperation. The presence of fixed neutron-absorbing material in the spent fuel storage, physical separation in the new fuel storage and the design of fuel handling systems precludes accidental criticality in accordance with Criterion 62. For further discussion, see the following section: Chapter / Section Title - 9.1 Fuel Storage and Handling 3.1.6.4 Criterion 63 - Monitoring Fuel and Waste Storage Criterion 63 Statement Appropriate systems shall be provided in the fuel storage and radioactive waste systems and associated handling areas to (1) detect conditions that may result in loss of residual heat removal capability and excessive radiation levels, and (2) initiate appropriate safety actions. Evalcation Against Crit. don 63 Fuel pool temperature and level are monitored as part of the Fuel and Auxiliary Pool Cooling System (FAPCS). High pool temperature or low skimmer surge tank level would signal the need for providing additional cooling. Area radiation monitors are provided as part of the Area Radiation Monitoring System, which monitors the operating / refueling floor for high radiation levels. The radwaste system has no active decay heat removal functions, since the decay heat from the activity in the inputs to radwaste is not sufficient to warrant concern. Radwaste building area radiation monitors are provided to protect against excessive personal exposure, and monitoring shipping container activity and surface radiation levels to meet appropriate waste form and transportation criteria. For further discussion, see the following subsections: Chapter / Section Title 5.4.8 Reactor Water Cleanup / Shutdown Cooling System 9.1.3 Fuel and Aimhary Pools Cooling System 9.2.6 Condensate Storage and Transfer System 3.1-60 Conformance with NRC General Design Criteria - Amendment 1
2SAS113 Rsv. A SBWR saadantsannyAnarrsis nepet 1 ( Chapter / l Section Title 11 Radioactive Waste Management System 12 Radiation Protection 3.1.6.5 Criterion 64-Monitoring Radioactivity Releases Criterion 64 Statement Means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation oflossef<oolant accident fluids, efDuent discharge paths, and the plant emirons for radioactivity that may be released from normal operations, including anticipated operational occurrences and from postulated accidents. Evaluation Against Criterion 64 Means have been provided for monitoring radioactivity releases resulting from normal and anticipated operational occurrences and from postulated accidents. The following O releases are monitored: a gaseous releases; and a liquid discharge. In addition, the containment atmosphere is monitored. [ f For further discussion of the means and equipment used for monitoring reactivity releases, see the following sections: Chapter / Section Tide r 5.2.5 Reactor Coolant Pressure Boundary leakage Detection l l 7.5 Information Systems Important to Safety l 11.2.3 EfBuent Monitoring and Sampling 11.5 Process and EfBuent Radiological Monitoring and Sampling Systems 3.1.7 COL License information . O
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l l l Conformance with NRC GeneralDesign Criteria - Amendment 1 1 1-61 l l
25A5112 Rav. A SBWR steadas safetyAnalysis neport O 3.1.8 References None. O 4 4 l 1 0 1 1-62 Conformance with NRC General Design Criteria - Amendment 1
A i I i 25A5113 Rev. A l SBWR standsidsanrtyAnarrsisneport if% 3.2 Classification of Structures, Systems, and Components SBWR structures, systems, and components are categodzed as safety-related or non-safety-related. The safety-related structures, systems, and components are those relied upon to remain functional during or following a design basis event (as defined in 10CFR50.49) to ensure the: a integrity of the reactor coolant pressure boundarf, a capability to shut down the reactor and maintain it in a safe condition; and a capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines of 10CFR100. Safety-related structures, systems, and components conform to the quality assurance requirements ofAppendix B to 10CFR50. Non-safety-related structures, sptems, and cornponents have quality assurance requirements applied commensurate with the importance of the item's function. The quality assurance program is described in l Chapter 17. lf j Specific requirements for seismic design and quality group classifications are identified for these SBWR items commensurate with their safety classification. Table 3.2-1 l identifies these classifications for SBWR structures, sptems, and components. 3.2.1 Safety Classification Safety-related structures, systems, and components of the SBWR are designated Q in Table 3.2-1. The safety classification of SBWR structures, systems and components is consistentwith ANS-58.14. Table 3.2-2 outlines minimum design requirements for various safety classes. Where possible, reference is made to accepted industry codes and standards that define design requirements commensurate with the safety-related function (s) to be performed. Structures, systems and components that have no safety-related functions are classified as nonsdety-related and designated N. l 3.2.2 System Quality Group Classification 1 l NRC Regulatory Guide 1.26, Quality Group Classifications and Standards for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power Plants (Reference 3.2-2), describes a quality group classification method for fluid systems and relates it to industry codes. Items are classified by Quality Group A, B, C, or D, as b indicated in Table 3.2-3. Table 3.2-3 tabulates the design and fabdcation requirements for each quality group, as defined in Regulatory Guide 1.26. Classi6 cation of Structures. Systems, and Components - Amendment 1 3.21
i l 2SAS113 Rev. A j SBWR saadasssarty Analysis neport 1 O Table 3.2-1 shows the quality group classifications for SBWR components. Although not within the scope of the regulatory guide, the component supports and containment boundaries that are within the scope of ASME Code, Section III, are assigned quality group classifications in accordance with Table 3.2-2. The defmitions of the Quality Groups provided in Regulatory Guide 1.26 are not directly applicable to the SBWR design. The following definitions in this section, which are based on Section 6 of ANS Standard 58.14, are consistent with the definitions in Regulatory Guide 1.26 but have been modified to be applicable to the SBWR design. 3.2.2.1 Quality Group A Quality Group A (QGA) applies to pressure-retaining portions and supports of mechanicalitems that form part of the reactor coolant pressure boundary (RCPB) and whose failure could cause a loss of reactor coolant in excess of the reactor coolant normal makeup capability. These items are designed to meet the ASME Boiler and Pressure Vessel Code, Section III. Remaining portions of the RCPB are classified in accordance with Subsection 3.2.2.2. 3.2.22 Quality Group B Quality Group B (QGB) applios to pressure-retainmg portions and supports of containment and other mechanical items, requirements for which are within the scope of ASME Boiler and Pressure Vessel Code, Section III. These items are not assigned to QGA and are relied upon to accomplish one or more of the following safety-related functions: a Maintain pressure integrity of RCPB items that are not QGA. m Dming or following design basis accidents whose consequences could result in potentral offsite exposures comparable to the guidelines of 10CFR100. These items include those that: 1
- Maintain pressure integrity of the containment, containment isolation, or i extension of containment. l l
1
- Maintain pressure integrity ofitems that are (1) exterior to the containment; 1 (2) communicate with the RCPB or contamment interior; and (3) are not isolated normally, cannot be automatically isolated, or are not isolated following design basis accident or transient.
- Maintain pressure integrity ofitems that provide emergency negative reactivity insertion (scram).
12-2 Classification of Structures, Systems, and Components - Amendment 1
l 2SAS113 R1v. A
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c 3.2.2.3 Quality Group C 1 i l 3 Quality Group C (QGC) applies to pressure-retaining portions and supports ofitems ! i that are not assigned to QGA or QGB, but (1) are within the scope of the codes and I standards defined on Table 3.2-3, and (2) are relied upon to accomplish safety-related functions. ! j 3.2.2.4 Quality Group D i
- Quality Group D (QGD) applies to pressure-retauung portions and supports ofitems that are not assigned to QGA or QGB, or QGC but (1) are within the scope of the codes and ,;tandards defined on Table 3.2-3, and (2) are subject to one or more significant i licensing requirements or commitments. These items include those that
a Process, extract, encase, or store radioactive waste. m Monitor radioactive effluents to ensure that release rates or total releases are within limits established for normal operation and design basis transients. 1 1 l v a Resist failure that could prevent any QGA, QGB, or QGC items from perfonning a safety-related function V 1 m Protect items necessary to attain or maintain safe shutdown following fire. 3.2.3 Seismic Classification i SBWR structures that must remain integral while systems and components (including l their foundations and supports) that must remain functional or retain their pressure l integrity in the event of a safe shutdown earthquake (SSE) are designated Seismic l Category I. These include all safety-related items and all fuel storage racks. The Seismic Category I structures, systems, and components are designed to withstand the appropriate seismic loads (as discussed in Section 3.7) in combination with other appropriate loads without loss of function or pressure integrity. The seismic classifications indicated in Table 3.2-1 are consistent with the guidelines of Regulatory Guide 1.29 (Reference 3.2-3). Structures, systems, and components that perform no safety-related function, but whose structural failure or interaction could degrade the functioning of a Seismic Category I item to an unacceptable level of safety or could result in incapacitating injury to occupants of the Main Control Room, are designated Seismic Category II. These items are designed to structurally withstand the effects of an SSE. V Structures, systems, and components that are not categorized as Seismic Category I or II are designated Seismic Category NS. Classification of Structures, Systems, and Components - Amendment 1 3.2-3
2SA5113RCv. A SBWR standardsaretyAnalysis neport O 3.2.4 COL License Information None. 3.2.5 References 3.2-1 ANS 58.14, Safety and Pressure Integrity Classification Criteria for Light Water Reactors. 3.2-2 NRC Regulatory Guide 1.26, Quality Group Classifications and Standards for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power Plants. 3.2-3 NRC Regulatory Guide 1-29, Seismic Design Classification. O O 3.24 Classification of Structures, Systems, and Components - Amendment 1
(m r\ 9 Table 3.2-1 Classification Summary # 3 9 f Safety QA Seismic Qualig Req.8Categorys $ g Principal Component1 Desig.2 Location 8 Group Notes W g B NUCLEAR STEAM SUPPLY SYSTEMS k B11 Reactor Pressure Vessel System , E j 1. Reactor pressure vessel O CV A B I { o
- 2. Reactor vessel appurtenances-reactor coolant pressure boundary Q CV A B I 3- (RCPB) portions E 3. Control rod drive (CRD) housing and Q CV A B 1 9 in-core housing b
g 4. Control rc,Js O CV - B I
$ 5. Standby Liquid Control System O CV B B l l (SLCS) header and spargers g b oi l 6. Reactor vessel support skirt and Q CV A B I 3 & stabilizer p B
i 7. Other safety-related reactor Q CV - B I b Internals, including core support structures (Subsection 3.9.5)
- 8. Reactor internals - non-safety- N CV -
E 11 related components (Subsection 3.9.5) B21 Nuclear Boiler System (NBS)
- 1. Level instrumentation condensing Q CV A B i n.
I chambers E
- 2. Safety / relief valves (SRVs) ant' O CV A B i depressurization valves (DP\'
- 3. Safety / relief discharge piping Q CV C B- 1- -
(including supports) E-E e a
. . = _ _ _ _ _ _ _ _ _ . _ -- ___ _ _ _ _ _ _ _ - - - __
- t Table 3.2-1 Classification Summary (Continued) m tg Safety QA Seismic Principal Component 1 Qualig Req.s Category Desig.2 Location 8 Group s Notes
- 4. Nitrogen accumulators (for ADS and O CV C B I manual actuation of SRVs)
- 5. Piping and valves (including Q CV,RB A B I supports) for main steam (MSL) and feedwater (FW) lines up to and including the outermost containment isolation valves G. Piping (including supports) for MSL Q RB B B I from outermost isolation valve to and including seismic interface restraint and FW from outermost p isolation valve to the seismic 3 interface restraint including the y
@ shutoff valve 5 k 7. Piping (including supports) for MSL N RB,TB B B NS Main Steam Lines -The main steam lines from k 2, from the seismic interface restraint the containment outboard isolation valves and .I E to the turbine stop valve all branch lines 2-1/2 inches in diameter and
$ larger, up to and including the first valve
$ (including lines and valve supports) are
'$ designed by the use of an appropriate dynamic 1 seismic system analysis to withstand the safe ij shutdown earthquake (SSE) design loads in 7 combination with other appropriate loads, E, within the limits specified for Class 2 pipe in the I
& ASME Code, Section Ill. The mathematical h 4 model for the dynamic seismic analyses of the il j main steam lines and branch line piping includes the turbine stop valves and piping to the turbine g
g ( l casing. The dynamic input loads for design of g h the main steam lines are derived from a time g
!' history model analysis or an equivalent method as described in Section 3.7.
g.
=
$ 4
- R O O O
t
'D (v
r J . O(~ , p Table 3.2-1 Classification Summary (Continued) to a- W g Safety Oualig OA Seismic 3 Group Req.8 Categorys Desig.2 Location Notes g
- g. Principal Componenti g k 8. Piping and valves (including N RB D B l
!? supports) from FW shutoff valve to
[c the seismic interface restraint j 9. Pipe whip restraints - MSlJFW if Q CV,RB - B NS Pipe Whip Restraints-The need for pipe whip m needed restraints on the main steam line (MSL) and 3 y feedwater (FW) piping is determined by a " leak-J before-break" evaluation. This is described , g further in Subsection 3.8.3. h 10. Main steam drain piping and valves Q CV,RB A B i (7) j (including supports) within y outermost containment isolation g valves [ g
- 11. RPV head vent piping and valves (including supports) to the main Q CV A B 1 f
g steam line and to the second y g isolation valve s a b
; 12. Piping (including supports) for main N RB B B NS steam drains beyond outermost MSL isolation valves up to and including first drain isolation valve
- 13. Piping and valves (including N RB,TB D E NS supports) for main steam drains beyond outermost MSL isolation I valves downstream of first drain [
isolation valve it
- 14. Piping (including supports) for Q CV,SE B B I (7) h instrumentation up to and including 4 first instrument isolation valve A'
~ ,
t Table 3.2-1 Classification Summary (Continued) (4 to 03 Safety Quality QA Seismic Desig.2 Location 8 Group4 Req.8 Categorys Notes Principal Componenti -
- 15. Piping and valves (including N CV,SE D E NS (7) supports) for instrumentation downstream of first instrument isolation valve
- 16. Other mechanical modules with O CV,SE, -
B l safety-related function CE
- 17. Other electrical modules, cable, and Q CV,SE, - B 1 instrumentation with safety-related CE function B32 Isolation Condenser System (ICS) p 1. Piping and valves (including Q CV A B I 3 supports)inside containment y pI between reactor and the G
.? containment penetration 3 = :o 2, 2. Isolation condenser and piping Q RB B B i R h Y outside containment fi j
- 3. Vent piping and valves (including supports) to suppression pool Q CV,RB B B l m
3 4. Electrical modules and cable with Q CV,RB, - B I y safety-related function SE g { C CONTROL AND INSTRUMENT SYSTEMS 9 C11 Rod Control and information N SE,R B, - E NS $ j System (RC&lS) CE B. k
!3 C12 Control Rod Drive System (CRD) 4 1 1. CRD primary pressure boundary Q CV A B I >
b E
] 2. CRD internals O CV - B l {
- 3. Hydraulic control unit Q SE - B l (8)
- a O O e -
O ! J U} p Table 3.2-1 Classification Summary (Continued) M a W g Safety Qualig OA Seismic j
- g. Principal Component1 Desig.2 Location 3 Group Req.8 Category8 Notes !
(
!?
- 4. Piping including supports -insert line Q CV,SE B .B i 2
S 5. High pressure makeup piping Q SE B B 1 CRD piping classification is consistent with j including supports, the check valve, piping to which it connects. , e and the injection valve. I g 8. Piping and valves with no safety- N SE D E NS (7)
- related function (pump suction, (g pump discharge, drive header, and other piping not part of hydraulic j control unit)
- 7. CRD water pumps N SE D NS f E j 8. Fine motion drive motor N CV -
E NS fg 9. Electrical modules and cable with safety-related function O CV,SE, CE
- ' B I ::
y
! 10. ATWS equipment associated with N- SE -
E NS Anticipated Tiransients Without Scrcm (ATWS) $ 0 the Alternate Rod insert (ARI) Equipment- A quality assurance program that functions. meets or exceeds the guidance of NRC Generic Letter 85-06 is applied to all non-safety-related ATWS equipment. C21 Leek Detection and Isolation System (LD8tlS)
- 1. Electrical modules (temperature O CV,SE, -
B 1 g sensors, pressure transmitters, etc.)
- CE' g and cable with safety-related function
{
- 2. Other electrical modules and cable N CV,SE, -
E NS with no safety-related function .CE-C31 Feedwater Control System (FWCS) N CV,TB, - E NS I' . SE,CE,
- EB R {
o 6 R j
t; Table 3.2-1 Classification Summary (Continued) fa s gg
- i. Safety Quality QA Seismic Principal Component 1 Desig.2 Location 8 Group' Req.8 Category 8 Notes l C41 Standby Liquid Control System (SLCS)
- 1. Standby liquid control accumulator Q SE B B I including supports
- 2. Valves - injection Q SE A B I
- 3. Piping and valves (including Q CV,SE A B 1 (7) supports) between injection valves and reactor vessel
- 4. Piping and valves (including Q SE B B i (7) supports) upstream of injection p valves and downstream of automatic
- N2 makeup valve IR A
*i
$ S. N2 gas bottles and associated piping N RB,SE -
E NS E
$' up to automatic N2 makeup valve P h 6. Electrical modules and cable with Q SE,CE -
B I ( j safety-related function E B 7. Electrical modules and cable - others N SE.CE - E NS Anticipated Transients Without Scram (ATWS)
# Equipment- A quality assurance program that m
i meets or exceeds the guidance of NRC Generic li Letter 85-08 is applied to all non-safety-related
$= ATWS equipment.
