ML19084A286

DCA Chapter 9 Phase 2 Safety Evaluation Report with Open Items - Public

ML19084A286
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Site:	NuScale
Issue date:	03/25/2019
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NuScale Chapter 9
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Introduction Summary of Application Code of Federal Regulations If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

Regulatory Basis ***

• • Technical Evaluation 9.1.1.4.1 Fuel Assembly Modeling 9.1.1.4.2 Storage Rack Modeling 9.1.1.4.3 Storage Rack Materials 9.1.1.4.3.1 Conformance with Industry Standards 9.1.1.4.3.2 Stainless Steel Materials 9.1.1.4.3.3 Neutron Absorber Material 9.1.1.4.3.3.1 Neutron Absorber Material-Fabrication and Licensing Basis 9.1.1.4.3.3.2 Neutron Absorber Material-Monitoring Program 9.1.1.4.4 Computational Methods and Data 9.1.1.4.5 Computational Method Validation

9.1.1.4.6 Bias and Uncertainty Analysis 9.1.1.4.7 Normal Conditions 9.1.1.4.8 Abnormal Conditions 9.1.1.4.9 Tier 1 Information Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Conditions and Limitations Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis

Technical Evaluation 9.1.2.4.1 GDC 2, Design Bases for Protection Against Natural Phenomena 9.1.2.4.2 GDC 4, "Environmental and Dynamic Effects Design Bases" 9.1.2.4.3 GDC 5, "Sharing of Structures, Systems, and Components" 9.1.2.4.4 GDC 61, "Fuel Storage and Handling and Radioactivity Control"

9.1.2.4.5 GDC 63, "Monitoring Fuel and Waste Storage"

9.1.2.4.6 As Low As Reasonably Achievable Principle Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis ****

• Technical Evaluation 9.1.3.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena" 9.1.3.4.2 GDC 4, "Environmental and Dynamic Effects Design Bases" 9.1.3.4.3 GDC 5, "Sharing of Structures, Systems, and Components" 9.1.3.4.4 GDC 61, "Fuel Storage and Handling and Radioactivity Control" *****

The Pool Cleanup System 9.1.3.4.5 GDC 63, "Monitoring Fuel and Waste Storage" 9.1.3.4.6 10 CFR 20.1101, "Radiation Protection Programs" Initial Test Program Inspections Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis *****Technical Evaluation 9.1.4.4.1 GDC 2, "Design Basis for Protection Against Natural Phenomena"

[O]verhead cranes may be operating at the time that an earthquake occurs. Therefore, the cranes should be designed to retain control of and hold the load, and the bridge and trolley should be designed to remain in place on their respective runways with their wheels prevented from leaving the tracks during a seismic event-.

9.1.4.4.2 GDC 5, "Sharing of Structures, Systems, and Components" 9.1.4.4.3 GDC 61, "Fuel Storage and Handling and Radioactivity Control" Protection against Personnel Radiation Exposure Protection against Radioactivity Releases 9.1.4.4.3 GDC 62, "Prevention of Criticality in Fuel Storage and Handling" Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Conclusion Introduction Summary of Application Regulatory Basis

• • • Technical Evaluation ****

9.1.5.4.1 GDC 1, "Quality Standards and Records" 9.1.5.4.2 GDC 2, "Design Basis for Protection Against Natural Phenomena" 9.1.5.4.3 GDC 4, "Environmental and Dynamic Effects Design Bases" The RBC is operated to move an NPM between its installed operating position in the reactor pool to the refueling pool and back. Travel paths are determined and attributes entered into the RBC control system. Each task is specified and scheduled by the crane operator. Figure 9.1.5-1 shows the safe load paths. Heavy load exclusion zones and safe load paths are defined in operating procedures and equipment drawings. Heavy load exclusion zones are marked in the plant areas where the load cannot be handled. This restriction reduces the probability of a heavy load drop that could result in safe shutdown equipment damage or result in a release of radioactive material that could cause unacceptable radiation exposures. The position control system assists in the alignment of the RBC with the NPM for engagement with the RBC prior to performing lifting operations. Heavy load exclusion zones are dependent on whether or not there is a load on the RBC.

