ML16054A446
| ML16054A446 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 01/26/2016 |
| From: | Northern States Power Co, Xcel Energy |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16054A376 | List:
|
| References | |
| L-MT-16-004 | |
| Download: ML16054A446 (63) | |
Text
Revision 22USAR APPENDIX EMONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 1APPENDIX EPLANT COMPARATIVE EVALUATION WITHTHE PROPOSED AEC 70 DESIGNCRITERIAI/mabTABLE OF CONTENTSSectionPageE.1Summary Description1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2Criterion - Conformance1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2.1Group I - Overall Plant Requirements1. . . . . . . . . . . . . . . . . . . . . . . E.2.2Group II - Protection by Multiple Fission Products Barriers11. . . . . E.2.3Group III - Nuclear and Radiation Controls15. . . . . . . . . . . . . . . . . . E.2.4Group IV - Reliability and Testability of Protection Systems21. . . . E.2.5Group V - Reactivity Control27. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2.6Group VI - Reactor Coolant Pressure Boundary31. . . . . . . . . . . . . . E.2.7Group VII - Engineered Safety Features34. . . . . . . . . . . . . . . . . . . . . E.2.8Group VIII - Fuel and Waste Storage Systems55. . . . . . . . . . . . . . . E.2.9Group IX - Plant Effluents60. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-E.TOCDoc Type:Admin Initials:Date:9703 Revision 22USAR E.1MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 1APPENDIX EPLANT COMPARATIVE EVALUATION WITHTHE PROPOSED AEC 70 DESIGNCRITERIAI/mabE.1Summary DescriptionThis appendix contains a comparative evaluation of the design basis of the MonticelloNuclear Generating Plant, Unit 1, with the 70 General Design Criteria for Nuclear Power Plant Construction Permits proposed by the Atomic Energy Commission for public comment in July, 1967.The comparative evaluation is made with each of the nine groups of criteria sent out inthe July 1967 AEC release. As to each group, there is a statement of Northern States Power Companys current understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.Following a restatement of the 70 proposed criteria is complete list of references tolocations in this USAR where there is discussed subject matter relating to the intent of the particular criteria.Based on its current understanding of the intent of the 70 proposed-criteria, theapplicant believes that the Monticello Nuclear Generating Plant, Unit 1, is in conformance with the intent of such proposed criteria.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-E.1Doc Type:Admin Initials:Date:9703 Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 61APPENDIX EPLANT COMPARATIVE EVALUATION WITHTHE PROPOSED AEC 70 DESIGNCRITERIAI/jlkE.2Criterion - ConformanceE.2.1Group I - Overall Plant RequirementsThe intent of the current draft of the proposed criteria for this group is to identifyand record the adequacy of the quality control and assurance programs, the applicable codes or standards, the standards of design, fabrication and erection, and to assure protection against appropriate environmental phenomena. Test Procedures, and inspection acceptance levels of the reactor facility's essentialcomponents and systems are also identified. The influence of this sharing ofcommon reactor facility components and systems along with the fire and explosion protection for all equipment is also to establish and described.It is concluded that the design of this plant is in conformance with the criteria ofGroup I based on NSP's current understanding of the intent of these criteria.The reactor facility's essential components and systems are designed,fabricated, erected, and perform in accordance with the specified quality standards which are, as a minimum, in accordance with applicable codes and regulations. These components and systems as well as applicable codes and standards have been identified in the report. Specific sections are included inthe reference letter list following this group's discussion. Where components orsystem design exceeds code requirements it has been noted. A quality control and assurance program has been established to assure compliance with acceptable quality control specifications and procedures. These programs as well as applicable tests and inspections have been identified. Specific sectionsare included in the reference list. In planning and executing these control andassurance programs, particular attention was given to the quality control specifications and to their compliance by those systems, components, and structures which are important to the plant safety. (Criterion 1) The plant equipment which is important to safety is designed to permit safe plant operationand to accommodate all design basis accidents for all appropriate environmentalphenomena at the site without loss of their capability, taking into consideration historical data and suitable margins for uncertainties. (Criterion 2) Further design allowances are provided to minimize the occurrence of fire and explosions and their effects by the use of noncombustible and fire resistantmaterials through the plant. (Criterion 3) Records of design, fabrication, andconstruction for this facility are to be stored or maintained either under the applicant's control or available to the applicant for inspection. (Criterion 5) This reactor facility consists of a single BWR generating unit. (Criterion 4)
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 61I/jlkReferences to applicable sections of the USAR are given below for the individualcriteria of this group.Criterion 1 - Quality Standards (Category A) Those systems and components ofreactor facilities which are essential to prevention of accidents which could affectthe public health and safety or to mitigation to their consequences shall be identified and then designed, fabricated, and erected to quality standards thatreflect the importance of the safety function to be performed. Where generallyrecognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assuranceprograms, test procedures, and inspection acceptance levels to be used shall beidentified. A showing of sufficiency and applicability of codes, standard, quality assurance programs, test procedures, and acceptance levels used is required.Conformance 1 - Quality Standards (Category A)a.GeneralSection 1.2.1Principal Design Criteria - General CriteriaSection 1.3.1.3Summary Design Description and SafetyAnalysis - GeologySection 1.3.1.4Summary Design Description and Safety Analysis - HydrologySection 1.3.1.5Summary Design Description and SafetyAnalysis - Regional and Site MeteorologySection 1.3.1.6Summary Design Description and SafetyAnalysis - Seismology and Design ResponseSpectrumSection 1.3.1.7Summary Design Description and SafetyAnalysis - Site Environmental Monitoring ProgramSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemSection 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and Handling Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 61I/jlkSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power SystemsSection 1.3.9Summary Design Description and Safety Analysis - Plant Shielding, Access Control, and Radiation Protection ProceduresSection 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste ControlSystemsSection Appendix CQuality Assurance Programb.Containment BarriersSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetyFuelSection 1.3.2Summary Design Description and Safety Analysis - Reactor SystemSection 3.4.4Fuel Mechanical Characteristics - Surveillance and TestingFuel CladdingSection 3.2.3Thermal and Hydraulic Characteristics - DesignCriteria and Safety LimitsSection 3.4.1Fuel Mechanical Characteristics - Design BasisSection 3.4.2Fuel Mechanical Characteristics - Description ofFuel AssembliesSection 3.4.3Fuel Mechanical Characteristics - Design EvaluationSection 3.4.4Fuel Mechanical Characteristics - Surveillance and Testing Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 4 of 61I/jlkReactor Coolant SystemSection 4Reactor Coolant SystemPrimary Containment SystemSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.2Primary Containment System - DescriptionSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.2.4Primary Containment System - Inspection and TestingSecondary Containment SystemSection 5.3.2Secondary Containment System - Design BasisSection 5.3.5Secondary Containment System - Performance AnalysisStandby Gas Treatment SystemSection 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsPlant Elevated Release PointSection 9.3Gaseous Radwaste Systemc.Plant Engineered SafeguardsSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 6.1Plant Engineered Safeguards - SummaryDescriptionControl Rod Velocity LimitersSection 6.4.3Control Rod Velocity Limiters - PerformanceAnalysisSection 6.4.4Control Rod Velocity Limiters - Inspection and Testing Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 5 of 61I/jlkControl Rod Drive Housing SupportsSection 6.5.3Control Rod Drive Housing Supports -Performance AnalysisSection 6.5.4Control Rod Drive Housing Supports - Inspectionand TestingReactor Standby Liquid Flow Control SystemSection 6.6.3Standby Liquid Control System - PerformanceAnalysisSection 6.6.4Standby Liquid Control System - Inspection andTrainingMain Steam Line Flow RestrictorsSection 6.3.3Main Steam Line Flow Restrictions -Performance AnalysisSection 6.3.4Main Steam Line Flow Restrictions - Inspectionand TestingEmergency Core Cooling Systems (ECCS)Section 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.5.3Automatic Depressurization System (ADS) -Performance AnalysisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -
Performance AnalysisSection 6.2.3.3Residual Heat Removal System (RHR) -
Performance AnalysisSection 6.2.6Emergency Core Cooling System (ECCS) -
ECCS Performance EvaluationPlant Structures and ShieldingSection 12.2Plant Principal Structures and FoundationsSection 12.3Shielding and Radiation Protection Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 6 of 61I/jlkCriterion 2 - Performance Standards (Category A) Those systems andcomponents of reactor facilities which are essential to prevention of accidentswhich could affect the public health and safety or to mitigation to theirconsequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, andother local site effects. The design bases so established shall reflect: (a)appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and surrounding area and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.Conformance 2 - Performance Standards (Category A)a.GeneralSection 1.2.1Principal Design Criteria - General CriteriaSection 1.3.1.3Summary Design Description and SafetyAnalysis - GeologySection 1.3.1.4Summary Design Description and SafetyAnalysis - HydrologySection 1.3.1.5Summary Design Description and Safety Analysis - Site and Regional MeteorologySection 1.3.1.6Summary Design Description and SafetyAnalysis - Seismology and Design ResponseSpectraSection 1.3.1.7Summary Design Description and SafetyAnalysis - Site Environmental MonitoringProgramSection 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power SystemsSection 1.3.9Summary Design Description and Safety Analysis - Plant Shielding, Access Control, and Radiation Protection ProceduresSection 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste Control SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant Safety Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 7 of 61I/jlkSection 2.3MeteorologySection 2.4HydrologySection 2.5Geology and Soil InvestigationSection 2.6SeismologySection 2.7Radiation Environmental Monitoring Program(REMP)Section 2.8Ecological and Biological Studiesb.Containment BarriersSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemFuel CladdingSection 1.3.6Summary Design Description and SafetyAnalysis - Plant Fuel Storage and HandlingSystemsSection 3.2.1Thermal and Hydraulic Characteristics - Design BasisSection 3.2.3Thermal and Hydraulic Characteristics -Design Criteria and Safety LimitsSection 3.3.1Nuclear Characteristics - Design BasisSection 3.3.3Nuclear Characteristics - Nuclear Design CharacteristicsSection 3.4.1Fuel Mechanical Characteristics - Design BasisSection 3.4.3Fuel Mechanical Characteristics - DesignEvaluationSection 3.5.1Reactivity Control Mechanical Characteristics -Design BasisSection 3.5.5Reactivity Control Mechanical Characteristics -
Operation and Performance Analysis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 8 of 61I/jlkReactor Coolant SystemSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 4 - CompleteReactor Coolant SystemPrimary Containment SystemSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.4Primary Containment System - Inspection and TestingSection 12.2.1.1Plant Principal Structures and Foundations -Safety CategoriesSection Appendix ADesign Bases - Seismic Design and AnalysisSection 12.2.1.6Plant Principal Structures and Foundations -Wind LoadsSecondary Containment SystemSection 5.3.2Secondary Containment System - Design BasisSection 5.3.5Secondary Containment System - Performance AnalysisSection 12.2.1.1Plant Principal Structures and Foundations -
Safety CategoriesSection 12.2.1.6Plant Principal Structures and Foundations -
Wind LoadsSection 12.2.1.7Plant Principal Structures and Foundations -FloodingStandby Gas Treatment SystemSection 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 12.2.1.2Plant Principal Structures and Foundations -
Class I Structures and Equipment Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 9 of 61I/jlkPlant Elevated Release PointSection 9.3Gaseous Radwaste Systemc.Plant Engineered SafeguardsSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemControl Rod Velocity LimitersSection 6.4.1Control Rod Velocity Limiters - Design BasisSection 6.4.3Control Rod Velocity Limiters - Performance AnalysisControl Rod Drive Housing SupportsSection 6.5.1Control Rod Drive Housing Supports - Design BasisSection 6.5.3Control Rod Drive Housing Supports -
Performance AnalysisReactor Standby Liquid Flow Control SystemSection 6.6.1Standby Liquid Control System - Design BasisSection 6.6.3Standby Liquid Control System - Performance AnalysisMain Steam Line Flow RestrictorsSection 6.3.1Main Steam Line Flow Restrictions - Design BasisSection 6.3.3Main Steam Line Flow Restrictions -
Performance AnalysisEmergency Core Cooling Systems (ECCS)Section 6.2.1.1Emergency Core Cooling Systems (ECCS) -
ECCS Design Basis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 10 of 61I/jlkSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.5.3Automatic Depressurization System (ADS) -
Performance AnalysisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -
Performance AnalysisSection 6.2.3.3Residual Heat Removal System (RHR) -Performance AnalysisSection 6.2.6Emergency Core Cooling Systems (ECCS) -ECCS Performance EvaluationPlant Structures and ShieldingSection 12.2Plant Principal Structures and FoundationsSection 12.3Shielding and Radiation ProtectionCriterion 3 - Fire Protection (Category A) The reactor facility shall be designed(a) to minimize the probability of events such as fires and explosions and (b) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical through the facility,particularly in areas containing critical portions of the facility such ascontainment, control room, and components of engineered safety features.Conformance 3 - Fire Protection (Category A)Section 1.2.1Principal Design Criteria - General CriteriaSection 10.3.1Plant Service Systems - Fire Protection SystemsCriterion 4 - Sharing of Systems (Category A) Reactor facilities shallnot sharesystems or components unless it is shown safety is not impaired by the sharing.
Conformance 4 - Sharing of Systems (Category A) This Plant is a single unitand does not share any system, component, or equipment with any other facility.
Criterion 5 - Records Requirements (Category A) Records of design, fabrication,and construction of essential components of the plant shall be maintained by thereactor operator (NSP) or under its control throughout the life of the reactor.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 11 of 61I/jlkConformance 5 - Records Requirements (Category A)Section Appendix CQuality Assurance ProgramSection 13.4Operational ProceduresSection 13.5Operational Records and ReportingRequirementsE.2.2Group II - Protection by Multiple Fission Products BarriersThe intent of the current draft of the proposed criteria for this group is to assurethat the plant has been provided with multiple barriers to protect against or tomitigate the effects of fission products prior to being released to the site environs and to establish that these barriers will remain intact under all operational transients caused by a single reactor operator error or equipment malfunction. It is the further intent of this group that proper barriers are made available for thedesign basis accidents.It is concluded that design of this plant is in conformance with the Criteria ofGroup II Based on NSP's understanding of the intent of these criteria.The plant containment barriers are the basic features which minimize release ofradioactive materials and associated doses. A boiling water reactor provides seven means of containing and/or mitigating the release of fission products; (a) the high density ceramic UO2 fuel, (b) the high integrity Zircaloy cladding, (c) thereactor vessel and its connected piping and isolation valves, (d) the drywell-suppression chamber primary containment, (e) the reactor building (secondary containment), (f) the reactor building standby gas treatment system utilizing high efficiency absolute and charcoal filters, and (g) the plant main stack. The primary containment system is designed, fabricated, and erected toaccommodate without failure, the pressures and temperatures resulting from orsubsequent to double-ended rupture or equivalent failure of any coolant pipe within the primary containment. The reactor building, encompassing the primary containment system, provides secondary containment when the primary containment is closed and in service, and provides primary containment whenthe primary containment is open for refueling operations. The two containmentsystems and such other associated engineered safety systems as may be necessary are designed and maintained so that off-site doses resulting from postulated design basis accidents are below the values stated in 10CFR100.(Criterion 10) The reactor core is designed so there is no inherent tendency forsudden divergent oscillation of operating characteristics of divergent power transient in any mode of plant operation. (Criterion 6, 7) The basis of the reactor core design, in combination with the plant equipment characteristics, nuclear instrumentation system, and the reactor protection system is, to providemargins to ensure that fuel damage will not occur in normal operation oroperational transient caused by single reactor operator error or equipment malfunction. (Criterion 6, 7) The reactor core is designed so that the overall01081199 Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 12 of 61I/jlkpower coefficient in the power operating range is not positive. (Criterion 8) Thereactor coolant system is designed to carry its dead weight and specified liveloads, separately or concurrently, such as pressure and temperature stress,vibrations, seismic loads as appropriately prescribed for the plant. Provisions are made to control or shutdown the reactor coolant system in the event of a malfunction of the operating equipment or excessive leakage of the coolant from the system. The reactor vessel and support structure are designed, within thelimits of applicable criteria for low probability accident conditions, to withstandthe forces that would be created by a full area flow from any vessel nozzle to the containment atmosphere with the reactor vessel at design pressure concurrent with the plant design earthquake loads. (Criterion 9)References to applicable sections of the USAR are given below for the individualcriteria of this group.Criterion 6 - Reactor Core Design (Category A) The reactor core shall bedesigned to function throughout its design lifetime, without exceeding acceptablefuel damage limits which have been stipulated and justified. The core design,together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of aturbine generator set, isolation of the reactor from its primary heat sink, and lossof off-site power.Conformance 6 - Reactor Core Design (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling SystemsSection 3.2Thermal and Hydraulic CharacteristicsSection 3.3Nuclear CharacteristicsSection 3.4Fuel Mechanical CharacteristicsSection 3.5Reactivity Control Mechanical CharacteristicsSection 4Reactor Coolant System Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 13 of 61I/jlkSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.2.5Reactor Auxiliary Systems - Reactor CoreIsolation Cooling System (RCIC)Section 14.4.3Transient Events Analyzed for Core Reload -
Rod Withdrawal ErrorCriterion 7 - Suppression of Power Oscillations (Category B) The core design,together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.Conformance 7 - Suppression of Power Oscillations (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreCriterion 8 - Overall Power Coefficient (Category B) The reactor shall bedesigned so that the overall power coefficient in the power operating range shallnot be positive.Conformance 8 - Overall Power Coefficient (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreSection 3.2Thermal and Hydraulic CharacteristicsSection 3.5Reactivity Control Mechanical CharacteristicsCriterion 9 - Reactor Coolant Pressure Boundary (Category A) The reactorcoolant pressure boundary shall be designed and constructed so as to have anexceedingly low probability of gross rupture or significant leakage throughout its design lifetime.Conformance 9 - Reactor Coolant Pressure Boundary (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 4 CompleteReactor Coolant SystemSection 7.4Reactor Vessel Instrumentation Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 14 of 61I/jlkCriterion 10 - Containment (Category A) Containment shall be provided. Thecontainment structure shall be designed to sustain the initial effects of grossequipment failures, such as a large coolant boundary area, without loss ofrequired integrity and, together with other engineered safety features as may be necessary to retain for as long as the situation requires the functional capability to protect the public.Conformance 10 - Containment (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.3.3Summary Design Description and SafetyAnalysis - Plant Containment SystemSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby CoolingSystemsSection 4 CompleteReactor Coolant SystemSection 5.1Containment System - Summary DescriptionSection 6.2Emergency Core Cooling Systems (ECCS)Section 6.4Control Rod Velocity LimitersSection 6.5Control Rod Drive Housing SupportsSection 6.6Standby Liquid Control SystemSection 5.2.1Primary Containment System - Design CriteriaSection 5.3.2Secondary Containment System - Design BasisSection 12 CompletePlant Structures and ShieldingSection 14.1.1Summary Description - General Safety DesignBasisSection 14.1.5Summary Description - Design Basis for Accidents Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 15 of 61I/jlkE.2.3Group III - Nuclear and Radiation ControlsThe intent of the current draft of the proposed criteria for this group is to identifyand define the instrumentation and control systems, necessary for maintainingthe plant in a safe operational status. This, also includes determining the adequacy of radiation shielding, effluent monitoring, and fission process controls, and providing for the effective sensing of abnormal conditions and initiation of engineered safety features.It is concluded that the design of this plant is in conformance with the criteria ofGroup III based on NSP's current understanding of the intent of these criteria.The plant is provided with a centralized main control room having adequateshielding, fire protection, air conditioning and facilities to permit access and continuous occupancy under 10CFR20 dose limits during all design basisaccident situations. However, if it is necessary to evacuate the main controlroom the design does not preclude the capability to bring the plant to a safe-cold shutdown from outside the main control room. (Criterion 11) The necessary plant controls, instrumentation, and alarms for safe and orderly operation are located in the main control room. These include such controls andinstrumentation as the reactor coolant system leakage detection system.(Criterion 11, 13, 16) The performance of the reactor core and the indication of power level are continuously monitored by the in-core nuclear instrumentation system. (Criterion 13) The reactor protection system, independent from the plant process control systems, overrides all other controls to initiate any requiredsafety action. The reactor protection system automatically initiates appropriateaction whenever the plant conditions approach pre-established operational limits.