P
=
C51 Neutron Monitoring System (NMS) h 4 1. Detector and tube assembly- Q CV B B 1 E'. E
]= primary pressure boundary f (g
3 y 2. Detector and tube assembly - Q CV C B 1 3 internals
- 3. Electrical modules and cable - Q CV,SE, -
B l j. 3 SRNM, LPRM, and APRM CE,RB as s 4
; E O O O
F O O l p Table 3.2-1 Classification Summary (Continued) m a a g Safety Quality QA Seismic Desig.2 Location3 Group" Req.8 Categorys g g, Principal Componenti Notes g CV,SE, NS i k C61 Remote Shutdown System (RSS) N E Sp This system controls components RB,CE that are included under C12, G31, a P21, P41, R10 & U40 5 y C62 Multiplexing System l
- 1. Electrical modules and cable with safety-related function Q SE.RB, CE
- B l E 2. Other electrical modules and cables N SE,RB, -
E NS 9 with no safety-related function CE,RW 4 SE,CE, B 8 C71 Reactor Protection System (RPS) Q - I g TB,RB l C74 Safety System Logic and Control Q SE,CE - B I $ D (SSLC) 3 w 2 E C82 Automatic Power Regulator System N CE - E NS @
! (APRS) [
C85 Steam Bypass end Pressure Control N CE - E NS System (SB&PC) C91 Process Computer System N CE & ALL - E NS D RADIATION MONITORING SYSTEMS P D11 Process Radiation Monitoring System (PRMS)
- 1. Radiation monitors, sensors, and Q CV,SE, - B i it other electrical modules and cable CE with safety-related function
- 2. Fission product monitoring piping Q CV,SE B B 1 D and valves (including supports) {R-forming part of the containment g boundary {
- a
l l t Table 3.2-1 Classification Summary (Continued) m gg s ! " Safety Quality QA Seismic
- Principal Component' Desig.2 Location8 Group4 Req.' Categorys Notes
- 3. Fission product monitoring system N CV,SE, -
E NS Process Radiation Monitoring System -Special (other portions) CE seismic qualification and quality assurance requirements are applied to the fission products monitoring system.
- 4. Other electrical modules and cable N ALL -
E NS with no safety-related function D21 Area Radiation Monitoring System N ALL, - E NS (ARMS) except CV D23 Containment Atmospheric Q CV,SE, - B l g Monitoring System (CAMS) CE I5 E CORE COOLING SYSTEMS 4 E k E50 Gravity-Driven Cooling System 8 (GDCS) g j 1. Piping and valves (including Q CV A B i ( g supports) connected with the reactor vessel, including the squib valves, 3 e and up to and including the check {N 2. valves upstream of the squib valves Piping and valves (including Q CV C B l
$ supports) from the check valves E upstream of the squib valves to the E suppression pool and GDCS pools $
{ Q CV C B l
% 3. Piping and valves (including
! supports) from the GDCS pools to h the lower drywell 4
{ E y 4. Electrical modules and cable Q CV,SE, - B 1 CE j, { i f
- s O O O
O O O p Table 3.2-1 Classification Summary (Continued) to a III g Safety Ouality OA 3 Group4 Req.5 Categorys Seismic g g. Principal Componenti Desig.2 Location Notes g k F REACTOR SERVICING EQUIPMENT f4 F11 Fuel Servicing Equipment N RB - E NS F12 Miscellaneous Servicing Equipment N RB - E NS NS (enF13 Reactor Pressure Vessel Servicing Equipment N RB - E NS f F14 RPVIntemal Servicing Equipment N RB - E E F15 Refueling Equipment 9 ll j 1. Fuel Handling platform N RB - E
- 2. Refueling Machine N RB E ll f
4 NS f 3. Refueling bellows N CV - E k f F16 Fuel Storage Facility g.
- 1. Fuel storage racks - new and spent N RB -
E I g h P. 2. Defective fuel storage devices N RB - E I F17 Under RPV Servicing Equipment N CV - E NS F21 CRD Maintenance Facility N RB - E NS F32 Fuel Cask Cleaning Facidty N RB - E NS F41 Plant Startup and Test Equipment N CV - E NS P= F42 Fuel Transfer System (FTS) N RB - E NS F43 Loose Parts Monitoring System N CV,RB,SE - E NS it. (LPMS) F51 In-Service inspection Equipment N CV,RB SE - E NS , E. c R a a
i t Table 3.2-1 Classificatior. Summary (Continued) m a
" Safety Quality Q# Selsmic W
Principal Component1 Desig.2 Location 3 Group' Req) Logory s Notes G DECAY HEAT REMOVAL NETWORK G21 Fuel and Auxiilary Pools Cooling System (FAPCS)
- 1. Independent line (including piping, O OO,RB C B l valves, and supports) for safety-related makeup
- 2. Piping and valves including supports Q CV,SE B B l between containment isolation valves (including valves)
- 3. Piping and valves including supports Q CV,SE BQ B l p from suppression pool suction a strainer to !nbcerd conteinment- y
!cct:tica vc!ve locked onen, k
[h maintenance valve inside drvwell 3
- n St. 4. Pioina of S/P suction line including Q CV.SE B B 1 i h
E valve and sucoorts from locked h ocen maintenance valve inside j drvwell to land including) outboard e isolation valve i 5. Piping and valves including supports N CV D - 11 g s between inboard containment E isolation valves and their g termination points inside = o containment, for
-GDCS pools saction line
{
!2 -GDCS pools return line h
-wetwell spray line 4
[{ -drywell spray line
-supproccion pee! return !!ne (
D t;- it. --re: der :!! & heed cavity drein i, 9 -- O -- O
p) c O'M l'h U p Table 3.2-1 Classification Summary (Continued) fa a W
% Safety Desig.2 Location Qualig QA Seismic 8 Group Req.s Categorys g
- h. Principal Component 1 Notes y o 6. Pioina and valves includina suonorts N QL Q = 1 h between inboard containment
[ isolation valves and their termina-g tion ooints inside containment. for P - suooression cool raturn ime {y. - reactor well and head cavity drain line
'h 7. Interconnecting piping between N CV D -
11 g GDCS pools O j 8. Piping and valves including supports NQ RB -R B I g between low pressure coolant a injection gate valve (including i> valve) up to the interface with Reactor Water Cleanup / Shutdown M h l Cooling System 3 f 9. All other mechanical modules and N SE,00, D E NS f h R piping, including normal makeup RB system components
- 10. Electrical modules and cables with Q R B,CE, -
B I safety-related function CV
- 11. Electrical modules and cables with N RB,CE -
E NS non-safety-related function P G31 Reactor Water Cleanup / Shutdown Cooling System (RWCU/SDC) =
- 1. Piping including supports and valves O CV,SE A B 1 (7) it, within and including outermost &
containment isolation valves on [ pump suction g
- 2. Piping including supports and valves Q SE B B l (7) fW from feedwater lines to and RWCU/SDC piping classification is consistent g including shutoff valves with piping to which it connects. {
h R
y Table 3.2-1 Classification Summary (Continued) to L 10
- Safety Seismic Qualig QA 3 Group Req.5 Categorys Principal Componenti Desig.2 Location Notes
- 3. Vessels including supports N RB C E NS (demineralizer)
- 4. Regenerative heat exchangers N RB C E NS (including supports) carrying reactor water
- 5. Cleanup recirculation pump, motors N RB C E NS
- 6. Other piping including supports and N SE,RB, C E NS (7) valves between containment TB isolation valves and shutoff valves at feedwater line connections (including lines to main condenser f and to radwaste management system) g i 7. Nonregenerative heat exchanger N RB C E NS 3 tube side and piping (including t' (g supports and valves) carrying
[ j process water c g 8. Nonregenerative heat exchanger N RB D E NS w shell and piping (including supports
} and valves) carrying cooling water NS 3 9. Sample station N RB D E k 10. Electrical modules and cable with Q RB,CE -
B : I safety-related function $ 3 { N RB,CE E NS
- 11. Electrical modules and cable with no -
!e safety-related function k 4
a w 2 a 3= 0 0 -. . - _ - _ _ - - - _ - _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ . 0
O O O j n Table 3.2-1 Classification Summary (Continued) (4 ! ii ID
% Safety Qualig QA Seismic
- h. Principal Component 1 Desig.2 Location 3 Group Req.5 Category8 Notes O
k H CONTROL PANELS E H10 Main Control Room Panels 5
- 1. Panels, electrical modules, and cable O CE B I Control Panels- Panels and associated
} # with safety-related function structures that support or house safety-related
{ g mechanical or electrical components are safety-related. h 2. Panels, electrical modules, and cable N CE - E 11 g with no safety-related function O j H11 MCR Equipment Room Panels O 2 1. Panels, electrical modules, and cable Q CE - B I Control Panels - Panels and associated
$ with safety-related function structures that support or house safety-related %
[ mechanical or electrical components are safety-related.
$ w S. 2. Panels, electrical modules, and cable N CE -
E 11 P f with no safety-related function [ H14 Radweste Control Room Panels N RW - E NS Radweste Management Systems- A quality assurance program meeting the guidance of NRC Regulatory Guide 1.143 is applied to radioactive waste management systems during design and construction. H21 Local Panels and Racks g
- 1. Panels, electrical modules, and cable - Q ALL -
B I Control Panels - Panels and associated h with safety-related function structures that support or house safety-related 31. mechanical or electrical components are safety- { related. 5
- 2. Panels, electrical modules, and cable N ALL -
E NS with no safety-related funct'on g{ n' J NUCLEAR FUEL 0 4= h a
t Table 3.2-1 Classification Summary (Continued) m a
- 03 Safety Quality QA Seismic Principal Componenti Desig.2 Location 3 Group4 Req.8 Categorys Notes
! J10 Core and Fuel Services No physical items to be classified J11 Nuclear Fuel O CV,SE - B l J12 Fuel Channel O CV,SE - B I K RADIOACTIVE WASTE MANAGEMENT SYSTEMS K10 Liquid Waste Management System (LWS)
- 1. Mechanical modules (including N RB,RW D E NS Radwaste Management Systems- A quality supports) (see assurance program meeting the guidance of note) NRC Regulatory Guide 1.143 is applied to n radioactive waste management systems during e design and construction. On fluid containing k portions this requires Quality Group D plus other E
g requirements. e i .: l 2. Electrical modules and cabling N RB,RW (see E NS Same as above. P y note) ( h K20 Solid Waste Management System j (SWS) { o
- 1. Mechanical modules (including supports)
N RB,RW D (see E NS See note for K10 item 1. 3 note) e E 2. Electrical modules and cabling N RB,RW (see E NS See note for K10 item 1. R
& note) h K30 Offges System (OG) N TB D E NS Offgas System- A quality assurance program ! (see meeting the guidance of NRC Regulatory Guide y note) 1.143 is applied to radioactive waste 3
3 management systems during design and g g construction. On fluid containing portions this 3 i requires Quality Group D plus other g g requirements. q b $ O _ O - O
/3 (' M C'\
C/ V U p Table 3.2-1 Classification Summary (Continued) (4 a til g Principal Component 1 Safety 8 Qualig QA Desig.2 Location Group Req.8 Seismic Category8 Notes g g R m E S 9 m 3* N POWER CYCLE SYSTEMS b g N11 Turbine Main Steam System (TMSS)
)n 1. Branch line of MSL including supports between the second N RB,TB B E NS Main Steam Lines -The main steam lines from j the containment outboard isolation valves and 3 isolation valve and the turbine stop all branch lines 2-1/2 inches in diameter and W valve from branch point at MSL to larger, up to and including the first valve fg and including the first valve in the branch line (including lines and valve supports) are designed by the use of an appropriate dynamic ,
8 seismic system analysis to withstand the safe C E. shutdown earthquake (SSE) design loads in P combination with other appropriate loads, $, within the limits specified for Class 2 pipe in the ASME Code, Section 111.The mathematical model for the dynamic seismic analyses of the main steam lines and branch line piping includes the turbine stop valves and piping to the turbine casing. The dynamic input loads for design of the main steam lines are derived frorn a time g history model analysis or an equivalent method = as described in Section 3.7. k
- 2. Other mechanical and electrical N TB D E NS g modules (
N21 Condensate and Foodwater System (C&FS)
- 1. Main feedwaterline (MFL) beyond N TB D E NS g.
seismic interface restraint no U 4
- s R
t; Table 3.2-1 Classification Summary (Continued) to 4 - Up Safety Quall OA Seismic Principal Component1 Desig.2 Location 8 Group] Req.5 Categorys Notes N22 Heater Drain and Vent System N TB - E NS (HDVS) N25 Condensate Purification System N TB - E NS (CPS) N31 Main Turbine N TB - E NS N32 Turbine Control System (TCS) N TB D E NS (9) N33 Turbine Gland Seal System (TGSS) N TB D E NS N34 Turbine Lubricating Oil System N TB - E NS (TLOS) p N35 Moisture Separator N TB - E NS k N36 Extraction System N TB E NS { P h N37 Turbine Bypass System (TBS) N TB B B NS Main Steam Lines -The main steam lines from i:t the containment outboard isolation valves and P h all branch lines 2-1/2 inches in diameter and [ [E larger, up to and including the first valve (including lines and valve supports), are
$ designed by the use of an appropriate dynamic 2 seismic system analysis to withstand the safe ) shutdown earthquake (SSE) design loads in 3 combination with other appropriate loads, g within the limits specified for Class 2 pipe in the g g ASME Code, Section Ill. The mathematical model for the dynamic seismic analyses of the g
g g o main steam lines and branch line piping includes
!it the turbine stop valves and piping to the turbine casing. The dynamic input loads for design of I
the main steam lines are derived from a time D fi history model analysis or an equivalent method h t as described in Section 3.7. { 4 i R l O - . - _ _ - _ _ - - - - _ - _ _ _ _ _ -.- -_ -_-.-- . O O . _ _ . - _ _ - _ _ . _ _ - _ _
n p Vp U. U
- p Table 3.2-1 Classification Summary (Continued) m
- 8 9 s Safety Desig.2 Location Quality QA Seismic 8 Group4 Req.8 Categorys Notes g
- h. Principal Component 1 g
( N39 Turbine Auxillary Steam System N TB - E NS Lo (TASS) N41 Generator N TB - E NS N42 Hydrogen Gas Cooling System N TB - E NS
} (HGCS) f N43 Generator Cooling System (GCS) N TB - E NS (p N44 Generator Sealing Oil System (GSOS)
N TB - E NS fg N45 Hydrogen and Carbon Dioxide Bulk Storage System N 00 - E NS i NS f* N51 Exciter N TB - E g N61 Main Condenser and Auxiliaries N TB - E NS $ w N71 Circulating Water System (CIRC) N TB D E NS y TB NS h a N39 Turbine Auxiliary Steam System N - E (TASS) P STATION AUXIUARY SYSTEMS P10 Makeup Water System (MWS) N OO,RW, D E NS SE,RB P21 Reactor Component Cooling Water g System (RCCWS) =
- 1. Piping and valves (including Q CV,SE B B I it.
supports) forming part of the containment boundary
- 2. Piping and valves inside N CV D E ll containment
- 3. Other mechanical and electrical N RB,SE D E NS
$ modules k u n
i t Table 3.2-1 Classification Summary (Continued) m b Safety % Quality QA Seismic Principal Component Desig.2 Location 8 Group' Req.s Category8 Notes P22 Turbine Component Cooling Water N TB D E NS System (TCCWS) P25 Chilled Water System (CWS) N TB,RB, D E NS CE P30 Condensate Storage and Transfar System (CS&TS)
- 1. Mechanical modules, including N OO,RB D E NS piping, valves, and condensate RW,TB storage tank
- 2. Electrical modules and cable N RB -
E NS f P32 Oxygen injection System (OIS) N TB - E NS , f P33 Process Sampling System (PSS) N RB,00, D E NS (7) h g- TB,RW, a l
~
SE P s
@ P41 Plant Service Water System (PSWS) NS h b
- 1. Mechanical and electrical modules, N SF,00, D E NS 3
e including piping and valves RB (including supports)
-{
y P51 Service Air System (SAS) N ALL D E NS { PS2 instrument Air System (IAS) N ALL D E NS g
& P54 High Pressure Nitrogen Supply System (HPNSS) (
- 1. Piping and valves (including Q CV,SE B B 1
!8 supports) forming part of the h containment boundary 4 I
D y 2. Other non-safety-related N SE,RB D E NS !t. g mechanical modules j.