The travel path is chosen to accommodate this information. Repeatability, proper load path, and proper locations are ensured by semi-automatic crane operation. The COL applicant that references the NuScale Power Plant design certification will provide a description of the program governing heavy loads handling. The program should address:

• operating and maintenance procedures

• inspection and test plan

• personnel qualification and operator training The COL applicant that references the NuScale Power Plant design certification will provide a description of the program governing heavy loads handling. The program should address:

• operating and maintenance procedures

• inspection and test plan

• personnel qualification and operator training

• detailed description of the safe load paths for movement of heavy loads 9.1.5.4.4 GDC 5, "Sharing of Structures, Systems, and Components" Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion This section is relevant to light water reactor (LWR) active designs that incorporate a service water system serving as the final heat transfer loop between various heat sources and the plant ultimate heat sink (UHS). The NuScale Power Plant design does not have a service water system. A typical LWR service water system provides essential cooling to safety-related equipment and can also cool nonsafety-related auxiliary components used for normal plant operation. The NuScale Power Plant passive design does not rely on active systems such as a service water system to provide cooling to essential equipment. The NuScale Power Modules are partially submerged in the reactor pool portion of the plant UHS. This design configuration ensures passive heat transfer from essential systems and components directly to the UHS, with no intermediate heat transfer loop such as that provided by a typical LWR essential service water system.

• • Introduction ****Summary of Application Regulatory Basis

• Technical Evaluation 9.2.2.4.1 System Design Considerations Design Basis System Design and Operation 9.2.2.4.2 GDC 2, "Design Bases for Protection Against Natural Phenomena" 9.2.2.4.3 GDC 4, "Environmental and Dynamic Effects Design Bases" 9.2.2.4.4 GDC 5, "Sharing of Structures, Systems, and Components" 9.2.2.4.5 GDC 60 and 64 9.2.2.4.6 Compliance with 10 CFR 20.1406, "Minimization of Contamination" Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items

Conclusion Introduction Summary of Application Regulatory Basis

• Technical Evaluation 9.2.3.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena" Where SSC (or portions thereof) as determined in the as-built plant which are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II consistent with Section 3.2.1.2 and analyzed as described in Section 3.7.3.8.

9.2.3.4.2 GDC 5, "Sharing of Structures, Systems, and Components" 9.2.3.4.3 Compliance with 10 CFR 20.1406, "Minimization of Contamination" Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis

• • Technical Evaluation 9.2.4.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena"

9.2.4.4.2 GDC 5, "Sharing of Structures, Systems, and Components" 9.2.4.4.3 GDC 60, "Control of Releases of Radioactive Materials to the Environment" 9.2.4.4.4 10 CFR 20.1406, "Minimization of Contamination" Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items

Conclusion Introduction Summary of Application Regulatory Basis ****oo o**

• Technical Evaluation 9.2.5.4.1 Principal Design Criterion 44, "Cooling Water" General Design Criterion 44 versus Principal Design Criterion 44 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 the system safety function can be accomplished, assuming a single failure. Ultimate Heat Sink Cooling Capability During an event where loss of electric power occurs, the volume of water already in the pool provides the inventory for the necessary heat removal. Upon loss of power, the reactor pool cooling and SFP cooling systems shut down. The UHS water expands as it heats and eventually begins to boil. Heat continues to be removed from the pool through boiling and evaporation, removing enough heat to maintain the spent fuel and fuel in the NPMs sufficiently cool to prevent fuel damage. The design is such that UHS water boil-off will continue to remove heat from the power modules and spent fuel for greater than 30 days without the need for operator action, makeup water, or electric power.

9.2.5.4.2 GDC 2, "Design Bases for Protection Against Natural Phenomena" 9.2.5.4.3 GDC 4, "Environmental and Dynamic Effects Design Bases" The UHS is protected from the effects of turbine missiles, as described in Regulatory Guide 1.13, Regulatory Position C.3, without loss of the UHS safety functions specified in Section 9.2.5.1. Section 3.5.1 provides additional detail on protection from turbine missiles. The UHS is designed to withstand environmental and dynamic effects, including the effects of postulated missiles, pipe whip, and discharging fluids that may result from equipment failures and from events and conditions that may occur within the RXB but outside the UHS boundary. Additionally, the physical location of the UHS within the RXB ensures that the effects of equipment failures and events, and conditions that may occur outside the NPM have no reasonable likelihood of adversely impacting UHS safety functions.

