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Revision 33 to the Updated Final Safety Analysis Report, Appendix E, Plant Comparative Evaluation with the Proposed AEC 70 Design Criteria
	ML16054A446
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Site:	Monticello
Issue date:	01/26/2016
From:	; Northern States Power Co, Xcel Energy
To:	; Office of Nuclear Reactor Regulation
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ML16054A376	List: L-MT-16-004, Revision 33 to Updated Final Safety Analysis Report, Drawings NF-36058 Through NH-36243-1, Rev. 82 (Public); ML16054A388; ML16054A413; ML16054A414; ML16054A415; ML16054A416; ML16054A417; ML16054A418; ML16054A419; ML16054A420; ML16054A421; ML16054A422; ML16054A423; ML16054A424; ML16054A425; ML16054A426; ML16054A427; ML16054A428; ML16054A429; ML16054A432; ML16054A434; ML16054A435; ML16054A436; ML16054A438; ML16054A439; ML16054A440; ML16054A441; ML16054A442; ML16054A443; ML16054A444; ML16054A445; ML16054A446; ML16054A447; ML16054A448; ML16155A258; ML16155A259;
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Revision 22 USAR APPENDIX EMONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 1APPENDIX EPLANT COMPARATIVE EVALUATION WITH THE PROPOSED AEC 70 DESIGN CRITERIA I/mabTABLE OF CONTENTS Section Page E.1 Summary Description

1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

E.2Criterion - Conformance

1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

E.2.1Group I - Overall Plant Requirements

1. . . . . . . . . . . . . . . . . . . . . . .

E.2.2Group II - Protection by Multiple Fission Products Barriers11. . . . .

E.2.3Group III - Nuclear and Radiation Controls

15. . . . . . . . . . . . . . . . . .

E.2.4Group IV - Reliability and Testability of Protection Systems21. . . . E.2.5Group V - Reactivity Control

27. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

E.2.6Group VI - Reactor Coolant Pressure Boundary

31. . . . . . . . . . . . . .

E.2.7Group VII - Engineered Safety Features

34. . . . . . . . . . . . . . . . . . . . .

E.2.8Group VIII - Fuel and Waste Storage Systems

55. . . . . . . . . . . . . . .

E.2.9Group IX - Plant Effluents

60. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTP Assoc Ref:

SR:2yrs N Freq: USAR-MANARMS:USAR-E.TOCDoc Type:Admin Initials:Date:

9703 Revision 22 USAR E.1MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 1APPENDIX EPLANT COMPARATIVE EVALUATION WITH THE PROPOSED AEC 70 DESIGN CRITERIA I/mabE.1Summary Description This appendix contains a comparative evaluation of the design basis of the Monticello Nuclear 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 in the 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 to locations 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, the applicant believes that the Monticello Nuclear Generating Plant, Unit 1, is in

conformance with the intent of such proposed criteria.FOR ADMINISTRATIVE USE ONLYResp Supv:CNSTP Assoc Ref:

SR:2yrs N Freq: USAR-MANARMS:USAR-E.1Doc Type:Admin Initials:Date:

9703 Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 61APPENDIX EPLANT COMPARATIVE EVALUATION WITH THE PROPOSED AEC 70 DESIGN CRITERIA I/jlkE.2Criterion - ConformanceE.2.1Group I - Overall Plant Requirements The intent of the current draft of the proposed criteria for this group is to identify and 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 essential components and systems are also identified. The influence of this sharing of common 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 or system 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 sections are included in the reference list. In planning and executing these control and assurance 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 operation and 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 resistant materials through the plant. (Criterion 3) Records of design, fabrication, and construction 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 2 of 61 I/jlkReferences to applicable sections of the USAR are given below for the individual criteria of this group.

Criterion 1 - Quality Standards (Category A)

Those systems and components ofreactor facilities which are essential to prevention of accidents which could affect the public health and safety or to mitigation to their consequences shall be

identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized 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 assurance programs, 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.General Section 1.2.1Principal Design Criteria - General Criteria Section 1.3.1.3Summary Design Description and Safety Analysis - Geology Section 1.3.1.4Summary Design Description and Safety

Analysis - Hydrology Section 1.3.1.5Summary Design Description and Safety Analysis - Regional and Site Meteorology Section 1.3.1.6Summary Design Description and Safety Analysis - Seismology and Design Response Spectrum Section 1.3.1.7Summary Design Description and Safety Analysis - Site Environmental Monitoring

Program Section 1.3.4Summary Design Description and Safety

Analysis - Plant Auxiliary and Standby Cooling

Systems Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrumentation Control System Section 1.3.6Summary Design Description and Safety

Analysis - Plant Fuel Storage and Handling

Systems Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 3 of 61 I/jlk Section 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power Systems Section 1.3.9Summary Design Description and Safety

Analysis - Plant Shielding, Access Control, and

Radiation Protection Procedures Section 1.3.10Summary Design Description and Safety Analysis - Plant Radioactive Waste Control Systems Section Appendix CQuality Assurance Programb.Containment Barriers Section 1.2.4Principal Design Criteria - Plant Containment Section 1.3.3Summary Design Description and Safety

Analysis - Plant Containment SystemSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant Safety Fuel Section 1.3.2Summary Design Description and Safety

Analysis - Reactor System Section 3.4.4Fuel Mechanical Characteristics - Surveillance and Testing Fuel Cladding Section 3.2.3Thermal and Hydraulic Characteristics - Design Criteria and Safety Limits Section 3.4.1Fuel Mechanical Characteristics - Design Basis Section 3.4.2Fuel Mechanical Characteristics - Description of Fuel Assemblies Section 3.4.3Fuel Mechanical Characteristics - Design