! 5
- a O O O
% f d b g Table 3.2-1 Classification Summary (Continued) en 8
Safety Quality QA Seismic - W g Desig.2 Location 8 Group4 Req.8 Categorys Notes
- g. Principal Component 1
( 50
- 3. Other non-safety-related electrical modules N SE,R B, CE E NS h 4. Nitrogen storage bottles N RB - E NS a
P62 Auxiliary Boiler System (ABS) N OL - E NS
}
3 P63 Hot Water System (HWS) N ALL - E NS
; P73 Hydrogen Water Chemistry System N TB -
E NS g (HWC) 1 [
]
P91 Post Accident Sampling System (PASS) N SE,RB D E NS P95 Iron injection System tilS) N TB - E NS f. f I R STATION ELECTRICAL SYSTEMS R10 Electrical Power Distribution 3 E System (EPDS) P ; s . Q h - 1. Main transformers N 00 - E NS 3
- 2. Main generators N TB -
E NS t R11 Unit Auxiliary Transformers N RB - E NS R13 Isolated Phase Bus N 00 - E NS ' R21 Non-Segregated Phase Bus N OO,EB - E NS ' P R22 Metal Clad Switch Gear N SE,RB, - E NS g_ EB R23 Power Centers N- SE,RB, - E NS EB R24 Motor Control Centers N SE,R B, - E NS EB,CE R31 Raceway System $ b
- . _. . _ . _ _ _ _ _ _ _ _ . _ - _ _ _ _ _ - __l
t Table 3.2-1 Classification Summary (Continued) m k # Safety Quality QA Seismic Principal Component1 Desig.2 Location 3 Group4 Req.8 Categorys Notes
- 1. Conduit, cable trays and supports O CV,SE, - B l with safety-related function CE,RB
- 2. Other electrical modules with no N CV,SE, -
E NS safety function CE,RB EB,OL R33 Cable
- 1. Cable and supports with safety- O CV,SE, -
B I related function CE,RB
- 2. Other cable and supports with no N CV,SE, -
E NS safety function CE,RB p EB,OL R34 Plant Grounding System O OO - B 1 h $- R35 Electrical Penetrations Q CV,SE - B I 3
= :=
R R40 Standby AC Power Supply N EB - E NS I La b Q R42 Direct Current Power Supply E a 1. Electrical modules and cable with O CV,SE, - B i m safety-related function CE,RB
- 2. Other electrical modules and cable N CV,SE, -
E NS 3 with no safety function CE,RB, { EB,O L, g 9 3 $ R46 Vital (Uninterruptible) AC Power E. l= Supply 8 1. Electrical modules and cable with Q CV,SE, - B I [ safety-related function CE,RB D
- 2. Other electrical modules and cable N CV,SE, -
E NS h with no safety function CE,RB, $ i OL j
- =
0 - - 0 0
3 O O p Table 3.2-1 Classification Summary (Continued) e a ID m Safety QA Seismic p Principal Component1 Qualig Req.8 Category' Desig.2 Location 3 Group Notes g. ( R47 Instrumentation and Control Power Supply
- 1. Electrical modules and cable with no N CV,SE, - E NS safety function CE,RB R51 Communication System N ALL - E NS 3 R52 Lighting and Servicing Power j Supply
( 1. Lighting N ALL - E NS
- 2. Emergency lighting in control room O CE -
B l g S POWER TRANSMISSION SYSTEMS $ S11 Main Transformer - Not in scope. y L S21 Switch Yard Not in scope. k T CONTAINMENT AND ENVIRONMENTAL CONTROL SYSTEMS T10 Containment System f ~ [
- 1. Upper and lower drywell airlocks O CV,SE -
B l and equipment hatches, lower drywell access tunnels, suppression chamber access hatch, and safety-related instrumentation T11 Containment Vessel g
- 1. Drywell head . O CV B B 1 $
E
- 2. RCCV Q CV B B l
- 3. Reactor pedestal O CV B B I
- 4. Portion of basemat under Q CV B B l D RCCV/ pedestal f"
=
. T12 Containmentintomal Structures 4 { k
l y Tabis 3.2-1 Classification Summary (Continued) m gg
- Safety Seismic Qualig Req.8 8 Group QA Category s Principal Component i Desig.2 Location Notes
- 1. Reactor vessel stabilizer truss O CV - B 1
- 2. Support structures for safety-related Q CV - B I piping, including supports and equipment
- 3. Reactor shield wall O CV - B l
- 4. Diaphragm floor Q CV - B l
- 5. GDCS pool O CV - B i T15 Passive Containment Cooling Q CV,RB B B I System (PCCS) p T31 Containment Atmospheric Control System (CACS) h 1. Piping and valves (including supports) forming part of the Q SE B B 1 f:
*f 8 containment boundary s*
9., g 2. Electrical modules and cables with O SE,CE - B 1 [ g safety-related function
$ 3. Other mechanical modules N SE,RB, - E NS
(including nitrogen storage tanks, OO l and vaporizers), piping, valves, and S electrical modules and cables with
% no safety function
" P T41 Drywell Cooling System (DCS) N CV,CE - E il E
B
!e T49 Flammability Control System (FCS) $
4 l 1. Igniter Assembly Q CV - B i g b l 2. Electrical modules and cables with Q SE - B 1 gm. R-il safety-related function
- a 3=
0 0 - - 0
. _ _ .- __ . _ _ ~ . - _ - _ . . - - . _ . . _ _ _ . -_ _ . . _ . . . . . _ . . _ _ _ _. - _ _ _ . _ _ _ _ . O O O p Table 3.2-1 Classification Summary (Continued) m a 5 Safety Quality QA Seismic W k Principal Component i Desig.2 Location 3 Group' Req.s Category s Notes 8 o T53 Suppression Pool Temperature Q CV,SE, - B I m Monitoring System (SPTMS) CE,RB U STRUCTURES AND SERVICING SYSTEMS a
-" U24 Turbine Pedestal N TB -
E NS 3 U31 Cranes, Holsts, and Elevators 3
- 1. Cranes - reactor building, refueling N RB -
E ll Cranes-The reactor building, refueling bridge, k bridge, fuel handling jib and fuel handling jib cranes are designed to p maintain their position and hold up their loads j under conditions of an SSE. O Upper and lower drywell servicing N NS lu 2. hoists and cranes CV - E y L 3. Main steam tunnel servicing hoists N OL - E NS k
$ and cranes 3
- 4. Special service rooms hoists and N SE,RB, -
E NS k
?. cranes TB,RW h
- 5. Elevator N SC,RB -
E NS U36 Electrical Equipment Building HVAC N EB - E NS U37 Service Building HVAC N SB - E NS U38 Radweste Building HVAC N RW - E NS U39 Turbine Building HVAC N TB - E NS U40 Reactor Building HVAC it,
- 1. Ducts, valves, and dampers Q SE - B I h (including supports) forming part of 4 the safety envelope boundary
- 2. Other ducts, valves, and dampers N SE,CE -
E NS 1 (including supports) $
$ 4 4 1
t Table 3.2-1 Classification Summary (Continued) (a 4 tig
- Safety Quality QA Selsmic Desig.2 Location8 Group' Req.5 Category s Notes Principal Component 1
- 3. Electrical modules and cable with O SE,CE -
B I safety-related function
- 4. Main control room bottled air Q CE,00 -
B l system
- 5. Other non-safety-related equipment N SE,RB, -
E NS CE U41 Other Bu lding HVAC N OL - E NS U42 Potable and Sanitary Water System N ALL - E NS except CV & SE f U43 Fire Protection System (FPS)
- 1. Piping and valves including supports N ALL D E NS Fire Protection System - A quality assurance pf p program meeting the guidance of NRC Branch C
- P Technical Position CMEB 9.5-1 (NUREG-0800) is y applied to the protection system. Also, special [
g seismic qualification requirements are applied. j 2. Pumps N SF D E NS Same as above. { 3. Pump motors N SF - E NS Same as above. j 4. Electrical modules and cables N ALL - E NS Same as above. e 9 5. CO2 actuation modules N RB,TB, - E NS Same as above.
" CE I o &
$ 6. Sprinklers N SE,CE D E NS Same as above. E.
- 7. Foam, preaction or deluge N RB,TP -
E NS Same as above. I U44 Sanitary Waste Discharge System N ALL - E NS g y except g CV & SE }g. m
$ U45 Plumbing and Drainage System N ALL -
E NS {
- R i
._ . .. .._. ~. _ . _ . . . . .m . . _ . _ _ _ _ _ _ _ _ _ _ . _ . . _ . _ . _ . _ _ . ._ ____._ ._ _ _._ _ - _ . _ . _ . - _ . _ _ _ . _
f' f% l \ Q) ^] i l-l p Table 3.2-1 Classification Summary (Continued) th a IB Safety Quality QA Seismic 3 p
- e. Principal Component i Desig.2 Location 8 Group4 Req.8 Category8 Notes g 8
g U50 Equipment and Floor Drain System m
- 1. Piping and valves forming part of the Q CV,SE B B l g containment boundary c
M 2. Drain piping and valves including N ALL D E NS { supports NS
- 3. Other mechanical and electrical N ALL - E h
g modules h USS Other Building' Structures N OL - E NS f% U71 Reactor Building Structure O/N SE,RB, CE B 1 a I f U72 Turbine Building Structure N TB - E NS N NS
>$ U73 Stack RB -
E : w E U74 Radweste Building Structure N RW - E NS Radweste Management Systems- A quality
! assurance program meetlag the guidance of i(
2 NRC Regulatory Guide 1.143 is applied to ! radioactive waste management systems during design and construction. U75 Service Building Structure N SB - E NS U80 Electrical Equipment Building N EB - E NS Structure g U81 Seismic Monitoring System N ALL - E NS $ E W INTAKE STRUCTURE AND SERVICING EQUIPMENT W12 Intake and Discharge Structures Not in Scope W24 CoolingTower W32 Screen Cleaning Facility Not in Scope Not in Scope { j, t W33 Screens, Racks, and Rakes Not in Scope 61
l l t Table 3.2-1 Classification Summary (Continued) m Safety Quality QA Seismic Desig.2 Location 3 Group4Req.s Categorys Notes Principal Component1 W41 Intake Structure Power Supply Not in Scope Y YARD STRUCTURES AND Not in Scope EQUIPMENT Y12 Roads and Walkways Not in Scope Y21 Warehouse Not in Scope Y41 Station Water System Not in Scope Y46 Cathod:c Protection System Not in Scope Y47 Meteorological Observation System Not in Scope YS1 Yard Miscellaneous Drain System Not in Scope 9 00 3 Y52 Oil Storage and Transfer System N - E - Y53 Chemical Storage and Transfer Not in Scope
$~ System y h Y71 Piping Duct N OL -
E - [ Y72 Cable Duct N OL - E - h M Y86 Site Security N ALL - E - Notes; { c: 3 g (1) Principal components: A module is an assembly ofinterconnected components that constitute an identifiable device or i piece of equipment. For example, electrical modules include sensors, power supplies, and signal processors; and I E mechanical modules include turbines, strainers, and orifices. b
$ il Y
(2) Safety designation: Qor N are designations for safety-related or non-safety-related as discussed in Subsection 3.2.1. s [ (3) Location codes: I l
& CV = Containment vessel RW= Radwaste building [
SB = Services building f SE = Safety envelope outside containment j
=
0 0 - - 0
,- ( ,
( wJ
) (
wJ I (J l p CE = Control room envelope CP = Circulating water pump house (4 RB = Reactor building outside safety SF = Service water and fire building IU R{ envelope and control room envelope $ 2 { 00 = Outdoors onsite OL = Any other location TB = Turbine building EB = Electrical equipment building st. E 5 (4) Quality group classifications: A, B, C, or D are quality groups defined in Regulatog Guide 1.26, as discussed in l Subsection 3.2.2. The principal components are classified, designed, and constructed in accordance with the
'm requirements identified in Tables 3.2-2 and 3.2-3. The designation " " indicates that the quality groups A through 0 f are not applicable to the associated principal component.
[p (5) Quality assurance requirements:The designation "B" indicates that the quality assurance requirements of 10CFR50, Appendix B, are applied in accordance with the quality assurance program described in Chapter 17. The designation j "E" indicates that quality assurance requirements are applied, commensurate with the importance of the item's function. R
$ (6) Seismic category: The designations "I" or "II" indicate that the design requirements of Seismic Category I or II structures %
[ and equipment are applied as described in Subsection 3.2.3 and Section 3.7, Seismic Design. Structures and equipment $
! that are not designated "I" or "II" are designated "NS". 3 E P f
~
(7) Small Piping and Instrument Lines- Lines one inch and smaller that are part of the reactor coolant pressure boundary are QGB and meet the requirements of the ASME Code, Section III, Class 2 and Seismic Category I, with the exceptions [ noted below: Instrument lines that are connected to the reactor coolant pressure boundary and are used to actuate or monitor safety-related systems are QGB from the outer isolation valve or the process shutoff valve (root valve) to the sensing instrumentation. Instrument lines that are connected to the reactor coolant pressure boundary and are not used to y actuate and monitor safety-related systems are non-safety-related and Quality Group D from the outer isolation valve or g the process shutoff valve (root valve) to the sensing instrumentation. Other instrument lines meet the following il requirements: @ k u Through the root valve: the lines are the same classification as the system to which they are attached. g a i Beyond the root valve, if used to actuate a safety-related system: the lines are the same classification as the system to C which they are attached. d S I
I l t a Beyond the root valve, if not used to actuate a safety-related system: the lines may be Quality Group D. (p) l i 6 03 Sample lines from the outer isolation vahe or the process root valve through the remainder of the sampling system may E be Quality Group D. D Safety-related instrument lines comply with the guidance of NRC Regulatory Guide 1.151. (8) Hydraulic Control Unit for Control Rod Drive System -The hydraulic control unit (HCU) is a factory-assembled, engineered module ofvalves, tubing, piping, and stored water that controls two control rod drives by the application of pressure and flow to accomplish rapid insertion for reactor scram. Although the HCU is field installed as a unit and connected to process piping, many ofits internal parts differ markedly from process piping and components because of the more complex functions of the HCU.Thus, although the codes and standards invoked by the different quality groups (A, B, C, and D) apply to the interfaces between the HCU and its connections to conventional piping components (e.g., pipe nipples, fittings, hand valves, etc.), they are not considered g g applicable to the specialty parts (e.g., solenoid valves, pneumatic components, and instruments). g k S
- g. However, the design and construction specifications for the HCU do invoke such codes and standards as can be 1:1 (W reasonably applied to indhidual parts in developing required quality levels. For example: (1) all welds are inspected [
h using liquid penetrant, (2) all socket welds are inspected for gaps between the pipe and socket bottom, (3) all welding h is performed by qualified welders, and (4) all work is perfonned in accordance with written procedures. Quality ji Group D is generally applicable because the codes and standards invoked by that group permit the use ofmanufacturer's y standards and proven design techniques that are not explicitly defined within the codes for Quality Groups A, B, or C.
$ This is supplemented by appropriate quality control (QC) techniques.
a m 8. 9 (9) Turbine Control System integrity.
'Ihe turbine stop valve is designed to withstand the SSE and maintain its pressure-retaining (&
4 it
! All cast pressure-retaining parts of a size and configuration for which volumetric methods are effective are examined by y radiographic methods by qualified personnel. Ultrasonic examination to equivalent standards is used as an alternative , > to radiographic methods. Examination procedures and acceptance standards are at least equivalent to those defined in [ ! Paragraph 136.4, Nonboiler External Piping, ANSI B31.1. j. > =
4 4 5 !i O O O
i p The fo t' e ing qualifications are met with respect to the certification requirements: g
.t 9 -[ (a) The manufacturer of the turbine stop vahes, turbine control valves, turbine bypass valves, and main steam lines -[ from turbine control valve to turbine casing uses quality control procedures at least equivalent to those defined m -
g GE Publication GEZ-4982A, General Electric Large Steam Turbine Generator Quality Control Program. - f
,g (b) A certification obtained from the manufacturer of these valves and steam lines demonstrates that the quality control program as defined has been accomplished. =
1 The following requirements are applied in addition to the Quality Group D requirements: !
!i % (a) All longitudinal and circumferential butt weldjoints are radiographed (or ultrasonically tested to equivalent standards). Where size or configuration does not permit effective volumetric examination, magnetic particle or ,
~ g liquid penetrant examination may be substituted. Examination procedures and acceptance st andards are at least g - equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI B31.1.
=
8 (b) All fillet and socket welds, and all structural attachment welds to pressure-retaining materials are examined by g ,
> either magnetic particle or liquid penetrant methods. Examination procedures and acceptance standards are at 5 f least equivalent to those specified as supplementary types of examinations, Paragraph 136.4 in ANSI B31.1.
l
~
(c) All inspection records are maintained for the life of the plant. These records include data pertaining to qualification ofinspection personnel, examination procedures, and examination results.- [ i . i 6
._--.,__.--=-=-.r...-.r1..,w-r
_ vm.- - - ,._.r.r-,...-wrr.w~.r., <..re r---.-,w w < rov vi1,w-=v--re~-+,,rs us,~er#- -,+m...-.,,-w,,ww - vv- ~*1.->s. , - - , , - -. - - - . - - - - . . - - - _ . - . - - - . . . . - .
i I i 25AS113R v. A SBWR standantsafetyAasirsis neport O l Table 3.2-2 Minimum Safety Designation Requirements Minimum Design Requirements l Safety Electrical l Designation Quality Group Seismic Category Classification Quality Assurance l l Q A, B, or C I Class 1ET 10CFRSO A,ng;r h B l N C or D* ll or NS Non-Class 1E 3 l
- Seismic Category I structures, systems, and components meet the design and analysis requirements of Section 3.7. Some safety-related items (e.g., pipe whip restraints) have no safety-related function in the event of an SSE and are not Seismic Category 1.
t Safety-related electrical equipment and instrumentation meet the design requirements of IEEE Class 1E (as well as Seismic Category I). Some non-safety-related electrical equipment and instrumentation are optionally designed to IEEE Class 1E requirements as noted in Table 3.2-1.