The system acts specifically to initiate the emergency core and containment cooling systems as required. (Criterion 12, 13, 14, 15) The plant radiation and process monitoring systems are provided for monitoring significant parametersfrom specific plant process systems and specific areas including the planteffluents to the site environs and to provide alarms and signals for appropriate corrective actions. (Criterion 17, 18)Reference to applicable sections of the USAR are given below for the individualcriteria of this group.Criterion 11 - Control Room (Category B) The facility shall be provided with acontrol room from which action to maintain safe operational status of the plantcan be controlled. Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control room orother areas as necessary to shut down and maintain safe control to the facilitywithout radiation exposures of personnel in excess of 10CFR20 limits. It shall be possible to shut the reactor down and maintain it in a safe condition if access the control room is lost due to fire or other causes.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 16 of 61I/jlkConformance 11 - Control Room (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.8Principal Design Criteria - Plant Shielding and Access ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation and Control SystemsSection 1.3.9Summary Design Description and SafetyAnalysis - Plant Shielding, Access Control, andRadiation Protection ProceduresSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant SafetySection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.6Plant Protection SystemSection 7.7Turbine-Generator System Instrumentation and ControlSection 12.3.3Shielding and Radiation Protection -
Performance AnalysisCriterion 12 - Instrumentation and Control Systems (Category B)Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges.Conformance 12 - Instrumentation and Control Systems (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and SafetyAnalysis - Plant Instrumentation Control SystemsSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 7Plant Instrumentation and Control Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 17 of 61I/jlkSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5Plant Radiation Monitoring SystemsSection 7.6Plant Protection SystemSection 7.7Turbine-Generator System Instrumentation andControlSection 7.8NUMAC Rod Worth Minimizer and Plant Process ComputerCriterion 13 - Fission Process Monitors and Controls (Category B) Means shallbe provided for monitoring and maintaining control over the fission process throughout core life and for all conditions that can reasonably be anticipated tocause variation in reactivity of the core, such as indication of position of controlrods and concentration of soluble reactivity control poisons.Conformance 13 - Fission Process Monitors and Controls (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemsSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.6Plant Protection SystemSection 7.8NUMAC Rod Worth Minimizer and Plant Process Computer Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 18 of 61I/jlkCriterion 14 - Core Protection Systems (Category B) Core protection systemstogether with associated equipment, shall be designed to act automatically toprevent or to suppress conditions that could result in exceeding acceptable fueldamage limits.Conformance 14 -Core Protection Systems (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation and Control SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant SafetySection 3.3Nuclear CharacteristicsSection 3.4Fuel Mechanical CharacteristicsSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.2Emergency Core Cooling System (ECCS)Section 6.3Main Steam Line Flow RestrictionsSection 6.4Control Rod Velocity LimitersSection 6.5Control Rod Drive Housing SupportsSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.6Plant Protection SystemSection 7.8NUMAC Rod Worth Minimizer and Plant Process Computer Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 19 of 61I/jlkSection 8 CompletePlant Electrical SystemsSection 14 CompletePlant Safety AnalysisCriterion 15 - Engineered Safety Features Protection Systems (Category B)Protection systems shall be provided for sensing accident situations andinitiating the operation of necessary engineered safety features.Conformance 15 - Engineered Safety Features Protection Systems (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control SystemsSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 6 CompletePlant Engineered SafeguardsSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5Plant Radiation Monitoring SystemsSection 7.6Plant Protection SystemSection 7.7Turbine-Generator Systems Instrumentation andControlSection 7.8NUMAC Rod Worth Minimizer and Plant Process ComputerCriterion 16 - Monitoring Reactor Coolant Pressure Boundary (Category B)Means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage.Conformance 16 - Monitoring Reactor Coolant Pressure Boundary (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 20 of 61I/jlkSection 5.2Primary Containment SystemSection 7.1Plant Instrumentation and Control Systems -Summary DescriptionSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.6Plant Protection SystemCriterion 17 - Monitoring Radioactivity Releases (Category B) Means shall beprovided for monitoring the containment atmosphere, the facility effluentdischarge paths, and the facility environs, for radioactivity that could be released from normal operations, from anticipated transients, and from accidentconditions.Conformance 17 - Monitoring Radioactivity Releases (Category B)Section 1.2.7Principal Design Criteria - Plant RadioactiveWaste DisposalSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control SystemsSection 5.3.4.1Secondary Containment System - Standby Gas Treatment System (SGTS)Section 7.5Plant Radiation Monitoring SystemsSection 7.6.1Plant Protection System - Reactor Protection SystemSection 9.2Liquid Radwaste SystemSection 9.3Gaseous Radwaste SystemSection 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsSection 10.3.7Plant Service Systems - Plant Process Sampling SystemSection 14.1.5Summary Description - Design Basis for Accidents Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 21 of 61I/jlkCriterion 18 - Monitoring Fuel and Waste Storage (Category B) Monitoring andalarm instrumentation shall be provided for fuel and waste storage and handlingareas for conditions that might contribute to loss of continuity in decay heatremoval and to radiation exposures.Conformance 18 - Monitoring Fuel and Waste Storage (Category B)Section 7.5Plant Radiation Monitoring SystemsSection 7.6.1Plant Protection System - Reactor ProtectionSystemSection 9.2.1Liquid Radwaste System - Design BasisSection 9.2.2.1Liquid Radwaste System - GeneralSection 9.2.2.3Liquid Radwaste System - Instrumentation andControl of the Liquid RadwasteSection 9.3.1Gaseous Radwaste System - Design BasisSection 9.3.3Gaseous Radwaste System - PerformanceAnalysisSection 9.4.1Solid Radwaste System - Design BasisSection 9.4.3Solid Radwaste System - Performance AnalysisSection 10.2.1.1Reactor Auxiliary Systems - Design BasisSection 10.2.1.2Reactor Auxiliary Systems - DescriptionSection 10.2.2.1Reactor Auxiliary Systems - Design BasisSection 10.2.2.3Reactor Auxiliary Systems - Performance AnalysisE.2.4Group IV - Reliability and Testability of Protection SystemsThe intent of the current draft of the proposed criteria for this group is to identifyand establish the functional reliability, in-service testability, redundancy, physicaland electrical independence and separation, and fail-safe design of the reactor protection instrumentation and control systems.It is concluded that the design of this plant is in conformance with the criteria ofGroup IV based on NSP's current understanding of the intent of these criteria.The reactor protection system automatically overrides the plant normaloperational control system (that is, functions independently) to initiate Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 22 of 61I/jlkappropriate action whenever the plant conditions monitored (neutron flux,containment, and vessel pressure, etc.) by the system approach pre-establishedlimits. (Criterion 22) By means of a dual channel protection system withcomplete redundancy in each channel, no loss of the protection systems can occur by either component failure or removal from service. The reactor protection system acts to shutdown the reactor, close primary containment isolation valves and initiates the operation of the emergency core andcontainment cooling systems. The reactor protection system is designed so thata credible plant transient or accident is sensed by different parametric measurements (e.g., loss of coolant accident is detected by high drywell pressure and low-low reactor level monitors). (Criterion 20) Components of the redundant subsystems can be removed from service for testing andmaintenance without negating the ability of the protection system to perform itsprotection functions (even when subjected to a single event, multiple failure incident) upon receipt of the appropriate signals. (Criterion 19, 20, 21) The design of the reactor protection system is such as to facilitate maintenance and trouble shooting while the reactor is at power operation without impeding theplant's operation or impairing its safety function. System faults are annunciatedin the main control room. (Criterion 25) The system electrical power requirements are supplied from independent, redundant sources. (Criterion 24)
The system circuits are isolated to preclude a circuit fault from inducing a fault in another circuit and to reduce the likelihood that adverse conditions, which mightaffect system reliability (1 of 2 x 2), will encompass more than one circuit. Thesystem sensors are electrically and physically separated with both sensors in any one trip channel not allowed to occupy the same local area or to be connected to the same power source or process measurement line. The system internal wiring or external cable routing arrangement are such as to negate anyexternal influence (a fire or accident) on the systems performance. (Criterion 23,24) A failure of any one reactor protection system input or subsystem component will produce a trip in one of two channels, a situation insufficient to produce a reactor scram but readily available to perform its protective functionupon another trip (either by failure or by exceeding the preset trip). (Criterion 26)This reactor protection system design includes allowance for single reactor operator error and equipment malfunction and still performs its intended function.
(Criterion 21) References to applicable sections of the USAR are given below for the individual criteria of this group.Criterion 19 - Protection Systems Reliability (Category B) Protection systemsshall be designed for high functional reliability and in-service testabilitycommensurate, with the safety functions to be performed.Conformance 19 - Protection Systems Reliability (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.1Summary Design Description and Safety Analysis - Plant Site and Environs Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 23 of 61I/jlkSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5.2Plant Radiation Monitoring systems - ProcessRadiation Monitoring SystemsSection 7.6Plant Protection SystemSection 11.2Turbine-Generator SystemSection 14.1.5Summary Description - Design Basis for AccidentsCriterion 20 - Protection Systems Redundancy and Independence (Category B)Redundancy and independence designed into protection systems shall be sufficient to assure that no single failure or removal from service of any component or channel of a system will result in loss of the protection function.The redundancy provided shall include, as a minimum, two channels ofprotection for each protection function to be served. Different principles shall be used where necessary to achieve true redundant instrumentation components.Conformance 20 - Protection Systems Redundancy and Independence(Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control SystemsSection 7.1Plant Instrumentation and Control Systems -
Summary DescriptionSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5.2Plant Radiation Monitoring Systems - Process Radiation Monitoring SystemSection 7.6Plant Protection SystemSection 11.2Turbine-Generator SystemSection 14.1.5Summary Description - Design Basis for Accidents Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 24 of 61I/jlkCriterion 21 - Single Failure Definition (Category B) Multiple failures from asingle event shall be treated as a single failure.Conformance 21 - Single Failure Definition (Category B)Section 7.2Reactor Control SystemsSection 7.6Plant Protection SystemSection 14.4Transient Events Analyzed for Core ReloadCriterion 22 - Separation of Protection and Control Instrumentation Systems(Category B) Protection systems shall be separated from control instrumentationsystems to the extent that failure or removal from service of any controlinstrumentation system component or channel, or of those common to control instrumentation and protection circuitry, leaves intact a system satisfying requirements for protection channels.Conformance 22 - Separation of Protection and Control Instrumentation Systems(Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control SystemsSection 7.4.2Reactor Vessel Instrumentation - DescriptionSection 7.4.3Reactor Vessel Instrumentation - Inspection and TestingSection 7.6.3Plant Protection System - Primary Containment Isolation SystemCriterion 23 - Protection Against Multiple Disability for Protection Systems(Category B) The effects of adverse conditions to which redundant channels orprotection systems might be exposed in common, either under normal conditions or those of an accident, shallnot result in loss of the protection function.Conformance 23 - Protection Against Multiple Disability for Protection Systems(Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 25 of 61I/jlkSection 5.2.1.3Primary Containment System -ContainmentPenetrationsSection 7.1Plant Instrumentation and Control Systems -
Summary DescriptionSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5Plant Radiation Monitoring SystemsSection 7.6Plant Protection SystemSection 11.2Turbine-Generator SystemCriterion 24 - Emergency Power for Protection Systems (Category B) In theevent of the loss of all off-site power, sufficient alternate sources of power shallbe provided to permit the required functioning of the protection systems.Conformance 24 - Emergency Power for Protection Systems (Category B)Section 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power SystemsSection 7 CompletePlant Instrumentation and Control SystemsSection 8.3Auxiliary Power SystemSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.3.8Plant Service Systems - Plant Communication SystemSection 10.3.9Plant Service Systems - Plant Lighting SystemCriterion 25 - Demonstration of Functional Operability of Protection System(Category B) Means shall be included for testing protection systems while thereactor is in operation to demonstrate that no failure or loss of redundancy has occurred.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 26 of 61I/jlkConformance 25 - Demonstration of Functional Operability of Protection System(Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and SafetyAnalysis - Plant Instrument Control SystemsSection 7.3.5.5Nuclear Instrumentation System - Inspection andTestingSection 7.4.3Reactor Vessel Instrumentation - Inspection and TestingSection 7.5.2.1Plant Radiation Monitoring Subsystem - GeneralSection 7.5.2.4.2Plant Radiation Monitoring Systems - DescriptionSection 7.6.1.4Plant Protection System - Inspection and TestingSection 7.6.3.4Plant Protection System - Inspection and TestingSection 10.3.1.4Plant Service Systems - Inspection and TestingSection 10.3.2.4Plant Service Systems - Plant Heating,Ventilating and Air Conditioning SystemsSection 10.3.9Plant Service Systems - Plant Lighting SystemSection 10.4Plant Cooling SystemsCriterion 26 - Protection Systems Fail-Safe Design (Category B) The protectionsystems shall be designed to fail into safe state or into a state established astolerable on a defined basis if conditions such as disconnection of the system,loss of energy (e.g., electric power, instrument air), or adverse environments(e.g., extreme heat or cold, fire, steam, or water) are experienced.Conformance 26 - Protection Systems Fail-Safe Design (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 27 of 61I/jlkSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power SystemsSection 3.5.1Reactivity Control Mechanical Characteristics -
Design BasisSection 3.5.5Reactivity Control Mechanical Characteristics -
Operation and Performance AnalysisSection 7.6Plant Protection SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.3Plant Service SystemsSection 10.4Plant Cooling SystemE.2.5Group V - Reactivity ControlThe intent of the current draft of the proposed criteria for this group is to establish the reactor core reactivity insertion and withdrawal rate limitations and the means to control the plant operations within these limits.It is concluded that the design of this plant is in conformance with the criteria ofGroup V based on NSP's current understanding of the intent of these criteria.The plant design contains two independent reactivity control systems of differentprinciples. Control of reactivity is operationally provided by a combination of movable control rods, fixed control devices or curtains, and reactor coolant recirculation system flow. These subsystems accommodate fuel burnup, load changes, and long term reactivity changes. Reactor shutdown by the control roddrive system is sufficiently rapid to prevent violation of fuel damage limits for alloperating transients. A reactor standby liquid control system is provided as a redundant, independent shutdown system to cover emergencies in the operational reactivity control system described above. This system is designed to shut down the reactor in about two hours. (Criterion 27, 28)The reactor core is designed to have (a) a reactivity response which regulates ordamps changes in power level and spatial distributions of power productions to a level consistent with safe and efficient operation, (b) a negative reactivityfeedback consistent with the requirements of overall plant nuclear-hydrodynamicstability, and (c) have a strong negative reactivity feedback under severe power transient conditions. (Criterion 27, 31) The operational reactivity control system is designed such that under conditions of normal operation sufficient reactivity compensation is always available to make the reactor adequately subcriticalfrom its most reactive condition, and means are provided for continuousregulation of the reactor core excess reactivity and reactivity distribution.