Evaluation Section 3.4.4Fuel Mechanical Characteristics - Surveillance and Testing Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 4 of 61 I/jlkReactor Coolant System Section 4Reactor Coolant System Primary Containment System Section 5.2.1Primary Containment System - Design Criteria Section 5.2.2Primary Containment System - Description Section 5.2.3Primary Containment System - Performance Analysis Section 5.2.4Primary Containment System - Inspection and Testing Secondary Containment System Section 5.3.2Secondary Containment System - Design Basis Section 5.3.5Secondary Containment System - Performance

AnalysisStandby Gas Treatment System Section 5.3.4.1Secondary Containment System - Standby GasTreatment System (SGTS)

Section 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning Systems Plant Elevated Release Point Section 9.3Gaseous Radwaste Systemc.Plant Engineered Safeguards Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 6.1Plant Engineered Safeguards - Summary DescriptionControl Rod Velocity Limiters Section 6.4.3Control Rod Velocity Limiters - Performance Analysis Section 6.4.4Control Rod Velocity Limiters - Inspection and Testing Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 5 of 61 I/jlkControl Rod Drive Housing Supports Section 6.5.3Control Rod Drive Housing Supports -

Performance Analysis Section 6.5.4Control Rod Drive Housing Supports - Inspectionand Testing Reactor Standby Liquid Flow Control System Section 6.6.3Standby Liquid Control System - Performance Analysis Section 6.6.4Standby Liquid Control System - Inspection andTraining Main Steam Line Flow Restrictors Section 6.3.3Main Steam Line Flow Restrictions -

Performance Analysis Section 6.3.4Main Steam Line Flow Restrictions - Inspectionand Testing Emergency Core Cooling Systems (ECCS)

Section 6.2.4.3High Pressure Coolant Injection System (HPCI) -

Performance Analysis Section 6.2.5.3Automatic Depressurization System (ADS) -

Performance Analysis Section 6.2.2.3Reactor Core Spray Cooling System (CSCS) -

Performance Analysis Section 6.2.3.3Residual Heat Removal System (RHR) -

Performance Analysis Section 6.2.6Emergency Core Cooling System (ECCS) -

ECCS Performance Evaluation Plant Structures and Shielding Section 12.2Plant Principal Structures and Foundations Section 12.3Shielding and Radiation Protection Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 6 of 61 I/jlkCriterion 2 - Performance Standards (Category A)

Those systems and components of reactor facilities which are essential to prevention of accidentswhich could affect the public health and safety or to mitigation to their consequences 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.General Section 1.2.1Principal Design Criteria - General Criteria Section 1.3.1.3Summary Design Description and Safety Analysis - Geology Section 1.3.1.4Summary Design Description and Safety Analysis - Hydrology Section 1.3.1.5Summary Design Description and Safety

Analysis - Site and Regional Meteorology Section 1.3.1.6Summary Design Description and Safety Analysis - Seismology and Design Response Spectra Section 1.3.1.7Summary Design Description and Safety Analysis - Site Environmental Monitoring Program Section 1.3.8Summary Design Description and Safety

Analysis - Plant Electrical Power Systems Section 1.3.9Summary Design Description and Safety

Analysis - Plant Shielding, Access Control, and

Radiation Protection Procedures Section 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 7 of 61 I/jlk Section 2.3Meteorology Section 2.4Hydrology Section 2.5Geology and Soil Investigation Section 2.6Seismology Section 2.7Radiation Environmental Monitoring Program (REMP)Section 2.8Ecological and Biological Studiesb.Containment Barriers Section 1.3.3Summary Design Description and Safety

Analysis - Plant Containment System Fuel Cladding Section 1.3.6Summary Design Description and Safety Analysis - Plant Fuel Storage and Handling Systems Section 3.2.1Thermal and Hydraulic Characteristics - Design

Basis Section 3.2.3Thermal and Hydraulic Characteristics -Design

Criteria and Safety Limits Section 3.3.1Nuclear Characteristics - Design Basis Section 3.3.3Nuclear Characteristics - Nuclear Design

Characteristics Section 3.4.1Fuel Mechanical Characteristics - Design Basis Section 3.4.3Fuel Mechanical Characteristics - Design Evaluation Section 3.5.1Reactivity Control Mechanical Characteristics -

Design Basis Section 3.5.5Reactivity Control Mechanical Characteristics -

Operation and Performance Analysis Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 8 of 61 I/jlkReactor Coolant System Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 4 - CompleteReactor Coolant System Primary Containment System Section 5.2.1Primary Containment System - Design Criteria Section 5.2.4Primary Containment System - Inspection and Testing Section 12.2.1.1Plant Principal Structures and Foundations -

Safety Categories Section Appendix ADesign Bases - Seismic Design and Analysis Section 12.2.1.6Plant Principal Structures and Foundations -

Wind Loads Secondary Containment System Section 5.3.2Secondary Containment System - Design Basis Section 5.3.5Secondary Containment System - Performance

Analysis Section 12.2.1.1Plant Principal Structures and Foundations -

Safety Categories Section 12.2.1.6Plant Principal Structures and Foundations -

Wind Loads Section 12.2.1.7Plant Principal Structures and Foundations -

FloodingStandby Gas Treatment System Section 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 9 of 61 I/jlk Plant Elevated Release Point Section 9.3Gaseous Radwaste Systemc.Plant Engineered Safeguards Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling

Systems Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrumentation Control SystemControl Rod Velocity Limiters Section 6.4.1Control Rod Velocity Limiters - Design Basis Section 6.4.3Control Rod Velocity Limiters - Performance

Analysis Control Rod Drive Housing Supports Section 6.5.1Control Rod Drive Housing Supports - Design

Basis Section 6.5.3Control Rod Drive Housing Supports -

Performance Analysis Reactor Standby Liquid Flow Control System Section 6.6.1Standby Liquid Control System - Design Basis Section 6.6.3Standby Liquid Control System - Performance

Analysis Main Steam Line Flow Restrictors Section 6.3.1Main Steam Line Flow Restrictions - Design

Basis Section 6.3.3Main Steam Line Flow Restrictions -

Performance Analysis Emergency Core Cooling Systems (ECCS)

Section 6.2.1.1Emergency Core Cooling Systems (ECCS) -

ECCS Design Basis Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 10 of 61 I/jlk Section 6.2.4.3High Pressure Coolant Injection System (HPCI) -

Performance Analysis Section 6.2.5.3Automatic Depressurization System (ADS) -

Performance Analysis Section 6.2.2.3Reactor Core Spray Cooling System (CSCS) -

Performance Analysis Section 6.2.3.3Residual Heat Removal System (RHR) -

Performance Analysis Section 6.2.6Emergency Core Cooling Systems (ECCS) -

ECCS Performance Evaluation Plant Structures and Shielding Section 12.2Plant Principal Structures and Foundations Section 12.3Shielding and Radiation Protection Criterion 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 as containment, control room, and components of engineered safety features.

Conformance 3 - Fire Protection (Category A)

Section 1.2.1Principal Design Criteria - General Criteria Section 10.3.1Plant Service Systems - Fire Protection Systems Criterion 4 - Sharing of Systems (Category A) Reactor facilities shallnot share systems 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORTPage 11 of 61 I/jlkConformance 5 - Records Requirements (Category A)

Section Appendix CQuality Assurance Program Section 13.4Operational Procedures Section 13.5Operational Records and Reporting RequirementsE.2.2Group II - Protection by Multiple Fission Products Barriers The intent of the current draft of the proposed criteria for this group is to assure that 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 the design 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 of radioactive 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 UO 2 fuel, (b) the high integrity Zircaloy cladding, (c) the reactor 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 to accommodate without failure, the pressures and temperatures resulting from or subsequent 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 when the primary containment is open for refueling operations. The two containment systems 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 for sudden 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 provide margins to ensure that fuel damage will not occur in normal operation or operational transient caused by single reactor operator error or equipment

malfunction. (Criterion 6, 7) The reactor core is designed so that the overall01081199 Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 12 of 61 I/jlkpower coefficient in the power operating range is not positive. (Criterion 8) The reactor 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 the limits of applicable criteria for low probability accident conditions, to withstand the 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 individual criteria of this group.

Criterion 6 - Reactor Core Design (Category A)

The reactor core shall be designed to function throughout its design lifetime, without exceeding acceptable fuel 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 a turbine 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 Core Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 1.3.4Summary Design Description and Safety

Analysis - Plant Auxiliary and Standby Cooling

Systems Section 3.2Thermal and Hydraulic Characteristics Section 3.3Nuclear Characteristics Section 3.4Fuel Mechanical Characteristics Section 3.5Reactivity Control Mechanical Characteristics Section 4Reactor Coolant System Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 13 of 61 I/jlk Section 8.4Plant Standby Diesel Generator Systems Section 8.5D-C Power Supply Systems Section 8.6Reactor Protection System Power Supplies Section 10.2.5Reactor Auxiliary Systems - Reactor Core Isolation Cooling System (RCIC)

Section 14.4.3Transient Events Analyzed for Core Reload -

Rod Withdrawal Error Criterion 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 shall not be positive.Conformance 8 - Overall Power Coefficient (Category B)

Section 1.2.2Principal Design Criteria - Reactor Core Section 3.2Thermal and Hydraulic Characteristics Section 3.5Reactivity Control Mechanical Characteristics Criterion 9 - Reactor Coolant Pressure Boundary (Category A)

The reactor coolant pressure boundary shall be designed and constructed so as to have an exceedingly 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 Core Section 4 CompleteReactor Coolant System Section 7.4Reactor Vessel Instrumentation Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 14 of 61 I/jlkCriterion 10 - Containment (Category A)

Containment shall be provided. Thecontainment structure shall be designed to sustain the initial effects of gross equipment failures, such as a large coolant boundary area, without loss of required 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 Core Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 1.2.4Principal Design Criteria - Plant Containment Section 1.3.3Summary Design Description and Safety Analysis - Plant Containment System Section 1.3.4Summary Design Description and Safety

Analysis - Plant Auxiliary and Standby Cooling Systems Section 4 CompleteReactor Coolant System Section 5.1Containment System - Summary Description Section 6.2Emergency Core Cooling Systems (ECCS)

Section 6.4Control Rod Velocity Limiters Section 6.5Control Rod Drive Housing Supports Section 6.6Standby Liquid Control System Section 5.2.1Primary Containment System - Design Criteria Section 5.3.2Secondary Containment System - Design Basis Section 12 CompletePlant Structures and Shielding Section 14.1.1Summary Description - General Safety Design Basis Section 14.1.5Summary Description - Design Basis for

Accidents Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 15 of 61 I/jlkE.2.3Group III - Nuclear and Radiation Controls The intent of the current draft of the proposed criteria for this group is to identify and define the instrumentation and control systems, necessary for maintaining the 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 adequate shielding, 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 control room 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 and instrumentation 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 required safety action. The reactor protection system automatically initiates appropriate action 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 parameters from 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 individual criteria of this group.Criterion 11 - Control Room (Category B)