- Some non-safeyrelated structures, systerns, and components are optionally designed to Quality Group C or D requirements, as designated in Table 3.2-1. Non-safety-related structures, systems, and components that are not assigned a quality group are designed to requirements of applicable industry codes and standards (see Subsection 3.2.1). j f Safety-related structures, systems, and components meet the quality assurance requirements of 10CFR50, Appendix B, as described in Chapter 17.
O 3.2-34 Classirocation of Structures. Systems, and Components - Amendment 1
/T d b 'v l 9 Table 3.2-3 Quality Group Designations- Codes and Industry Standards #
It W
- [
g- ASME Pressure Vessels and ASME Section ill
]
4 l Quality Group Section 111 Code Heat Pipes, Valves, Storage Tanks Storage Tanks Component Containment y Classification Classes Exchangers and Piping 0-15 psig Atmospheric Supports Boundary k A 1 NCA and NB NCA and NB - - NCA and NF - f TEMA C { n B 2 NCA and NB TEMA C NCA and NC NCA and NC NCA and NC NCA and NF - S
$ CC' and MC - - - - - NCA, CC*, and il NE C 3 NCA and ND NCA and ND NCA and ND NCA and ND NCA and NF -
3 TEMA C E k D - ASME B31.10 for API-620 or API-650 AWWA- - - 2 Sect. Vill piping & valves t equivalent
- D100 ANSI 3 E Division 1 B96.1 or y TEMA C equivalent
{
-[
- RCCV is designed to Subsection CC in ASME Section ill, Div. 2.
t For pumpe classified in Quality Group D of the ASME Boiler and Pressure Veeeel Code, Section Vill, Division 1, are used as a guide in determining the well thicknees for pressure-retaining parte and in eizing the cover bolting.
- Tanks are designed to meet the intent of API, AWWA, ard/or ANSI B96.1 etanderde, se applicable.
V 5 a. I t i tr se
? 4 l 1
-- - -ma_ - _sa1 <o-- m,, a a_ ,__.a-aa ,,_ ,_ ,y a , _ _ ,
i O 9 1 d a 4 4 O, f s i l I I l i l i I
\
l 4 W W O
2SA5113 Rev. A SBWR stenkntseveryAnar1 sis neport l l 3.3 Wind and Tornado Loadings SBWR Standard Plant structures which are Seismic Category I and II are designed for tomado and extreme wind phenomena. 3.3.1 Wind Loadings ! 3.3.1.1 Design Wind Velocity Seismic Category I and 11 structures are designed to withstand a design wind velocity described in SST Sub= eden 2.2.1 SSAR Subsection 2.3.1.2: an de= tion of10 m (33 , fer:) abcce gade -ith a recu~ence i-te.md of 1^^ ye =. Refer to Subsection 3.3.3 for interface requirement. 3.3.1.2 Determination of Applied Forces The design wind velocity is converted to velocity pressure in accordance with Reference 3.S1 using the formula: qz = 0.00256Kz(IV)2 {3,y)) (
- s where
l Kz = the velocity pressure exposure coefficient, which depends upon the type of exposure and height (z) above ground per Table 6 of Reference 3.S1 creened te r. etic unie; I = the importance factor, which depends on the type of exposure; appropriate values ofI are listed in Table 3.3-1; I V = design wind velocity.mp.h;and t j qz = velocity pressure in-MPa(psf). The velocity pressure (qz) distribution with height for exposure types C and D of Reference 3.S1 are given in Table S.S2b for the reactor building. Table 3.S3 cives correction factors to correct loads for building heichts other than the 39.5m (129 ft) reactor building. The design wind pressures and forces for buildings, components and cladding, and other structures at various heights above the ground are obtained,in accordance with Table 4 of Reference 3.S1 by multiplying the velocity pressure by the appropriate pressure coefficients and gust factors. Gust factors are in accordance with Table 8 of Reference S.SI. Appropriate pressure coefficients are in accordance with Figures 2, Sa, ! Sb,4, and Tables 9 and 11 through 16 of Reference S.SI. Reference 3.S2 is used to ! obtain the effective wind pressures for cases which Reference 3.S1 does not cover. Since Wind and Tomado Loadinga - Amendment 1 3.3-1
i 25AS113 Rev. A SBWR sunderdsoretyAnalysisseport O the Seismic Category I and II structures are not slender or flexible, vortex-shedding analysis is not required and the above wind loading is applied as a static load. 3.3.2 Tornado Loadings 3.3.2.1 Applicable Design Parameters The design basis tornado and applicable missiles are described in Sub:eceen 2.2.1 Subsection 2.3.1. Refer to Subsection 3.3.3 for COL License Information. 3.3.2.2 Determination of Forces on Structures The procedures of transforming the tornado loading into effective loads and the distribution across the structures are in accordance with Reference 3.?-3. The procedure for transforming the tornado-generated missile impact into an effective or equivalent static load on structures is given in Subsection 3.5.3. The loading combinations of the individual tornado loading components and the load factors are in accordance with Reference 3.3-3. The reactor building is not a vented structure. The exposed exterior roofs and walls of this structure are designed for the full pressure drop. Tornado dampers are provided on all air intake and exhaust openings. These dampers are designed to withstand the full negative pressure drop. 3.3.2.3 Effect of Failure of Structures or Components Not Designed forTornado Loads All safety-related system and components are protected within tornado-resistant structures. 3.3.3 COL Ucense Information . Site-Specific Design Basis Wind and Tornado The site-specific design basis wind pad tornado shall not exceed the wind given in Subsection 2.2.1 Subsection 2.3.1. Effect of Remainder of Plant Structures, Systems, and Components not Designed for Tornado Loads All remainders of plant structures, systems, and components not designed for tornado loads shall be analyzed for the site-specific loadings to ensure that their mode of failure will not affect the ability of the Seismic Category I and II SBWR Standard Plant structures, systems, and components to perfonn their intended functions (Subsection 3.3.2.3). 13-2 Wind and Tornado Loadings- Amendment 1
25AS113 Rzv. A SBWR senaderdsunrtyAnalysis Report \ 3.3.4 References 3.S1 ANSI sent.3 A38.2 ASCE Standard 7-1988. Minimum Design Loads for Buildings and Other Structures, Committee A. 58.1, American National Standards Institute. 3.S2 ASCE Paper No. 3269, Wind Forces on Structures, Transactions of the
.American Society of Civil Engineers, Vol.126, Part II.
3.S3 Bechtel Topical Report BC-TOP-SA, Revision 3, Tornado and Extreme Wind Design Criteria for Nuclear Power Plants. f ( Wind and Tomado Londinge - Amendment 1 gy
l , 2SA5193 Rev, h i SBWR standard sanry Analysis neport l O' Table 3.3-1 Importance Factor (1) for Wind Loads l E erre C Non-Safety-Related Structures Egerre O Safetv-Related Structures 1 1.00 1.11 Noteu
*)" :: v:! :: ef(') cre b : d up n T:b!: 5 ef Def --ner ?.? ? b f. :r: .~ed!S- te r:90'* 'he '^^ ye:r I return peded of the d--!;n r%d v:!::b; verre: the E? ye:r :: turn peded b::! ef Reference ?.? '. l
- 2) Exp^ ure ::tegede are 2: d:E: d in Se*jen E.E.? cf . fer:nce ?.? '.
O O.l l l 134 Wind and Tomado Loadinge - Amendment 1
i l l 2SA5113 R:v. A SBWR standant safety Analysis neport O V Tab!e 3.3-2 Ve!:::t/ Pcener: Dietribu'! n :nd Ge:t F:de : :t Ver?::: M:!;ht Table 3.3-2a Desian Pressure Distribution at Various Heiahts for Safetv-Related Structures (Importance Factor = 1.11) Windwerd Wall Leeward Wall Pressure Side Wall Suction Roof Suction Suction Height Zone z 0.8Ghqz 0.7Ghqh 0.7Ghqh 0.5Ghqh m(ft) Pa(psf) Pa(psf) Pa(psf) Pa(psf) Exposure Type C 0-4.57 (0-15) 1544 (32) 2507 (52) 2507 (52) 1791 (37) 6.10(20) 1641 (34) 2450(51) 2450 (51) 1750 (37) 7.62 (25) 1727 (36) 2412 (50) 2412 (50) 1723 (36) 9.14 (30) 1805 (38) 2393 (50) 2393(50) 1710 (36) 12.19 (40) 1906 (40) 2336(49) 2336(49) 1669 (35) 1524(50) 1999 (42) 2298 (48) 2298 (48) 1642 (34) 1829(60) 2088 (44) 2279 (48) 2279 (48)- 1628 (34) 2134(70) 2157 (45) 2260 (47) 2260 (47) 1615 (34) 2438(80) 2225 (47) 2241 (47) 2241 (47) 1601 (33) (q) 27.43 (90) 2292 (48) 2222 (46) 2222 (46) 1587 (33) 30.48 (100) 2340 (49) 2203 (46) 2203 (46) 1574 (33) 36.58 (120) 2438 (51) 2184 (46) 2184 (46) 1560(33) 42.67 (140) 2533 (53) 2165 (45) 2165 (45) 1547 (32) 48.77 (160) 2610(55) 2146 (45) 2146 (45) 1533 (32) Exposure Type D ; 0-4.57 (0-15) 2017 (42) 2707 (57) 2707 (57) 1993 (40) 6.10(20) 2117 (44) 2683 (56) 2683 (56) 1917 (40) 7.62 (25) 2181 (46) 2660 (56) 2660 (56) 1900 (40) 9.14 (30) 2243 (47) 2636 (55) 2636 (55) 1883 (39) 12.19 (40) 2369 (50) 2613 (55) 2613 (55) 1866 (39) 1524(50) 2444 (51) 2589 (54) 2589 (54) 1849 (39) 1829(60) 2518 (53) 2565 (54) 2565 (54) 1832 (38) 21.34 (70) 2574 (54) 2542 (53) 2542 (53) 1816 (38) 2438(80) 2637 (55) 2542 (53) 2542 (53) 1816 (38) 27.43 (90) 2675(56) 2518(53) 2518 (53) 1799 (38) 30.48(100) 2737 (57) 2518(53) 2518 (53) 1799 (38) 36.58 (120) 2805 (59) 2495(52) 2429 (52) 1782 (37) 42.67 (140) 2870(60) 2471 (52) 2491 (52) 1765 (37) 48.77(160) 2947 (60) 2471(52) 2471 (52) 1765 (37) Wind and Tomodo Loadings - Amendment 1 3.3-5
l l l l 25A5113 Rev. A SBWR standard saretyAna!ysis neport Table 3.3-2b Desian Pressure Distribution at Various Heiahts for Non-Safc4tv-Related Structures (ImDortance Factor = 1.00) Windward Wall 1 eeward Wall Pressure Side Wall Suction Roof Suction Suction Height Zone : 0.8Ghqz 0.7Ghqh 0.7Ghqh 0.5Ghqh l m(ft) Pa(psf) Pa(psf) Pa(psf) Pa(psf) Exposure Type C 0-4.57 (0-15) 1253(26) 2035(43) 2035(43) 1454(30) 6.10 (20) 1332 (28) 1989 (42) 1989 (42) 1421 (30) 7.62 (25) 1401 (29) 1958 (41) 1958 (41) 1399 (29) I 9.14 (30) 1465(31) 1943 (41) 1943 (41) 1388 (29) 12.19 (40) 1547 (32) 1869 (40) 1869 (40) 1355 (28) 15.24(50) 1622 (34) 1865 (39) 1865 (39) 1332 (28) 1829(60) 1694 (35) 1850 (39) 1850(39) 1321 (28) 2134(70) 1751 (37) 1835 (38) 1835(38) 1310 (27) 2438(80) 1806(38) 1819 (38) 1819 (38) 1299 (27) 27.43(90) 1860(39) 1804(38) 1804(38) 1288 (27) 30.48(100) 1899 (40) 1778 (37) 1778 (37) 1277 (27) 36.58(120) 1979 (41) 1773 (37) 1773 (37) 1266(26) 42.67(140) 2056 (43) 1758 (37) 1758 (37) 1255 (26) 48.77 (160) 2118 (44) 1742 (36) 1742 (36) 1240 (26) Exposure Type D 0-4.57 (0-15) 1637 (34) 2197 (46) 2197 (46) 1569 (33) 6.10 (20) 1718 (36) 2178 (46) 2178 (46) 1556 (33) 7.62 (25) 1770(37) 2159 (45) 2159 (45) 1542 (32) 9.14 (30) 1821(38) 2140 (45) 2140 (45) 1528 (32) 12.19 (40) 1923 (40) 2120 (44) 2120(44) 1515 (32) 15.24 (SA) 1984 (41) 2101 (44) 2101 (44) 1501 (31) 1829(60) 2043 (43) 2082 (43) 2082 (44) 1487 (31) 2134(70) 2089 (44) 2063 (43) 2063 (43) 1474 (31) 2438(80) 2140(45) 2063 (43) 2063 (43) 1474 (31) 27.43(90) 2171 (45) 2044 (43) 2044 (43) 1460 (31) 30.48(100) 2222 (46) 2044 (43) 2044 (43) 1460 (31) 36.58(120) 2276(48) 2025 (42) 2025(42) 1446 (30) 42.67(140) 2330(49) 2006 (42) 2006(42) 1443 (30) 48.77(160) 2392 (50) 2006 (42) 2006(42) 1433 (30) O 3.3-6 VMnd and Tornado Loadings - Amendment 1
25A5113 Rev. A SBWR smadan(sanyAndysis neper 'OO , i l Notes for Tables 3.S2a and 3.S2b: m Windward wall desien pressure is positive. side and leeward walls and roof pressures are negative. M For Exoosure C: Add 551 pa (11.5 osn to windward wall and subtract 551 na (11.5 osi) from side and leeward walls and ref for Condition I buildines. ! Add 551 pa (11.5 osn to windward wall and subtract 1653 na (34.5 osD from side and leeward walls and roof for Contlition II buildines. For Exoosure D: I Add 683 ca (14.25 osn to windward wall and subtract 683 na (14.25 osi) from side and leeward walls and roof for Condition I buildines. l (N. Add 683 oa (14.25 psn to windward wall and subtract 2048 ca (42.75 osi) from side and leeward walls and roof for Condition II buildines. I i i Condition I and Condition II buildines are defined oer Table 9 of oararraoh , 2.2.1a ofASCE Standard 7-1988. { g This table is based on the 39.5 meter (130 foot) high Reactor Building. loads ; must be adiusted for differer s t height buildines. , I l v Wind and Torr. ado Londings - Amendment 1 3.3-7 l
2 SASS 13 Rev. A SBWR standardsakryAnalysis neport O Table 3.3-3 Factor fazn/az130) to Adiust Table 2 Loads for Buildina Heiahts ! Other Than 39.5 Meters (130 ft) (q ,"/q " ) (qz"/q ") i Exposure C Exposure D Building Height i = 1.00 and I = 1.11 I = 1.00 and I = 1.11 0-15 0.54 0.65 20 0.59 0.69 25 0.63 0.72 30 0.66 0.75 40 0.71 0.79 50 0.76 0.83 60 0.80 0.86 70 0.84 0.89 80 0.89 OS1 90 OSO OS3 100 0S3 OS5 120 OS8 OS8 130 1.00 1.00 140 1.02 1.02 160 1.06 1.04 l l l 1.%8 Wind and Tomado Loadings - Amendment 1
l i' 2SA5113 Rev. A j SBWR samrantsanyAurysis neport lO ,v 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping This section deals with the structures, systems, components, and equipment in the , SBWR Standard Flant. Subsections 3.6.1 and 3.6.2 describe the design bases and protective measures which ensure that (1) the containment, (2) safety-related systems, components and equipment, and (3) other safety-related structures are adequately protected from the consequences associated with a postulated rupture of high-energy piping or crack of moderate-energy piping both inside and outside the containment. Before delineating the criteria and assumptions used to evaluate the consequences of piping failures inside and outside of containment,it is necessary to define a pipe break event and a postulated piping failure: a Pipe Break Event-Any single postulated piping failure occurnng during normal l plant operation and any subsequent piping failure and/or equipment failure that occun as a direct consequence of the postulated piping failure. s a Postulated Piping Failure-Longitudinal or circumferential break or nipture postulated in high-energy fluid system piping or throughwall leakage crack postulated in moderate-energy fluid system piping.The terms used in this definition are explained in Subsection 3.6.2. , l Structures, systems, components and equipment that are required to shut down the , reactor and mitigate the consequences of a postulated piping failure, without off-site power, are defined as safety-related and are designed to Seismic Category I requirements. , 1 l The dynamic effects that may result frorL a postulated rupture of high-energy piping i include (1) missile generation, (2) pipe whipping, (3) pipe break reaction forces,
- (4) jet impingement forces, (5) compartment, subcompartment, and cavity pressurizations, (6) decompression waves within the ruptured pipes, and (7) seven types ofloadsidentified with LOCA. )
l l Subsection 3.6.3 and Appendix 3C describe the implementation of the leak-before-l break (LBB) evaluation procedures as permitted by the broad scope amendment to General Design Criteria 4 (GDC4) published in Reference 3.6-1. An LBB report shall be prepared with the stress report for the LBB<1nahfiable piping in accordance with the guidelines presented in Appendix 3C. The LBB-qualified piping will be excluded from 'Q pipe breaks, which are required to be postulated by Subsections 3.6.1 and 3.6.2, for Q design against their potential dynamic effects. However, such piping are included in postulation of pipe cracks for their effects as described in Subsections 3.6.1 and 3.6.2. I Protection Against Dynamic Fffects Associated with the Postulated Rupture of Piping- Amendment 1 3.6-1 l l
25AS113 Rov. A standardsarery Analysis neport SBWR O 3.6.1 Postulated Piping Failures in Fluid Systems inside and Outside of Containment 1 i This subsection sets forth the design bases, description, and safety evaluation for determining the effects of postulated piping failures inside the containment, and for including necessary pmtective measures. 3.6.1.1 Design Bases Criteria j Pipe break event protection conforms to 10CFR50 Appendix A, General Design Criterion 4, Environmental and Missile Design Bases. The design bases for this protection are in compliance with NRC Branch Technical Position (BTP) ASB 3-1 and MEB 3-1 included in Subsections 3.6.1 and 3.6.2, respectively, of NUREG0800 (Standard Review Plan). MEB 3-1 describes an acceptable basis for selecting the design locations and , orientations of postulated breaks and cracks in fluid systems piping. Standard Review f Plan Subsections 3.6.1 and 3.6.2 describe acceptable measures that could be taken for protection against the breaks and cmcks and for restraint against pipe whip that may resultfrom breaks. The design of the containment structure, component arrangement, pipe runs, pipe whip restraints, and compartmentahrntion are done in consonance with the acknowledgment of protection against dynamic effects associated with a pipe break event. Analytically sized and positioned pipe whip restraints are engineered to preclude damage based on the pipe break evaluation. Objectives Protection against pipe break event dynamic effects is provided to fulfill the following objectives: 1 m Assure that the reactor can be shut down safely and maintained in a safe shutdown condition and that the consequences of the postulated piping failure are mitigated j to acceptable limits without off site power. l m Assure that containment integrity is maintained. m Assure that the radiological doses of a postulated piping failure remain below the limits of10CFR100. Assumptions l The following assumptions are used to determine the protection requirements: 16-2 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 l I
25A5113 R:v. A SBWR standardsaktyAnalysis neport A
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i e Pipe break events may occur during normal plant conditions (i.e., reactor startup, operation at power, normal hot standby (Reference 3.6-1)or reactor cooldown to a cold shutdown conditions but excluding test modes). m A pipe break event may occur simultaneously with a seismic event; however, a seismic event does not initiate a pipe break event. This applies to Seismic Categog I and non-Seismic CategoryI piping. m A single active component failure (SAG) is assumed in systems used to mitigate consequences of the postulated piping failure and to shut down the reactor, except as noted in item below. A SAG is the malfunction or loss of function of a component of electrical or fluid systems. The failure of an active component of a fluid system is considered to be a loss of component function as a result of mechanical, hydraulic, or electrical malfunction but not the loss of component structural integrity. The direct consequences of a SAG are considered to be a part of the single active failure. The single active component failure is assumed to occur in addition to the postulated piping failure and any direct consequences of the piping failure.