(Criterion 29, 30) This system is also designed to be capable of compensating Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 28 of 61I/jlkfor positive and negative reactivity changes resulting from nuclear coefficients,fuel depletion, and fission product transients and buildup. (Criterion 29) Thesystem design is such that control rod worths, and the rate at which reactivitycan be added, are limited to assure that credible reactivity accidents cannot cause a transient capable of damaging the reactor coolant system, disrupt the reactor core, its support structures, or other vessel internals sufficiently to impair the emergency core cooling systems effectiveness, if needed. Acceptable fueldamage limits will not be exceeded for any reactivity transient resulting from asingle equipment malfunction or reactor operator error. (Criterion 29, 31, 32)References to applicable sections of the USAR are given below for individualcriteria of this group.Criterion 27 - Redundancy of Reactivity Control (Category A) At least twoindependent reactivity control systems, preferable of different principles, shall beprovided.Conformance 27 - Redundancy of Reactivity Control (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.3.1Nuclear Characteristic - Design BasisSection 3.3.3.3Nuclear Characteristic - Reactivity ControlSection 3.3.3.4Nuclear Characteristic - Control Rod WorthSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6.3Standby Liquid Control System - PerformanceAnalysisSection 7.2Reactor Control SystemsSection 8.4Plant Standby Diesel Generator SystemsCriterion 28 - Reactivity Hot Shutdown Capability (Category A) At least two ofthe reactivity control systems provided shall independently be capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes, sufficiently fast to prevent exceeding acceptable fuel damage limits.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 29 of 61I/jlkConformance 28 - Reactivity Hot Shutdown Capability (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.3.1Nuclear Characteristic - Design BasisSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsCriterion 29 - Reactivity Shutdown Capability (Category A) At least one of thereactivity control systems provided shall be capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceedingly acceptable fuel damage limits. Shutdown margins greater than the maximum worth of the most efficient control rod when fully withdrawn shall be provided.Conformance 29 - Reactivity Shutdown Capability (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsCriterion 30 - Reactivity Holddown Capability (Category B) At least one of thereactivity control systems provided shall be capable of making and holding thecore subcritical under any conditions with appropriate margins for contingencies.Conformance 30 - Reactivity Holddown Capability (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.3.3.3Nuclear Characteristic - Reactivity Control Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 30 of 61I/jlkSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsCriterion 31 - Reactivity Control Systems Malfunction (Category B) Thereactivity control systems shall be capable of sustaining any single malfunction,such as unplanned continuous withdrawal (not ejection) of a control rod, without causing a reactivity transient which could result in exceeding acceptable fuel damage limits.Conformance 31 - Reactivity Control Systems Malfunction (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.2Thermal and Hydraulic CharacteristicsSection 3.3Nuclear CharacteristicSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.4Control Rod Velocity LimitersSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsCriterion 32 - Maximum Reactivity Worth of Control Rods (Category A) Limits,which include considerable margin, shall be placed on the maximum reactivityworth of control rods or elements and on rates at which reactivity can beincreased to ensure that the potential effects of a sudden or large change ofreactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling.Conformance 32 - Maximum Reactivity Worth of Control Rods (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 3.3.3.3Nuclear Characteristic - Reactivity ControlSection 3.3.3.4Nuclear Characteristic - Control Rod Worth Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 31 of 61I/jlkSection 3.4Fuel Mechanical CharacteristicsSection 3.5Reactivity Control Mechanical CharacteristicsSection 4 CompleteReactor Coolant SystemSection 6.4Control Rod Velocity LimitersSection 6.5Control Rod Drive Housing SupportsSection 7.8NUMAC Rod Worth Minimizer and Plant ProcessComputerSection 14.1.5Summary Description - Design Basis for AccidentsE.2.6Group VI - Reactor Coolant Pressure BoundaryThe intent of the current draft of the proposed criteria for this group is to establish the reactor coolant pressure boundary design requirements and toidentify the means used to satisfy these design requirements.It is concluded that the design of this plant is in conformance with the criteria ofGroup VI based on NSP's current understanding of the intent of these criteria.The inherent safety features of the reactor core design in combination withcertain engineered safety features (control rod velocity limiters and control rod housing supports, etc.) and the plant operational reactivity control system aresuch that the consequences of the most severe potential nuclear excursionaccident, caused by a single component failure within the reactivity control system (control rod drop accident) cannot result in damage (either by motion or rupture) to the reactor coolant system. (Criterion 33) The ASME and USASI Codes are used as the established and acceptable criteria for design,fabrication, and operation of components of the reactor primary pressuresystem. The reactor primary system is designed and fabricated to meet the following as a minimum: (Criterion 34)(1)Reactor Vessel - ASME Boiler and Pressure Vessel Code, SectionIII, Nuclear Vessels, Subsection A(2)Pumps - ASME Boiler and Pressure Vessel Code,Section III,Nuclear Vessels, Subsection C(3)Piping and Valves - USASI-B-31.1, Code for Pressure, Power PipingProtection against the brittle fracture or other failure modes of the reactor coolant pressure boundary system components is provided for all potential service Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 32 of 61I/jlkloading temperatures. Control is exercised in the selection of materials andfabrication and design of equipment and components. It is intended that NDTtesting be performed on all ferritic materials in the reactor coolant pressureboundary with appropriate modifications for material thickness of individual components. (Criterion 35)The reactor coolant system will be given a final hydrostatic test at 1560 psig inaccordance with Code requirements prior to initial reactor startup. A hydrostatic test, not to exceed system operating pressure, will be made on the reactor coolant system following each removal and replacement of the reactor vesselhead. The reactor primary system will be checked for leaks and abnormalconditions will be corrected before reactor startup. The minimum vessel temperature during hydrostatic test shall at least be 60° F above the calculated NDT temperature prior to pressurizing the vessel. Extensive quality control assurance programs are being so followed during the entire fabrication of thereactor coolant system. (Criterion 36) Vessel material surveillance samples arelocated within the reactor primary vessel to enable periodic monitoring of material properties with exposure. The program will include specimens of the base metal, heat affected zone metal, and standards specimens. Leakage from the reactor coolant system is monitored during reactor operation. (Criterion 36)References to applicable sections of the USAR are given on the following pagefor the individual criteria of this group.Criterion 33 - Reactor Coolant Pressure Boundary Capability (Category A)The reactor coolant pressure boundary shall be capable of accommodatingwithout rupture and with only limited allowance for energy absorption through plastic deformation, the static and dynamic loads imposed on any boundary component as a result of any inadvertent and sudden release of energy to thecoolant. As a design reference, this sudden release shall be taken as that whichwould result from a sudden reactivity insertion such as rod ejection (unless prevented by positive mechanical means), rod dropout, or cold water addition.Conformance 33 - Reactor Coolant Pressure Boundary Capability (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 3.3.3.3Nuclear Characteristic - Reactivity ControlSection 3.3.3.4Nuclear Characteristic - Control Rod WorthSection 3.4Fuel Mechanical CharacteristicsSection 3.5Reactivity Control Mechanical Characteristics Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 33 of 61I/jlkSection 4 CompleteReactor Coolant SystemSection 6.4Control Rod Velocity LimitersSection 6.5Control Rod Drive Housing SupportsSection 7.8NUMAC Rod Worth Minimizer and Plant ProcessComputerSection 14.1.5Summary Description - Design Basis for AccidentsCriterion 34 - Reactor Coolant Pressure Boundary Rapid Propagation FailurePrevent (Category A)The reactor coolant pressure boundary shall be designed to minimize the probability of rapidly propagating type failures. Consideration shall be given (a) to the notch-toughness properties if materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loading, (c) to the quality control specified for materials and componentfabrication to limit flaw sizes, and (d) to the provisions for control over servicetemperature and irradiation effects which may require operational restrictions.Conformance 34 - Reactor Coolant Pressure Boundary Rapid PropagationFailure Prevention (Category A)Section Appendix CQuality Assurance ProgramSection 4 CompleteReactor Coolant SystemCriteria 35 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention(Category A)Under conditions where reactor coolant pressure boundary system componentsconstructed of Ferritic materials may be subjected to potential loadings, such as a reactivity-induced loading, service temperatures shall be at least 120° F above the nil ductility transition (NDT) temperature of the component material if theresulting energy is expected to be absorbed within the elastic strain energyrange.Conformance 35 - Reactor Coolant Pressure Boundary Brittle FracturePrevention (Category A)Section 4.2.3Reactor Vessel - Design EvaluationSection 4.3.1Recirculation System - Design CriteriaSection 4.3.3Recirculation System - Performance Evaluation Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 34 of 61I/jlkSection 4.4.3Reactor Pressure Relief System - PerformanceAnalysisCriteria 36 - Reactor Coolant Pressure Boundary Surveillance (Category A)Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance by appropriate means to assess the structural and leak tight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance programconforming with ASTM-E-185-66 shall be provided.Conformance 36 - Reactor Coolant Pressure Boundary Surveillance (CategoryA)Section 4.2.1Reactor Vessel - Design BasisSection 4.3.1Recirculation System - Design BasisSection 4.3.4Recirculation System - Inspection and TestingSection 4.4.4Reactor Pressure Relief System - Inspectionand TestingE.2.7Group VII - Engineered Safety FeaturesThe intent of the current draft of the proposed criteria for this group is (a) to identify the engineered safety features (ESF), (b) to examine each ESF for independency, redundancy, capability, testability, inspectability, and reliability, (c) to determine the suitability of each ESF for its intended duty, and (d) justify that each ESFs capability-scope envelopes all the anticipated and crediblephenomena associated with the plant operational transients or design basisaccidents being considered.It is concluded that the design of the plant is in conformance with the criteria ofGroup VII based on NSP's current understanding of the intent of these criteria.The normal plant control systems maintain plant variables within narrowoperating limits. These systems are thoroughly engineered and backed up asignificant amount of experience in system design and operation. Even if an improbable maloperation or equipment failure including a reactor coolant boundary break up to and including the circumferential rupture of any pipe in that boundary assuming an unobstructed discharge from both sides allows variablesto exceed their operating limits, an extensive system of engineered safetyfeatures (ESF) limit the transient and the effects to levels well below those which are of public safety concern.These engineered safety features (ESF) include the normal protection systems(reactor core, reactor coolant system, plant containment system, plant and Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 35 of 61I/jlkreactor control systems, reactor protection system, other instrumentation andprocess systems, etc.); those which offer additional protection against areactivity excursion (reactor standby liquid control system, control rod velocitylimiters, and control rod housing support, etc.); those which act to reduce the consequences of design basis accidents (main steam line flow restrictors, etc.);
and those which provide emergency core and standby containment cooling in the event of a loss of normal cooling (emergency core cooling systems (ECCS),residual heat removal system (RHRS), high pressure coolant injection system(HPCIS), automatic depressurization system (ADS), and the standby coolant supply system). (Criterion 37)The engineered safety features are designed to provide high reliability and readytestability. Specific provisions are made in each ESF to demonstrate operability and performance capabilities. (Criterion 38) Components of the ESF which are required to function after design basis accidents or incidents are designed to withstand the most severe forces and credible environmental effects, includingmissiles from plant equipment failures anticipated from the events, withoutimpairment of their performance capability. (Criterion 40, 42, 43)Sufficient off-site and redundant, independent and testable standby auxiliarysources of electrical power are provided to attain prompt shutdown and continued maintenance of the plant in a safe condition under all credible circumstances. The capacity of the power sources are adequate to accomplish all required engineered safety features functions under all postulated designbasis accident conditions (Criterion 39).The emergency core cooling systems (ECCS) are designed such that at leasttwo different ECCSs of different phenomena are provided to prevent clad meltover the entire spectrum of postulated breaks. Such capability is available evenwith the loss of all off-site AC power. The ECCS (individual systems) themselves are designed to various levels of component redundancy such that no single active component failure in addition to the accident will negate the necessary emergency core cooling capability (Criterion 41, 44). To further assure that theECCS will function properly, if needed, specific provisions have been made fortesting the sequential operability and functional performance of each individual system (Criterion 46, 47, 48). Design provisions have also been made to enable physical and visual inspection of the ECCS components (Criterion 45).The primary containment structure, including access openings and penetrations,is designed to withstand the peak transient pressure and temperatures which could occur due to the postulated design basis loss-of-coolant design accident.The containment design includes considerable allowance for energy additionfrom metal-water or other chemical reactions beyond conditions that would occur with normal operation of Emergency Core Cooling Systems (ECCS). The primary containment has a metal-water reaction capability approximately 55% (at 2 hr) which is 500 times the calculated metal water reaction for the design basisloss-of-coolant accident (Criterion 49). Plates, structural member, forgings, andpipe associated with the drywell have an initial NDT temperature of approximately 0°F when tested in accordance with the appropriate code for the Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 36 of 61I/jlkmaterials. It is intended that the drywell will not be pressurized or subjected tosubstantial stress at temperatures below 30° F. Provisions are made for theremoval of heat from within the plant containment system and to isolate thevarious process system lines as may be necessary to maintain the integrity of the plant containment systems as long as necessary following the various postulated design basis accidents. The plant containment is designed and maintained so that the off-site doses resulting from the postulated design basisaccident will be below the values stated in 10CFR 100 (Criterion 50, 51, 54). Allpipes or ducts, which penetrate the primary containment and which connect to the reactor coolant system or to the drywell, are provided with at least two isolation valves in series (Criterion 53). The plant design provides for preoperational pressure and leak rate testing of the primary containment system,and include the capability for leak testing at design pressure after the plant hascommenced operation (Criterion 54, 55). Provisions are also made for demonstrating the functional performance of the plant containment system isolation valves and leak testing of selected penetrations (Criterion 56, 57).The pressure suppression pool and the containment spray cooling systemprovide two different means to rapidly condense the steam portion of the flow from the postulated design basis loss-of-coolant accident so that the peak transient pressure shall be substantially less than the primary containmentdesign pressure (Criterion 52). Demonstration of operability and the ability totest the functional performance and inspect the containment spray/cooling system are provided (Criterion 58, 59, 60, 61). The secondary containment standby gas treatment system is designed such that means are provided for periodic testing of the system performance including tracer injection andsampling (Criterion 64). The system may be physically inspected and itsoperability demonstrated (Criterion 62, 63, 65).References to applicable sections of the USAR are given below for the individualcriteria of this group.Criterion 37 - Engineered Safety Features Basis for Design (Category A)Engineered safety features shall be provided in the facility to back up the safetyprovided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assumingunobstructed discharge from both ends.Conformance 37 - Engineered Safety Features Basis for Design (Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand Control Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 37 of 61I/jlkSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.3Summary Design Description and SafetyAnalysis - Plant Containment SystemSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby CoolingSystemsSection 1.3.5Summary Design Description and SafetyAnalysis - Plant Instrumentation Control SystemsSection 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power SystemsSection 5 CompleteContainment SystemSection 6 CompletePlant Engineered SafeguardsSection 7 CompletePlant Instrumentation and Control SystemsSection 8 CompletePlant Electrical SystemsSection 10.3.8Plant Service Systems - Plant Communication SystemSection 10.3.9Plant Service Systems - Plant Lighting SystemSection 14.1.5Summary Description - Design Basis for AccidentsCriterion 38 - Reliability and Testability of Engineered Safety Features(Category A)All engineered safety features shall be designed to provide high functional reliability and ready testability. In determining the suitability of a facility for a proposed site, the degree of reliance upon and acceptance of the inherent and engineered safety afforded by the systems, including engineering safety features, will be influenced by the known and the demonstrated performancecapability and reliability of the systems, and by the extent to which the operabilityof such systems can be tested and inspected where appropriate during the life of the plant.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 38 of 61I/jlkConformance 38 - Reliability and Testability of Engineered Safety Features(Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.3Summary Design Description and SafetyAnalysis - Plant Containment SystemSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemsSection 5 CompleteContainment SystemSection 6 CompletePlant Engineered SafeguardsSection 7 CompletePlant Instrumentation and Control SystemsSection 8 CompletePlant Electrical SystemsSection 10.3.8Plant Service Systems - Plant Communication SystemSection 10.3.9Plant Service Systems - Plant Lighting SystemCriterion 39 - Emergency Power for Engineered Safety Features (Category A)Alternate power systems shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioningrequired of the engineered safety features. As a minimum, the on-site powersystem and the off-site power system shall each, independently, provide this capacity assuming a failure of a single active component in each power system.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 39 of 61I/jlkConformance 39 - Emergency Power for Engineered Safety Features(Category A)Section 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power SystemsSection 8.2Transmission SystemSection 8.3Auxiliary Power SystemSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesCriterion 40 - Missile Protection (Category A)Protection for engineered safety features shall be provided against dynamiceffects and missiles that might result from the plant equipment failures.Conformance 40 - Missile Protection (Category A)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.3.5Secondary Containment System - PerformanceAnalysisSection 12 CompletePlant Structures and ShieldingCriterion 41 - Engineered Safety Features Performance Capability (Category A)Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 40 of 61I/jlkConformance 41 - Engineered Safety Features Performance Capability(Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby CoolingSystemsSection 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power SystemsSection 5.2.1Primary Containment System - Design CriteriaSection 5.3.2Secondary Containment System - Design BasisSection 6.2.1.1Emergency Core Cooling System (ECCS) -
ECCS Design BasisSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.5.3Automatic Depressurization System (ADS) -Performance AnalysisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -
Performance AnalysisSection 6.2.3.3Residual Heat Removal System (RHR) -
Performance AnalysisSection 6.2.6Emergency Core Cooling System (ECCS) -
ECCS Performance EvaluationSection 6.3Main Steam Line Flow RestrictionsSection 6.4Control Rod Velocity Limiters Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 41 of 61I/jlkSection 6.5Control Rod Drive Housing SupportsSection 6.6Standby Liquid Control SystemSection 8.2Transmission SystemSection 8.3Auxiliary Power SystemsSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.3.4Plant Service Systems - Plant Instrumentationand Service Air SystemsSection 10.3.8Plant Service Systems - Plant Communication SystemSection 10.3.9Plant Service Systems - Plant Lighting SystemSection 14.1.5Summary Description - Design Basis for AccidentsCriterion 42 - Engineered Safety Features Components Capability (Category A)Engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident.Conformance 42 - Engineered Safety Features Components Capability(Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 3.6Other Reactor Vessel InternalsSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - Performance Analysis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 42 of 61I/jlkSection 6 CompletePlant Engineered SafeguardsSection 7.4Reactor Vessel InstrumentationSection 7.6Plant Protection SystemSection 12 CompletePlant Structures and ShieldingSection 14.1.5Summary Description - Design Basis AccidentAnalysisCriterion 43 - Accident Aggravation Prevention (Category A)Engineered safety features shall be designed so that any action of the engineered safety features which might accentuate the adverse affects of the loss of normal cooling avoided.Conformance 43 - Accident Aggravation Prevention (Category A)Section 5.2.3Primary Containment System - PerformanceAnalysisSection 6.2.1.1Emergency Core Cooling System (ECCS) -
ECCS Design BasisSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.5.3Automatic Depressurization System (ADS) -Performance AnalysisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -
Performance AnalysisSection 6.2.3.3Residual Heat Removal System (RHR) -Performance AnalysisSection 6.2.6Emergency Core Cooling System (ECCS) -ECCS Performance EvaluationSection 6.3.1Main Steam Line Flow Restrictions - Design BasisSection 6.4.1Control Rod Velocity Limiters - Design BasisSection 6.5.1Control Rod Drive Housing Supports - Design Basis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 43 of 61I/jlkSection 6.6.1Standby Liquid Control System - Design BasisCriterion 44 - Emergency Core Cooling System Capability (Category A)At least two emergency core cooling systems, preferably of different designprinciples, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction tonegligible amounts of all sizes of breaks in the reactor coolant pressureboundary, including the double-ended rupture of the largest pipe. The performance of each emergency core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shallnot share active components and shallnot share other features or components unless it can bedemonstrated that (a) the capability of the shared feature or components toperform its required function can be readily ascertained during reactor operation, (b) failure of the shared feature or component does not initiate a loss-of-coolant accident, and (c) capability of the shared feature or component to perform its required function is not impaired by the effects of a loss-of-coolant accident andis not lost during the entire period this function is required following the accident.Conformance 44 - Emergency Core Cooling Systems Capability (Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby Cooling SystemsSection 6.2.1.2Emergency Core Cooling System (ECCS) -