The facility shall be provided with a control room from which action to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit

access, even under accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control to the facility without 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 16 of 61 I/jlkConformance 11 - Control Room (Category B)

Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.2.8Principal Design Criteria - Plant Shielding and

Access Control Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrumentation and Control

Systems Section 1.3.9Summary Design Description and Safety Analysis - Plant Shielding, Access Control, and Radiation Protection ProceduresSection 1.3.11Summary Design Description and Safety

Analysis - Summary Evaluation of Plant Safety Section 7.2Reactor Control Systems Section 7.3Nuclear Instrumentation System Section 7.6Plant Protection System Section 7.7Turbine-Generator System Instrumentation and

Control Section 12.3.3Shielding and Radiation Protection -

Performance Analysis Criterion 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 Instrumentation and Control Section 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant Safety Section 7Plant Instrumentation and Control Systems Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 17 of 61 I/jlk Section 7.2Reactor Control Systems Section 7.3Nuclear Instrumentation System Section 7.4Reactor Vessel Instrumentation Section 7.5Plant Radiation Monitoring Systems Section 7.6Plant Protection System Section 7.7Turbine-Generator System Instrumentation and Control Section 7.8NUMAC Rod Worth Minimizer and Plant Process

Computer Criterion 13 - Fission Process Monitors and Controls (Category B)

Means shall be provided for monitoring and maintaining control over the fission process

throughout core life and for all conditions that can reasonably be anticipated to cause variation in reactivity of the core, such as indication of position of control rods and concentration of soluble reactivity control poisons.

Conformance 13 - Fission Process Monitors and Controls (Category B)

Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrumentation Control Systems Section 3.5Reactivity Control Mechanical Characteristics Section 6.6Standby Liquid Control System Section 7.2Reactor Control Systems Section 7.3Nuclear Instrumentation System Section 7.4Reactor Vessel Instrumentation Section 7.6Plant Protection System Section 7.8NUMAC Rod Worth Minimizer and Plant Process

Computer Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 18 of 61 I/jlkCriterion 14 - Core Protection Systems (Category B)

Core protection systems together with associated equipment, shall be designed to act automatically to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

Conformance 14 -Core Protection Systems (Category B)

Section 1.2.2Principal Design Criteria - Reactor Core Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.3.4Summary Design Description and Safety

Analysis - Plant Auxiliary and Standby Cooling

Systems Section 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 Safety Section 3.3Nuclear Characteristics Section 3.4Fuel Mechanical Characteristics Section 3.5Reactivity Control Mechanical Characteristics Section 6.2Emergency Core Cooling System (ECCS)

Section 6.3Main Steam Line Flow Restrictions Section 6.4Control Rod Velocity Limiters Section 6.5Control Rod Drive Housing Supports Section 7.2Reactor Control Systems Section 7.3Nuclear Instrumentation System Section 7.6Plant Protection System Section 7.8NUMAC Rod Worth Minimizer and Plant Process

Computer Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 19 of 61 I/jlk Section 8 CompletePlant Electrical Systems Section 14 CompletePlant Safety Analysis Criterion 15 - Engineered Safety Features Protection Systems (Category B)

Protection systems shall be provided for sensing accident situations and initiating the operation of necessary engineered safety features.

Conformance 15 - Engineered Safety Features Protection Systems (Category B)

Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrument Control SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant Safety Section 6 CompletePlant Engineered Safeguards Section 7.2Reactor Control Systems Section 7.3Nuclear Instrumentation System Section 7.4Reactor Vessel Instrumentation Section 7.5Plant Radiation Monitoring Systems Section 7.6Plant Protection System Section 7.7Turbine-Generator Systems Instrumentation and Control Section 7.8NUMAC Rod Worth Minimizer and Plant Process

Computer Criterion 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 Instrumentation and Control Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrument Control Systems Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 20 of 61 I/jlk Section 5.2Primary Containment System Section 7.1Plant Instrumentation and Control Systems -

Summary Description Section 7.3Nuclear Instrumentation System Section 7.4Reactor Vessel Instrumentation Section 7.6Plant Protection System Criterion 17 - Monitoring Radioactivity Releases (Category B)

Means shall beprovided for monitoring the containment atmosphere, the facility effluent discharge paths, and the facility environs, for radioactivity that could be released

from normal operations, from anticipated transients, and from accident conditions.

Conformance 17 - Monitoring Radioactivity Releases (Category B)

Section 1.2.7Principal Design Criteria - Plant RadioactiveWaste Disposal Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrument Control Systems Section 5.3.4.1Secondary Containment System - Standby Gas Treatment System (SGTS)

Section 7.5Plant Radiation Monitoring Systems Section 7.6.1Plant Protection System - Reactor Protection

System Section 9.2Liquid Radwaste System Section 9.3Gaseous Radwaste System Section 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning Systems Section 10.3.7Plant Service Systems - Plant Process Sampling

System Section 14.1.5Summary Description - Design Basis for

Accidents Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 21 of 61 I/jlkCriterion 18 - Monitoring Fuel and Waste Storage (Category B)

Monitoring and alarm instrumentation shall be provided for fuel and waste storage and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposures.Conformance 18 - Monitoring Fuel and Waste Storage (Category B)