The UHS is below grade and contains two heat sources (i.e., stored spent fuel and power modules). The RXB is serviced with a non-safety related heating ventilation and air conditioning system (see Section 9.4.2) that controls the environment. The resident heat sources in the UHS, the fact that it is below grade, and the controlled environment within the RXB prevents the UHS from reaching freezing temperatures. A Safe Shutdown Earthquake (SSE) event can generate waves in the UHS. An analysis of sloshing determined that an SSE generates a maximum wave height of less than three feet. The top of the normal SFP water level range is at 94 building elevation or six feet lower than the operating floor at the 100 building elevation. The UHS pool level provides approximately six feet of freeboard space from the normal pool level operating level for accommodation of sloshing waves or overfill conditions. Normal pool level control is monitored to initiate draining excess volume to the PSCS storage tank.9.2.5.4.4 GDC 5, "Sharing of Structures, Systems, and Components" 9.2.5.4.5 GDC 45, "Inspection of Cooling Water System," and GDC 46, "Testing of Cooling Water System" functions, the UHS design permits the inspection of important components, such as the pool water level instrumentation, the pool liner, and the outside surfaces of the containment vessels. Section 6.6 provides additional information related to the inspection of the containment vessel exterior. Verification of the pool water level ensures adequate water inventory to provide sufficient cooling for the necessary loads. Table 9.2.5-1 lists the minimum water levels. The integrity of the UHS is monitored by the pool leak detection system for evidence of liner leaks. The liner welds are inspected during power operation or shutdown for leak tightness. 9.2.5.4.6 Fuel Storages and Handling and Radioactivity Control *

• • 9.2.5.4.7 Leakage and Makeup The UHS pool liner has the function to prevent potential pool inventory leakage from the SFP, reactor pool, and RFP. The pool liner design prevents inventory leakage and ensures the cooling of the spent fuel assemblies and NPM is maintained during required modes of operation. - Leakage of UHS pool water is collected by the pool leakage detection system. - Any postulated leakage would be nonconsequential due to the size of the water inventory in the UHS pools. The UHS has a makeup line that is designed to meet Regulatory Guide 1.26, Quality Group C; Regulatory Guide 1.29 Seismic Category I; and American Society of Mechanical Engineers BPVC Section III requirements, and is protected from external natural phenomena. 9.2.5.4.8 Instrumentation The reactor pool cooling system and SFP cooling system temperature instrumentation is used to monitor the UHS. Temperature instrumentation located in the pool is seismic Category I. The UHS water level instrumentation is Seismic Category 1. Water level instrumentation in the UHS is powered under normal and off-normal operational scenarios by the plant electrical distribution system and is battery backed. Remote power connections for the electrical distribution system are provided to enable repowering the equipment from outside the plant. UHS primary and backup level instrument channels are qualified for temperature, humidity, and radiation levels consistent with the pool water at saturation conditions for an extended period. The UHS pool level instrumentation mounting protects it from natural phenomena. To ensure redundancy, instruments are physically separated and mounted at opposite ends of the pools. Since the UHS communicates with pool areas while the water is above the weir, this provides multiple areas to monitor pool level. The location for each of the instruments provides assurance that a single event will not cause damage to all of the level instruments.

The UHS level information is displayed in the main control room and remote shutdown station. Alarms alert the operator of these parameters during both normal and post-accident conditions. UHS Level instrumentation provides level information for post-accident monitoring. DCA Part 2, Tier 2, Figure 9.2.5-2 shows the relative location of the level instrumentation. Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis

• Technical Evaluation 9.2.6.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena" 9.2.6.4.2 GDC 5, "Sharing of Structures, Systems, and Components" 9.2.6.4.3 GDC 60, "Control of Releases of Radioactive Materials to the Environment" 9.2.6.4.4 10 CFR 20.1406, "Minimization of Contamination" Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis

• • Technical Evaluation 9.2.7.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena"

[W]here SSC (or portions thereof) as determined in the as-built plant which are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II consistent with Section 3.2.1.2 and analyzed as described in Section 3.7.3.8. 9.2.7.4.2 GDC 4, "Environmental and Dynamic Effects Design Bases" 9.2.7.4.3 GDC 5, "Sharing of Structures, Systems, and Components" 9.2.7.4.4 GDC 60, "Control of Releases of Radioactive Materials to the Environment" 9.2.7.4.5 10 CFR 20.1406, "Minimization of Contamination" Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis ******

• Technical Evaluation 9.2.8.4.1 GDC 1, "Quality Standards and Records" 9.2.8.4.2 GDC 2, "Design Basis for Protection Against Natural Phenomena" -General Design Criteria (GDC) 2, 4, and 5 were considered in the design of the CHWS. The CHWS is not required to function during or after a natural phenomenon event or other events that result in the generation of missiles, pipe whipping, or discharging fluids. Portions of the system that are in proximity to Seismic Category I SSC are designed to Seismic Category II standards. ... See Section 9.2.8.3 for the safety evaluation. The safety, risk significance, seismic, quality, and other design classifications for CHWS structures, systems and components are provided in Table 3.2-1. [T]he CHWS is designated as a Seismic Category III system. However, it is possible that the final layout of the plant will result in some of its components located such that a seismic event could cause them to damage a Seismic I component-.