Description and Function of ECCSSection 6.2.2.1Reactor Core Spray Cooling System (CSCS) -
Design BasisSection 6.2.3.1Residual Heat Removal System (RHR) -
Design BasisSection 6.2.4.1High Pressure Coolant Injection System (HPCI) -
Design BasisSection 6.2.5.1Automatic Depressurization System (ADS) -
Design BasisSection 6.2.6Emergency Core Cooling System (ECCS) -
ECCS Performance Evaluation Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 44 of 61I/jlkSection 14.1.5Summary Description - Design Basis forAccidentsCriterion 45 - Inspection of Emergency Core Cooling Systems (Category A)Design provisions shall be made to facilitate physical inspection of all critical parts of the emergency core cooling systems, including reactor vessel internalsand water injection nozzles.Conformance 45 - Inspection of Emergency Core Cooling Systems (Category A)Section 3.6.1Other Reactor Vessel Internals - Design BasisSection 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and TestingSection 6.2.3.4Residual Heat Removal System (RHR) -
Inspection and TestingSection 6.2.4.4High Pressure Coolant Injection System (HPCI) -
Inspection and TestingSection 6.2.5.4Automatic Depressurization System (ADS) -
Inspection and TestingCriterion 46 - Testing of Emergency Core Cooling Systems Components(Category A)Design provisions shall be made so that active components of the emergencycore cooling systems, such as pumps and valves, can be tested periodically foroperability and require functional performance.Conformance 46 - Testing of Emergency Core Cooling Systems Components(Category A)Section 6.2.1.1Emergency Core Cooling System (ECCS) -ECCS Design BasisSection 6.2.2.1Reactor Core Spray Cooling System (CSCS) -
Design BasisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -Performance AnalysisSection 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and TestingSection 6.2.4.1 High Pressure Coolant Injection System (HPCI)-
Design Basis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 45 of 61I/jlkSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.4.4High Pressure Coolant Injection System (HPCI) -
Inspection and TestingSection 6.2.3.1Residual Heat Removal System (RHR) -
Design BasisSection 6.2.3.3Residual Heat Removal System (RHR) -Performance AnalysisSection 6.2.3.4Residual Heat Removal System (RHR) -Inspection and TestingSection 6.2.5.1Automatic Depressurization System (ADS) -
Design BasisSection 6.2.5.3Automatic Depressurization System (ADS) -
Performance AnalysisSection 6.2.5.4Automatic Depressurization System (ADS) -Inspection and TestingCriterion 47 - Testing of Emergency Core Cooling Systems (Category A)A capability shall be provided to test periodically the delivery capability of theemergency core cooling systems at a location as close to the core as is practical.Conformance 47 - Testing of Emergency Core Cooling Systems (Category A)Section 6.2.1.1Emergency Core Cooling System (ECCS) -ECCS Design BasisSection 6.2.2.1Reactor Core Spray Cooling System (CSCS) -
Design BasisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -Performance AnalysisSection 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and TestingSection 6.2.4.1 High Pressure Coolant Injection System (HPCI)-
Design BasisSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -
Performance Analysis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 46 of 61I/jlkSection 6.2.4.4High Pressure Coolant Injection System (HPCI) -Inspection and TestingSection 6.2.3.1Residual Heat Removal System (RHR) -
Design BasisSection 6.2.3.3Residual Heat Removal System (RHR) -
Performance AnalysisSection 6.2.3.4Residual Heat Removal System (RHR) -Inspection and TestingSection 6.2.5.1Automatic Depressurization System (ADS) -Design BasisSection 6.2.5.3Automatic Depressurization System (ADS) -
Performance AnalysisSection 6.2.5.4Automatic Depressurization System (ADS) -
Inspection and TestingCriterion 48 - Testing of Operational Sequence of Emergency Core CoolingSystem (Category A)A capability shall be provided to test under conditions as close to design aspractical the full operational sequence that would bring the emergency corecooling systems into action, including the transfer to alternate power sources.Conformance 48 - Testing of Operational Sequence of Emergency Core CoolingSystem (Category A)Section 6.2Emergency Core Cooling System (ECCS)Section 8 CompletePlant Electrical SystemsSection 8.2Transmission SystemSection 8.3Auxiliary Power SystemSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.4Plant Cooling System Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 47 of 61I/jlkCriterion 49 - Containment Design Basis (Category A)The containment structure, including access openings and penetrations, and anynecessary containment heat removal systems shall be designed so that the containment structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energyrelease following a loss-of-coolant accident, including a considerable margin foreffects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems.Conformance 49 - Containment Design Basis (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3Summary Design Description and Safety AnalysisSection 5.1Containment System - Summary DescriptionSection 5.2.3Primary Containment System - Performance AnalysisSection 5.2.4Primary Containment System - Inspection and TestingSection 5.3.2Secondary Containment System - Design BasisSection 5.3.5Secondary Containment System - Performance AnalysisSection 5.3.6Secondary Containment System - Inspection and TestingSection 6.2Emergency Core Cooling System (ECCS)Section 6.6Standby Liquid Control System Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 48 of 61I/jlkSection 10.2.5Reactor Auxiliary Systems - Reactor CoreIsolation Cooling System (RCIC)Section 14.1.5Summary Description - Design Basis forAccident AnalysisCriterion 50 - NDT Requirement for Containment Material (Category A)Principal load carrying components of ferritic materials exposed to the externalenvironment shall be selected so that their temperatures under normal operating and testing conditions are not less than 30° F above nil ductility transition (NDT) temperature.Conformance 50 - NDT Requirement for Containment Material (Category A)Section 5.2.2.2 - Primary Containment Construction Materials Criterion 51 - Reactor Coolant Pressure Boundary Outside Containment(Category A)If part of the reactor coolant pressure boundary is outside the containment,appropriate features as necessary shall be provided to protect the health and safety of the public in case of an accidental rupture in that part. Determination ofthe appropriateness of features such as isolation valves and additionalcontainment shall include consideration of the environmental and population conditions surrounding the site.Conformance 51 - Reactor Coolant Pressure Boundary Outside Containment(Category A)Section 1.2.1Principal Design Criteria - General CriteriaSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.2Summary Design Description and Safety Analysis - Reactor SystemSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.5Summary Design Description and SafetyAnalysis - Plant Instrumentation Control SystemsSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power System Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 49 of 61I/jlkSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 2.2Site DescriptionSection 5.2Primary Containment SystemSection 5.3Secondary Containment SystemSection 6.3Main Steam Line Floor RestrictionsSection 7.5.2Plant Radiation Monitoring Systems - Process Radiation Monitoring SystemSection 7.6.3Plant Protection System - Primary Containment Isolation SystemSection 14.1.5Summary Description - Design Basis for Accident AnalysisCriterion 52 - Containment Heat Removal Systems (Category A)Where active heat removal systems are needed under accident conditions toprevent exceeding containment design pressure, at least two systems,preferably of different principles, each with full capacity, shall be provided.Conformance 52 - Containment Heat Removal Systems (Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby CoolingSystemsSection 5.2Primary Containment SystemSection 6.2Emergency Core Cooling System (ECCS)Section 10.2Reactor Auxiliary Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 50 of 61I/jlkSection 10.4Plant Cooling SystemSection 14.1.5Summary Description - Design Basis forAccident AnalysisCriterion 53 - Containment Isolation Valves (Category A)Penetrations that require closure for the containment function shall be protectedby redundant valving and associated apparatus.Conformance 53 - Containment Isolation Valves (Category A)Section 5.2.1.3Primary Containment System - ContainmentPenetrationsSection 5.2.2.5.3Primary Containment System - Isolation SystemSection 5.2.3.7Primary Containment System - PenetrationsSection 5.2.3.6.2Primary Containment System - Isolation SystemSection 5.2.4Primary Containment System - Inspection and TestingSection 7.6.3Plant Protection System - Primary Containment Isolation SystemCriterion 54 - Containment Leakage Rate Testing (Category A)Containment shall be designed so that an integrated leakage rate testing can be conducted at design pressure after completion and installation of all penetrations and leakage rate measured over a sufficient period of time to verify its conformance with required performance.Conformance 54 - Containment Leakage Rate Testing (Category A)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.2.4Primary Containment System - Inspection and TestingSection 5.3.2Secondary Containment System - Design Basis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 51 of 61I/jlkSection 5.3.5Secondary Containment System - PerformanceAnalysisSection 5.3.6Secondary Containment System - Inspection and TestingCriterion 55 - Containment Periodic Leakage Rate Testing (Category A)The containment shall be designed so that integrated leakage rate testing can be done periodically at design pressure during plant lifetime.Conformance 55 - Containment Periodic Leakage Rate Testing (Category A)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.3.2Secondary Containment System - Design BasisCriterion 56 - Provisions for Testing of Penetrations (Category A)Provisions shall be made for testing penetrations which have resilient seals or expansion bellows to permit leak tightness to be demonstrated at design pressure at anytime.Conformance 56 - Provisions for Testing of Penetrations (Category A)Section 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.2.4Primary Containment System - Inspection andTestingSection 5.3.5Secondary Containment System - PerformanceAnalysisSection 5.3.6Secondary Containment System - Inspection and Testing Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 52 of 61I/jlkCriteria 57 - Provisions for Testing of Isolation Valves (Category A)Capability shall be provided for testing functional operability of valves andassociated apparatus essential to the containment function for establishing thatno failure has occurred and for determining that valve leakage does not exceedacceptable limits.Conformance 57 - Provisions for Testing of Isolation Valves (Category A)Section 7.6.3.1Plant Protection System - Design BasisSection 7.6.3.3Plant Protection System - Performance AnalysisSection 7.6.3.4Plant Protection System - Inspection and TestingSection 7.5.2Plant Radiation Monitoring Systems - ProcessRadiation Monitoring SystemCriterion 58 - Inspection of Containment Pressure-Reducing System(Category A)Design provisions shall be made to facilitate the periodic physical inspection ofall important components of the containment pressure-reducing systems, suchas, pumps, valves, spray nozzles, torus, and sumps.Conformance 58 - Inspection of Containment Pressure-Reducing System(Category A)Section 5.2.4Primary Containment System - Inspection andTestingSection 6.2Emergency Core Cooling System (ECCS)Criterion 59 - Testing of Containment Pressure-Reducing Systems Components(Category A)The containment pressure-reducing systems shall be designed so that active components such as pumps and valves can be tested periodically for operability and required functional performance.Conformance 59 - Testing of Containment Pressure-Reducing SystemsComponents (Category A)Section 6.2.1.1Emergency Core Cooling System (ECCS) -Design BasisSection 6.2Emergency Core Cooling System (ECCS)
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 53 of 61I/jlkCriterion 60 - Testing of Containment Spray Systems (Category A)A capability shall be provided to test periodically the delivery capability of thecontainment spray system at a position as close to the spray nozzle as ispractical.Conformance 60 - Testing of Containment Spray Systems (Category A)Section 6.2.1.1Emergency Core Cooling System (ECCS) -Design BasisSection 6.2Emergency Core Cooling System (ECCS)Criterion 61 - Testing of Operational Sequence of ContainmentPressure-Reducing Systems (Category A)A capability shall be provided to test under conditions as close to the design as practical the full operational sequence that would bring the containmentpressure-reducing systems into action, including the transfer to alternate powersources.Conformance 61 - Testing of Operational Sequence of ContainmentPressure-Reducing Systems (Category A)Section 5.2 CompletePrimary Containment SystemSection 7.6.3.3Plant Protection System - Performance AnalysisSection 7.6.3.4Plant Protection System - Inspection and TestingSection 6.2.1.1Emergency Core Cooling System (ECCS) -Design BasisSection 6.2Emergency Core Cooling System (ECCS)Section 8 CompletePlant Electrical SystemsCriterion 62 - Inspection of Air Cleanup Systems (Category A)Design provisions shall be made to facilitate physical inspection of all critical parts of containment air cleanup systems such as ducts, filters, fans, and dampers.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 54 of 61I/jlkConformance 62 - Inspection of Air Cleanup Systems (Category A)Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 5.3.5Secondary Containment System - Performance AnalysisSection 5.3.6Secondary Containment System - Inspection and TestingSection 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsCriterion 63 - Testing of Air Cleanup Components (Category A)Design provisions shall be made so that active components of the air cleanup systems, such as fans, dampers, can be tested periodically for operability and required functional performance.Conformance 63 - Testing of Air Cleanup Components (Category A)Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 5.3.5Secondary Containment System - PerformanceAnalysisSection 5.3.6Secondary Containment System - Inspection and TestingSection 10.3.2Plant Service Systems - Plant Heating,Ventilating and Air Conditioning SystemsCriterion 64 - Testing of Air Cleanup Systems (Category A)A capability shall be provided for insitu periodic testing and surveillance of the aircleanup systems to ensure (a) filter bypass paths have not developed and (b) filter and trapping materials have not deteriorated beyond acceptable limits.Conformance 64 - Testing of Air Cleanup Systems (Category A)Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 5.3.5Secondary Containment System - PerformanceAnalysis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 55 of 61I/jlkSection 5.3.6Secondary Containment System - Inspection andTestingSection 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsCriterion 65 - Testing of Operational Sequence Air Cleanup Systems(Category A)A capability shall be provided to test under conditions close to design as practical the full operational sequence that would bring the air cleanup systemsto action, including the transfer to alternate power sources and the design airflow delivery capability.Conformance 65 - Testing of Operational Sequence Air Cleanup Systems(Category A)Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 5.3.5Secondary Containment System - Performance AnalysisSection 5.3.6Secondary Containment System - Inspection andTestingSection 7.5.2Plant Radiation Monitoring Systems - ProcessRadiation Monitoring SystemSection 7.6.1Plant Protection System - Reactor Protection SystemSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsE.2.8Group VIII - Fuel and Waste Storage SystemsThe intent of the current draft of the proposed criteria for this group is to establish the safe fuel and waste storage systems design and to identify the means used to satisfy these requirements.It is concluded that the design of this plant is in conformance with criteria ofGroup VIII based on NSP's current understanding of the intent of these criteria.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 56 of 61I/jlkAppropriate plant fuel handling and storage facilities are provided to precludeaccidental criticality and to provide sufficient cooling for spent fuel. (Criterion 66,67) The new fuel storage vault racks (located inside the secondary containmentreactor building) are top entry, and are designed to prevent an accidental critical array, even in the event the vault becomes flooded. Vault drainage is provided to prevent possible water collection. (Criterion 66) The handling and storage of spent fuel, which takes place entirely within the reactor building (which providescontainment), is done in the spent fuel storage pool. The pool has provisions tomaintain water clarity, temperature control, and instrumentation to monitor water level. Water depth in the pool will be such as to provide sufficient shielding for normal reactor building occupancy (10 CFR 20) by operating personnel. The storage racks in which spent fuel assemblies are placed are designed andarranged to ensure subcriticality in the storage pool. (Criterion 66, 67, 68, 69)The spent fuel pool cooling and demineralizer system is designed to maintain the pool water temperature (decay heat removal) to control water clarity (safe fuel movement), and to reduce water radioactivity (shielding and effluent release control). (Criterion 66, 67, 68) Accessible portions of the reactor and radwastebuildings shall have sufficient shielding to maintain dose rates within 10 CFR 20.(Criterion 68) The radwaste building is designed to preclude accidental release of radioactive materials to the environs. (Criterion 69) The spent fuel storage pool and racks are designed and constructed such that all credible missiles as a result of a design basis tornado and tornado itself, will not have radiologicaleffects exceeding 10 CFR 100 guideline limitations.References to applicable sections of the USAR are given below for the individualcriteria of this group. (Criterion 67, 69)Criterion 66 - Prevention of Fuel Storage Critically (Category B)Critically in new and spent storage shall be prevented by physical systems orprocesses. Such means as geometrically safe configurations shall beemphasized over procedural controls.Conformance 66 - Prevention of Fuel Storage Critically (Category B)Section 1.2.9Principal Design Criteria - Plant Fuel Handlingand StorageSection 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and Handling SystemsSection 6.6.3Standby Liquid Control System - Performance AnalysisSection 10.2.1.1Reactor Auxiliary Systems - Design BasisSection 10.2.1.2Reactor Auxiliary Systems - Description01081199 Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 57 of 61I/jlkCriterion 67 - Fuel and Waste Storage Decay Heat (Category B)Reliable decay heat removal systems shall be designed to prevent damage tothe fuel in storage facilities that could result in radioactivity release to plantoperating areas or the public environs.Conformance 67 - Fuel and Waste Storage Decay Heat (Category B)Section 1.2.7Principal Design Criteria - Plant RadioactiveWaste DisposalSection 1.2.9Principal Design Criteria - Plant Fuel Handling and StorageSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling systemSection 1.3Summary Design Description and Safety AnalysisSection 6.2.1.2Emergency Core Cooling System (ECCS) -
Description and Function of ECCSSection 10.2.1Reactor Auxiliary Systems - Fuel Storage andFuel Handling SystemsSection 10.2.2Reactor Auxiliary Systems - Spent Fuel PoolCooling and Demineralizer SystemSection 10.2.3Reactor Auxiliary Systems - Reactor Cleanup Demineralizer SystemSection 10.2.4Reactor Auxiliary Systems - Reactor Shutdown Cooling SystemSection 12 CompletePlant Structures and ShieldingCriterion 68 - Fuel and Waste Storage Radiation Shielding (Category B)Shielding for radiation protection shall be provided in the design of spent fuel andwaste storage facilities as required to meet requirements of 10 CFR 20.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 58 of 61I/jlkConformance 68 - Fuel and Waste Storage Radiation Shielding (Category B)Section 1.2.8Principal Design Criteria - Plant Shielding andAccess ControlSection 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and HandlingSystemsSection 1.3.9Summary Design Description and SafetyAnalysis - Plant Shielding, Access Control, and Radiation Protection ProceduresSection 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste Control SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant SafetySection 12.3Shielding And Radiation ProtectionSection 9.2.1Liquid Radwaste System - Design BasisSection 9.2.3Liquid Radwaste System - Performance AnalysisSection 9.3.1Gaseous Radwaste System - Design BasisSection 9.3.3Gaseous Radwaste System - Performance AnalysisSection 9.4.1Solid Radwaste System - Design BasisSection 9.4.3Solid Radwaste System - Performance AnalysisSection 10.2.1.1Reactor Auxiliary Systems - Design BasisSection 10.2.1.2Reactor Auxiliary Systems - DescriptionSection 10.2.1.3Reactor Auxiliary Systems - PerformanceAnalysisCriterion 69 - Protection Against Radioactivity Release from Spent Fuel andWaste Storage (Category B)Containment of fuel and waste storage shall be provided if accidents could leadto release of undue amounts of radioactivity to the public environs.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 59 of 61I/jlkConformance 69 - Protection Against Radioactivity Release from Spent Fuel andWaste Storage (Category B)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.8Principal Design Criteria - Plant Shielding andAccess ControlSection 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and Handling SystemsSection 1.3.9Summary Design Description and SafetyAnalysis - Plant Shielding, Access Control, andRadiation Protection ProceduresSection 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste Control SystemsSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 5.1Containment System - Summary DescriptionSection 5.3Secondary Containment SystemSection 9 CompletePlant Radioactive Waste Control SystemsSection 10.2.1Reactor Auxiliary Systems - Fuel Storage andFuel Handling SystemsSection 10.2.2Reactor Auxiliary Systems - Spent Fuel PoolCooling and Demineralizer SystemSection 1.2.7Principal Design Criteria - Plant RadioactiveWaste DisposalSection 1.2.8Principal Design Criteria - Plant Shielding and Access ControlSection 14.7.6.4.2Refueling Accident Analysis - RadiologicalConsequencesSection 14.7.4Accident Evaluation Methodology - Fuel LoadingError Accident Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 60 of 61I/jlkE.2.9Group IX - Plant EffluentsThe intent of the current draft of the proposed criterion for this group is toestablish the plant effluent release limits and to identify the means of controlling the releases within these guide limits.It is concluded that the design of this plant is in conformance with the criteria ofGroup IX based on NSP's current understanding of the intent of these criteria.The plant radioactive waste control systems (which include the liquid, gaseousand solid radwaste sub-systems) are designed to limit the off-site radiation exposure to levels below doses set forth in 10 CFR 20. The plant engineeredsafety systems (including the containment barriers) are designed to limit theoff-site dose under various postulated "design basis" accidents to levels significantly below the limits of 10 CFR 100. The air ejector off-gas system is designed with sufficient holdup retention capacity so that during normal plant operation the controlled release of radioactive materials does not exceed theestablished release limits at the elevated plant stack. (Criterion 70)References to applicable sections of the USAR are given for the individualcriteria of this group.Criterion 70 - Control of Release of Radioactivity to the Environment(Category B)The facility design shall include those means necessary to maintain control overthe plant radioactive effluents, whether gaseous, liquid, or solid. Appropriateholdup capacity shall be provided for retention of gaseous, liquid, or solideffluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10 CFR 20 requirements for normaloperations and for any transient situation that might reasonably be anticipated tooccur and (b) on the basis of 10 CFR 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where highpopulation densities or very large cities can be affected by the radioactiveeffluents.Conformance 70 - Control of Release of Radioactivity to the Environment(Category B)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.7Principal Design Criteria - Plant RadioactiveWaste DisposalSection 1.2.8Principal Design Criteria - Plant Shielding and Access Control Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 61 of 61I/jlkSection 1.3.9Summary Design Description and SafetyAnalysis - Plant Shielding, Access Control, andRadiation Protection ProceduresSection 1.3.10Summary Design Description and SafetyAnalysis - Plant Radioactive Waste Control SystemsSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 2.2Site DescriptionSection 5 CompleteContainment SystemSection 12 CompletePlant Structures and ShieldingSection 7.5Plant Radiation Monitoring SystemsSection 8 CompletePlant Electrical SystemsSection 9 CompletePlant Radioactive Waste Control SystemsSection 10.3.6Plant Service Systems - Plant Equipment andFloor Drainage SystemsSection 10.3.7Plant Service Systems - Plant Process SamplingSystemSection 11.3.2Main Condenser System - Main Condenser GasRemoval SystemSection 13 CompletePlant OperationsSection 14 CompletePlant Safety Analysis Revision 22USAR APPENDIX EMONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 1APPENDIX EPLANT COMPARATIVE EVALUATION WITHTHE PROPOSED AEC 70 DESIGNCRITERIAI/mabTABLE OF CONTENTSSectionPageE.1Summary Description1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2Criterion - Conformance1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2.1Group I - Overall Plant Requirements1. . . . . . . . . . . . . . . . . . . . . . . E.2.2Group II - Protection by Multiple Fission Products Barriers11. . . . . E.2.3Group III - Nuclear and Radiation Controls15. . . . . . . . . . . . . . . . . . E.2.4Group IV - Reliability and Testability of Protection Systems21. . . . E.2.5Group V - Reactivity Control27. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E.2.6Group VI - Reactor Coolant Pressure Boundary31. . . . . . . . . . . . . . E.2.7Group VII - Engineered Safety Features34. . . . . . . . . . . . . . . . . . . . . E.2.8Group VIII - Fuel and Waste Storage Systems55. . . . . . . . . . . . . . . E.2.9Group IX - Plant Effluents60. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-E.TOCDoc Type:Admin Initials:Date:9703 Revision 22USAR E.1MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 1APPENDIX EPLANT COMPARATIVE EVALUATION WITHTHE PROPOSED AEC 70 DESIGNCRITERIAI/mabE.1Summary DescriptionThis appendix contains a comparative evaluation of the design basis of the MonticelloNuclear Generating Plant, Unit 1, with the 70 General Design Criteria for Nuclear Power Plant Construction Permits proposed by the Atomic Energy Commission for public comment in July, 1967.The comparative evaluation is made with each of the nine groups of criteria sent out inthe July 1967 AEC release. As to each group, there is a statement of Northern States Power Companys current understanding of the intent of the criteria in that group and a discussion of the plant design conformance with the intent of the group of criteria.Following a restatement of the 70 proposed criteria is complete list of references tolocations in this USAR where there is discussed subject matter relating to the intent of the particular criteria.Based on its current understanding of the intent of the 70 proposed-criteria, theapplicant believes that the Monticello Nuclear Generating Plant, Unit 1, is in conformance with the intent of such proposed criteria.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTPAssoc Ref:SR:2yrsNFreq:USAR-MANARMS:USAR-E.1Doc Type:Admin Initials:Date:9703 Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 1 of 61APPENDIX EPLANT COMPARATIVE EVALUATION WITHTHE PROPOSED AEC 70 DESIGNCRITERIAI/jlkE.2Criterion - ConformanceE.2.1Group I - Overall Plant RequirementsThe intent of the current draft of the proposed criteria for this group is to identifyand record the adequacy of the quality control and assurance programs, the applicable codes or standards, the standards of design, fabrication and erection, and to assure protection against appropriate environmental phenomena. Test Procedures, and inspection acceptance levels of the reactor facility's essentialcomponents and systems are also identified. The influence of this sharing ofcommon reactor facility components and systems along with the fire and explosion protection for all equipment is also to establish and described.It is concluded that the design of this plant is in conformance with the criteria ofGroup I based on NSP's current understanding of the intent of these criteria.The reactor facility's essential components and systems are designed,fabricated, erected, and perform in accordance with the specified quality standards which are, as a minimum, in accordance with applicable codes and regulations. These components and systems as well as applicable codes and standards have been identified in the report. Specific sections are included inthe reference letter list following this group's discussion. Where components orsystem design exceeds code requirements it has been noted. A quality control and assurance program has been established to assure compliance with acceptable quality control specifications and procedures. These programs as well as applicable tests and inspections have been identified. Specific sectionsare included in the reference list. In planning and executing these control andassurance programs, particular attention was given to the quality control specifications and to their compliance by those systems, components, and structures which are important to the plant safety. (Criterion 1) The plant equipment which is important to safety is designed to permit safe plant operationand to accommodate all design basis accidents for all appropriate environmentalphenomena at the site without loss of their capability, taking into consideration historical data and suitable margins for uncertainties. (Criterion 2) Further design allowances are provided to minimize the occurrence of fire and explosions and their effects by the use of noncombustible and fire resistantmaterials through the plant. (Criterion 3) Records of design, fabrication, andconstruction for this facility are to be stored or maintained either under the applicant's control or available to the applicant for inspection. (Criterion 5) This reactor facility consists of a single BWR generating unit. (Criterion 4)
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 2 of 61I/jlkReferences to applicable sections of the USAR are given below for the individualcriteria of this group.Criterion 1 - Quality Standards (Category A) Those systems and components ofreactor facilities which are essential to prevention of accidents which could affectthe public health and safety or to mitigation to their consequences shall be identified and then designed, fabricated, and erected to quality standards thatreflect the importance of the safety function to be performed. Where generallyrecognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assuranceprograms, test procedures, and inspection acceptance levels to be used shall beidentified. A showing of sufficiency and applicability of codes, standard, quality assurance programs, test procedures, and acceptance levels used is required.Conformance 1 - Quality Standards (Category A)a.GeneralSection 1.2.1Principal Design Criteria - General CriteriaSection 1.3.1.3Summary Design Description and SafetyAnalysis - GeologySection 1.3.1.4Summary Design Description and Safety Analysis - HydrologySection 1.3.1.5Summary Design Description and SafetyAnalysis - Regional and Site MeteorologySection 1.3.1.6Summary Design Description and SafetyAnalysis - Seismology and Design ResponseSpectrumSection 1.3.1.7Summary Design Description and SafetyAnalysis - Site Environmental Monitoring ProgramSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemSection 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and Handling Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 3 of 61I/jlkSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power SystemsSection 1.3.9Summary Design Description and Safety Analysis - Plant Shielding, Access Control, and Radiation Protection ProceduresSection 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste ControlSystemsSection Appendix CQuality Assurance Programb.Containment BarriersSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetyFuelSection 1.3.2Summary Design Description and Safety Analysis - Reactor SystemSection 3.4.4Fuel Mechanical Characteristics - Surveillance and TestingFuel CladdingSection 3.2.3Thermal and Hydraulic Characteristics - DesignCriteria and Safety LimitsSection 3.4.1Fuel Mechanical Characteristics - Design BasisSection 3.4.2Fuel Mechanical Characteristics - Description ofFuel AssembliesSection 3.4.3Fuel Mechanical Characteristics - Design EvaluationSection 3.4.4Fuel Mechanical Characteristics - Surveillance and Testing Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 4 of 61I/jlkReactor Coolant SystemSection 4Reactor Coolant SystemPrimary Containment SystemSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.2Primary Containment System - DescriptionSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.2.4Primary Containment System - Inspection and TestingSecondary Containment SystemSection 5.3.2Secondary Containment System - Design BasisSection 5.3.5Secondary Containment System - Performance AnalysisStandby Gas Treatment SystemSection 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsPlant Elevated Release PointSection 9.3Gaseous Radwaste Systemc.Plant Engineered SafeguardsSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 6.1Plant Engineered Safeguards - SummaryDescriptionControl Rod Velocity LimitersSection 6.4.3Control Rod Velocity Limiters - PerformanceAnalysisSection 6.4.4Control Rod Velocity Limiters - Inspection and Testing Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 5 of 61I/jlkControl Rod Drive Housing SupportsSection 6.5.3Control Rod Drive Housing Supports -Performance AnalysisSection 6.5.4Control Rod Drive Housing Supports - Inspectionand TestingReactor Standby Liquid Flow Control SystemSection 6.6.3Standby Liquid Control System - PerformanceAnalysisSection 6.6.4Standby Liquid Control System - Inspection andTrainingMain Steam Line Flow RestrictorsSection 6.3.3Main Steam Line Flow Restrictions -Performance AnalysisSection 6.3.4Main Steam Line Flow Restrictions - Inspectionand TestingEmergency Core Cooling Systems (ECCS)Section 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.5.3Automatic Depressurization System (ADS) -Performance AnalysisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -
Performance AnalysisSection 6.2.3.3Residual Heat Removal System (RHR) -
Performance AnalysisSection 6.2.6Emergency Core Cooling System (ECCS) -
ECCS Performance EvaluationPlant Structures and ShieldingSection 12.2Plant Principal Structures and FoundationsSection 12.3Shielding and Radiation Protection Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 6 of 61I/jlkCriterion 2 - Performance Standards (Category A) Those systems andcomponents of reactor facilities which are essential to prevention of accidentswhich could affect the public health and safety or to mitigation to theirconsequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, andother local site effects. The design bases so established shall reflect: (a)appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and surrounding area and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design.Conformance 2 - Performance Standards (Category A)a.GeneralSection 1.2.1Principal Design Criteria - General CriteriaSection 1.3.1.3Summary Design Description and SafetyAnalysis - GeologySection 1.3.1.4Summary Design Description and SafetyAnalysis - HydrologySection 1.3.1.5Summary Design Description and Safety Analysis - Site and Regional MeteorologySection 1.3.1.6Summary Design Description and SafetyAnalysis - Seismology and Design ResponseSpectraSection 1.3.1.7Summary Design Description and SafetyAnalysis - Site Environmental MonitoringProgramSection 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power SystemsSection 1.3.9Summary Design Description and Safety Analysis - Plant Shielding, Access Control, and Radiation Protection ProceduresSection 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste Control SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant Safety Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 7 of 61I/jlkSection 2.3MeteorologySection 2.4HydrologySection 2.5Geology and Soil InvestigationSection 2.6SeismologySection 2.7Radiation Environmental Monitoring Program(REMP)Section 2.8Ecological and Biological Studiesb.Containment BarriersSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemFuel CladdingSection 1.3.6Summary Design Description and SafetyAnalysis - Plant Fuel Storage and HandlingSystemsSection 3.2.1Thermal and Hydraulic Characteristics - Design BasisSection 3.2.3Thermal and Hydraulic Characteristics -Design Criteria and Safety LimitsSection 3.3.1Nuclear Characteristics - Design BasisSection 3.3.3Nuclear Characteristics - Nuclear Design CharacteristicsSection 3.4.1Fuel Mechanical Characteristics - Design BasisSection 3.4.3Fuel Mechanical Characteristics - DesignEvaluationSection 3.5.1Reactivity Control Mechanical Characteristics -Design BasisSection 3.5.5Reactivity Control Mechanical Characteristics -
Operation and Performance Analysis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 8 of 61I/jlkReactor Coolant SystemSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 4 - CompleteReactor Coolant SystemPrimary Containment SystemSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.4Primary Containment System - Inspection and TestingSection 12.2.1.1Plant Principal Structures and Foundations -Safety CategoriesSection Appendix ADesign Bases - Seismic Design and AnalysisSection 12.2.1.6Plant Principal Structures and Foundations -Wind LoadsSecondary Containment SystemSection 5.3.2Secondary Containment System - Design BasisSection 5.3.5Secondary Containment System - Performance AnalysisSection 12.2.1.1Plant Principal Structures and Foundations -
Safety CategoriesSection 12.2.1.6Plant Principal Structures and Foundations -
Wind LoadsSection 12.2.1.7Plant Principal Structures and Foundations -FloodingStandby Gas Treatment SystemSection 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 12.2.1.2Plant Principal Structures and Foundations -
Class I Structures and Equipment Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 9 of 61I/jlkPlant Elevated Release PointSection 9.3Gaseous Radwaste Systemc.Plant Engineered SafeguardsSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemControl Rod Velocity LimitersSection 6.4.1Control Rod Velocity Limiters - Design BasisSection 6.4.3Control Rod Velocity Limiters - Performance AnalysisControl Rod Drive Housing SupportsSection 6.5.1Control Rod Drive Housing Supports - Design BasisSection 6.5.3Control Rod Drive Housing Supports -
Performance AnalysisReactor Standby Liquid Flow Control SystemSection 6.6.1Standby Liquid Control System - Design BasisSection 6.6.3Standby Liquid Control System - Performance AnalysisMain Steam Line Flow RestrictorsSection 6.3.1Main Steam Line Flow Restrictions - Design BasisSection 6.3.3Main Steam Line Flow Restrictions -
Performance AnalysisEmergency Core Cooling Systems (ECCS)Section 6.2.1.1Emergency Core Cooling Systems (ECCS) -
ECCS Design Basis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 10 of 61I/jlkSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.5.3Automatic Depressurization System (ADS) -
Performance AnalysisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -
Performance AnalysisSection 6.2.3.3Residual Heat Removal System (RHR) -Performance AnalysisSection 6.2.6Emergency Core Cooling Systems (ECCS) -ECCS Performance EvaluationPlant Structures and ShieldingSection 12.2Plant Principal Structures and FoundationsSection 12.3Shielding and Radiation ProtectionCriterion 3 - Fire Protection (Category A) The reactor facility shall be designed(a) to minimize the probability of events such as fires and explosions and (b) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical through the facility,particularly in areas containing critical portions of the facility such ascontainment, control room, and components of engineered safety features.Conformance 3 - Fire Protection (Category A)Section 1.2.1Principal Design Criteria - General CriteriaSection 10.3.1Plant Service Systems - Fire Protection SystemsCriterion 4 - Sharing of Systems (Category A) Reactor facilities shallnot sharesystems or components unless it is shown safety is not impaired by the sharing.
Conformance 4 - Sharing of Systems (Category A) This Plant is a single unitand does not share any system, component, or equipment with any other facility.
Criterion 5 - Records Requirements (Category A) Records of design, fabrication,and construction of essential components of the plant shall be maintained by thereactor operator (NSP) or under its control throughout the life of the reactor.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 11 of 61I/jlkConformance 5 - Records Requirements (Category A)Section Appendix CQuality Assurance ProgramSection 13.4Operational ProceduresSection 13.5Operational Records and ReportingRequirementsE.2.2Group II - Protection by Multiple Fission Products BarriersThe intent of the current draft of the proposed criteria for this group is to assurethat the plant has been provided with multiple barriers to protect against or tomitigate the effects of fission products prior to being released to the site environs and to establish that these barriers will remain intact under all operational transients caused by a single reactor operator error or equipment malfunction. It is the further intent of this group that proper barriers are made available for thedesign basis accidents.It is concluded that design of this plant is in conformance with the Criteria ofGroup II Based on NSP's understanding of the intent of these criteria.The plant containment barriers are the basic features which minimize release ofradioactive materials and associated doses. A boiling water reactor provides seven means of containing and/or mitigating the release of fission products; (a) the high density ceramic UO2 fuel, (b) the high integrity Zircaloy cladding, (c) thereactor vessel and its connected piping and isolation valves, (d) the drywell-suppression chamber primary containment, (e) the reactor building (secondary containment), (f) the reactor building standby gas treatment system utilizing high efficiency absolute and charcoal filters, and (g) the plant main stack. The primary containment system is designed, fabricated, and erected toaccommodate without failure, the pressures and temperatures resulting from orsubsequent to double-ended rupture or equivalent failure of any coolant pipe within the primary containment. The reactor building, encompassing the primary containment system, provides secondary containment when the primary containment is closed and in service, and provides primary containment whenthe primary containment is open for refueling operations. The two containmentsystems and such other associated engineered safety systems as may be necessary are designed and maintained so that off-site doses resulting from postulated design basis accidents are below the values stated in 10CFR100.(Criterion 10) The reactor core is designed so there is no inherent tendency forsudden divergent oscillation of operating characteristics of divergent power transient in any mode of plant operation. (Criterion 6, 7) The basis of the reactor core design, in combination with the plant equipment characteristics, nuclear instrumentation system, and the reactor protection system is, to providemargins to ensure that fuel damage will not occur in normal operation oroperational transient caused by single reactor operator error or equipment malfunction. (Criterion 6, 7) The reactor core is designed so that the overall01081199 Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 12 of 61I/jlkpower coefficient in the power operating range is not positive. (Criterion 8) Thereactor coolant system is designed to carry its dead weight and specified liveloads, separately or concurrently, such as pressure and temperature stress,vibrations, seismic loads as appropriately prescribed for the plant. Provisions are made to control or shutdown the reactor coolant system in the event of a malfunction of the operating equipment or excessive leakage of the coolant from the system. The reactor vessel and support structure are designed, within thelimits of applicable criteria for low probability accident conditions, to withstandthe forces that would be created by a full area flow from any vessel nozzle to the containment atmosphere with the reactor vessel at design pressure concurrent with the plant design earthquake loads. (Criterion 9)References to applicable sections of the USAR are given below for the individualcriteria of this group.Criterion 6 - Reactor Core Design (Category A) The reactor core shall bedesigned to function throughout its design lifetime, without exceeding acceptablefuel damage limits which have been stipulated and justified. The core design,together with reliable process and decay heat removal systems, shall provide for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and for transient situations which can be anticipated, including the effects of the loss of power to recirculation pumps, tripping out of aturbine generator set, isolation of the reactor from its primary heat sink, and lossof off-site power.Conformance 6 - Reactor Core Design (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling SystemsSection 3.2Thermal and Hydraulic CharacteristicsSection 3.3Nuclear CharacteristicsSection 3.4Fuel Mechanical CharacteristicsSection 3.5Reactivity Control Mechanical CharacteristicsSection 4Reactor Coolant System Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 13 of 61I/jlkSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.2.5Reactor Auxiliary Systems - Reactor CoreIsolation Cooling System (RCIC)Section 14.4.3Transient Events Analyzed for Core Reload -
Rod Withdrawal ErrorCriterion 7 - Suppression of Power Oscillations (Category B) The core design,together with reliable controls, shall ensure that power oscillations which could cause damage in excess of acceptable fuel damage limits are not possible or can be readily suppressed.Conformance 7 - Suppression of Power Oscillations (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreCriterion 8 - Overall Power Coefficient (Category B) The reactor shall bedesigned so that the overall power coefficient in the power operating range shallnot be positive.Conformance 8 - Overall Power Coefficient (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreSection 3.2Thermal and Hydraulic CharacteristicsSection 3.5Reactivity Control Mechanical CharacteristicsCriterion 9 - Reactor Coolant Pressure Boundary (Category A) The reactorcoolant pressure boundary shall be designed and constructed so as to have anexceedingly low probability of gross rupture or significant leakage throughout its design lifetime.Conformance 9 - Reactor Coolant Pressure Boundary (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 4 CompleteReactor Coolant SystemSection 7.4Reactor Vessel Instrumentation Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 14 of 61I/jlkCriterion 10 - Containment (Category A) Containment shall be provided. Thecontainment structure shall be designed to sustain the initial effects of grossequipment failures, such as a large coolant boundary area, without loss ofrequired integrity and, together with other engineered safety features as may be necessary to retain for as long as the situation requires the functional capability to protect the public.Conformance 10 - Containment (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.3.3Summary Design Description and SafetyAnalysis - Plant Containment SystemSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby CoolingSystemsSection 4 CompleteReactor Coolant SystemSection 5.1Containment System - Summary DescriptionSection 6.2Emergency Core Cooling Systems (ECCS)Section 6.4Control Rod Velocity LimitersSection 6.5Control Rod Drive Housing SupportsSection 6.6Standby Liquid Control SystemSection 5.2.1Primary Containment System - Design CriteriaSection 5.3.2Secondary Containment System - Design BasisSection 12 CompletePlant Structures and ShieldingSection 14.1.1Summary Description - General Safety DesignBasisSection 14.1.5Summary Description - Design Basis for Accidents Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 15 of 61I/jlkE.2.3Group III - Nuclear and Radiation ControlsThe intent of the current draft of the proposed criteria for this group is to identifyand define the instrumentation and control systems, necessary for maintainingthe plant in a safe operational status. This, also includes determining the adequacy of radiation shielding, effluent monitoring, and fission process controls, and providing for the effective sensing of abnormal conditions and initiation of engineered safety features.It is concluded that the design of this plant is in conformance with the criteria ofGroup III based on NSP's current understanding of the intent of these criteria.The plant is provided with a centralized main control room having adequateshielding, fire protection, air conditioning and facilities to permit access and continuous occupancy under 10CFR20 dose limits during all design basisaccident situations. However, if it is necessary to evacuate the main controlroom the design does not preclude the capability to bring the plant to a safe-cold shutdown from outside the main control room. (Criterion 11) The necessary plant controls, instrumentation, and alarms for safe and orderly operation are located in the main control room. These include such controls andinstrumentation as the reactor coolant system leakage detection system.(Criterion 11, 13, 16) The performance of the reactor core and the indication of power level are continuously monitored by the in-core nuclear instrumentation system. (Criterion 13) The reactor protection system, independent from the plant process control systems, overrides all other controls to initiate any requiredsafety action. The reactor protection system automatically initiates appropriateaction whenever the plant conditions approach pre-established operational limits.