Section 7.5Plant Radiation Monitoring Systems Section 7.6.1Plant Protection System - Reactor Protection System Section 9.2.1Liquid Radwaste System - Design Basis Section 9.2.2.1Liquid Radwaste System - General Section 9.2.2.3Liquid Radwaste System - Instrumentation and Control of the Liquid Radwaste Section 9.3.1Gaseous Radwaste System - Design Basis Section 9.3.3Gaseous Radwaste System - Performance Analysis Section 9.4.1Solid Radwaste System - Design Basis Section 9.4.3Solid Radwaste System - Performance Analysis Section 10.2.1.1Reactor Auxiliary Systems - Design Basis Section 10.2.1.2Reactor Auxiliary Systems - Description Section 10.2.2.1Reactor Auxiliary Systems - Design Basis Section 10.2.2.3Reactor Auxiliary Systems - Performance

AnalysisE.2.4Group IV - Reliability and Testability of Protection Systems The intent of the current draft of the proposed criteria for this group is to identifyand establish the functional reliability, in-service testability, redundancy, physical and 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 normal operational control system (that is, functions independently) to initiate Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 22 of 61 I/jlkappropriate action whenever the plant conditions monitored (neutron flux, containment, and vessel pressure, etc.) by the system approach pre-established limits. (Criterion 22) By means of a dual channel protection system with complete 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 and containment 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 and maintenance without negating the ability of the protection system to perform its protection 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 annunciated in 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. The system 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 any external 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 function upon 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 systems shall be designed for high functional reliability and in-service testability commensurate, with the safety functions to be performed.

Conformance 19 - Protection Systems Reliability (Category B)

Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.3.1Summary Design Description and Safety

Analysis - Plant Site and Environs Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 23 of 61 I/jlk Section 7.2Reactor Control Systems Section 7.3Nuclear Instrumentation System Section 7.4Reactor Vessel Instrumentation Section 7.5.2Plant Radiation Monitoring systems - Process Radiation Monitoring Systems Section 7.6Plant Protection SystemSection 11.2Turbine-Generator System Section 14.1.5Summary Description - Design Basis for

Accidents Criterion 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 Instrumentation and Control Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrument Control Systems Section 7.1Plant Instrumentation and Control Systems -

Summary Description Section 7.3Nuclear Instrumentation System Section 7.4Reactor Vessel Instrumentation Section 7.5.2Plant Radiation Monitoring Systems - Process

Radiation Monitoring System Section 7.6Plant Protection SystemSection 11.2Turbine-Generator System Section 14.1.5Summary Description - Design Basis for

Accidents Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 24 of 61 I/jlkCriterion 21 - Single Failure Definition (Category B)

Multiple failures from a single event shall be treated as a single failure.

Conformance 21 - Single Failure Definition (Category B)

Section 7.2Reactor Control Systems Section 7.6Plant Protection System Section 14.4Transient Events Analyzed for Core Reload Criterion 22 - Separation of Protection and Control Instrumentation Systems (Category B)

Protection systems shall be separated from control instrumentation systems to the extent that failure or removal from service of any control instrumentation 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 Instrumentation and Control Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrument Control Systems Section 7.4.2Reactor Vessel Instrumentation - Description Section 7.4.3Reactor Vessel Instrumentation - Inspection and Testing Section 7.6.3Plant Protection System - Primary Containment

Isolation System Criterion 23 - Protection Against Multiple Disability for Protection Systems (Category B) The effects of adverse conditions to which redundant channels or protection 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 Instrumentation and Control Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrument Control Systems Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 25 of 61 I/jlk Section 5.2.1.3Primary Containment System -Containment Penetrations Section 7.1Plant Instrumentation and Control Systems -

Summary Description Section 7.3Nuclear Instrumentation System Section 7.4Reactor Vessel Instrumentation Section 7.5Plant Radiation Monitoring Systems Section 7.6Plant Protection SystemSection 11.2Turbine-Generator System Criterion 24 - Emergency Power for Protection Systems (Category B)

In theevent of the loss of all off-site power, sufficient alternate sources of power shall be 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 Power Section 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power Systems Section 7 CompletePlant Instrumentation and Control Systems Section 8.3Auxiliary Power System Section 8.4Plant Standby Diesel Generator Systems Section 8.5D-C Power Supply Systems Section 8.6Reactor Protection System Power Supplies Section 10.3.8Plant Service Systems - Plant Communication

System Section 10.3.9Plant Service Systems - Plant Lighting System Criterion 25 - Demonstration of Functional Operability of Protection System (Category B)

Means shall be included for testing protection systems while the reactor is in operation to demonstrate that no failure or loss of redundancy has

occurred.

Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 26 of 61 I/jlkConformance 25 - Demonstration of Functional Operability of Protection System (Category B)

Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.3.5Summary Design Description and Safety Analysis - Plant Instrument Control Systems Section 7.3.5.5Nuclear Instrumentation System - Inspection andTesting Section 7.4.3Reactor Vessel Instrumentation - Inspection and Testing Section 7.5.2.1Plant Radiation Monitoring Subsystem - General Section 7.5.2.4.2Plant Radiation Monitoring Systems - Description Section 7.6.1.4Plant Protection System - Inspection and Testing Section 7.6.3.4Plant Protection System - Inspection and Testing Section 10.3.1.4Plant Service Systems - Inspection and Testing Section 10.3.2.4Plant Service Systems - Plant Heating,Ventilating and Air Conditioning Systems Section 10.3.9Plant Service Systems - Plant Lighting System Section 10.4Plant Cooling Systems Criterion 26 - Protection Systems Fail-Safe Design (Category B)

The protection systems shall be designed to fail into safe state or into a state established as tolerable 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 Instrumentation and Control Section 1.2.6Principal Design Criteria - Plant Electrical Power Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrument Control Systems Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 27 of 61 I/jlk Section 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power Systems Section 3.5.1Reactivity Control Mechanical Characteristics -

Design Basis Section 3.5.5Reactivity Control Mechanical Characteristics -

Operation and Performance Analysis Section 7.6Plant Protection Systems Section 8.6Reactor Protection System Power Supplies Section 10.3Plant Service Systems Section 10.4Plant Cooling SystemE.2.5Group V - Reactivity Control The 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 different principles. 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 all operating 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 or damps changes in power level and spatial distributions of power productions to a level consistent with safe and efficient operation, (b) a negative reactivity feedback 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 subcritical from its most reactive condition, and means are provided for continuous regulation of the reactor core excess reactivity and reactivity distribution.