Where SSC (or portions thereof) as determined in the as-built plant which are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II consistent with Section 3.2.1.2 and analyzed as described in Section 3.7.3.8.

9.2.8.4.3 GDC 4, "Environmental and Dynamic Effects Design Bases" 9.2.8.4.4 GDC 5, "Sharing of Structures, Systems, and Components" Consistent with GDC 5, the components in the CHWS do not have a safety function or a function for shutting a unit down or maintaining a NPM in a safe shutdown condition. Operation of the CHWS does not interfere with the ability to operate or shut down a unit. 9.2.8.4.5 GDC 44, "Cooling Water" 9.2.8.4.6 GDC 45, "Inspection of Cooling Water System," and GDC 46, "Testing of Cooling Water System" Consistent with GDC 45, the CHWS is not safety-related and does not perform any safety-related functions during normal operations, anticipated operational occurrences, and accident conditions; therefore, no specific provisions are included in the design of the CHWS for periodic inspection. Consistent with GDC 46, the CHWS is not safety-related and therefore periodic pressure and functional testing of the system is not required-.

9.2.8.4.7 10 CFR 20.1406, "Minimization of Contamination" The CHWS is designed to be a closed loop, non-radioactive system that is precluded from connection to systems that contain, or could contain, radioactive contamination. The design of the CHWS takes into account the requirements of 10 CFR 20.1406(a) as discussed in Section 12.3. The CHWS is designed to be a closed loop, non-radioactive system. The design of the CHWS provides protection against the spread of contamination in accordance with 10 CFR 20.1406(a) as discussed in Section 12.3. The CHWS provides water to condensers in the LWRS and gas coolers in the GRWS. Design features prevent radioactive contaminants in the LRWS and GRWS from entering the CHWS:

• The shell side of the LRWS condensers is exhausted by a vacuum pump thus maintaining the LRWS pressure at a vacuum and eliminating any driving force into the CHWS.

• In the GRWS gas coolers, CHWS pressure is higher than GRWS pressure eliminating any driving force into the CHWS. Consistent with the Regulatory Guide 4.21, Minimization of Contamination and Radioactive Waste Generation: Life-Cycle Planning, the design of the CHWS and its interfaces with other systems prevents contamination from entering the CHWS.

Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis ****Technical Evaluation Where SSC (or portions thereof) as determined in the as-built plant which are identified as Seismic Category III in this table could, as the result of a seismic event, adversely affect Seismic Category I SSC or result in incapacitating injury to occupants of the control room, they are categorized as Seismic Category II, consistent with Section 3.2.1.2, and analyzed as described in Section 3.7.3.8.

Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Initial Test Program Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis

• • • Technical Evaluation Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis **

Technical Evaluation

• • Post-Accident Sampling

Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis ***

• Technical Evaluation 9.3.3.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena" 9.3.3.4.2 GDC 4, "Environmental and Dynamic Effects Design Bases" .9.3.3.4.3 GDC 5, "Sharing of Structures, Systems, and Components" 9.3.3.4.4 GDC 64, "Monitoring Radioactivity Releases" 9.3.3.4.5 GDC 60, "Control of Releases of Radioactive Materials to the Environment" 9.3.3.4.6 Compliance with 10 CFR 20.1406, "Minimization of Contamination" Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis ******

• • • • Technical Evaluation 9.3.4.4.1 GDC 1, "Quality Standards and Records" 9.3.4.4.2 GDC 2, "Design Bases for Protection Against Natural Phenomena" 9.3.4.4.3 GDC 5, "Sharing of Structures, Systems, and Components" 9.3.4.4.4 GDC 14, "Reactor Coolant Pressure Boundary"

When measured water chemistry parameters are outside the specified range, corrective actions are taken to bring the parameter back within the acceptable range and within the time period specified in the EPRI water chemistry guidelines.

9.3.4.4.5 GDC 29, "Protection Against Anticipated Operational Occurrences" 9.3.4.4.6 GDC 33, "Reactor Coolant Makeup"

• • 9.3.4.4.7 GDC 60, "Control of Releases of Radioactive Materials to the Environment" and Leakage Detection 9.3.4.4.8 GDC 61, "Fuel Storage and Handling and Radioactivity Control" 9.3.4.4.9 Minimization of Contamination Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Combined License Information Items Conclusion Introduction Summary of Application *

Regulatory Basis ****

Technical Evaluation 9.3.6.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena" 9.3.6.4.2 GDC 5, "Sharing of Structures, Systems, and Components" 9.3.6.4.3 GDC 34, "Residual Heat Removal" Suitable redundancy...shall be provided to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available), the system safety function can be accomplished, assuming a single failure. The system(s) that can be used to take the reactor from normal operating conditions to cold shutdown shall satisfy the following functional requirements: A. The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-grade systems.