The system acts specifically to initiate the emergency core and containment cooling systems as required. (Criterion 12, 13, 14, 15) The plant radiation and process monitoring systems are provided for monitoring significant parametersfrom specific plant process systems and specific areas including the planteffluents to the site environs and to provide alarms and signals for appropriate corrective actions. (Criterion 17, 18)Reference to applicable sections of the USAR are given below for the individualcriteria of this group.Criterion 11 - Control Room (Category B) The facility shall be provided with acontrol room from which action to maintain safe operational status of the plantcan be controlled. Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control room orother areas as necessary to shut down and maintain safe control to the facilitywithout radiation exposures of personnel in excess of 10CFR20 limits. It shall be possible to shut the reactor down and maintain it in a safe condition if access the control room is lost due to fire or other causes.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 16 of 61I/jlkConformance 11 - Control Room (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.8Principal Design Criteria - Plant Shielding and Access ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation and Control SystemsSection 1.3.9Summary Design Description and SafetyAnalysis - Plant Shielding, Access Control, andRadiation Protection ProceduresSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant SafetySection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.6Plant Protection SystemSection 7.7Turbine-Generator System Instrumentation and ControlSection 12.3.3Shielding and Radiation Protection -
Performance AnalysisCriterion 12 - Instrumentation and Control Systems (Category B)Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges.Conformance 12 - Instrumentation and Control Systems (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and SafetyAnalysis - Plant Instrumentation Control SystemsSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 7Plant Instrumentation and Control Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 17 of 61I/jlkSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5Plant Radiation Monitoring SystemsSection 7.6Plant Protection SystemSection 7.7Turbine-Generator System Instrumentation andControlSection 7.8NUMAC Rod Worth Minimizer and Plant Process ComputerCriterion 13 - Fission Process Monitors and Controls (Category B) Means shallbe provided for monitoring and maintaining control over the fission process throughout core life and for all conditions that can reasonably be anticipated tocause variation in reactivity of the core, such as indication of position of controlrods and concentration of soluble reactivity control poisons.Conformance 13 - Fission Process Monitors and Controls (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemsSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.6Plant Protection SystemSection 7.8NUMAC Rod Worth Minimizer and Plant Process Computer Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 18 of 61I/jlkCriterion 14 - Core Protection Systems (Category B) Core protection systemstogether with associated equipment, shall be designed to act automatically toprevent or to suppress conditions that could result in exceeding acceptable fueldamage limits.Conformance 14 -Core Protection Systems (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation and Control SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant SafetySection 3.3Nuclear CharacteristicsSection 3.4Fuel Mechanical CharacteristicsSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.2Emergency Core Cooling System (ECCS)Section 6.3Main Steam Line Flow RestrictionsSection 6.4Control Rod Velocity LimitersSection 6.5Control Rod Drive Housing SupportsSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.6Plant Protection SystemSection 7.8NUMAC Rod Worth Minimizer and Plant Process Computer Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 19 of 61I/jlkSection 8 CompletePlant Electrical SystemsSection 14 CompletePlant Safety AnalysisCriterion 15 - Engineered Safety Features Protection Systems (Category B)Protection systems shall be provided for sensing accident situations andinitiating the operation of necessary engineered safety features.Conformance 15 - Engineered Safety Features Protection Systems (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control SystemsSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 6 CompletePlant Engineered SafeguardsSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5Plant Radiation Monitoring SystemsSection 7.6Plant Protection SystemSection 7.7Turbine-Generator Systems Instrumentation andControlSection 7.8NUMAC Rod Worth Minimizer and Plant Process ComputerCriterion 16 - Monitoring Reactor Coolant Pressure Boundary (Category B)Means shall be provided for monitoring the reactor coolant pressure boundary to detect leakage.Conformance 16 - Monitoring Reactor Coolant Pressure Boundary (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 20 of 61I/jlkSection 5.2Primary Containment SystemSection 7.1Plant Instrumentation and Control Systems -Summary DescriptionSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.6Plant Protection SystemCriterion 17 - Monitoring Radioactivity Releases (Category B) Means shall beprovided for monitoring the containment atmosphere, the facility effluentdischarge paths, and the facility environs, for radioactivity that could be released from normal operations, from anticipated transients, and from accidentconditions.Conformance 17 - Monitoring Radioactivity Releases (Category B)Section 1.2.7Principal Design Criteria - Plant RadioactiveWaste DisposalSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control SystemsSection 5.3.4.1Secondary Containment System - Standby Gas Treatment System (SGTS)Section 7.5Plant Radiation Monitoring SystemsSection 7.6.1Plant Protection System - Reactor Protection SystemSection 9.2Liquid Radwaste SystemSection 9.3Gaseous Radwaste SystemSection 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsSection 10.3.7Plant Service Systems - Plant Process Sampling SystemSection 14.1.5Summary Description - Design Basis for Accidents Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 21 of 61I/jlkCriterion 18 - Monitoring Fuel and Waste Storage (Category B) Monitoring andalarm instrumentation shall be provided for fuel and waste storage and handlingareas for conditions that might contribute to loss of continuity in decay heatremoval and to radiation exposures.Conformance 18 - Monitoring Fuel and Waste Storage (Category B)Section 7.5Plant Radiation Monitoring SystemsSection 7.6.1Plant Protection System - Reactor ProtectionSystemSection 9.2.1Liquid Radwaste System - Design BasisSection 9.2.2.1Liquid Radwaste System - GeneralSection 9.2.2.3Liquid Radwaste System - Instrumentation andControl of the Liquid RadwasteSection 9.3.1Gaseous Radwaste System - Design BasisSection 9.3.3Gaseous Radwaste System - PerformanceAnalysisSection 9.4.1Solid Radwaste System - Design BasisSection 9.4.3Solid Radwaste System - Performance AnalysisSection 10.2.1.1Reactor Auxiliary Systems - Design BasisSection 10.2.1.2Reactor Auxiliary Systems - DescriptionSection 10.2.2.1Reactor Auxiliary Systems - Design BasisSection 10.2.2.3Reactor Auxiliary Systems - Performance AnalysisE.2.4Group IV - Reliability and Testability of Protection SystemsThe intent of the current draft of the proposed criteria for this group is to identifyand establish the functional reliability, in-service testability, redundancy, physicaland electrical independence and separation, and fail-safe design of the reactor protection instrumentation and control systems.It is concluded that the design of this plant is in conformance with the criteria ofGroup IV based on NSP's current understanding of the intent of these criteria.The reactor protection system automatically overrides the plant normaloperational control system (that is, functions independently) to initiate Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 22 of 61I/jlkappropriate action whenever the plant conditions monitored (neutron flux,containment, and vessel pressure, etc.) by the system approach pre-establishedlimits. (Criterion 22) By means of a dual channel protection system withcomplete redundancy in each channel, no loss of the protection systems can occur by either component failure or removal from service. The reactor protection system acts to shutdown the reactor, close primary containment isolation valves and initiates the operation of the emergency core andcontainment cooling systems. The reactor protection system is designed so thata credible plant transient or accident is sensed by different parametric measurements (e.g., loss of coolant accident is detected by high drywell pressure and low-low reactor level monitors). (Criterion 20) Components of the redundant subsystems can be removed from service for testing andmaintenance without negating the ability of the protection system to perform itsprotection functions (even when subjected to a single event, multiple failure incident) upon receipt of the appropriate signals. (Criterion 19, 20, 21) The design of the reactor protection system is such as to facilitate maintenance and trouble shooting while the reactor is at power operation without impeding theplant's operation or impairing its safety function. System faults are annunciatedin the main control room. (Criterion 25) The system electrical power requirements are supplied from independent, redundant sources. (Criterion 24)
The system circuits are isolated to preclude a circuit fault from inducing a fault in another circuit and to reduce the likelihood that adverse conditions, which mightaffect system reliability (1 of 2 x 2), will encompass more than one circuit. Thesystem sensors are electrically and physically separated with both sensors in any one trip channel not allowed to occupy the same local area or to be connected to the same power source or process measurement line. The system internal wiring or external cable routing arrangement are such as to negate anyexternal influence (a fire or accident) on the systems performance. (Criterion 23,24) A failure of any one reactor protection system input or subsystem component will produce a trip in one of two channels, a situation insufficient to produce a reactor scram but readily available to perform its protective functionupon another trip (either by failure or by exceeding the preset trip). (Criterion 26)This reactor protection system design includes allowance for single reactor operator error and equipment malfunction and still performs its intended function.
(Criterion 21) References to applicable sections of the USAR are given below for the individual criteria of this group.Criterion 19 - Protection Systems Reliability (Category B) Protection systemsshall be designed for high functional reliability and in-service testabilitycommensurate, with the safety functions to be performed.Conformance 19 - Protection Systems Reliability (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.1Summary Design Description and Safety Analysis - Plant Site and Environs Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 23 of 61I/jlkSection 7.2Reactor Control SystemsSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5.2Plant Radiation Monitoring systems - ProcessRadiation Monitoring SystemsSection 7.6Plant Protection SystemSection 11.2Turbine-Generator SystemSection 14.1.5Summary Description - Design Basis for AccidentsCriterion 20 - Protection Systems Redundancy and Independence (Category B)Redundancy and independence designed into protection systems shall be sufficient to assure that no single failure or removal from service of any component or channel of a system will result in loss of the protection function.The redundancy provided shall include, as a minimum, two channels ofprotection for each protection function to be served. Different principles shall be used where necessary to achieve true redundant instrumentation components.Conformance 20 - Protection Systems Redundancy and Independence(Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control SystemsSection 7.1Plant Instrumentation and Control Systems -
Summary DescriptionSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5.2Plant Radiation Monitoring Systems - Process Radiation Monitoring SystemSection 7.6Plant Protection SystemSection 11.2Turbine-Generator SystemSection 14.1.5Summary Description - Design Basis for Accidents Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 24 of 61I/jlkCriterion 21 - Single Failure Definition (Category B) Multiple failures from asingle event shall be treated as a single failure.Conformance 21 - Single Failure Definition (Category B)Section 7.2Reactor Control SystemsSection 7.6Plant Protection SystemSection 14.4Transient Events Analyzed for Core ReloadCriterion 22 - Separation of Protection and Control Instrumentation Systems(Category B) Protection systems shall be separated from control instrumentationsystems to the extent that failure or removal from service of any controlinstrumentation system component or channel, or of those common to control instrumentation and protection circuitry, leaves intact a system satisfying requirements for protection channels.Conformance 22 - Separation of Protection and Control Instrumentation Systems(Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control SystemsSection 7.4.2Reactor Vessel Instrumentation - DescriptionSection 7.4.3Reactor Vessel Instrumentation - Inspection and TestingSection 7.6.3Plant Protection System - Primary Containment Isolation SystemCriterion 23 - Protection Against Multiple Disability for Protection Systems(Category B) The effects of adverse conditions to which redundant channels orprotection systems might be exposed in common, either under normal conditions or those of an accident, shallnot result in loss of the protection function.Conformance 23 - Protection Against Multiple Disability for Protection Systems(Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 25 of 61I/jlkSection 5.2.1.3Primary Containment System -ContainmentPenetrationsSection 7.1Plant Instrumentation and Control Systems -
Summary DescriptionSection 7.3Nuclear Instrumentation SystemSection 7.4Reactor Vessel InstrumentationSection 7.5Plant Radiation Monitoring SystemsSection 7.6Plant Protection SystemSection 11.2Turbine-Generator SystemCriterion 24 - Emergency Power for Protection Systems (Category B) In theevent of the loss of all off-site power, sufficient alternate sources of power shallbe provided to permit the required functioning of the protection systems.Conformance 24 - Emergency Power for Protection Systems (Category B)Section 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power SystemsSection 7 CompletePlant Instrumentation and Control SystemsSection 8.3Auxiliary Power SystemSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.3.8Plant Service Systems - Plant Communication SystemSection 10.3.9Plant Service Systems - Plant Lighting SystemCriterion 25 - Demonstration of Functional Operability of Protection System(Category B) Means shall be included for testing protection systems while thereactor is in operation to demonstrate that no failure or loss of redundancy has occurred.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 26 of 61I/jlkConformance 25 - Demonstration of Functional Operability of Protection System(Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.5Summary Design Description and SafetyAnalysis - Plant Instrument Control SystemsSection 7.3.5.5Nuclear Instrumentation System - Inspection andTestingSection 7.4.3Reactor Vessel Instrumentation - Inspection and TestingSection 7.5.2.1Plant Radiation Monitoring Subsystem - GeneralSection 7.5.2.4.2Plant Radiation Monitoring Systems - DescriptionSection 7.6.1.4Plant Protection System - Inspection and TestingSection 7.6.3.4Plant Protection System - Inspection and TestingSection 10.3.1.4Plant Service Systems - Inspection and TestingSection 10.3.2.4Plant Service Systems - Plant Heating,Ventilating and Air Conditioning SystemsSection 10.3.9Plant Service Systems - Plant Lighting SystemSection 10.4Plant Cooling SystemsCriterion 26 - Protection Systems Fail-Safe Design (Category B) The protectionsystems shall be designed to fail into safe state or into a state established astolerable on a defined basis if conditions such as disconnection of the system,loss of energy (e.g., electric power, instrument air), or adverse environments(e.g., extreme heat or cold, fire, steam, or water) are experienced.Conformance 26 - Protection Systems Fail-Safe Design (Category B)Section 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 27 of 61I/jlkSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power SystemsSection 3.5.1Reactivity Control Mechanical Characteristics -
Design BasisSection 3.5.5Reactivity Control Mechanical Characteristics -
Operation and Performance AnalysisSection 7.6Plant Protection SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.3Plant Service SystemsSection 10.4Plant Cooling SystemE.2.5Group V - Reactivity ControlThe intent of the current draft of the proposed criteria for this group is to establish the reactor core reactivity insertion and withdrawal rate limitations and the means to control the plant operations within these limits.It is concluded that the design of this plant is in conformance with the criteria ofGroup V based on NSP's current understanding of the intent of these criteria.The plant design contains two independent reactivity control systems of differentprinciples. Control of reactivity is operationally provided by a combination of movable control rods, fixed control devices or curtains, and reactor coolant recirculation system flow. These subsystems accommodate fuel burnup, load changes, and long term reactivity changes. Reactor shutdown by the control roddrive system is sufficiently rapid to prevent violation of fuel damage limits for alloperating transients. A reactor standby liquid control system is provided as a redundant, independent shutdown system to cover emergencies in the operational reactivity control system described above. This system is designed to shut down the reactor in about two hours. (Criterion 27, 28)The reactor core is designed to have (a) a reactivity response which regulates ordamps changes in power level and spatial distributions of power productions to a level consistent with safe and efficient operation, (b) a negative reactivityfeedback consistent with the requirements of overall plant nuclear-hydrodynamicstability, and (c) have a strong negative reactivity feedback under severe power transient conditions. (Criterion 27, 31) The operational reactivity control system is designed such that under conditions of normal operation sufficient reactivity compensation is always available to make the reactor adequately subcriticalfrom its most reactive condition, and means are provided for continuousregulation of the reactor core excess reactivity and reactivity distribution.