(Criterion 29, 30) This system is also designed to be capable of compensating Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 28 of 61 I/jlkfor positive and negative reactivity changes resulting from nuclear coefficients, fuel depletion, and fission product transients and buildup. (Criterion 29) The system design is such that control rod worths, and the rate at which reactivity can 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 fuel damage 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 individual criteria of this group.

Criterion 27 - Redundancy of Reactivity Control (Category A)

At least twoindependent reactivity control systems, preferable of different principles, shall be provided.Conformance 27 - Redundancy of Reactivity Control (Category A)

Section 1.2.2Principal Design Criteria - Reactor Core Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 3.3.1Nuclear Characteristic - Design Basis Section 3.3.3.3Nuclear Characteristic - Reactivity Control Section 3.3.3.4Nuclear Characteristic - Control Rod Worth Section 3.5Reactivity Control Mechanical Characteristics Section 6.6.3Standby Liquid Control System - Performance Analysis Section 7.2Reactor Control Systems Section 8.4Plant Standby Diesel Generator Systems Criterion 28 - Reactivity Hot Shutdown Capability (Category A)

At least two of the 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 29 of 61 I/jlkConformance 28 - Reactivity Hot Shutdown Capability (Category A)

Section 1.2.2Principal Design Criteria - Reactor Core Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 3.3.1Nuclear Characteristic - Design Basis Section 3.5Reactivity Control Mechanical Characteristics Section 6.6Standby Liquid Control System Section 7.2Reactor Control Systems Criterion 29 - Reactivity Shutdown Capability (Category A)

At least one of the reactivity 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 Core Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 3.5Reactivity Control Mechanical Characteristics Section 6.6Standby Liquid Control System Section 7.2Reactor Control Systems Criterion 30 - Reactivity Holddown Capability (Category B)

At least one of the reactivity control systems provided shall be capable of making and holding the core subcritical under any conditions with appropriate margins for contingencies.

Conformance 30 - Reactivity Holddown Capability (Category B)

Section 1.2.2Principal Design Criteria - Reactor Core Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 3.3.3.3Nuclear Characteristic - Reactivity Control Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 30 of 61 I/jlk Section 3.5Reactivity Control Mechanical Characteristics Section 6.6Standby Liquid Control System Section 7.2Reactor Control Systems Criterion 31 - Reactivity Control Systems Malfunction (Category B)

The reactivity 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 Core Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 3.2Thermal and Hydraulic Characteristics Section 3.3Nuclear Characteristic Section 3.5Reactivity Control Mechanical Characteristics Section 6.4Control Rod Velocity Limiters Section 6.6Standby Liquid Control System Section 7.2Reactor Control SystemsCriterion 32 - Maximum Reactivity Worth of Control Rods (Category A)

Limits, which include considerable margin, shall be placed on the maximum reactivity worth 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 of reactivity 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 Core Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 3.3.3.3Nuclear Characteristic - Reactivity Control Section 3.3.3.4Nuclear Characteristic - Control Rod Worth Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 31 of 61 I/jlk Section 3.4Fuel Mechanical Characteristics Section 3.5Reactivity Control Mechanical Characteristics Section 4 CompleteReactor Coolant System Section 6.4Control Rod Velocity Limiters Section 6.5Control Rod Drive Housing Supports Section 7.8NUMAC Rod Worth Minimizer and Plant Process Computer Section 14.1.5Summary Description - Design Basis for

AccidentsE.2.6Group VI - Reactor Coolant Pressure Boundary The intent of the current draft of the proposed criteria for this group is to

establish the reactor coolant pressure boundary design requirements and to identify 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 with certain engineered safety features (control rod velocity limiters and control rod

housing supports, etc.) and the plant operational reactivity control system are such that the consequences of the most severe potential nuclear excursion accident, 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 pressure system. 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 Piping Protection against the brittle fracture or other failure modes of the reactor coolant

pressure boundary system components is provided for all potential service Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 32 of 61 I/jlkloading temperatures. Control is exercised in the selection of materials and fabrication and design of equipment and components. It is intended that NDT testing be performed on all ferritic materials in the reactor coolant pressure boundary 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 in accordance 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 vessel head. The reactor primary system will be checked for leaks and abnormal conditions 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 are located 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 page for the individual criteria of this group.