,.m
\ m Where the postulated piping failare is assumed to occur in one of two or more llb redundant trains of a dual-purpose moderate-energy safety-related system (i.e., one l required to operate during normal plant conditions as well as to shut down the reactor and mitigate the consequences of the piping failure), single active failure of components in the other train or trains of that system only are not assumed, provided the system is designed to Seismic Category I standards, is powered from both off-site and on-site sources, and is constructed, operated, and inspected to quality assurance, testing and inservice inspection standards appropriate for nuclear j safety-related systems. l l
e If a pipe break event involves a failure of nonSeismic Category I piping, the pipe break event must not result in failure of safety-related systems, components and I equipment to shut down the reactor and mitigate the consequences of the pipe break event considering a SAG. m Ifloss of off-site power is a direct consequence of the pipe break event (e.g., trip of l the turbine-generator producing a power surge which,in turn, trips the main l breaker), then a loss of off-site power occurs in a mechanistic time sequence with a ! SACF. Otherwise, off site power is assumed available with a SAG. l m A whipping pipe is not capable of rupturing impacted pipes of equal or greater nominal pipe diameter, but may develop throughwall cracks in equal or larger l(p) l
%j nominal pipe sizes with thinner wall thickness.
i Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 3.6-3 l
l 25A5113 Rev. A SBWR senmtantsaktyAnalysis Report 9 m All available systems, including those actuated by operator actions, are available to mitigate the consequences of a failure. Injudging the availability ofsystems, account is taken of the postulated failure and its direct consequences such as unit trip and loss of off4ite power, and of the assumed SACF and its direct consequences. The feasibility of carrying out operator actions arejudged on the basis of ample time and adequate access to equipment being available for the proposed actions. m Although a pipe break event outside the containment may require a cold shutdown, up to eight hours in hot standby is allowed in order for plan t penonnel to assess the situation and make repain. m Pipe whip occurs in the plane defined by the piping geometry and causes movement in the direction of thejet reaction. If unrestrained, a whipping pipe with a constant energy source forms a plastic hinge ana rotates about the nearest rigid restraint, anchor, or wall penetration. If unrestrained, a whipping pipe without a constant energy source (i.e., a break at a closed valve with only one side subject to pressure) is not capable of forming a plastic hinge and rotating provided its movement can be defined and evaluated. m The fluid internal energy associated with the pipe break reaction can take into account any line restrictions (e.g., flow limiter) between the pressure source and break location and absence of energy reservoirs, as applicable. Approach To comply with the objectives previously described, the safety-related systems, components, and equipment are identified. The safety-related systems, components, and equipment, or portions thereof, are identified in Table 3.6-1 for piping failures postulavd inside the containment and in Table 3.6-2 for outside the contamment. 3.6.1.2 Description The lines identi6ed as high-energy per Subsection 3.6.2.1 are listed in Table 3.6-3 for inside the containment and in Table 3.6-4 for outside the containment. Pressure response analyses are performed for the subcompartments containing high-energy i piping. A detailed discussion of the line breaks selected, vent paths, room volumes, analytical methods, pressure results, etc., is provided in Section 6.2. The effects of pipe whip, jet impingement, spraying, and flooding on required function of safety-related systems, components, and equipment, or portions thereof, inside and outside the containment, are considered. In panicular, there are no high-energy lines near the control room. As such, there are no effects upon the habitability of the control room by a piping failure in the control room or elsewhere either from pipe whip, jet impingement, or transport of steam. Further discussion on control room habitability systems is prosided in Section 6.4. 164 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping - Amendment 1
25AS113 R:v. A l SBWR standantsareryAnalysis aeport fy l '( V) 3.6.1.3 Safety Evaluation General An analysis of pipe break events is performed to identify diose safety-related systems, components, and equipment that provide protective actions required to mitigate, to acceptable limits, the consequences of the pipe break event. i i Pipe break events involving high-energy fluid systems are evaluated for the effects of pipe whip, jet impingement, flooding, room pressurization, and other environmental effects such as temperature. Pipe break events invohing moderate <nergy fluid systems l are evaluated for wetting from spray, flooding, and other emironmental effects. By means of the design features such as separation, barriers, and pipe whip restraints, a discussion of which follows, adequate protection is provided against the effects of pipe l break events for safety-related items to an extent that their ability to shut down the plant I safely or mitigate the consequences of the postulated pipe failure would not be l impaired. GeneralProtection Methods The direct effects associated with a particular postulated break or crack must be
- Qp) mechanistically consistent with the failure. Thus, actual pipe dimensions, piping layouts, material properties, and equipment arrangements are considered in defining the following specific measure for protection agamst actual pipe movement and other associated consequences of postulated failures
a Protection against the dynamic effects of pipe failures is provided in the form of pipe whip restraints, equipment shields, and physical separation of piping, equipment, and instrumentation. a The precise method chosen depends largely upon limitations placed on the designer such as accessibility, maintenance, and proximity to other pipes. Protection Methods by Separation l The plant arrangement provides physical separation to the extent practicable to j maintain the independence of redundant safety-related systems (including their auxiliaries) in order to prevent the loss of safety function caused by any single , postulated event. Redundant trains (e.g., A and B tratns) and dhisions are located in l separate compartments to the extent possible. Physical separation between redundant l safety-related systems with their related auxiliary supporting features, therefore, is the l basic protective measure incorporated in the design to protect against the dynamic j effects of postulated pipe failures. I
/m\ ,
l () Because of the complexities of several divisions being adjacent to high-energy lines in the drywell specific break locations are determined in accordance with Subsection 3.6.2.1 for possible spatial separation. Care is taken to avoid concentrating Protection Against Dynamic Effects Associated with the Postuisted Rupture of Piping- Amendment 1 3.6-5 l
2SA5113 Rev. A SBWR standardsareryAnalysisneport O safety-related equipment in the break exclusion zone allowed according to Subsection 3.6.2.1. Ifspatial separation requirements (distance and/or arrangement to prevent damage) cannot be met based on the postulation of specific breaks, then barrien, enclosures, shields, or restraints are provided. These methods of protection are discussed below. For other areas where physical separation is not practical, the following highenergyline separation analysis (HELSA) evaluation is done to determine which high-energy lines meet the spatial separation requirement and which lines require further protection: a For the HELSA evaluation, no particular break points are identified. Cubicles or areas through which the high-energy lines pass are examined in total. Breaks are postulated at any point in the piping system. m Safety-related systems, components, and equipment at a distance greater than 30 feet from any high energy piping are considered as meeting spatial separation requirements. No damage is assumed to occur on account ofjet impingement since the impingement force becomes negligible beyond 30 feet. Likewise, a 30-foot evaluation zone is established for pipe breaks to assure protection against potential damage from a whipping pipe. Assurance that 30 feet represents the maximum free length is made in the piping layout. m Safety-related systems, components, and equipment at a distance less than 30 feet from any high-energy piping are evaluated to see if damage could occur to more than one safety-related division, preventing safe shutdown of the plant. If damage occurred to only one division of a redundant system, the requirement for redundant separation is met. Other redundant divisions are available for safe shutdown of the plant and no further evaluation is performed. m If damage could occur to more than one division of a redundant safety-related system within 30 feet of any high energy piping, other protection in the form of barriers, shields, or enclosures is used. Pipe whip restraints are used if protection from whipping pipe is not possible by barriers and shields. These methods of protection are discussed below. Barriers, Shields, andEnclosures Protection requirements are met through the protection afforded by the walls, floors, columns, abutments, and foundations in many cases. Where adequate protection is not already present because of spatial separation or existing plant features, additional barriers, deflectors, or shields are identified as necessary to meet the functional protection requirements. 16-6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1
2SA5113 Rev. A l SBWR standantsannyAnstrsis neport i Barriers or shields that are identified as necessary by the use of specific break locations in the drywell are designed for the specific loads associated with the particular break location. The MSIVs and the feedwater isolation and check valves located inside the tunnel shall i be designed for the effects of a line break. The details of how the MSIV and feedwater isolation and check valves functional capabilities are protected against the effects of these postulated pipe failures will be provided by the COL applicant (Subsection 3.6.4). Barriers or shields that are identified as necessary by the HELSA evaluation (i.e., based on no specific break locations) are designed for worst. case loads. The closest high-energy pipe location and resultant loads are und to size the barriers. Pipe Whip Restraints Pipe whip restraints are used where pipe break protection requirements could not be ! satisfied using spatial separation, barriers, shields, or enclosures alone. Restraints are located based on the specific break locations determined in accordance with Subsection 3.6.2.1. After the restraints are located, the piping and safety-related systems are evaluated forjet impingement and pipe whip. For those cases wherejet impingement damage could still occur, barriers, shields, or enclosures are utilized. L The design criteria for restraints are given in Subsection 3.6.2.3. l Speci6c Protection Measures a Non-safety-related systems and system components are not required for the safe shutdown of the reactor, nor are they required for the limitation of the off4ite : release in the event of a pipe rupture. However, while none of this equipment is needed during or following a pipe break event, pipe whip protection is considered l where a resulting failure of a non-safety-related system or component could initiate ] or escalate the pipe break event in a safety-related system or component, or in another non-safety-related system whose failure could affect a safety-related system. m For high energy piping systems penetrating through the containment, isolation valves are located as close to the containment as possible, a The pressure, water level, and flow sensor instrumentation for those safety-related systems, which are required to function following a pipe rupture, are protected. s High<nergy fluid system pipe whip restraints and protective measures are designed so that a postulated break in one pipe could not, in turn, lead to a rupture of other !p nearby pipes or components if the secondary rupture could result in consequences Q that would be considered unacceptable for the initial postulated break. Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping - Amendment 1 3.6-7 I l l
25A5113 Rev. A SBWR standantsakryAnalysisReport 1 O m For any postulated pipe rupture, the structural integrity of the containment j structure is maintained. In addition, for those postulated ruptures classified as a loss of reactor coolant, the design leaktightness of the containment fission product barrier is maintained. m Safety / relief valves (SRVs) are located and restrained so that a pipe failure would not prevent depressurization. m Protection for the FMCRD scram insert lines is not required, since the motor operation of the FMCRD can adequately insert the control rods even with a complete loss ofinsertlines (Subsection 3.6.2.1.3). e The escape of steam, water, combustible or corrosive fluids, gases, and heat in the event of a pipe rupture do not preclude:
- accessibility to any areas required to cope with the postulated pipe rupture;
- habitability of the control room; or
- the ability of safety-related instrumentation, electric power supplies, components, and controls to perform their safety-related function.
3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping Information concerning break and crack location cdteria and methods of analysis for dynamic effects is presented in this subsection. The location criteria and methods of analysis are needed to evaluate the dynamic effects associated with postulated breaks and cracks in high- and moderate-energy fluid system piping inside and outside of t_hc pdmary containment. This information provides des the basis for the requirements for the protection of safety < elated structures, systems, and components defined in the introduction of Section 3.6. 3.6.2.1 Criteria Used to Define Break and Crack Location and Configuration The following subsections establish the criteria for the location and configuration of pestulated breaks and cracks. De6nition of High-Energy Fluid Systems High<nergy fluid systems are defined to be those systems or portions of systems that, during normal plant conditions (as defined in Subsection 3.6.1.1), are either in operation or are mamtained pressurized under conditions where either or both of the following are met: a maximum operating temperature exceeds 200 F; or l 3.64 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 1
25AS113 Rev. A SBWR saadantsawyAnarrsisnerat C s a maximum operating pressure exceeds 275 psig. De6nition of Moderate-Energy Fluid Systems Moderate-energy fluid systems are defined to be those systems or portions of systems that, during normal plant conditions (as defined in Subsection 3.6.1.1), are either in operation or are maintained pressurized (above atmospheric pressure) under conditions where both of the following are met: s maximum operating temperature is 200 F or less; and a mnvimum operating pressure is 275 psig or less. l Piping systems are classified as moderate-energy systems when they operate as high- , energy piping for only short operational periods in performing their system function but, for the major operational period, qualify as moderate-energy fluid systems. An operational period is considered short if the total fraction of time that the system ' operates within the pressure-temperature conditions specified for high-energy fluid systems is less than 2% of the total time that the system operates as a moderatunergy fluid system. O ,Q Postulated Pipe Breaks and Cracks > A postulated pipe break is defined as a sudden gross failure of the pressure boundary ' either in the form of a complete circumferential severance (guillotine break) or a ' sudden longitudinal split without pipe severance, and is postulated for high-energy fluid systems only. For moderate-energy fluid systems, pipe failures are limited to postulation of cracks in piping and branch runs. These cracks affect the surrounding environmental conditions only and do the first restraint may be permitted higher ! stresses provided a plastic hinge is not result in whipping of the cracked pipe. High-l energy fluid systems are conservative environmental conditions in a confined area where high-end moderate-energy fluid systems are located. The following high-energy piping systems (or portions of systems) are considered as potential candidates for a postulated pipe treak during normal plant conditions and l are analyzed for potential damage resulting from dynamic effects: e all piping which is part of the reactor coolant pressure boundary and subject to reactor pressure continuously during station operation; e all piping which is beyond the second isolation valve but subject to reactor pressure continuously during station operation; and a all other piping systems or portions of piping systems considered high-energy systems. o Protection Against Dynamic Effecto Associated with the Postulated Rupture of Piping- Amendment 1 36-9
25A5113 Rev. A SBWR standard sarety Analysisneport O Portions of piping systems that are isolated from the source of the high-energy fluid during nortnal plant conditions are exempted from consideration of postulated pipe breaks. This includes portions of piping systems beyond normally closed valves. Pump and valve bodies are also exempted from consideration of pipe break because of their greater wall thickness. 3.6.2.1.1 Locations of Postulated Pipe Breaks Postulated pipe locations are selected as follows: Piping Meeting Separation Requirements Based on the HELSA evaluation described in Subsection 3.6.1.3, the high-energy lines which meet the spatial separation requirements are generally not identified with particular break points. Breaks are postulated at all possible points in such high-energy piping systems. However, in some systems break points are particularly specified according to the following subsections if special protection de3 ices such as barriers or restraints are provided. Piping in Containment Penetration Areas No pipe breaks or cracks are postulated in those portions of piping from dit. contamment wanpenetration to and including the inboard or outboard isolation valves which meet the following requirements in addition to the requirement of the ASME Code, Section III, Subarticle NE.1120: a The following design stress and fatigue limits are not exceeded: For ASME Code, Section III, Cha 1 Piping
- The maximum stress range between any two load sets (including the zero load set) does not exceed 2.4 S , and is calculated by Equation 10 in NB-3653, ASME Code, Section III.