(Criterion 29, 30) This system is also designed to be capable of compensating Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 28 of 61I/jlkfor positive and negative reactivity changes resulting from nuclear coefficients,fuel depletion, and fission product transients and buildup. (Criterion 29) Thesystem design is such that control rod worths, and the rate at which reactivitycan be added, are limited to assure that credible reactivity accidents cannot cause a transient capable of damaging the reactor coolant system, disrupt the reactor core, its support structures, or other vessel internals sufficiently to impair the emergency core cooling systems effectiveness, if needed. Acceptable fueldamage limits will not be exceeded for any reactivity transient resulting from asingle equipment malfunction or reactor operator error. (Criterion 29, 31, 32)References to applicable sections of the USAR are given below for individualcriteria of this group.Criterion 27 - Redundancy of Reactivity Control (Category A) At least twoindependent reactivity control systems, preferable of different principles, shall beprovided.Conformance 27 - Redundancy of Reactivity Control (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.3.1Nuclear Characteristic - Design BasisSection 3.3.3.3Nuclear Characteristic - Reactivity ControlSection 3.3.3.4Nuclear Characteristic - Control Rod WorthSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6.3Standby Liquid Control System - PerformanceAnalysisSection 7.2Reactor Control SystemsSection 8.4Plant Standby Diesel Generator SystemsCriterion 28 - Reactivity Hot Shutdown Capability (Category A) At least two ofthe reactivity control systems provided shall independently be capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes, sufficiently fast to prevent exceeding acceptable fuel damage limits.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 29 of 61I/jlkConformance 28 - Reactivity Hot Shutdown Capability (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.3.1Nuclear Characteristic - Design BasisSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsCriterion 29 - Reactivity Shutdown Capability (Category A) At least one of thereactivity control systems provided shall be capable of making the core subcritical under any condition (including anticipated operational transients) sufficiently fast to prevent exceedingly acceptable fuel damage limits. Shutdown margins greater than the maximum worth of the most efficient control rod when fully withdrawn shall be provided.Conformance 29 - Reactivity Shutdown Capability (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsCriterion 30 - Reactivity Holddown Capability (Category B) At least one of thereactivity control systems provided shall be capable of making and holding thecore subcritical under any conditions with appropriate margins for contingencies.Conformance 30 - Reactivity Holddown Capability (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.3.3.3Nuclear Characteristic - Reactivity Control Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 30 of 61I/jlkSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsCriterion 31 - Reactivity Control Systems Malfunction (Category B) Thereactivity control systems shall be capable of sustaining any single malfunction,such as unplanned continuous withdrawal (not ejection) of a control rod, without causing a reactivity transient which could result in exceeding acceptable fuel damage limits.Conformance 31 - Reactivity Control Systems Malfunction (Category B)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 3.2Thermal and Hydraulic CharacteristicsSection 3.3Nuclear CharacteristicSection 3.5Reactivity Control Mechanical CharacteristicsSection 6.4Control Rod Velocity LimitersSection 6.6Standby Liquid Control SystemSection 7.2Reactor Control SystemsCriterion 32 - Maximum Reactivity Worth of Control Rods (Category A) Limits,which include considerable margin, shall be placed on the maximum reactivityworth of control rods or elements and on rates at which reactivity can beincreased to ensure that the potential effects of a sudden or large change ofreactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals sufficiently to impair the effectiveness of emergency core cooling.Conformance 32 - Maximum Reactivity Worth of Control Rods (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 3.3.3.3Nuclear Characteristic - Reactivity ControlSection 3.3.3.4Nuclear Characteristic - Control Rod Worth Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 31 of 61I/jlkSection 3.4Fuel Mechanical CharacteristicsSection 3.5Reactivity Control Mechanical CharacteristicsSection 4 CompleteReactor Coolant SystemSection 6.4Control Rod Velocity LimitersSection 6.5Control Rod Drive Housing SupportsSection 7.8NUMAC Rod Worth Minimizer and Plant ProcessComputerSection 14.1.5Summary Description - Design Basis for AccidentsE.2.6Group VI - Reactor Coolant Pressure BoundaryThe intent of the current draft of the proposed criteria for this group is to establish the reactor coolant pressure boundary design requirements and toidentify the means used to satisfy these design requirements.It is concluded that the design of this plant is in conformance with the criteria ofGroup VI based on NSP's current understanding of the intent of these criteria.The inherent safety features of the reactor core design in combination withcertain engineered safety features (control rod velocity limiters and control rod housing supports, etc.) and the plant operational reactivity control system aresuch that the consequences of the most severe potential nuclear excursionaccident, caused by a single component failure within the reactivity control system (control rod drop accident) cannot result in damage (either by motion or rupture) to the reactor coolant system. (Criterion 33) The ASME and USASI Codes are used as the established and acceptable criteria for design,fabrication, and operation of components of the reactor primary pressuresystem. The reactor primary system is designed and fabricated to meet the following as a minimum: (Criterion 34)(1)Reactor Vessel - ASME Boiler and Pressure Vessel Code, SectionIII, Nuclear Vessels, Subsection A(2)Pumps - ASME Boiler and Pressure Vessel Code,Section III,Nuclear Vessels, Subsection C(3)Piping and Valves - USASI-B-31.1, Code for Pressure, Power PipingProtection against the brittle fracture or other failure modes of the reactor coolant pressure boundary system components is provided for all potential service Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 32 of 61I/jlkloading temperatures. Control is exercised in the selection of materials andfabrication and design of equipment and components. It is intended that NDTtesting be performed on all ferritic materials in the reactor coolant pressureboundary with appropriate modifications for material thickness of individual components. (Criterion 35)The reactor coolant system will be given a final hydrostatic test at 1560 psig inaccordance with Code requirements prior to initial reactor startup. A hydrostatic test, not to exceed system operating pressure, will be made on the reactor coolant system following each removal and replacement of the reactor vesselhead. The reactor primary system will be checked for leaks and abnormalconditions will be corrected before reactor startup. The minimum vessel temperature during hydrostatic test shall at least be 60° F above the calculated NDT temperature prior to pressurizing the vessel. Extensive quality control assurance programs are being so followed during the entire fabrication of thereactor coolant system. (Criterion 36) Vessel material surveillance samples arelocated within the reactor primary vessel to enable periodic monitoring of material properties with exposure. The program will include specimens of the base metal, heat affected zone metal, and standards specimens. Leakage from the reactor coolant system is monitored during reactor operation. (Criterion 36)References to applicable sections of the USAR are given on the following pagefor the individual criteria of this group.Criterion 33 - Reactor Coolant Pressure Boundary Capability (Category A)The reactor coolant pressure boundary shall be capable of accommodatingwithout rupture and with only limited allowance for energy absorption through plastic deformation, the static and dynamic loads imposed on any boundary component as a result of any inadvertent and sudden release of energy to thecoolant. As a design reference, this sudden release shall be taken as that whichwould result from a sudden reactivity insertion such as rod ejection (unless prevented by positive mechanical means), rod dropout, or cold water addition.Conformance 33 - Reactor Coolant Pressure Boundary Capability (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 3.3.3.3Nuclear Characteristic - Reactivity ControlSection 3.3.3.4Nuclear Characteristic - Control Rod WorthSection 3.4Fuel Mechanical CharacteristicsSection 3.5Reactivity Control Mechanical Characteristics Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 33 of 61I/jlkSection 4 CompleteReactor Coolant SystemSection 6.4Control Rod Velocity LimitersSection 6.5Control Rod Drive Housing SupportsSection 7.8NUMAC Rod Worth Minimizer and Plant ProcessComputerSection 14.1.5Summary Description - Design Basis for AccidentsCriterion 34 - Reactor Coolant Pressure Boundary Rapid Propagation FailurePrevent (Category A)The reactor coolant pressure boundary shall be designed to minimize the probability of rapidly propagating type failures. Consideration shall be given (a) to the notch-toughness properties if materials extending to the upper shelf of the Charpy transition curve, (b) to the state of stress of materials under static and transient loading, (c) to the quality control specified for materials and componentfabrication to limit flaw sizes, and (d) to the provisions for control over servicetemperature and irradiation effects which may require operational restrictions.Conformance 34 - Reactor Coolant Pressure Boundary Rapid PropagationFailure Prevention (Category A)Section Appendix CQuality Assurance ProgramSection 4 CompleteReactor Coolant SystemCriteria 35 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention(Category A)Under conditions where reactor coolant pressure boundary system componentsconstructed of Ferritic materials may be subjected to potential loadings, such as a reactivity-induced loading, service temperatures shall be at least 120° F above the nil ductility transition (NDT) temperature of the component material if theresulting energy is expected to be absorbed within the elastic strain energyrange.Conformance 35 - Reactor Coolant Pressure Boundary Brittle FracturePrevention (Category A)Section 4.2.3Reactor Vessel - Design EvaluationSection 4.3.1Recirculation System - Design CriteriaSection 4.3.3Recirculation System - Performance Evaluation Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 34 of 61I/jlkSection 4.4.3Reactor Pressure Relief System - PerformanceAnalysisCriteria 36 - Reactor Coolant Pressure Boundary Surveillance (Category A)Reactor coolant pressure boundary components shall have provisions for inspection, testing, and surveillance by appropriate means to assess the structural and leak tight integrity of the boundary components during their service lifetime. For the reactor vessel, a material surveillance programconforming with ASTM-E-185-66 shall be provided.Conformance 36 - Reactor Coolant Pressure Boundary Surveillance (CategoryA)Section 4.2.1Reactor Vessel - Design BasisSection 4.3.1Recirculation System - Design BasisSection 4.3.4Recirculation System - Inspection and TestingSection 4.4.4Reactor Pressure Relief System - Inspectionand TestingE.2.7Group VII - Engineered Safety FeaturesThe intent of the current draft of the proposed criteria for this group is (a) to identify the engineered safety features (ESF), (b) to examine each ESF for independency, redundancy, capability, testability, inspectability, and reliability, (c) to determine the suitability of each ESF for its intended duty, and (d) justify that each ESFs capability-scope envelopes all the anticipated and crediblephenomena associated with the plant operational transients or design basisaccidents being considered.It is concluded that the design of the plant is in conformance with the criteria ofGroup VII based on NSP's current understanding of the intent of these criteria.The normal plant control systems maintain plant variables within narrowoperating limits. These systems are thoroughly engineered and backed up asignificant amount of experience in system design and operation. Even if an improbable maloperation or equipment failure including a reactor coolant boundary break up to and including the circumferential rupture of any pipe in that boundary assuming an unobstructed discharge from both sides allows variablesto exceed their operating limits, an extensive system of engineered safetyfeatures (ESF) limit the transient and the effects to levels well below those which are of public safety concern.These engineered safety features (ESF) include the normal protection systems(reactor core, reactor coolant system, plant containment system, plant and Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 35 of 61I/jlkreactor control systems, reactor protection system, other instrumentation andprocess systems, etc.); those which offer additional protection against areactivity excursion (reactor standby liquid control system, control rod velocitylimiters, and control rod housing support, etc.); those which act to reduce the consequences of design basis accidents (main steam line flow restrictors, etc.);
and those which provide emergency core and standby containment cooling in the event of a loss of normal cooling (emergency core cooling systems (ECCS),residual heat removal system (RHRS), high pressure coolant injection system(HPCIS), automatic depressurization system (ADS), and the standby coolant supply system). (Criterion 37)The engineered safety features are designed to provide high reliability and readytestability. Specific provisions are made in each ESF to demonstrate operability and performance capabilities. (Criterion 38) Components of the ESF which are required to function after design basis accidents or incidents are designed to withstand the most severe forces and credible environmental effects, includingmissiles from plant equipment failures anticipated from the events, withoutimpairment of their performance capability. (Criterion 40, 42, 43)Sufficient off-site and redundant, independent and testable standby auxiliarysources of electrical power are provided to attain prompt shutdown and continued maintenance of the plant in a safe condition under all credible circumstances. The capacity of the power sources are adequate to accomplish all required engineered safety features functions under all postulated designbasis accident conditions (Criterion 39).The emergency core cooling systems (ECCS) are designed such that at leasttwo different ECCSs of different phenomena are provided to prevent clad meltover the entire spectrum of postulated breaks. Such capability is available evenwith the loss of all off-site AC power. The ECCS (individual systems) themselves are designed to various levels of component redundancy such that no single active component failure in addition to the accident will negate the necessary emergency core cooling capability (Criterion 41, 44). To further assure that theECCS will function properly, if needed, specific provisions have been made fortesting the sequential operability and functional performance of each individual system (Criterion 46, 47, 48). Design provisions have also been made to enable physical and visual inspection of the ECCS components (Criterion 45).The primary containment structure, including access openings and penetrations,is designed to withstand the peak transient pressure and temperatures which could occur due to the postulated design basis loss-of-coolant design accident.The containment design includes considerable allowance for energy additionfrom metal-water or other chemical reactions beyond conditions that would occur with normal operation of Emergency Core Cooling Systems (ECCS). The primary containment has a metal-water reaction capability approximately 55% (at 2 hr) which is 500 times the calculated metal water reaction for the design basisloss-of-coolant accident (Criterion 49). Plates, structural member, forgings, andpipe associated with the drywell have an initial NDT temperature of approximately 0°F when tested in accordance with the appropriate code for the Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 36 of 61I/jlkmaterials. It is intended that the drywell will not be pressurized or subjected tosubstantial stress at temperatures below 30° F. Provisions are made for theremoval of heat from within the plant containment system and to isolate thevarious process system lines as may be necessary to maintain the integrity of the plant containment systems as long as necessary following the various postulated design basis accidents. The plant containment is designed and maintained so that the off-site doses resulting from the postulated design basisaccident will be below the values stated in 10CFR 100 (Criterion 50, 51, 54). Allpipes or ducts, which penetrate the primary containment and which connect to the reactor coolant system or to the drywell, are provided with at least two isolation valves in series (Criterion 53). The plant design provides for preoperational pressure and leak rate testing of the primary containment system,and include the capability for leak testing at design pressure after the plant hascommenced operation (Criterion 54, 55). Provisions are also made for demonstrating the functional performance of the plant containment system isolation valves and leak testing of selected penetrations (Criterion 56, 57).The pressure suppression pool and the containment spray cooling systemprovide two different means to rapidly condense the steam portion of the flow from the postulated design basis loss-of-coolant accident so that the peak transient pressure shall be substantially less than the primary containmentdesign pressure (Criterion 52). Demonstration of operability and the ability totest the functional performance and inspect the containment spray/cooling system are provided (Criterion 58, 59, 60, 61). The secondary containment standby gas treatment system is designed such that means are provided for periodic testing of the system performance including tracer injection andsampling (Criterion 64). The system may be physically inspected and itsoperability demonstrated (Criterion 62, 63, 65).References to applicable sections of the USAR are given below for the individualcriteria of this group.Criterion 37 - Engineered Safety Features Basis for Design (Category A)Engineered safety features shall be provided in the facility to back up the safetyprovided by the core design, the reactor coolant pressure boundary, and their protection systems. As a minimum, such engineered safety features shall be designed to cope with any size reactor pressure boundary break up to and including the circumferential rupture of any pipe in that boundary assumingunobstructed discharge from both ends.Conformance 37 - Engineered Safety Features Basis for Design (Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand Control Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 37 of 61I/jlkSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.3Summary Design Description and SafetyAnalysis - Plant Containment SystemSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby CoolingSystemsSection 1.3.5Summary Design Description and SafetyAnalysis - Plant Instrumentation Control SystemsSection 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power SystemsSection 5 CompleteContainment SystemSection 6 CompletePlant Engineered SafeguardsSection 7 CompletePlant Instrumentation and Control SystemsSection 8 CompletePlant Electrical SystemsSection 10.3.8Plant Service Systems - Plant Communication SystemSection 10.3.9Plant Service Systems - Plant Lighting SystemSection 14.1.5Summary Description - Design Basis for AccidentsCriterion 38 - Reliability and Testability of Engineered Safety Features(Category A)All engineered safety features shall be designed to provide high functional reliability and ready testability. In determining the suitability of a facility for a proposed site, the degree of reliance upon and acceptance of the inherent and engineered safety afforded by the systems, including engineering safety features, will be influenced by the known and the demonstrated performancecapability and reliability of the systems, and by the extent to which the operabilityof such systems can be tested and inspected where appropriate during the life of the plant.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 38 of 61I/jlkConformance 38 - Reliability and Testability of Engineered Safety Features(Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.3Summary Design Description and SafetyAnalysis - Plant Containment SystemSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemsSection 5 CompleteContainment SystemSection 6 CompletePlant Engineered SafeguardsSection 7 CompletePlant Instrumentation and Control SystemsSection 8 CompletePlant Electrical SystemsSection 10.3.8Plant Service Systems - Plant Communication SystemSection 10.3.9Plant Service Systems - Plant Lighting SystemCriterion 39 - Emergency Power for Engineered Safety Features (Category A)Alternate power systems shall be provided and designed with adequate independency, redundancy, capacity, and testability to permit the functioningrequired of the engineered safety features. As a minimum, the on-site powersystem and the off-site power system shall each, independently, provide this capacity assuming a failure of a single active component in each power system.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 39 of 61I/jlkConformance 39 - Emergency Power for Engineered Safety Features(Category A)Section 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power SystemsSection 8.2Transmission SystemSection 8.3Auxiliary Power SystemSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesCriterion 40 - Missile Protection (Category A)Protection for engineered safety features shall be provided against dynamiceffects and missiles that might result from the plant equipment failures.Conformance 40 - Missile Protection (Category A)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.3.5Secondary Containment System - PerformanceAnalysisSection 12 CompletePlant Structures and ShieldingCriterion 41 - Engineered Safety Features Performance Capability (Category A)Engineered safety features such as emergency core cooling and containment heat removal systems shall provide sufficient performance capability to accommodate partial loss of installed capacity and still fulfill the required safety function. As a minimum, each engineered safety feature shall provide this required safety function assuming a failure of a single active component.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 40 of 61I/jlkConformance 41 - Engineered Safety Features Performance Capability(Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby CoolingSystemsSection 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power SystemsSection 5.2.1Primary Containment System - Design CriteriaSection 5.3.2Secondary Containment System - Design BasisSection 6.2.1.1Emergency Core Cooling System (ECCS) -
ECCS Design BasisSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.5.3Automatic Depressurization System (ADS) -Performance AnalysisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -
Performance AnalysisSection 6.2.3.3Residual Heat Removal System (RHR) -
Performance AnalysisSection 6.2.6Emergency Core Cooling System (ECCS) -
ECCS Performance EvaluationSection 6.3Main Steam Line Flow RestrictionsSection 6.4Control Rod Velocity Limiters Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 41 of 61I/jlkSection 6.5Control Rod Drive Housing SupportsSection 6.6Standby Liquid Control SystemSection 8.2Transmission SystemSection 8.3Auxiliary Power SystemsSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.3.4Plant Service Systems - Plant Instrumentationand Service Air SystemsSection 10.3.8Plant Service Systems - Plant Communication SystemSection 10.3.9Plant Service Systems - Plant Lighting SystemSection 14.1.5Summary Description - Design Basis for AccidentsCriterion 42 - Engineered Safety Features Components Capability (Category A)Engineered safety features shall be designed so that the capability of each component and system to perform its required function is not impaired by the effects of a loss-of-coolant accident.Conformance 42 - Engineered Safety Features Components Capability(Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 3.6Other Reactor Vessel InternalsSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - Performance Analysis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 42 of 61I/jlkSection 6 CompletePlant Engineered SafeguardsSection 7.4Reactor Vessel InstrumentationSection 7.6Plant Protection SystemSection 12 CompletePlant Structures and ShieldingSection 14.1.5Summary Description - Design Basis AccidentAnalysisCriterion 43 - Accident Aggravation Prevention (Category A)Engineered safety features shall be designed so that any action of the engineered safety features which might accentuate the adverse affects of the loss of normal cooling avoided.Conformance 43 - Accident Aggravation Prevention (Category A)Section 5.2.3Primary Containment System - PerformanceAnalysisSection 6.2.1.1Emergency Core Cooling System (ECCS) -
ECCS Design BasisSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.5.3Automatic Depressurization System (ADS) -Performance AnalysisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -
Performance AnalysisSection 6.2.3.3Residual Heat Removal System (RHR) -Performance AnalysisSection 6.2.6Emergency Core Cooling System (ECCS) -ECCS Performance EvaluationSection 6.3.1Main Steam Line Flow Restrictions - Design BasisSection 6.4.1Control Rod Velocity Limiters - Design BasisSection 6.5.1Control Rod Drive Housing Supports - Design Basis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 43 of 61I/jlkSection 6.6.1Standby Liquid Control System - Design BasisCriterion 44 - Emergency Core Cooling System Capability (Category A)At least two emergency core cooling systems, preferably of different designprinciples, each with a capability for accomplishing abundant emergency core cooling, shall be provided. Each emergency core cooling system and the core shall be designed to prevent fuel and clad damage that would interfere with the emergency core cooling function and to limit the clad metal-water reaction tonegligible amounts of all sizes of breaks in the reactor coolant pressureboundary, including the double-ended rupture of the largest pipe. The performance of each emergency core cooling system shall be evaluated conservatively in each area of uncertainty. The systems shallnot share active components and shallnot share other features or components unless it can bedemonstrated that (a) the capability of the shared feature or components toperform its required function can be readily ascertained during reactor operation, (b) failure of the shared feature or component does not initiate a loss-of-coolant accident, and (c) capability of the shared feature or component to perform its required function is not impaired by the effects of a loss-of-coolant accident andis not lost during the entire period this function is required following the accident.Conformance 44 - Emergency Core Cooling Systems Capability (Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby Cooling SystemsSection 6.2.1.2Emergency Core Cooling System (ECCS) -