After the passive RHR system or main steam system effected the initial shutdown, a non-safety-grade reactor shutdown cooling system will be available to bring the plant to cold shutdown conditions for inspection and repair. EPRI stated that "these non-safety systems are required to be highly reliable...and there is no single failure of these systems or their support systems which would result in inability to terminate use of the passive safety grade system and achieve cold shutdown if desired. Water Reactor Utility Requirements Document (Reference 5.4-3) was determined to be acceptable by the Nuclear Regulatory Commission as documented in SECY-94-084. Per SECY-94-084 and NUREG-1242, Volume 3, Part 2, transition of a passive plant from safe shutdown conditions to cold shutdown conditions may be reached using nonsafety-related systems. The nonsafety-related containment flood and drain system is used to flood the containment to allow passive long term decay heat removal via convection and conduction to the reactor pool via the RCS, RPV shell, flooded containment, and CNV shell. The NuScale Power proposed technical specifications describe five operating modes: Operations, Hot Shutdown, Safe Shutdown, Transition and Refueling as shown in Part 4, Volume 1, Table 1.1-1. Of these modes, none are directly analogous to the legacy operating mode of "cold shutdown". If an accident scenario occurs, the reactor module will be brought from Mode 1, "Operations,"

to Mode 3, "Safe Shutdown," utilizing either the decay heat removal system, emergency core cooling system, or normal non-safety means such as the feedwater system and condenser.... If additional coolant is needed to cool the nuclear power module (NPM) from normal operating conditions to conditions equivalent to cold shutdown in a conventional plant (i.e., RCS temperature less than 200 degrees F), inventory can be added to the NPM through either containment flooding and drain system (CFDS) or the chemical volume and control system (CVCS), as described in FSAR Sections 9.3.4 and 9.3.6. The CFDS and CVCS are separate systems and share no components, allowing the function of adding inventory to the NPM while meeting the intent of the single failure proof criteria.

9.3.6.4.4 GDC 60, "Control of Releases of Radioactive Materials to the Environment

• • The NuScale design supports exemption from 10 CFR 50.34(f)(2)(xiv)(E) as applied to the CES. The CES will have radioactivity monitors on the discharge path of the system; however, they will not trigger a safety-related containment isolation signal in the event of a high radioactivity signal. Rather, upon a high radioactive effluent signal on the gaseous discharge path, the system automatically diverts the gaseous effluent from the [reactor building ventilation system (RBVS)] to the [gaseous radioactive waste system (GRWS)] using nonsafety-related signals and equipment. Refer to Part 7, Chapter 13 for further details. Initial Test Program Inspections, Tests, Analyses, and Acceptance Criteria Technical Specifications Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis

• Technical Evaluation 9.4.1.4.1 GDC 2, Design bases for protection against natural phenomena

9.4.1.4.2 GDC 4, Environmental and dynamic effects design bases 9.4.1.4.3 GDC 5, Sharing of structures, systems, and components 9.4.1.4.4 GDC 19, Control room 9.4.1.4.5 10 CFR 20.1406, Minimization of Contamination

9.4.1.4.6 GDC 60, Control of releases of radioactive materials to the environment 9.4.1.4.7 10 CFR 50.63 Loss of all alternating current power The SBO duration for passive plant designs is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, which is consistent with Nuclear Regulatory Commission policy provided by SECY-94-084 and SECY 132 and the associated staff requirements memorandums. Passive plants are required to demonstrate that safety related functions can be performed without reliance on AC power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the initiating event. The relevant guidelines of Regulatory Guide (RG) 1.155 are applied as they pertain to compliance with 10 CFR 50.63 for the passive NuScale design.

The analysis results show that a safe and stable shutdown is achieved, and that the reactor is cooled and containment integrity is maintained for the 72-hour duration with no operator actions. The core remains subcritical for the duration of the event. The reactor coolant inventory ensures that the core remains covered without the need for makeup systems.