Criterion 33 - Reactor Coolant Pressure Boundary Capability (Category A)

The reactor coolant pressure boundary shall be capable of accommodating without 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 the coolant. As a design reference, this sudden release shall be taken as that which would 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 Core Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 3.3.3.3Nuclear Characteristic - Reactivity Control Section 3.3.3.4Nuclear Characteristic - Control Rod Worth Section 3.4Fuel Mechanical Characteristics Section 3.5Reactivity Control Mechanical Characteristics Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 33 of 61 I/jlk Section 4 CompleteReactor Coolant System Section 6.4Control Rod Velocity Limiters Section 6.5Control Rod Drive Housing Supports Section 7.8NUMAC Rod Worth Minimizer and Plant Process Computer Section 14.1.5Summary Description - Design Basis for

Accidents Criterion 34 - Reactor Coolant Pressure Boundary Rapid Propagation Failure Prevent (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 component fabrication 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 Propagation Failure Prevention (Category A)

Section Appendix CQuality Assurance Program Section 4 CompleteReactor Coolant System Criteria 35 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention (Category A)

Under conditions where reactor coolant pressure boundary system components constructed 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 the resulting energy is expected to be absorbed within the elastic strain energy range.Conformance 35 - Reactor Coolant Pressure Boundary Brittle Fracture Prevention (Category A)

Section 4.2.3Reactor Vessel - Design Evaluation Section 4.3.1Recirculation System - Design Criteria Section 4.3.3Recirculation System - Performance Evaluation Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 34 of 61 I/jlk Section 4.4.3Reactor Pressure Relief System - Performance Analysis Criteria 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 program conforming with ASTM-E-185-66 shall be provided.

Conformance 36 - Reactor Coolant Pressure Boundary Surveillance (Category A)Section 4.2.1Reactor Vessel - Design Basis Section 4.3.1Recirculation System - Design Basis Section 4.3.4Recirculation System - Inspection and Testing Section 4.4.4Reactor Pressure Relief System - Inspectionand TestingE.2.7Group VII - Engineered Safety Features The 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 credible phenomena associated with the plant operational transients or design basis accidents 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 narrow operating limits. These systems are thoroughly engineered and backed up a significant 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 variables to 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 35 of 61 I/jlkreactor control systems, reactor protection system, other instrumentation andprocess systems, etc.); those which offer additional protection against a reactivity excursion (reactor standby liquid control system, control rod velocity limiters, 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, including missiles 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 auxiliary sources 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 design basis 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 melt over 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 for testing 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 addition from 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, and pipe associated with the drywell have an initial NDT temperature of

approximately 0

°F when tested in accordance with the appropriate code for the Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 36 of 61 I/jlkmaterials. It is intended that the drywell will not be pressurized or subjected to substantial stress at temperatures below 30

° F. Provisions are made for the removal of heat from within the plant containment system and to isolate the various 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 basis accident will be below the values stated in 10CFR 100 (Criterion 50, 51, 54). All pipes 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 has commenced 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 containment design pressure (Criterion 52). Demonstration of operability and the ability to test 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 and sampling (Criterion 64). The system may be physically inspected and its operability demonstrated (Criterion 62, 63, 65).

References to applicable sections of the USAR are given below for the individual criteria 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 assuming unobstructed discharge from both ends.

Conformance 37 - Engineered Safety Features Basis for Design (Category A)

Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 1.2.4Principal Design Criteria - Plant Containment Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 37 of 61 I/jlk Section 1.2.6Principal Design Criteria - Plant Electrical Power Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 1.3.3Summary Design Description and Safety Analysis - Plant Containment System Section 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling Systems Section 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control Systems Section 1.3.8Summary Design Description and Safety

Analysis - Plant Electrical Power Systems Section 5 CompleteContainment System Section 6 CompletePlant Engineered Safeguards Section 7 CompletePlant Instrumentation and Control Systems Section 8 CompletePlant Electrical Systems Section 10.3.8Plant Service Systems - Plant Communication

System Section 10.3.9Plant Service Systems - Plant Lighting System Section 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 performance capability and reliability of the systems, and by the extent to which the operability of such systems can be tested and inspected where appropriate during the life of

the plant.

Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 38 of 61 I/jlkConformance 38 - Reliability and Testability of Engineered Safety Features (Category A)

Section 1.2.2Principal Design Criteria - Reactor Core Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 1.2.4Principal Design Criteria - Plant Containment Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 1.3.3Summary Design Description and Safety Analysis - Plant Containment System Section 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling

Systems Section 1.3.5Summary Design Description and Safety

Analysis - Plant Instrumentation Control Systems Section 5 CompleteContainment System Section 6 CompletePlant Engineered Safeguards Section 7 CompletePlant Instrumentation and Control Systems Section 8 CompletePlant Electrical Systems Section 10.3.8Plant Service Systems - Plant Communication

System Section 10.3.9Plant Service Systems - Plant Lighting System Criterion 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 functioning required 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 39 of 61 I/jlkConformance 39 - Emergency Power for Engineered Safety Features (Category A)

Section 1.2.6Principal Design Criteria - Plant Electrical Power Section 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power Systems Section 8.2Transmission System Section 8.3Auxiliary Power System Section 8.4Plant Standby Diesel Generator Systems Section 8.5D-C Power Supply Systems Section 8.6Reactor Protection System Power Supplies Criterion 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 Containment Section 5.2.1Primary Containment System - Design Criteria Section 5.2.3Primary Containment System - Performance Analysis Section 5.3.5Secondary Containment System - Performance Analysis Section 12 CompletePlant Structures and Shielding Criterion 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 40 of 61 I/jlkConformance 41 - Engineered Safety Features Performance Capability (Category A)

Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 1.2.4Principal Design Criteria - Plant Containment Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.2.6Principal Design Criteria - Plant Electrical Power Section 1.3.3Summary Design Description and Safety

Analysis - Plant Containment System Section 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling Systems Section 1.3.8Summary Design Description and Safety

Analysis - Plant Electrical Power Systems Section 5.2.1Primary Containment System - Design Criteria Section 5.3.2Secondary Containment System - Design Basis Section 6.2.1.1Emergency Core Cooling System (ECCS) -