Description and Function of ECCSSection 6.2.2.1Reactor Core Spray Cooling System (CSCS) -
Design BasisSection 6.2.3.1Residual Heat Removal System (RHR) -
Design BasisSection 6.2.4.1High Pressure Coolant Injection System (HPCI) -
Design BasisSection 6.2.5.1Automatic Depressurization System (ADS) -
Design BasisSection 6.2.6Emergency Core Cooling System (ECCS) -
ECCS Performance Evaluation Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 44 of 61I/jlkSection 14.1.5Summary Description - Design Basis forAccidentsCriterion 45 - Inspection of Emergency Core Cooling Systems (Category A)Design provisions shall be made to facilitate physical inspection of all critical parts of the emergency core cooling systems, including reactor vessel internalsand water injection nozzles.Conformance 45 - Inspection of Emergency Core Cooling Systems (Category A)Section 3.6.1Other Reactor Vessel Internals - Design BasisSection 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and TestingSection 6.2.3.4Residual Heat Removal System (RHR) -
Inspection and TestingSection 6.2.4.4High Pressure Coolant Injection System (HPCI) -
Inspection and TestingSection 6.2.5.4Automatic Depressurization System (ADS) -
Inspection and TestingCriterion 46 - Testing of Emergency Core Cooling Systems Components(Category A)Design provisions shall be made so that active components of the emergencycore cooling systems, such as pumps and valves, can be tested periodically foroperability and require functional performance.Conformance 46 - Testing of Emergency Core Cooling Systems Components(Category A)Section 6.2.1.1Emergency Core Cooling System (ECCS) -ECCS Design BasisSection 6.2.2.1Reactor Core Spray Cooling System (CSCS) -
Design BasisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -Performance AnalysisSection 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and TestingSection 6.2.4.1 High Pressure Coolant Injection System (HPCI)-
Design Basis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 45 of 61I/jlkSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -Performance AnalysisSection 6.2.4.4High Pressure Coolant Injection System (HPCI) -
Inspection and TestingSection 6.2.3.1Residual Heat Removal System (RHR) -
Design BasisSection 6.2.3.3Residual Heat Removal System (RHR) -Performance AnalysisSection 6.2.3.4Residual Heat Removal System (RHR) -Inspection and TestingSection 6.2.5.1Automatic Depressurization System (ADS) -
Design BasisSection 6.2.5.3Automatic Depressurization System (ADS) -
Performance AnalysisSection 6.2.5.4Automatic Depressurization System (ADS) -Inspection and TestingCriterion 47 - Testing of Emergency Core Cooling Systems (Category A)A capability shall be provided to test periodically the delivery capability of theemergency core cooling systems at a location as close to the core as is practical.Conformance 47 - Testing of Emergency Core Cooling Systems (Category A)Section 6.2.1.1Emergency Core Cooling System (ECCS) -ECCS Design BasisSection 6.2.2.1Reactor Core Spray Cooling System (CSCS) -
Design BasisSection 6.2.2.3Reactor Core Spray Cooling System (CSCS) -Performance AnalysisSection 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and TestingSection 6.2.4.1 High Pressure Coolant Injection System (HPCI)-
Design BasisSection 6.2.4.3High Pressure Coolant Injection System (HPCI) -
Performance Analysis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 46 of 61I/jlkSection 6.2.4.4High Pressure Coolant Injection System (HPCI) -Inspection and TestingSection 6.2.3.1Residual Heat Removal System (RHR) -
Design BasisSection 6.2.3.3Residual Heat Removal System (RHR) -
Performance AnalysisSection 6.2.3.4Residual Heat Removal System (RHR) -Inspection and TestingSection 6.2.5.1Automatic Depressurization System (ADS) -Design BasisSection 6.2.5.3Automatic Depressurization System (ADS) -
Performance AnalysisSection 6.2.5.4Automatic Depressurization System (ADS) -
Inspection and TestingCriterion 48 - Testing of Operational Sequence of Emergency Core CoolingSystem (Category A)A capability shall be provided to test under conditions as close to design aspractical the full operational sequence that would bring the emergency corecooling systems into action, including the transfer to alternate power sources.Conformance 48 - Testing of Operational Sequence of Emergency Core CoolingSystem (Category A)Section 6.2Emergency Core Cooling System (ECCS)Section 8 CompletePlant Electrical SystemsSection 8.2Transmission SystemSection 8.3Auxiliary Power SystemSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.4Plant Cooling System Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 47 of 61I/jlkCriterion 49 - Containment Design Basis (Category A)The containment structure, including access openings and penetrations, and anynecessary containment heat removal systems shall be designed so that the containment structure can accommodate without exceeding the design leakage rate the pressures and temperatures resulting from the largest credible energyrelease following a loss-of-coolant accident, including a considerable margin foreffects from metal-water or other chemical reactions that could occur as a consequence of failure of emergency core cooling systems.Conformance 49 - Containment Design Basis (Category A)Section 1.2.2Principal Design Criteria - Reactor CoreSection 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling SystemsSection 1.3Summary Design Description and Safety AnalysisSection 5.1Containment System - Summary DescriptionSection 5.2.3Primary Containment System - Performance AnalysisSection 5.2.4Primary Containment System - Inspection and TestingSection 5.3.2Secondary Containment System - Design BasisSection 5.3.5Secondary Containment System - Performance AnalysisSection 5.3.6Secondary Containment System - Inspection and TestingSection 6.2Emergency Core Cooling System (ECCS)Section 6.6Standby Liquid Control System Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 48 of 61I/jlkSection 10.2.5Reactor Auxiliary Systems - Reactor CoreIsolation Cooling System (RCIC)Section 14.1.5Summary Description - Design Basis forAccident AnalysisCriterion 50 - NDT Requirement for Containment Material (Category A)Principal load carrying components of ferritic materials exposed to the externalenvironment shall be selected so that their temperatures under normal operating and testing conditions are not less than 30° F above nil ductility transition (NDT) temperature.Conformance 50 - NDT Requirement for Containment Material (Category A)Section 5.2.2.2 - Primary Containment Construction Materials Criterion 51 - Reactor Coolant Pressure Boundary Outside Containment(Category A)If part of the reactor coolant pressure boundary is outside the containment,appropriate features as necessary shall be provided to protect the health and safety of the public in case of an accidental rupture in that part. Determination ofthe appropriateness of features such as isolation valves and additionalcontainment shall include consideration of the environmental and population conditions surrounding the site.Conformance 51 - Reactor Coolant Pressure Boundary Outside Containment(Category A)Section 1.2.1Principal Design Criteria - General CriteriaSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.5Principal Design Criteria - Plant Instrumentationand ControlSection 1.2.6Principal Design Criteria - Plant Electrical PowerSection 1.3.2Summary Design Description and Safety Analysis - Reactor SystemSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.5Summary Design Description and SafetyAnalysis - Plant Instrumentation Control SystemsSection 1.3.8Summary Design Description and SafetyAnalysis - Plant Electrical Power System Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 49 of 61I/jlkSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 2.2Site DescriptionSection 5.2Primary Containment SystemSection 5.3Secondary Containment SystemSection 6.3Main Steam Line Floor RestrictionsSection 7.5.2Plant Radiation Monitoring Systems - Process Radiation Monitoring SystemSection 7.6.3Plant Protection System - Primary Containment Isolation SystemSection 14.1.5Summary Description - Design Basis for Accident AnalysisCriterion 52 - Containment Heat Removal Systems (Category A)Where active heat removal systems are needed under accident conditions toprevent exceeding containment design pressure, at least two systems,preferably of different principles, each with full capacity, shall be provided.Conformance 52 - Containment Heat Removal Systems (Category A)Section 1.2.3Principal Design Criteria - Reactor Core CoolingSection 1.2.4Principal Design Criteria - Plant ContainmentSection 1.3.2Summary Design Description and SafetyAnalysis - Reactor SystemSection 1.3.3Summary Design Description and Safety Analysis - Plant Containment SystemSection 1.3.4Summary Design Description and SafetyAnalysis - Plant Auxiliary and Standby CoolingSystemsSection 5.2Primary Containment SystemSection 6.2Emergency Core Cooling System (ECCS)Section 10.2Reactor Auxiliary Systems Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 50 of 61I/jlkSection 10.4Plant Cooling SystemSection 14.1.5Summary Description - Design Basis forAccident AnalysisCriterion 53 - Containment Isolation Valves (Category A)Penetrations that require closure for the containment function shall be protectedby redundant valving and associated apparatus.Conformance 53 - Containment Isolation Valves (Category A)Section 5.2.1.3Primary Containment System - ContainmentPenetrationsSection 5.2.2.5.3Primary Containment System - Isolation SystemSection 5.2.3.7Primary Containment System - PenetrationsSection 5.2.3.6.2Primary Containment System - Isolation SystemSection 5.2.4Primary Containment System - Inspection and TestingSection 7.6.3Plant Protection System - Primary Containment Isolation SystemCriterion 54 - Containment Leakage Rate Testing (Category A)Containment shall be designed so that an integrated leakage rate testing can be conducted at design pressure after completion and installation of all penetrations and leakage rate measured over a sufficient period of time to verify its conformance with required performance.Conformance 54 - Containment Leakage Rate Testing (Category A)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.2.4Primary Containment System - Inspection and TestingSection 5.3.2Secondary Containment System - Design Basis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 51 of 61I/jlkSection 5.3.5Secondary Containment System - PerformanceAnalysisSection 5.3.6Secondary Containment System - Inspection and TestingCriterion 55 - Containment Periodic Leakage Rate Testing (Category A)The containment shall be designed so that integrated leakage rate testing can be done periodically at design pressure during plant lifetime.Conformance 55 - Containment Periodic Leakage Rate Testing (Category A)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.3.2Secondary Containment System - Design BasisCriterion 56 - Provisions for Testing of Penetrations (Category A)Provisions shall be made for testing penetrations which have resilient seals or expansion bellows to permit leak tightness to be demonstrated at design pressure at anytime.Conformance 56 - Provisions for Testing of Penetrations (Category A)Section 5.2.1Primary Containment System - Design CriteriaSection 5.2.3Primary Containment System - PerformanceAnalysisSection 5.2.4Primary Containment System - Inspection andTestingSection 5.3.5Secondary Containment System - PerformanceAnalysisSection 5.3.6Secondary Containment System - Inspection and Testing Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 52 of 61I/jlkCriteria 57 - Provisions for Testing of Isolation Valves (Category A)Capability shall be provided for testing functional operability of valves andassociated apparatus essential to the containment function for establishing thatno failure has occurred and for determining that valve leakage does not exceedacceptable limits.Conformance 57 - Provisions for Testing of Isolation Valves (Category A)Section 7.6.3.1Plant Protection System - Design BasisSection 7.6.3.3Plant Protection System - Performance AnalysisSection 7.6.3.4Plant Protection System - Inspection and TestingSection 7.5.2Plant Radiation Monitoring Systems - ProcessRadiation Monitoring SystemCriterion 58 - Inspection of Containment Pressure-Reducing System(Category A)Design provisions shall be made to facilitate the periodic physical inspection ofall important components of the containment pressure-reducing systems, suchas, pumps, valves, spray nozzles, torus, and sumps.Conformance 58 - Inspection of Containment Pressure-Reducing System(Category A)Section 5.2.4Primary Containment System - Inspection andTestingSection 6.2Emergency Core Cooling System (ECCS)Criterion 59 - Testing of Containment Pressure-Reducing Systems Components(Category A)The containment pressure-reducing systems shall be designed so that active components such as pumps and valves can be tested periodically for operability and required functional performance.Conformance 59 - Testing of Containment Pressure-Reducing SystemsComponents (Category A)Section 6.2.1.1Emergency Core Cooling System (ECCS) -Design BasisSection 6.2Emergency Core Cooling System (ECCS)
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 53 of 61I/jlkCriterion 60 - Testing of Containment Spray Systems (Category A)A capability shall be provided to test periodically the delivery capability of thecontainment spray system at a position as close to the spray nozzle as ispractical.Conformance 60 - Testing of Containment Spray Systems (Category A)Section 6.2.1.1Emergency Core Cooling System (ECCS) -Design BasisSection 6.2Emergency Core Cooling System (ECCS)Criterion 61 - Testing of Operational Sequence of ContainmentPressure-Reducing Systems (Category A)A capability shall be provided to test under conditions as close to the design as practical the full operational sequence that would bring the containmentpressure-reducing systems into action, including the transfer to alternate powersources.Conformance 61 - Testing of Operational Sequence of ContainmentPressure-Reducing Systems (Category A)Section 5.2 CompletePrimary Containment SystemSection 7.6.3.3Plant Protection System - Performance AnalysisSection 7.6.3.4Plant Protection System - Inspection and TestingSection 6.2.1.1Emergency Core Cooling System (ECCS) -Design BasisSection 6.2Emergency Core Cooling System (ECCS)Section 8 CompletePlant Electrical SystemsCriterion 62 - Inspection of Air Cleanup Systems (Category A)Design provisions shall be made to facilitate physical inspection of all critical parts of containment air cleanup systems such as ducts, filters, fans, and dampers.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 54 of 61I/jlkConformance 62 - Inspection of Air Cleanup Systems (Category A)Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 5.3.5Secondary Containment System - Performance AnalysisSection 5.3.6Secondary Containment System - Inspection and TestingSection 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsCriterion 63 - Testing of Air Cleanup Components (Category A)Design provisions shall be made so that active components of the air cleanup systems, such as fans, dampers, can be tested periodically for operability and required functional performance.Conformance 63 - Testing of Air Cleanup Components (Category A)Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 5.3.5Secondary Containment System - PerformanceAnalysisSection 5.3.6Secondary Containment System - Inspection and TestingSection 10.3.2Plant Service Systems - Plant Heating,Ventilating and Air Conditioning SystemsCriterion 64 - Testing of Air Cleanup Systems (Category A)A capability shall be provided for insitu periodic testing and surveillance of the aircleanup systems to ensure (a) filter bypass paths have not developed and (b) filter and trapping materials have not deteriorated beyond acceptable limits.Conformance 64 - Testing of Air Cleanup Systems (Category A)Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 5.3.5Secondary Containment System - PerformanceAnalysis Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 55 of 61I/jlkSection 5.3.6Secondary Containment System - Inspection andTestingSection 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsCriterion 65 - Testing of Operational Sequence Air Cleanup Systems(Category A)A capability shall be provided to test under conditions close to design as practical the full operational sequence that would bring the air cleanup systemsto action, including the transfer to alternate power sources and the design airflow delivery capability.Conformance 65 - Testing of Operational Sequence Air Cleanup Systems(Category A)Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)Section 5.3.5Secondary Containment System - Performance AnalysisSection 5.3.6Secondary Containment System - Inspection andTestingSection 7.5.2Plant Radiation Monitoring Systems - ProcessRadiation Monitoring SystemSection 7.6.1Plant Protection System - Reactor Protection SystemSection 8.4Plant Standby Diesel Generator SystemsSection 8.5D-C Power Supply SystemsSection 8.6Reactor Protection System Power SuppliesSection 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsE.2.8Group VIII - Fuel and Waste Storage SystemsThe intent of the current draft of the proposed criteria for this group is to establish the safe fuel and waste storage systems design and to identify the means used to satisfy these requirements.It is concluded that the design of this plant is in conformance with criteria ofGroup VIII based on NSP's current understanding of the intent of these criteria.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 56 of 61I/jlkAppropriate plant fuel handling and storage facilities are provided to precludeaccidental criticality and to provide sufficient cooling for spent fuel. (Criterion 66,67) The new fuel storage vault racks (located inside the secondary containmentreactor building) are top entry, and are designed to prevent an accidental critical array, even in the event the vault becomes flooded. Vault drainage is provided to prevent possible water collection. (Criterion 66) The handling and storage of spent fuel, which takes place entirely within the reactor building (which providescontainment), is done in the spent fuel storage pool. The pool has provisions tomaintain water clarity, temperature control, and instrumentation to monitor water level. Water depth in the pool will be such as to provide sufficient shielding for normal reactor building occupancy (10 CFR 20) by operating personnel. The storage racks in which spent fuel assemblies are placed are designed andarranged to ensure subcriticality in the storage pool. (Criterion 66, 67, 68, 69)The spent fuel pool cooling and demineralizer system is designed to maintain the pool water temperature (decay heat removal) to control water clarity (safe fuel movement), and to reduce water radioactivity (shielding and effluent release control). (Criterion 66, 67, 68) Accessible portions of the reactor and radwastebuildings shall have sufficient shielding to maintain dose rates within 10 CFR 20.(Criterion 68) The radwaste building is designed to preclude accidental release of radioactive materials to the environs. (Criterion 69) The spent fuel storage pool and racks are designed and constructed such that all credible missiles as a result of a design basis tornado and tornado itself, will not have radiologicaleffects exceeding 10 CFR 100 guideline limitations.References to applicable sections of the USAR are given below for the individualcriteria of this group. (Criterion 67, 69)Criterion 66 - Prevention of Fuel Storage Critically (Category B)Critically in new and spent storage shall be prevented by physical systems orprocesses. Such means as geometrically safe configurations shall beemphasized over procedural controls.Conformance 66 - Prevention of Fuel Storage Critically (Category B)Section 1.2.9Principal Design Criteria - Plant Fuel Handlingand StorageSection 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and Handling SystemsSection 6.6.3Standby Liquid Control System - Performance AnalysisSection 10.2.1.1Reactor Auxiliary Systems - Design BasisSection 10.2.1.2Reactor Auxiliary Systems - Description01081199 Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 57 of 61I/jlkCriterion 67 - Fuel and Waste Storage Decay Heat (Category B)Reliable decay heat removal systems shall be designed to prevent damage tothe fuel in storage facilities that could result in radioactivity release to plantoperating areas or the public environs.Conformance 67 - Fuel and Waste Storage Decay Heat (Category B)Section 1.2.7Principal Design Criteria - Plant RadioactiveWaste DisposalSection 1.2.9Principal Design Criteria - Plant Fuel Handling and StorageSection 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling systemSection 1.3Summary Design Description and Safety AnalysisSection 6.2.1.2Emergency Core Cooling System (ECCS) -
Description and Function of ECCSSection 10.2.1Reactor Auxiliary Systems - Fuel Storage andFuel Handling SystemsSection 10.2.2Reactor Auxiliary Systems - Spent Fuel PoolCooling and Demineralizer SystemSection 10.2.3Reactor Auxiliary Systems - Reactor Cleanup Demineralizer SystemSection 10.2.4Reactor Auxiliary Systems - Reactor Shutdown Cooling SystemSection 12 CompletePlant Structures and ShieldingCriterion 68 - Fuel and Waste Storage Radiation Shielding (Category B)Shielding for radiation protection shall be provided in the design of spent fuel andwaste storage facilities as required to meet requirements of 10 CFR 20.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 58 of 61I/jlkConformance 68 - Fuel and Waste Storage Radiation Shielding (Category B)Section 1.2.8Principal Design Criteria - Plant Shielding andAccess ControlSection 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and HandlingSystemsSection 1.3.9Summary Design Description and SafetyAnalysis - Plant Shielding, Access Control, and Radiation Protection ProceduresSection 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste Control SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant SafetySection 12.3Shielding And Radiation ProtectionSection 9.2.1Liquid Radwaste System - Design BasisSection 9.2.3Liquid Radwaste System - Performance AnalysisSection 9.3.1Gaseous Radwaste System - Design BasisSection 9.3.3Gaseous Radwaste System - Performance AnalysisSection 9.4.1Solid Radwaste System - Design BasisSection 9.4.3Solid Radwaste System - Performance AnalysisSection 10.2.1.1Reactor Auxiliary Systems - Design BasisSection 10.2.1.2Reactor Auxiliary Systems - DescriptionSection 10.2.1.3Reactor Auxiliary Systems - PerformanceAnalysisCriterion 69 - Protection Against Radioactivity Release from Spent Fuel andWaste Storage (Category B)Containment of fuel and waste storage shall be provided if accidents could leadto release of undue amounts of radioactivity to the public environs.
Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 59 of 61I/jlkConformance 69 - Protection Against Radioactivity Release from Spent Fuel andWaste Storage (Category B)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.8Principal Design Criteria - Plant Shielding andAccess ControlSection 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and Handling SystemsSection 1.3.9Summary Design Description and SafetyAnalysis - Plant Shielding, Access Control, andRadiation Protection ProceduresSection 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste Control SystemsSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 5.1Containment System - Summary DescriptionSection 5.3Secondary Containment SystemSection 9 CompletePlant Radioactive Waste Control SystemsSection 10.2.1Reactor Auxiliary Systems - Fuel Storage andFuel Handling SystemsSection 10.2.2Reactor Auxiliary Systems - Spent Fuel PoolCooling and Demineralizer SystemSection 1.2.7Principal Design Criteria - Plant RadioactiveWaste DisposalSection 1.2.8Principal Design Criteria - Plant Shielding and Access ControlSection 14.7.6.4.2Refueling Accident Analysis - RadiologicalConsequencesSection 14.7.4Accident Evaluation Methodology - Fuel LoadingError Accident Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 60 of 61I/jlkE.2.9Group IX - Plant EffluentsThe intent of the current draft of the proposed criterion for this group is toestablish the plant effluent release limits and to identify the means of controlling the releases within these guide limits.It is concluded that the design of this plant is in conformance with the criteria ofGroup IX based on NSP's current understanding of the intent of these criteria.The plant radioactive waste control systems (which include the liquid, gaseousand solid radwaste sub-systems) are designed to limit the off-site radiation exposure to levels below doses set forth in 10 CFR 20. The plant engineeredsafety systems (including the containment barriers) are designed to limit theoff-site dose under various postulated "design basis" accidents to levels significantly below the limits of 10 CFR 100. The air ejector off-gas system is designed with sufficient holdup retention capacity so that during normal plant operation the controlled release of radioactive materials does not exceed theestablished release limits at the elevated plant stack. (Criterion 70)References to applicable sections of the USAR are given for the individualcriteria of this group.Criterion 70 - Control of Release of Radioactivity to the Environment(Category B)The facility design shall include those means necessary to maintain control overthe plant radioactive effluents, whether gaseous, liquid, or solid. Appropriateholdup capacity shall be provided for retention of gaseous, liquid, or solideffluents, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10 CFR 20 requirements for normaloperations and for any transient situation that might reasonably be anticipated tooccur and (b) on the basis of 10 CFR 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where highpopulation densities or very large cities can be affected by the radioactiveeffluents.Conformance 70 - Control of Release of Radioactivity to the Environment(Category B)Section 1.2.4Principal Design Criteria - Plant ContainmentSection 1.2.7Principal Design Criteria - Plant RadioactiveWaste DisposalSection 1.2.8Principal Design Criteria - Plant Shielding and Access Control Revision 25USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 61 of 61I/jlkSection 1.3.9Summary Design Description and SafetyAnalysis - Plant Shielding, Access Control, andRadiation Protection ProceduresSection 1.3.10Summary Design Description and SafetyAnalysis - Plant Radioactive Waste Control SystemsSection 1.3.11Summary Design Description and SafetyAnalysis - Summary Evaluation of Plant SafetySection 2.2Site DescriptionSection 5 CompleteContainment SystemSection 12 CompletePlant Structures and ShieldingSection 7.5Plant Radiation Monitoring SystemsSection 8 CompletePlant Electrical SystemsSection 9 CompletePlant Radioactive Waste Control SystemsSection 10.3.6Plant Service Systems - Plant Equipment andFloor Drainage SystemsSection 10.3.7Plant Service Systems - Plant Process SamplingSystemSection 11.3.2Main Condenser System - Main Condenser GasRemoval SystemSection 13 CompletePlant OperationsSection 14 CompletePlant Safety Analysis