The environmental conditions in the main control room during the SBO were evaluated. The control room remains habitable for the duration of the SBO event using the control room habitability system. The control room instrumentation to monitor the event mitigation and confirm the status of reactor cooling, reactor integrity, and containment integrity also remains available. The control room habitability system is described in Section 6.4. In a station blackout event, the CRE isolation dampers close to form part of the CRE. The CRHS then provides bottled air to the CRE. Along with the CRHS, the CRE isolation dampers ensure that a suitable operating environment is maintained to support operators and equipment in the MCR. 9.4.1.4.8 Technical Support Center (TSC) Accident analyses are performed for one emergency mode: that of uninterrupted power supply with continuous filtered airflow to the Technical Support Center (TSC) envelope for the event duration. In the event of immediate loss of power with control room habitability system (CRHS) activation, TSC personnel are evacuated and the TSC functions are transferred to an alternate site-specific location. With loss of power with CRHS activation, the TSC is evacuated since it is not serviced by the CRHS.

9.4.1.4.9 ITAAC *

9.4.1.4.10 Technical Reports Combined License Information Items

Conclusion Introduction Summary of Application The RBVS serves no safety-related function and is not credited for mitigating any design basis events. The system is designed to remove radioactive contaminants from the exhaust streams of RXB general area, the radioactive waste building (RWB) general area, and Annex Building (ANB). The exhaust from the RBVS is monitored for radioactivity contamination. The RBVS includes air filtration and utilizes automatic realignment of the Spent Fuel Pool (SFP) area subsystem to limit release of airborne radioactivity contaminants to the environment. RBVS exhaust paths are monitored for radioactivity releases. Regulatory Basis *

• Technical Evaluation

• • TheFSAR Table 3.2-1 identifies the Annex Building as conceptual design information with a classification of B2 (non-safety related and not risk significant). The supporting systems for the Annex Building, including the HVAC system, do not perform any safety-related or risk significant functions.

The annex building HVAC system exhausts to the RBVS, which, as described in FSAR Section 9.4.2, does not perform any safety-related or risk-significant functions. No sections of NUREG-0800 are applicable to the annex building HVAC system. For these reasons, the annex building HVAC system does not need to be described in detail in the FSAR. FSAR Section 9.4.2 has been modified to provide a brief description of the interaction of the annex building HVAC system with the reactor building HVAC system. 9.4.2.4.1 GDC 2, "Design bases for protection against natural phenomena"

9.4.2.4.2 GDC 5, "Sharing of structures, systems, and components" The Annex Building HVAC System (ABVS) is a shared system, supporting 12 NPMs that provides heating, ventilation, and air conditioning for the annex building. A fault in this system does not contribute to any initiating event frequency. An initiator is not modeled because the equipment in the annex building is not relied on for plant operations nor does it serve a role in accident mitigation.

9.4.2.4.3 GDC 60 Control of releases of radioactive materials to the environment RG 1.140, Regulatory Position C.2 RG 1.140, Regulatory Position C.3

9.4.2.4.4 GDC 61 Fuel storage and handling and radioactivity control 9.4.2.4.5 GDC 64 Monitoring radioactivity releases 9.4.2.4.6 10 CFR 20.1406 Minimization of contamination [t]he RBVS is designed to move air from areas of typically lower radioactive contamination to areas that potentially more contaminated. The RBVS is designed with ducting runs as short as practical and that do not have sudden directional changes. Interior and exterior duct surfaces are relative smooth. The SFP area exhaust subsystem removes air from this potentially contaminated area and filters the exhaust. 9.4.2.4.7 10 CFR 52.47(b)(1) ITAAC

• • • 9.4.2.4.8 High Energy Line Break (HELB) Outside Containment

The exhaust ductwork from the spent fuel pool area of the Reactor Building is credited as a passive relief path for pressure resulting from a high energy line break as a means to protect the Reactor Building and its contents. This information has been added to FSAR Section 9.4.2. This ductwork does not need to be designed to Seismic Category I standards, as the provisions of Branch Technical Position (BTP) 3-3 do not apply. Section A of BTP 3-3 states, in part, that the intent of the BTP is "that postulated piping failures in fluid systems should not cause a loss of function of essential safety-related systems and that nuclear plants should be able to withstand postulated failures of any fluid system piping outside containment, taking into account the direct results of such failure and the further failure of any single active component, with acceptable offsite consequences" (emphasis added). BTP 3-3 defines essential systems and components as those necessary to shut down the reactor and mitigate the consequences of a postulated pipe rupture without offsite power. The Reactor Building does not meet this definition of an essential safety-related system. As stated in NuScale Topical Report TR-0915-17565, Accident Source Term Methodology, no credit is taken for the Reactor Building for meeting offsite dose limits. Therefore, leakage from the Reactor Building due to a HELB will not result in unacceptable offsite consequences. For these reasons, the provisions of BTP 3-3 do not apply to the RBVS, and the ventilation path does not need to be designed to Seismic Category I standards. As a passive ventilation path for a HELB, the spent fuel pool exhaust ductwork of the RBVS provides a protective function for the safety-related, Seismic Category I Reactor Building. In accordance with the intent of RG 1.29, this portion of the RBVS is, therefore, designed to Seismic Category II standards. FSAR Table 3.2-1 and FSAR Section 9.4.2 have been revised to show that the ductwork and associated components of the spent fuel pool exhaust subsystem of the RBVS are designed to Seismic Category II standards. NuScale is currently re-evaluating the plant response to high energy line breaks as a result of RAI 9063. If the results of that analysis change the requirements for the passive ventilation path NuScale will update this response.