- The cumulative usage factor is less than 0.1.
- The maximum stress as calculated by Equation 9 in NB.3652 under the loadings resulting from a postulated piping failure beyond those portions of piping does not exceed the lesser of 2.25S, and 1.8Sy except that, following a failure outside containment, the pipe between the outboard isolation valve and the first rectraint may be permitted higher stress, provided a pbstic hinge is not formed and operability of the valves with such stresses is assured in accordance with the requirement identified in Section 3.9.3. Primary loads include those wilich are deflection limited by whip restraints.
16-10 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping - Amendnwnt 1
25AS113 Rev. A SBWR standantsurtyAnalysis Report p
\ l
'%.J For ASME Code, Section III, Oass 2 Piping
- The maximum stress as calculated by the sum of Equations 9 and 10 in Paragraph NC-3652, ASME Code, Section III, considering those loads and conditions thereof for which level A and level B stress limits are specified in the system's Design Specification (i.e., sustained loads, occasional loads, and thermal expansion), including an OBE event, does not exceed 0.8(1.8 S h + SA )-
The Shand S Aare allowable stresses at maximum (hot) temperature and allowable stress range for thermal expansion, respectively, as defined in Article NC-3600 of the ASME Code, Section ID.
- The maximum stress, as calculated by Equation 9 in NC-3653 under the loadings resulting from a postulated piping failure of fluid system piping beyond these portions of piping, does not exceed the lesser of 2.25 Sh and 1.8 Sy.
Primary loads include those which are deflection limited by whip restraints. The exceptions permitted above may also be applied provided that, when the piping between the outboard isolation valve and the restraint is constructed in accordance with the Power Piping Code ANSI B31.1, the piping is either of seamless construction with full radiography of all circumferential welds, or all longitudinal (v) and circumferential welds are fully radiographed. m Welded attachments, for pipe supports or other purposes, to these portions of piping are avoided except where detailed stress analyses, or tests, are performed to demonstrate compliance with the above mentioned code limits. m The number of circumferential and longitudinal piping welds and branch connections are minimned. Where penetration sleeves are used, the enclosed portion of fluid system piping is seamless construction and without circumferential welds unless specific access provisions are made to permit inservice volumetric eumination oflongitudinal and circumferential welds. m The length of these portions of piping are reduced to the minimum length practical. 1 l m The design of pipe anchon or restraints (e.g., connections to containment penetrations and pipe whip restraints) do not require welding directly to the outer surface of the piping (e.g., flued integrally forged pipe fittings may be used) except where such welds are 100% volumetrically examinable in service and a detal'ed stress analysis is performed to demonstrate compliance with the above mentioned 7 code limits. i s I bl a Sleeves provided for those portions of piping in the containment penetration areas are constructed in accordance with the rules of Oass MC, Subsection hT of the Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 16-11
2SA5113 Rev. A SBWR standardsafetyAnalysisReport 1 O ASME Code, Section III, where the sleeve is part of the containment boundary. In addition, the entire sleeve assembly is designed to meet the following requirements and tests:
- The design pressure and temperature are not less than the maximum operating pressure and temperature of the enclosed pipe under normal plant conditions.
- The Level C stress limits in NE-3220, ASME Code, Section III, are not exceeded under the loadings associated with containment design pressure and temperature in combination with the safe shutdown earthquake.
- The assemblies are subjected to a single pressure test at a pressure not less than its design pressure.
- The assemblies do not prevent the access required to conduct the insenice namination specified below, a A 100% volumetric inservice nammation of all pipe welds would be conducted during each inspection interval as defined in IWA-2400, ASME Code, Section XI.
ASME Code Section ill Class 1 Piping in Areas Other Than Containment Penetration With the exception of those portions of piping identified above, breaks in ASME Code, Section III, CIm 1 piping are postulated at the following locations in each piping and branch run: m At terminal ends. m At intermediate locations where the maximum stress range as calculated by Equation 10 in NB-3653, ASME Code, Section III exceeds 2.4S m . and rif4he-calculated maximum ::re= mnge ofEqua d
" -- d the stress range calculated by either both Equation 12 gr and Equation 13 in Paragraph NB-3653 exceeds 60uld =ce: Se limit of 2.4 Sm-m Atintermediate locations where the cumulative usage factor exceeds 0.1. As a result ofpiping reanalysis caused by differences between the design configuration and the as-built configuration, the highest stress or cumulative usage factor locations may be shifted; however, the initially determined intermediate break locations need not be changed unless one of the following conditions exists:
- The dynamic effects from the new (as-built) intermediate break locations are not mitigated by the original pipe whip restraints andjet shields.
- A change is required in pipe parameters, such as major differences in pipe size, wall thickness, and routing. J 3.6-12 Protection Against Dynamic Effec's Associated with the Postulated Rupture of Piping- Amendment 1 I
25A5113 Rev. A SBWR saadantsunny Analysis neport O V l ASME Code Section Ill Class 2 and 3 Piping in Areas Other Than Containment ; Pwetration . With the exceptions of those portions of piping identified above, breaks in ASME l Codes, Section III, Class 2 and 3 piping are postulated at the following locations in those ; j ponions of each piping and branch run: l . a At tenninal ends. l i l a At intermediate locations selected by one of the following criteria:
- At each pipe fitting (e.g., elbow, tee, cross, flange, and nonstandard fitting),
welded attachment, and valve. Where the piping contains no fittings, welded l attachments, or valves, at one location at each extreme of the piping run l adjacent to the protective structure.
- At each location where stresses calculated by the sum of Equations 9 and 10 in NC/ND-3653, ASME Code, Section III, exceed 0.8 times the sum of the stress limits given in NC/ND3653.
G As a result of piping reanalysis caused by differences between the design configuration and the as-built configuration, the highest stress locations may be shift-d; however, the initially determined intermediate break locations may be used unless a redesign of the piping resulting in a change in the pipe parameten (diameter, wall thickness, routing) is required, or the dynamic effects from the new (as-built) intennediate break location , are not mitigated by the original pipe whip restraints andjet shields. ; Non-ASME Class Piping Breaks in seismically analyzed non-ASME Class (not ASME Class 1,2, or 3) piping are postulated according to the same requirements for.ASME Class 2 and 3 piping above. I Separation and interaction requirements between seismically analyzed and non-seismically analyzed oinine are met as described in Subsection 3.7.3.8. . Separating Structure With High-Energy Lines If a stmcture separates a high-energy line from a safety-related component, the j separating structure is designed to withstand the consequences of the pipe break in the high<nergy line at locations that the aforementioned criteria require to be postulated. However, as noted in Subsection S.6.1.3, some structures that are identified as necessary by the HEISA evaluation (i.e., based on no specific break locations), are designed for wont-case loads. J
\
Y l l Protection Ageinst Dynamic Effects Associated with the Pontulated Rupture of Piping- Amendment 1 3.6-13 i
25AS113 Rev. A SBWR samtantsakryAnarysis neport O 3.6.2.1.2 Locations of Postulated Pipe Cracks Postulated pipe crack locations are selected as follows: Piping Meeting Separation Requirements l Based on the HEISA evaluation described in Subsection 3.6.1.3, the high- or moderate-energy lines which meet the separation requirements are not identified with particular crack locations. Cracks are postulated at all possible points that are necessary to demonstrate adequacy of separation or other means of protections provided for safety-related structures, systems and components. High-Energy Piping With the exception of those portions of piping identified above, leakage cracks are postulated for the most severe environmental effects as follows: a For ASME Code, Section IU Class 1 piping, at axial locations where the calculated stress range by Equation 10 and either Equation 12 or Equation 13 in NB-3653 exceeds 1.2 Sm-a For ASME Code, Section III Class 2 and 3 or non-ASME class piping, at axial locations where the calculated stress by the sum of Equations 9 and 10 in NC/ND-3653 exceeds 0.4 times the sum of the stress limits given in NC/ND-3653. m Non-ASME class piping which has not been evaluated to obtain stress information have leakage cracks postulated at axial locations that produce the most severe emironmental effects. Moderate-Energy Piping in Containment Penetration Areas Leakage cracks are not postulated in tho e portions of piping from the containment wall to and including the inboard or outboard isolation vahes, provided (1) they meet the requirements of the ASME Code, Section III, NE-1120, and (2) the stresses calculated by the sum of Equations 9 and 10 in ASME Code, Section III, NC-3653 do not exceed 0.4 times the sum of the stress limits given in NC-3653. Moderate Energy Piping in Areas Other Than Containment Penetration i a Irakage cracks are postulated in piping located adjacent to safety-related structures, systems or components, except:
- Where exempted above.
- For ASME Code, Section III, Class 1 piping the stress range calculated by Equation 10 and either Equation 12 or Equation 13 in NB-3653 is less than 1.2 Sm.
16-14 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping - Amendment 1
4 l j 25A5113 Rev. A SBWR suarantsannyAurysis sepan C (
- For ASME Code, Section Ill, Class 2 or 3 and non-ASME class piping, the stresses calculated by the sum of Equations 9 and 10 in NC/ND-3653 are less than 0.4 times the sum of the stress limits given in NC/ND-3653.
m leakage cracks, unless the piping system is exempted above, are postulated at axial and circumferential locations that result in the most severe emironmental consequences. m Leakage cracks are postulated in fluid system piping designed to nonseismic standards as necessary to meet the emironmental protection requirements of - Subsection 3.6.1.1. Moderate-Energy Piping in Proximity to High-Energy Piping Moderate-energy fluid system piping or portions thereof that are located within a compartment of confined area involving considerations for a postulated break in high-energy fluid system piping are acceptable without postulation of throughwall leakage cracks except where a postulated leakage crack in the moderate-energy fluid system piping results in more severe environmental conditions than the break in the proximate high-energy fluid system piping,in which case the provisions of this subsection are ; applied. 3.6.2.1.3 Types of Breaks and Cracks to be Postulated Pipe Breaks The following types of breaks are postulated in high-energy fluid system piping at the locations identified by the criteria specified in Subsection 3.6.2.1.1. m No breaks are postulated in piping having a nominal diameter less than or equal to ) i 1 inch. Instrument lines 1 in. and less nominal pipe or tubing size meet the provision of Regulatory Guide 1.11 (Table 3.2-1). Additionally, the 1-1/4 in. HCU fast scram lines do not require special protection measure because of the following reasons: The piping to the control rod drives from the hydraulic control units (HCUs) are located in the contaimnent under reactor vessel, and in the reactor building away from other safety-related equipment; therefore, should a line fail, it would not affect any safety-related equipment but only impact on other HCU lines. As discussed in Subsection 3.6.1.1, a whipping pipe will only rupture an impacted pipe of smaller nominal pipe size or cause a throughwall crack in the same nominal pipe size but with thinner wall thickness. f - The total amount of energy contained in the 1-1/4 in. piping between the Q] norTnally closed scram insert valve on the HCU module and the ball < heck valve in the control rod housing is small. In the event of a rupture of this line, the ball-check valve will close to prevent reactor vessel flow out of the break. Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amor.dment 1 1 6-15
l l SSA5113 Rev. A SBWR StandardSafetyAnalysis Report 9
- Even if a number of the HCU lines ruptured, the control rod insertion function would not be impaired, since the electrical motor of the fine motion control drive would drive in the control rods.
m Longitudmal breaks are postulated only in piping having a nominal diameter equal to or greater than 4 inches. m Circumferential breaks are only assumed at all terminal ends. m At each of the intermediate postulated break locations identified to exceed the stress and usage factor limits of the critena in Subsection 3.6.2.1.1, consideration is given to the occurrence of either a longitudinal or circumferential break. Framination of the state of stress in the vicinity of the postulated break location is used to identify the most probable type of break. If the maximum stress range in the longitudinal direction is greater than 1.5 times the maximum stress range in the circumferential direction, only the circumferential break is postulated. Conversely, if the maximum stress range in the circumferential direction is greater than 1.5 times the stress range in the longitudinal direction, only the longitudinal break is postulated. If no significant difference between the circumferential and longitudinal stresses is determined, then both types of breaks are considered. m Where breaks are postulated to occur at each intermediate pipe fitting, weld attachment, or valve without the benefit of stress calculations, only circumferential breaks are postulated. m For both longitudinal and circumferential breaks, after assessing the contnbution of upstream piping flexibility, pipe whip is assumed to occur in the plane defined by the piping geometry and configuration for circumferential breaks and out of plane for longitudinal breaks and to cause piping movement in the direction of thejet i reactions. Structural m embers, piping restraints, or piping stiffness as demonstrated l by inelastic limit analysis are considered in determining the piping movement limit l (alternatively, circumferential breaks are assumed to result in pipe severance and separation amounting to at least a one-diameter lateral displacement of the ruptured piping sections). m For a circumferential break, the dynamic force of thejet discharged at the break location is based upon the effective cross-sectional flow area of the pipe and on a calculated fluid pressure as modified by an analytically or experimentally determined thrust coefficient. Limited pipe displacement at the break location,line restrictions, flow limiten, positive pump controlled flow, and the absence of energy reservoirs are used, as applicable, in the reduction of thejet discharge. m Longitudiraal breaks in the form of avint split without pipe severance are postulated in the center of the piping at two diametrically opposed points (but not 3.G16 Protection Against Dynamic Effects Associated with the Postulated Ruptum of Piping - Amendment 1
i ! 2SA5113 Rev A l SBWR samtantsanyAnnoysisnoran i it ) O l l concurrently) located so th:a na reaction force is perpendicular to the plane of the piping configuration and producis out-of-plane bending. Alternatively, a single split is assumed at the section of highest tensile stress as determined by detailed stress analysis (e.g., finite element analysis). l e The dynamic force of the fluidjet discharge is based on a circular or elliptical
- (2D x 1/2D) break area equal to the effective cross-sectional flow area of the pipe at the break location and on a calculated fluid pressure modified by an analytically or experimentally determined thrust coefficient as detennined for a circumferential break at the same location. Line restrictions, flow limiters, positive pump controlled j flow, and the absence of energy reservoirs may be taken into account as applicable in the reduction ofjet discharge.
Pipe Cracks The following criteria are used to postulate throughwall leakage cracks in high- or moderate-energy fluid systems or portions of systems: ! m Cracks are postulated in moderate-energy fluid system piping and branch runs exceeding a nominal pipe size of 1 inch. m At axial locations determined per Subsection 3.6.2.1.2, the postulated cracks are oriented circumferentially to result in the most severe environmental ; consequences. I j s Crack openings are assumed as a circular orifice of area equal to that of a rectangle , l having dimensions one42alf-pipe-diameter in length and one-half-pipe-wall ! thickness in width. I t a The flow from the crack opening is assumed to result in an environment that wets all unprotected components within the compartment, with consequent flooding in the compartment and communicating compartments, based on a conservatively , estimated time period to effect corrective actions. . l ! 3.6.2.2 Analytic Methods to Define Blowdown Forcing Functions and Response Models l Analytic Methods to De6ne Blowdown Forcing Functions The rupture of a pressurized pipe causes the flow characteristics of the system to ; change, creating reaction forces which can dynamically excite the piping system. The ; reaction forces are a function of time and space and depend upon fluid state within the pipe prior to rupture, break flow area, frictional losses, plant system characteristics, l piping system, and other factors. The methods used to calculate the reaction forces for various piping systems are presented in the following subsections. l The criteria that are used for calculation of fluid blowdown forcing functions include: r Protection Agairst Dynamic Effecto Associated with the Postulated Rupture of Piping - Amendinent 1 3.6-17
25A5113 Rev. A SBWR senadardseneryAnarysis neport 9 m Circumferential breaks are assumed to result in pipe severance and separation amounting to at least a one-diameter lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural members, or piping stiffness as may be demonstrated by inelastic limit analysis (e.g., a plastic hinge in the piping is not developed under loading). s The dynamic force of thejet discharge at the break location is based on the cross-sectional flow area of the pipe and on a calculated fluid pressure as modified by ar . ally or experimemally-determined thrust coefficient. Line restrictions, flow lim > positive pump <ontrolled flow, and the absence of energy reservoirs are taken a accounts, as applicable,in the reduction ofjet discharge. m All breaks are assumed to attain full size within one mdhsecond after break initiation. Blowdown forcing functions are determined by the method specified in Appendix B of ANSI /ANS-58.2. Pipe Whip Oynamic Response Analyses The prediction of time-dependent and steady thrust reaction loads caused by blowdown of subcooled, saturated, and two-phase fluid from ruptured pipe is used in design and evaluation of dynamic effects of pipe breaks. A discussion of the analytical methods employed to compute these blowdown loads is given above. Following is a discussion of
' balytical methods used to account for this loading.
The cdteda used for performing the pipe whip dynamic response analyses include the following: a A pipe whip analysis is performed for each postulated pipe break. However, a given analysis can be used for more than one postulated break location if the blowdown forcing function, piping and restraint system geo metry, and piping and restraint system properties are conservative for other break locations. m The analysis includes the dynamic response of the pipe in question and the pipe whip restraints which transmit loading to the support structures. l l m The analytical model adequately represents the mass / inertia and stiffness properties of the system. m Pipe whipping is assumed to occur in the plane denned by the piping geometry and configuration and to cause pipe movement in the direction of thejet reaction. m Piping within the broken loop is no longer considered part of the RCPB. Plastic deformation in the pipe is considered as a potential energy absorber. Limits of strain are imposed which are similar to strain levels allowed in restraint plastic 3.6-18 Protection Agsinst Dynamic Effects Associated with the Postulated Rupture of Piping- Amendenent 1
l 25A5113 Rev. A SBWR Standard SafetyAnalysis Report li A i N.) , members. Piping systems are designed so that plastic instability does not occur in l the pipe at the design dynamic and static loads unless damage studies are performed which show the consequences do not result in direct damage to any safety-related system or component. m Components, such as vessel safe ends and valves which are attached to the broken
- piping system, do not serve a safety-related function, or failure of which would not further escalate the consequences of the accident are not designed to meet ASME Code-imposed limits for safety-related components under faulted loading.
However,if these components are required for safe shutdown or serve to protect the i structural integrity of a safety-related component, limits to meet the Code requirements for faulted conditions and limits to ensure required operability will be met. l An analysis for pipe whip restraint selection using the piping design analysis (PDA) computer program and a pipe break modeling program (ANSYS) are perfonned as described in Appendix SB,which predicts the r esponse ofa pipe subjected to the thrust force occurring after a pipe break. The program treats the situation in terms of generic fm pipe break configuration which involves a straight, uniform pipe fixed at one end and ('# ) subjected to a time-dependent thrust force at the other end. A typical restraint used to reduce the resulting defonnation is also included at a location between the two ends. Nonlinear and time-independent stress strain relationships are used to model the pipe and the restraint. Using a plastic-Innge concept, bending of the pipe is assumed to occur only at the fixed end and at the location supported by the restraint. Effects of pipe shear deflection are considered negligible. The pipe-bending moment-deflection (or rotation) relation used for these locations is obtained from a static nonlinear cantilever-beam analysis. Using the moment-rotation relation, nonlinear equations of motion of the pipe are formulated using energy considerations and the equations are numerically integrated in small time steps to yield time.-history of the pipe i motion. The piping stresses in the containment penetration areas are calculated by the ANSYS computer program, a program as described in Appendix 3B. The program is used to perform the non-linear analysis of a piping system for time varymg displacements and forces due to postulated pipe breaks. 3.6.2.3 Dynamic Analysis Methods to Verify integrity and Operability 3.6.2.3.1 Jet unpingement Analyses and Effects on Safety-Related Components p The methods used to evaluate thejet effects resulting from the postulated breaks of , Q high-energy piping are described in Appendices C and D of ANSI /ANS 58.2 and l presented in this subsection. Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 a6-19
25AS113 Ras A l SBWR Sundanf Sahrty Analysis Report l 9 The criteria used for evaluating the effects of fluidjets on safety-related structures, systems, and components are as follows: a Safety-related structures, systems, and components are not impaired so as to preclude safety 4 elated functions. For any given postulated pipe break and consequentjet, those safety-related structures, systems, and components needed to safely shut down the plant are identified. m Safety-related structures, systems, and components which are not necessary to safely shut down the plant for a given break are not protected from the consequences of the fluidjet. m Safe shutdown of the plant caused by postulated pipe ruptures within the RCPB is not aggravated by sequential failures of safety < elated piping and the required emergency cooling system performance is maintained. m Off4ite dose limits are in compliance with 10CFR100. m Postulated breaks resulting injet impingement loads are assumed to occur in high-energy lines at full (100%) power operation of the plant. m Throughwall leakage cracks are postulated in moderate-energy lines and are assumed to result in wetting and spraying of safety-related structures, systems, and components. m Reflectedjets are considered only when there is an obvious reflecting surface (such as a flat plate) which directs thejet onto safety-related equipment. Only the fint reflection is considered in evaluating potential targets. m Potential targets in thejet path are t.onsidered at the calculated final position of the l broken end of the ruptured pipe. This selection of potential targets is considered adequate due to the large number of breaks analyzed and the protection provided from the effects of these postulated breaks. The analytical methods used to determine which targets will be impinged upon by a l fluidjet and the correspondingjet impingement load include: a The direction of the fluidjet is based on the arrested position of the pipe during steady-state blowdown. m The impingingjet prt>ceeds along a straight path, e 'Ihe total impingement force acting on any cross 4ectional area of thejet is time and j distance invariant with a totaliaagnitude equivalent to the steady-state fluid l l 1 1620 Protection Against Dynamic Effects Associated with the PostulatedRupture of Piping - Amendment 1
i i 25A5113 Rtv. A l SBWR StandardSafetyAnalysis Report lIG') l blowdown force given in Subsection 3.6.2.2 and withjet characteristics shown in Mgure 3.41. m Thejet impingement force is uniformly distributed across the cross-sectional area
- of thejet and only the portion intercepted by the target is considered.
m The break opening is assumed to be a circular orifice of cross-sectional flow area equal to the effective flow area of the break. m Thejet impingement force is equal to the steady state value of the fluid blowdown l force calculated by the methods described in Subsection 3.6.2.2. e The distance ofjet travel is divided into two or three regions. Region 1 (Figure 3.G1, l items a, b, c) extends from the break to the asymptotic area. Within this region the ! discharging fluid flashes and undergoes expansion from the break area pressure to l the atmospheric pressure. In Region 2 thejet expands further. For partial-separation circumferential breaks, the area increases as thejet expands. In Region 3, thejet expands at a half angle of10 (Figure 3.61, items a and c). e The analytical model for estimating the asymptoticjet area for subcooled water and y saturated water assumes a constantjet area. For fluids discharging from a break which are below the saturation temperature at the corresponding room pressure or have a pressure at the break area equal to the room pressure, the free expansion does not occur. m The distance downstream from the break where the asymptotic area is reached (Region 2) is calculated for circumferential and longitudinal breaks. m Both longitudinal and fully separated circumferential breaks are treated similarly. The value ofIL/D used in the blowdown calculation is also used forjet impingement. m Circumferential breaks with partial (i.e., h<D/2) separation between the two ends of the broken pipe not significantly offset (i.e., no more than one pipe wall thickness lateral displacement) are more difficult to quantify. For these cases, the following assumptions are made. l
- Thejet is uniformly distributed around the periphery.
- Thejet cross-section at any cut through the pipe axis has the configuration depicted in Mgure 3.41, item b. Thejet regions are also shown.
(n) v Thejet force F j= total blowdown F.
- The pressure at any point intersected by thejet is:
Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 16-21
25A5113 R v. A SBWR standantsafety Anairsis neport O F P=lA 3 (3.41) a where AR= the total 360 area of thejet at a radius equal to the distance from the pipe centerline to the target
- The pressure of thejet is then multiplied by the area of the target submerged within thejet.
m Target loads are determined using the following procedures:
- For both the fully separated circumferential break and the longitudinal break, thejet is studied by determining target locations vs. asymptomatic distance and applying ANSI /ANS 58.2, Appendices C and D.
- For circumferential break limited separation, thejet is analyzed by using different equations of ANSI /ANS 58.2, Appendices C and D and determining respective target and asymptomatic locations.
- After determination of the total area of thejet at the target, thejet pressure is calculated by:
F P 3=d A, (3.6-2) where P= i incident pressure A= x area of the expandedjet at the target intersection. If the effective target area (A te ) is less than the expandedjet area (A te s Ax), the target is fully submerged in thejet and the impingement load is equal to (P 3 ) (A te ). If the effective target area is greater than the expandedjet area (Ate > Ax ), the target intercepts the entirejet and the impingement load is equal to (P3 ) (Ax) = Fj . The effective target area (Ate) for various geometries follows: a Flat Surface - For a case where a target with physical area Asis oriented at angle d with respect to thejet axis and with no flow reversal, the effective target area A et is: A,, = ( A,) (sin $) . (3.6-3) O 3.6-22 Protection Against Dynstnic Effects Associated with the PostulatedRupture of Piping - Amendment 1
l 1 l 2SA5113 R:v. A SBWR standant saretyAnalysis neport i ! v l 1 l 4 m Pipe Surface- As thejet hits the convex surf me of the pipe,its forward momentum 1 is decreased rather than stopped; therefore, the jet impingement load on the impacted area is expected to be reduced. For consenatism, no credit is taken for this reduction and the pipe is assumed to be impacted with the fullimpingement load. However, where shape factors arejustifiable, they may be used. The effective target area A1 , is: A, = (D g) (D) (3.6-4) where D= A diameter of thejet at the targetinterface D= pipe OD of target pipe for a fully submerged pipe. When the target (pipe) is larger than the area of thejet, the effective target area equals the expandedjet area n A ,, = A ,. (3.6-5)
! ) \ /
U m For all cases, thejet area (Ax ) is assumed to be uniform and the load is uniformly distributed on the impinged target area Aw. 3.6.2.3.2 Pipe Whip Effects on Safety-Related Components This subsection provides the criteria and methods used to evaluate the effects of pipe displacements on safety-related structures, systems, and components following a postulated pipe rupture. Pipe whip (displacement) effects on safety-related structures, systems, and components can be placed in two categories: (1) pipe displacement effects on components (nozzles, valves, tees, etc.) which are in the same piping run that the break occurs in; and (2) pipe l whip or controlled displacements onto external components such as building structure, j i other piping systems, cable trays, and conduits, etc. 1 3.6.2.3.3 Pipe Displacement Effects on Components in the Same Piping Run i The criteria for determining the effects of pipe displacements on inline components l are as follows: l l l m Components such as vessel safe ends and valves which are attached to the broken g piping system and do not serve a safety function or failure ofwhich would not l ('] further escalate the consequences of the accident need not be designed to meet ASME Code Section III-imposed limits for safety-related components under faulted loading. Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 3.6-23
2SA5113 Rev. A SBWR standerdsareryAnalysisReport O a If these components are required for safe shutdown or serve to protect the structural integrity of a safety-related component, limits to meet the ASME Code requirements for faulted conditions and limits to ensure required operability are met. 3.6.2.3.4 Pipe Displacement Effects on Safety-Related Structures, Other Systems, and Components The criteria and methods used to calculate the effects of pipe whip on external components consist of the following: a The effects on safety-related structures and barriers are evaluated in accordance with the barrier design procedures given in Subsection 3.5.3. m If the whipping pipe impacts a pipe of equal or greater nominal pipe diameter and equal or greater wall thickness, the whipping pipe does not rupture the impacted pipe. Otherwise, the impacted pipe is assumed to be ruptured. m If the whipping pipe impacts other components (valve actuators, cable trays, conduits, etc.),it is assumed that the impacted componentis unavailable to mitigate the consequences of the pipe break event. m Damage of unrestrained whipping pipe on safety-related structures, components, and systems other than the ruptured one is prevented by either separating high energy systems from the safety-related systems or providing pipe whip restraints. 3.6.2.3.5 Loading Combinations and Design Criteria for Pipe Whip Restraint Pipe whip restraints, as differentiated from piping supports, are designed to function and carryloads for an extremely low-probability gross failure in a piping system carrymg high-energy fluid. In the SBWR plant, the piping integrity does not depend on the pipe whip restraints for any piping design loading combination, including an earthquake, but shall remain functional following an earthquake up to and including the SSE (Subsection 3.2.1). When the piping integrity is lost because of a postulated break, the pipe whip restraint acts to limit the movement of the broken pipe to an acceptable j distance. The pipe whip restraints (i.e., those devices which serve only to control the l movement of a ruptured pipe following gross failure) will be subjected to once-m-a- ) lifetime loading. For the purpose of the pipe whip restraint design, the pipe break is considered to be a faulted condition (Subsection 3.9.3.1) and the structure to which the restraint is attached is also analyzed and designed accordingly. The pipe whip restraints ) are non-ASME Code components; however, the ASME Code requirements may be used in the design selectively to assure its safety-related function if ever needed. Other methods (i.e., testing) with a reliable database for design and sizing of pipe whip i restraints can also be used. 3.6-24 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendnwnt 1
2SA5113 Rev. A SBWR standantsareryAnairsisaeport p i ! Q ,) The pipe whip restraints utilize energy absorbing U-rods to attenuate the kine tic energy of a ruptured pipe. A typical pipe whip restraint is shown in Egure 3.f>.2. The principal feature of these restraints is that they are installed with several inches of annular clearance between them and the process pipe. This allows for installation of normal piping insulation and for unrestricted pipe thermal movements during plant operation. Select critical locations inside the pnmary containment are also monitored during hot functional testing to provide verification of adequate clearances prior to plant operation. The specific design objectives for the restraints are: a The restraints shall in no way increase the reactor coolant pressure boundary stresses by their presence during any normal mode of reactor operation or condition. s The restraint system shall function to stop the movement ofa pipe falhu e (gross loss of piping integrity) without allowing damage to critical components or missile development. m The restraints should provide minimum hindrance to inservice inspection of the
, process piping.
(-) ( For the purpose of design, the pipe whip restraints are designed for the following dynamic loads: a Blowdown thrust of the pipe section that impacts the restraint. s Dynamic inertia loads of the moving pipe section which is accelerated by the blowdown thrust and subsequent impact on the restraint. m Design characteristics of the pipe whip restraints are included and verified by the pipe whip dynamic analysis described in Subsection 3.6.2.2. s Since the pipe whip restraints are not contacted during normal plant operation, the postulated pipe rupture event is the only design loading condition. Strain rate effects and other material property variations have been considered in the design of the pipe whip restraints. The material properties utilized in the design have included one or more of the following methods: a Code minimum or specification yield and ultimate strength ulues for the affected components and structures are used for both the dynamic and steady-state events. s Not more than a 10% increase in minimum code or specification strength values is 7 ( ) used when designing components or structures for the dynamic event, and code minimum or specification yield and ultimate strength values are used for the steady-l'd state loads. l l Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 3.6-25 l l
25A5113 Rect. A SBWR sundantsareryAnairsis aeport O m Representative or actual test data values are used in the design of components and stmctures indudingjustifiably elevated strain rate-affected stress limits in excess of l I 10E 1 m Representative or actual test data are used for any affected component (s) and the i minimum code or specification values are used for the structures for the dynamic l and the steady-state events. l l 3.6.2.4 Guard Pipe Assembly Design l The SBWR does not require guard pipes. 3.6.3 Leak-Before-Break Evaluation Procedures Per Regulatory Guide 1.70, Revision 3, November 1978, the safety analysis (Section 3.6) has traditionally addressed the protection measures against dynamic effects associated with the non-mechanistic or postulated ruptures of piping.The dynamic effects are defined in the introduction to Section 3.6. Three forms of piping failure (full flow area circumferential and longitudinal breaks, and throughwall leakage crack) are postulated in accordance with Subsection 3.6.2 and Branch Technical Position MEB 3-1 of NUREG 0800 (Standard Review Plan) for their dynamic and environmental effects. However,in accordance with the modified Geneml Design Criterion 4 (GDC4), effective November 27,1987, (Reference 3.f>-1), the mechanistic leak-before-break (LBB) approach, justified by appropriate fracture mechanics techniques,is recognized as an acceptable procedure under certain conditions to exclude design against the dynamic effects from postulation of breaks in high energy piping.The LBB approach is not used to exclude postulation of cracks and associated effects as required in Subsection 3.6.2.2. It is required, as mentioned in Subsection 3.6.1, that a report shall be prepared demonstrating LBB qmbfication of the piping. The report shall be prepared in accordance with the guidelines presented in Appendix 3C in conjunction with the stress report of the piping. The qualified piping, referred to in this SSAR as the LBB-qualified piping, will be excluded from pipe breaks, which are required to be postulated by Subsections 3.6.1 and 3.6.2, for design against their potential dynamic effects. The following subsections describe (1) certain design bases where the LBB approach is not recognized by the NRC as applicable for exclusion of pipe breaks, and (2) certain conditions which limit the LBB applicability. Appendix 3C provides guidelines for LBB applications describing in detail the following necessary elements of an LBB report to be submitted by a COL applicant for NRC approval: (1) fracture mechanics methods, (2) leak rate prediction methods, (3) leak detection capabilities, and (4) typical special 4 considerations for LBB applicability. Also included in Appendix 3C is a list of candidate l piping systems for LBB qualification. The LBB application approach described in this ; i 3.6-26 Protection Against Dynamic Eflects Associated with the Postulated Rupture of Piping- Amendment 1 l 1
. . . , - - - . - . - . - ~~ .~ _-
. ~ - - -
4 k J ! 2SA5113 Rsv. A l } } SBWR saadantsannyAnarysisnepour 1 j subsection and Appendix 3C is consistent with that documented in Draft SRP 3.6.3 i (Reference 3.6-2) and NUREG-1061 (Reference 3.6-3). i !- 3.6.3.1 Scope of LBB Applicability l i ' } The LBB approach is not used to replace existing regulations or criteda pertaining to l- the design bases of the Emergency Core Cooling System (Section 6.3), containment ; i system (Section 6.2), or environmental qualification (Section 3.11). However, : l consistent with modi 6ed GDC4, the design bases for dynamic qualification of . l l ' mechanical and electrical equipment (Section 3.10) may exclude the dynamic load or { vibration effects resulting from postulation of breaks in the LBB. qualified piping. This ; is also reflected in a note to Table 3.9-2 for ASME components. The LBBqualmed { j piping may not be excluded from design bases for environmental qualmcation unless ; j the regulation permits it at the time of LBB qualification. For cladfication, it is noted - , j- that the LBB approach is not used to relax the design requirements of the containment t , system that includes the Containment Vessel (CV), vent system (vertical flow channels ; j and horizontal vent chscharges), drywell zones, suppression pool (wetwell), vacuum i j breakers, CV penetrations, and drywell head. } O 3.6.3.2 Conditions for LBB Applicability [ } f j The LBB approach is not applicable to piping systems where operating expedence has l indicated particular susceptibility to failure from the effects ofintergranular stress i corrosion cracking (IGSCC), waterhammer, thermal fatigue, or erosion. Necessary l preventive or mitigation measures are used and necessary analyses are performed, as , l discussed below, to avoid concerns for these effects. Other concerns, such as creep, j
- brittle cleavage-type failure, potential indirect source of pipe failure, and deviation of l as-built piping conGguration, are also addressed. ,
1 i j s Degradation by erosion, erosion / corrosion and erosion / cavitation caused by ! j unfavorable flow conditions and water chemistry is namined. The evaluation is ! l based on the industry experience and guidelines. Additionally, fabrication wall ; j thinning of elbows and other fittings is considered in the purchase speci6 cation to ; i assure that the code minimum wall requirements are met. These evaluations l l demonstrate that these mechanisms are not potential sources of pipe rupture. ! 1 i ] a The SBWR plant design involves operation below 700*F in ferritic steel piping and ; j below 800 F in austenitic steel piping. This assures that creep and creep-fatigue are j j not potential sources of pipe rupture. ! a The design also assures that the piping material is not susceptible to brittle cleavage-_ ! type failure over the full range of system operating temperatures (i.e., the matedal j !- is on the upper shelf). i
- i
] Protection Apoinst Dynamic Effects Annociated with the Poetuisted Rupture of Piping- Amendment 1 16-27 . d L l
. - - - . , . ~ - , - . .