Technical Reports **

Combined License Information Items Conclusion Introduction Summary of Application

• Regulatory Basis **** Technical Evaluation 9.4.3.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena" 9.4.3.4.2 GDC 5, "Sharing of Structures, Systems, and Components" 9.4.3.4.3 GDC 60, "Control of Releases of Radioactive Materials to the Environment" 9.4.3.4.4 10 CFR 20.1406, "Minimization of Contamination" Inspections, Tests, Analyses, and Acceptance Criteria Initial Test Program With regard to preoperation inspection and tests for this system, DCA Tier 2, Section 9.4.3.4, "Inspection and Testing," states that component fans, cooling coils, and electrical coils are factory tested and certified. A system air balance test and adjustment to design conditions are conducted in the course of the plant preoperational test program, as indicated in DCA Tier 2, Table 14.2-21, "Radioactive Waste Building HVAC System Test #21," which lists component tests for the functional verification of RWBVS fans, dampers, and instrumentation to ensure compliance with RWBVS design requirements. Combined License Information Items Conclusion Introduction Summary of Application ***

Regulatory Basis ****

• Technical Evaluation 9.4.4.4.1 GDC 2, "Design Bases for Protection Against Natural Phenomena"

9.4.4.4.2 GDC 5, "Sharing of Structures, Systems, and Components"

9.4.4.4.3 GDC 60, "Control of Releases of Radioactive Materials to the Environment"

The only potential source of radioactive material in the TGB is from a postulated steam generator tube failure. However, the main steam system and the Condensate and Feed-water System described in Chapter 10, have process radiation monitors to detect radioactive material introduced into the TGB.

Inspections, Tests, Analyses, and Acceptance Criteria Initial Test Program Components are tested and inspected prior to installation and as an integrated system following installation. System airflows are measured and adjustments are made to ensure compliance with design requirements. Preoperational testing of the TBVS is performed in accordance with the requirements of FSAR Tier 2 Section 14.2. Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis

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• Technical Evaluation 9.5.1.4.1 GDC 3, "Fire Protection"

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• • 9.5.1.4.2 Manual Fire Suppression 9.5.1.4.3 GDC 5, " 9.5.1.4.4 GDC 19, "Control Room" 9.5.1.4.5 GDC 23, "Protection System Failure Modes" 9.5.1.4.6 Enhanced Fire Protection Criteria To minimize fire as a significant contributor to the likelihood of severe accidents for advanced plants, the staff concludes that current NRC guidance must be enhanced. Therefore the evolutionary ALWR designers must ensure that safe shutdown can be achieved, assuming that all equipment in any one fire area will be rendered inoperable by fire and that re-entry into the fire area for repairs and operator actions is not possible. Because of its physical configuration, the control room is excluded from this approach, provided an independent alternative shutdown capability that is physically and electrically independent of the control room is included in the design. Evolutionary ALWRs must provide fire protection for redundant shutdown systems in the reactor containment building that will ensure, to the extent practicable, that one shutdown division will be free of fire damage. Additionally, the evolutionary ALWR designers must ensure that smoke, hot gases, or the fire suppressant will not migrate into other fire areas to the extent that they could adversely affect safe-shutdown capabilities, including

operator actions.

Inspections, Tests, Analyses, and Acceptance Criteria Initial Test Program Technical Specifications Combined License Information Items Conclusion Introduction ****

• • Summary of Application Regulatory Basis ****

• • • • Technical Evaluation 9.5.2.4.1 Compliance with 10 CFR Part 50, Appendix E, IV.E(9) A COL applicant that references the NuScale Power Plant design certification will provide a description of the offsite communication system, how that system interfaces with the onsite communications system, as well as how continuous communications capability is maintained to ensure effective command and control with onsite and offsite resources during both normal and emergency situations.

9.5.2.4.2 Compliance with 10 CFR 50.34(f)(2)(xxv), 10 CFR 50.47(b)(6), and 10 CFR 50.47(b)(8) 9.5.2.4.3 Compliance with 10 CFR 50.47(b)(6) and 10 CFR 50.47(b)(8) A]dequate provisions for communications are provided and maintained in the emergency facilities and control room to support the emergency response, including prompt communication among principal response organizations to emergency personnel and to the public.