1 1 2SA5113 Rov. A SBWR Standant SafetyAnalysis Report O' s The SBWR plant design specifies use of austenitic stainless steel piping made of material (e.g., nuclear grade or low carbon type) that is recognized as resistant to j IGSCC. The material of major high<nergy piping in the primary and secondary containments is carbon steel or ferritic steel, except for the austenitic stainless reactor water cleanup piping in the containment. m A systems evaluation of potential waterhammer is made to assure that pipe rupture caused by to this mechanism is unlikely. Waterhammer is a generic term including ) various unanticipated high frequency hydrodynamic events such as steam hammer and water slugging. To demonstrate that waterhammer is not a significant ; contributor to pipe rupture, reliance on historical frequency of waterhammer events in specific piping systems coupled with a review of operating procedures and conditions is used for this evaluation. The SBWR design includes features such as vacutun breakers andjockey pumps, coupled with improved operational procedures to reduce or eliminate the potential for waterhammer identified by past expedence. Certain anticipated waterhammer events, such as a closure of a nive, are accounted for in the Code design and analysis of the piping. m The systems enluation also addresses a poten tial for fatigue cracking or failure from thermal and mechanical induced fatigue. Based on past experience, the piping design avoids potential for significant mixing ofhigh and low temperature fluids or mechanical vibration. The startup and preoperational monitoring assures avoidance of detrimental mechanical vibration. m Based on experience and studies by Lawrence Livermore laboratory, potential sources ofindirect pipe rupture are extremely unlikely. Compliance with the snubber surveillance requirements of the technical specifications assures that snubber faihre rates are acceptably low. m Initial LBB evaluation is based on the design configuration and stress levels that are acceptably higher than those identified by the initial analysis. This evaluation is reconciled when the as-built configuration is documented and the Code stress enluation is reconciled. It is assured that the as-built configuration does not deviate significantly from the design configuration to invalidate the initial LBB enluation, or a new enluation coupled with necessary configuration modifications is made to assure applicability of the LBB procedure. j I O l 3.6-28 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping - Amendment 1
25A5113 Rsv. A l SBWR senadard sanonyAnalysis Report o 3.6.4 COL License Information Details of Pipe Break Analysis Results and Protection Methods The following shall be provided by the COL applicant: , l a A summary of the dynamic analyses applicable to high-energy piping systems in i accordance with Subsection 3.6.2.5 of Regulatory Guide 1.70. This shall include the following:
- Sketches of applicable piping systems showing the location, size and orientation of postulated pipe breaks and the location of pipe whip restraints andjet impingement barriers.
- A summary of the data developed to select postulated break locations including calculated stress intensities, cumulative usage facton and stress ranges as delineated in BTP MEB S-1.
1 m For failure in the moderate-energy piping systems, descriptions showing how safety-related systems are protected from the resultingjets, floodmg and other adverse environmental effects. V a Identification ofprotective measures provided against the effects of postulated pipe failures for protection of each of the systems listed in Tables 3.6-1 and 3.6-2. m The details of how the MSIV functional capability is protected against the effects of postulated pipe failures. l s Typical examples,if any, where protection for safety-related systems and , components against the dynamic effects of pipe failures include their enclosure in suitably designed structures or compartments (including any additional drainage system or equipment environmental qualification needs). m The details of how the feedwater line check and feedwater isolation valves functional capabilities are protected against the effects of postulated pipe failures. 3.6.5 References 3.6-1 Modification of General Design Criterion 4, Requirements for Protection Against Dynamic Effects of Postulated Pipe Rupture, Federal Register, Volume 52, No. 207, Rules and Regulations, pages 41288 to 41295, October 27,1987. 3.6-2 Standard Review Plan; Public Comments Solicited, Federal Register, Volume 52, No.167, Notices, pages 32626 to 32633, August 28,1987. 3.6-3 NUREG-1061, Volume 3, Evaluation of Potential for Pipe Breaks, Report of I the US NRC Piping Review Committee, November 1984. Protecti>n Against Dynamic Effects Associated with the Postulated Rupture of Piping- Amendment 1 a6-29
25A5113 Rev. A SBWR sanndardSaktyAnalysis Report O Table 3.6-1 Safety-Related Systems, Components, and Equipment for Postulated Pipe Failures inside Containment
- 1. Reactor Coolant Pressure Boundary (up to and including the outboard isolation valves)
- 2. Containment isolation System and Containment Boundary (including liner plate)
- 3. Reactor Protection System (SCRAM signals)
- 4. Control Rod Drive System (scram / rod insertion)
- 5. Flow restrictors (passive)
- 6. Flammability Control System
- 7. Passive Containment Cooling System
- 8. Gravity-Driven Cooling System
- 9. The following equipment / systems or portions thereof required to assure the proper operation of thoso safety-related items listed in items 1 through 8.
(a) Class 1E electrical systems , (b) Instrumentation Table 3.6 2 Safety-Related Systems, Components, and Equipment O for Postulated Pipe Failures Outside Containment
- 1. Containment isolation System and Containment Boundary (including liner plate)
- 2. Reactor Protection System (SCRAM signals)
- 3. Flow restrictors
- 4. Fuel and Auxiliary Pool Cooling,IC/PCCS make-up line only
- 5. Safety Envelope Boundary
- 6. The following equipment / systems or portions thereof required to assure the proper operation of those safety-related items listed in items 1 through 5.
(a) Room Coolers (b) Component Cooling System (c) instrumentation (d) Process Sampling System O 1 3.6-30 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping - Amendrnent 1
1 25A5113 Rev. A l SBWR saadantsafeyAnawsanon gs Table 3.6-3 High Energy Piping inside Containment
- 1. Feedwater System
- 2. Main Steam
- 3. CRD (to and from HCU) l
- 4. Reactor Water Cleanup (suction and discharge)
- 5. IC Steam Line (from RPV to IC)
- 6. Passive Containment Cooling System Steam Line (from RPV to outboard isolation valve)
- 7. Gravity-Driven Cooling System injection Line (from RPV to isolation valve)
Table 3.6-4 High Energy Piping Outside Containment
- 1. Reactor Water Clear.:'o (suction from RPV and discharge in Feedwater Line)
- 2. Feedwater Lines in Steam Tunnel
- 3. Main Steam Lines in Steam Tunnel
- 4. CRD (to and from HCU) ,
1 i l l 1 l l
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Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping - Amendment 1 3.6-31
l 25AS113 Rev. A l SBWR sonderd safetyAnalysis aeport i ASYMPTOTIC AREA A4 , BREAK AREA --
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23A5113 R:v. A SBWR StandardSafetyAnalysis Report
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( h j tv/ 3.7 Seismic Design For seismic design purposes, all structures, systems, and components of the Simplified Boiling Water Reactor (SBWR) standard plant are classified into Seismic Category I (GI), Seismic Category II (GII), or Non-Seismic (NS) in accordance with the requirements to withstand the effects of the Safe Shutdown Earthquake (SSE) as defined in Section 3.2. For those GI and GII structures, systems, and components in l the reactor building complex, the effects of other dynamic loads caused by reactor l building vibration (RBV) caused by suppression pool dynamics are also considered in the design. Although this section addresses seismic aspects of design and analysis in accordance with Regulatory Guide 1.70, the methods of this section are also applicable to RBV dynamic loadings, unless noted otherwise. l The safe shutdown earthquake (SSE) is that earthquake which is based upon an evaluation of the maximum earthquake potential considering the regional and local geology, seismology, :md specific characteristics oflocal subsurface material. It is that earthquake which produces the maximum vibratory ground motion for which Seismic Category I stmctures. systems. rj tems and components are designed to remain functional.These systems and components are those necessag to ensure the following:
/m i
f i V'/ a the integrity of the reactor coolant pressure boundary; a the capability to shut down the reactor and maintain it in a safe condition; and a the capability to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable to the guideline exposures of10CFR100. Seismic Category II (GII) includes all plant stmctures, systems, and components which perform no safety-related function, and whose continued function is not required, but whose structural failure or interaction could degrade the functioning of a Seismic Categog I stmcture, system, or component to an unacceptable safety level, or could result in incapacitating injury to occupants of the control room. Thus, this categog includes the structures, systems and components whose structural integrity, not their operational performance, is required. Seismic Categog II structures, systems, and components are designed and/or so physically arranged so that the SSE would not cause unacceptable structural interaction or failure. For fluid systems, this requires an appropriate level of pressure boundary integdty when located near sensitive equipment. Appropriate seismic ductility factors are selected for design to take credit for realistic amounts of energy dissipation in GII items. Seismic Category II (GII) items are those corresponding to positions C.2 and C.4 of Regulatog Guide 1.29. (n) V Non-seismic structures and equipment are those which do not fallinto Seismic Category I or II definitions. NS structures and equipment are designed for seismic requirements in accordance with the Uniform Building Code for Zone 2A. The building structures Seismic Design - Amendment 1 171
25AS?13 Rev. A SBWR standardsafetyAnalysis neport O are classified with an Importance Factor of1.25 as " Essential Buildings." Either of the methods permitted by Uniform Building Code (UBC), simplified analysis or dynamic analysis, is acceptable for determination of seismic loads on NS structures and equipment. The Operating Basis Earthquake (OBE) is not an SBWR design requirement. Consistent with the Draft Appendix S to 10CFR Part 50, the design requirements associated with the OBE, when the level of OBE ground motion is chosen to be one-third of the SSE ground motion, are satisfied without performing explicit response or design analyses. The effects oflow-level earthquakes (lesser magnitude than the SSE) on fatigue evaluation and plant shutdown criteria are addressed in Subsections 3.7.3.2 and 3.7.4.4, respectively. 3.7.1 Seismic Design Parameters 3.7.1.1 Design Ground Motion The peak ground acceleration (PGA) of the Safe Shutdown Earthquake (SSE) is 0.3g in the horizontal direction for the standard plant design. The PGAin the vertical direction is equal to the horizontal PGA. 3.7.1.1.1 Design Response Spectra The design response spectra are constmcted in accordance with RG 1.60 and specified at the finished grade in the free field. The 0.3gSSE design response spectra of vanous damping ratios are shown in Figures 3.7-1 and 3.7-2 for the horizontal and vertical motions, respectively. The horizontal response spectra are equally applicable to two orthogonal horizontal directions. 3.7.1.1.2 Design Time History Seismic input motions in the form of time histories are generated to envelop the design response spectra. The 0.3gSSE acceleration time histories for two horizontal components (H1 and H2) and vertical (VT) component are shown in Figures 3.7-3 through 3.7-5, respectively, together with corresponding velocity and displacement time histories. Each time history has a total duration of 22 seconds. 1 These time histories satisfy the spectrum-enveloping requirement stipulated in the NRC Standard Review Plan (SRP) 3.7.1.The computed response spectra of 2%,3%,4% and 7% damping are compared with the corresponding design RG 1.60 spectra in Hgures 3.7-6 through 3.7-9 for the H1 component,in Figures 3.7-10 through 3.7-13 for the H2 component, and in Figures 3.7-14 through 3.7-17 for the VT component. The response spectra are computed at frequency intervals suggested in Table 3.7.1-1 of SRP 3.7.1 plus three additional frequencies at 40,50, and 100 Hz. 174 Seismic Design- Amendment 1
25A5113 Rev. A l SBWR samtantsa.ryAnarrsisneven
/
'(O) 4 l The time histories of the two horizontal components also satisfy the Power Spectra l Density (PSD) requirement stipulated in Appendix A to SRP 3.7.1. The computed PSD I { functions envelop the target PSD of a maximum 0.3gacceleration with a wide margin in i the frequency range of 0.3 Hz to 24 Hz as shown in Figures 3.7-18 and 3.7-19 for the H1 i and H2 components, respectively. In these figures, the curve labeled as 80% of the target PSD is the minimum PSD requirement. i The time histories of three spatial components are checked for statistically independency. The cross <orrelation coefficient at zero time lag is 0.01351 between H1 and H2,0.07037 between H1 and VT, and 0.07367 between H2 andVT. All of them are i less than 0.16 as recommended in the reference of RG 1.92. Thus, H1, H2, and VT acceleration time histories are mutually statistically independent.
- 3.7.1.2 Percentage of Critical Damping Values i
j Damping values ofvarious structures and components are shown in Table 3.7-1 for SSE dynamic analysis. These damping values are consistent with RG 1.61 SSE damping. For ASME Section III, Dhision 1 Class 1,2, and 3, and ASME/ ANSI B31.1 piping systems, damping values of ASME Code Case N-411-1 may be used as permitted by RG 1.84, in place of RG 1.61 damping. ASME Code Case N-411-1 damping can not be used for G analyzing linear eneruv absorbing supoorts designed in accordance with ASME Code , Case N-420. The damping values shown in Table 3.7-1 are applicable to all modes of a structure or component constructed of the same material. Damping values for systems composed of subsystems with different damping properties are obtained from the procedures described in St bsection 3.7.2.13. 3.7.1.3 Supporting Media for Category I Structures The Seismic Category I structures have concrete mat foundations supported on soil, rock or compacted backfill.The embedment depth, dimensions of the structural foundation, and total structural height for each structure are given in Subsection 3.8.5.1. The soil conditions considered for the design of the standard plant are described in Appendix SA. ] 3.7.2 Seismic System Analysis I This section applies to building structures that constitute primary structural systems. The reactor pressure vessel (RPV) is not a primary structural component but, due to its strong dynamic interaction with supporting structure, is considered as part of the , primary system of the reactor building for the purpose of dynamic analysis. 3.7.2.1 Seismic Analysis Methods Analysis can be performed using any of the following methods: a time history method; j Seismic Design-Amendment 1 1 7-3
25A5113 Rev. A SBWR sondardsareryAnalysis Report O s response spectrum method;
- singly- or multi 4upported system with Uniform Support Motion (USM); or
- multi-supported system with Independent Support Motion (ISM); or a static coefficient method.
3.7.2.1.1 Time History Method The response of a multi-degree-of-freedom linear system subjected to external forces and/or uniform support excitations is represented by the following differential equations of motion in the matrix form: [M] {5} + [C] {d} + [K] {u} = {P} (3.7-1) where [M] = mass matrix [C] = damping matdx [K) = stiffness matrix {u} = column vector of time-dependent relative displacements {d} = column vector of time <lependent relative velocities {5} = column vector of time dependent relative accelerations {P} = column vector of time-dependent applied forces
= -[M] {i g} r 8UPPort excitation in which g{i l i* ' l"*"
vector of time <lependent support accelerations The above equation can be solved by modal superposition or direct integration in the time domain, or by the complex frequency response method in the frequency domam. For the time domain solution, the numerical integration time step is sufficiently small to accurately define the dynamic excitation and to render stability and convergency of the solution up to the highest frequency (or shortest period) of significance. For most of commonly used numerical integration methods (such as Newmark l Method and i Wilson B-method), the maximum time step is limited to one-tenth of the shortest pedod of significance. For the frequency domain solution, the dynamic excitation time history is digitized with time steps no larger than the inverse of two times the highest frequency 3.74 Seismic Design - Arnendment 1
25A5113 Rev. A SBWR sondedsawyAndrsisa n at l
\
s of significance and the frequency interval is selected to accurately define the tmnsfer functions at structural frequencies within the range of significance. l The modal superposition method is used when the equations of motion (Eqt.ation 3.7-1) can be decoupled using the transformation, {u} = [$]{q} (3.7-2) where [$) = mode shape matrix; often mass normalized, i.e., [$]T[M] [$) = [1] , (q) = column vector of normal or generalized coordinates Substituting Equation 3.7-2 into Equation 3.7-1 and premultiplying}}