TSC and OSC are equipped with voice communications such as private branch exchange, public address and general alarm system, plant radio, and sound powered telephone systems, which provide communications between the TSC and OSC and plant, local, and offsite emergency response facilities, the Nuclear Regulatory Commission, and local and state operations centers. 9.5.2.4.4 Compliance with General Design Criteria Consistent with GDCs 1 and 10 CFR 50.55a, COMS structures, systems, and components are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.

Recognized codes and standards are identified and evaluated to determine their applicability, adequacy, and sufficiency and supplemented or modified as necessary to assure a quality product in keeping with the required safety function. Consistent with GDC 2, portions of the COMS whose structural failure could adversely affect the function of Seismic Category I SSC are designed to Seismic Category II requirements in accordance with Section 3.2.1.2.

Consistent with GDC 3, the COMS systems are designed and located to minimize consistent with other safety requirements, the probability and effect of fires and explosions. Consistent with the requirements of Regulatory Guide 1.189 Position 4.1.7, the COMS is designed to provide effective communication between plant personnel in vital areas during fire conditions under maximum potential noise levels. COMS equipment is designed to operate reliably within the environment in which it is installed including environmental conditions such as temperature, humidity, radiation, and noise. Furthermore the COMS is designed to operate taking into account placement of barriers such as shield walls. COMS equipment is accessible to personnel for operation, inspection, maintenance, and testing. A failure in a COMS subsystem will not significantly impair the ability of the other COMS subsystems to perform, including in the event of an accident in one NuScale Power Module and an orderly shutdown and cooldown of the remaining NuScale Power Modules. the public address and general alarm system, private branch exchange, and plant radio systems-serve as a backup to one another in the event of system failure as a result of natural phenomena, environmental or dynamic effects, and fires. The three independent voice communications systems are designed and installed to provide assurance that any single event does not cause a complete loss of intra-plant communication. However the various independent and diverse communications systems located in the MCR significantly increase the overall command and control the reactor operators have over the plant by providing the ability to communicate and direct activities with operations, maintenance, health physics, firefighters, security, and rescue teams. The NuScale Power Plant has an independent plant radio system for security purposes. Other communications systems such as the public address and general alarm system and private branch exchange are available as alternate means, if necessary. Initial Test Program Compliance with 10 CFR 52.47(b)(1)

The Communications System (COMS) is not safety-related, does not protect safety-related components, and is classified as a B2 (nonrisk-significant, nonsafety-related) system. The COMS system does not have a Tier 1 design description because there are no design commitments for the COMS system.

However, Physical Security System ITAAC Nos. 03.16.11, 03.16.12, and 03.16.13 in FSAR Tier 1 Table 3.16-1 use a portion of the COMS to verify these ITAAC. Further, these ITAAC are verified, in part, by the Communications System Test #68 described in FSAR Tier 2 Table 14.2-68. Compliance with 10 CFR 73.45(e)(2)(iii), 10 CFR 73.45(g)(4)(i), 10 CFR 73.45(g)(4)(ii), 10 CFR 73.46(f), 10 CFR 73.55(e)(9)(vi)(B), and 10 CFR 73.55(j) Electromagnetic Interference and Radiofrequency Interference Compatibility The COMS does not adversely impact other plant systems with EMI and RFI and that the system is designed in consideration of the guidelines of Regulatory Guide 1.180, which identifies electromagnetic environment operating envelopes, design, installation, and test practices for addressing the effects of EMI, RFI, and power surges on instrumentation and controls systems and components.

Combined License Information Items Conclusion Introduction Summary of Application Lighting fixtures in the MCR, remote shutdown station, and in areas containing safety-related structures, systems, and components are mounted to meet Seismic Category II requirements to ensure that their failure does not reduce the functional reliability of safety-related equipment, result in incapacitating injury to control room occupants, or render the control room uninhabitable. 9.5.3.2.1 Normal Lighting System 9.5.3.2.2 Emergency Lighting System 9.5.3.2.3 Normal and Emergency Main Control Room Lighting Regulatory Basis

• Technical Evaluation

Initial Test Program Inspections, Tests, Analyses and Acceptance Criteria Combined License Information Items Conclusion Introduction Summary of Application Regulatory Basis *

• • • • Technical Evaluation

• • • • • • Inspections, Tests, Analyses, and Acceptance Criteria Initial Test Program Technical Specifications Combined License Information Items Conclusion