ECCS Design Basis Section 6.2.4.3High Pressure Coolant Injection System (HPCI) -

Performance Analysis Section 6.2.5.3Automatic Depressurization System (ADS) -

Performance Analysis Section 6.2.2.3Reactor Core Spray Cooling System (CSCS) -

Performance Analysis Section 6.2.3.3Residual Heat Removal System (RHR) -

Performance Analysis Section 6.2.6Emergency Core Cooling System (ECCS) -

ECCS Performance Evaluation Section 6.3Main Steam Line Flow Restrictions Section 6.4Control Rod Velocity Limiters Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 41 of 61 I/jlk Section 6.5Control Rod Drive Housing Supports Section 6.6Standby Liquid Control System Section 8.2Transmission System Section 8.3Auxiliary Power Systems Section 8.4Plant Standby Diesel Generator Systems Section 8.5D-C Power Supply Systems Section 8.6Reactor Protection System Power Supplies Section 10.3.4Plant Service Systems - Plant Instrumentation and Service Air Systems Section 10.3.8Plant Service Systems - Plant Communication

System Section 10.3.9Plant Service Systems - Plant Lighting System Section 14.1.5Summary Description - Design Basis for

Accidents Criterion 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 Cooling Section 1.2.4Principal Design Criteria - Plant Containment Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.2.6Principal Design Criteria - Plant Electrical Power Section 3.6Other Reactor Vessel Internals Section 5.2.1Primary Containment System - Design Criteria Section 5.2.3Primary Containment System - Performance

Analysis Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 42 of 61 I/jlk Section 6 CompletePlant Engineered Safeguards Section 7.4Reactor Vessel Instrumentation Section 7.6Plant Protection System Section 12 CompletePlant Structures and Shielding Section 14.1.5Summary Description - Design Basis Accident Analysis Criterion 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 - Performance Analysis Section 6.2.1.1Emergency Core Cooling System (ECCS) -

ECCS Design Basis Section 6.2.4.3High Pressure Coolant Injection System (HPCI) -

Performance Analysis Section 6.2.5.3Automatic Depressurization System (ADS) -

Performance Analysis Section 6.2.2.3Reactor Core Spray Cooling System (CSCS) -

Performance Analysis Section 6.2.3.3Residual Heat Removal System (RHR) -

Performance Analysis Section 6.2.6Emergency Core Cooling System (ECCS) -

ECCS Performance Evaluation Section 6.3.1Main Steam Line Flow Restrictions - Design

Basis Section 6.4.1Control Rod Velocity Limiters - Design Basis Section 6.5.1Control Rod Drive Housing Supports - Design

Basis Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 43 of 61 I/jlk Section 6.6.1Standby Liquid Control System - Design Basis Criterion 44 - Emergency Core Cooling System Capability (Category A)At least two emergency core cooling systems, preferably of different design principles, 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 to negligible 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 be demonstrated that (a) the capability of the shared feature or components to perform 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 and is 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 Cooling Section 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling

Systems Section 6.2.1.2Emergency Core Cooling System (ECCS) -

Description and Function of ECCS Section 6.2.2.1Reactor Core Spray Cooling System (CSCS) -

Design Basis Section 6.2.3.1Residual Heat Removal System (RHR) -

Design Basis Section 6.2.4.1High Pressure Coolant Injection System (HPCI) -

Design Basis Section 6.2.5.1Automatic Depressurization System (ADS) -

Design Basis Section 6.2.6Emergency Core Cooling System (ECCS) -

ECCS Performance Evaluation Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 44 of 61 I/jlk Section 14.1.5Summary Description - Design Basis for Accidents Criterion 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 internals and water injection nozzles.

Conformance 45 - Inspection of Emergency Core Cooling Systems (Category A)

Section 3.6.1Other Reactor Vessel Internals - Design Basis Section 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and Testing Section 6.2.3.4Residual Heat Removal System (RHR) -

Inspection and Testing Section 6.2.4.4High Pressure Coolant Injection System (HPCI) -

Inspection and Testing Section 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 emergency core cooling systems, such as pumps and valves, can be tested periodically for operability 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 Basis Section 6.2.2.1Reactor Core Spray Cooling System (CSCS) -

Design Basis Section 6.2.2.3Reactor Core Spray Cooling System (CSCS) -

Performance Analysis Section 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and Testing Section 6.2.4.1 High Pressure Coolant Injection System (HPCI)-

Design Basis Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 45 of 61 I/jlk Section 6.2.4.3High Pressure Coolant Injection System (HPCI) -

Performance Analysis Section 6.2.4.4High Pressure Coolant Injection System (HPCI) -

Inspection and Testing Section 6.2.3.1Residual Heat Removal System (RHR) -

Design Basis Section 6.2.3.3Residual Heat Removal System (RHR) -

Performance Analysis Section 6.2.3.4Residual Heat Removal System (RHR) -Inspection and Testing Section 6.2.5.1Automatic Depressurization System (ADS) -

Design Basis Section 6.2.5.3Automatic Depressurization System (ADS) -

Performance Analysis Section 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 the emergency 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 Basis Section 6.2.2.1Reactor Core Spray Cooling System (CSCS) -

Design Basis Section 6.2.2.3Reactor Core Spray Cooling System (CSCS) -

Performance Analysis Section 6.2.2.4Reactor Core Spray Cooling System (CSCS) -Inspection and Testing Section 6.2.4.1 High Pressure Coolant Injection System (HPCI)-

Design Basis Section 6.2.4.3High Pressure Coolant Injection System (HPCI) -

Performance Analysis Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 46 of 61 I/jlk Section 6.2.4.4High Pressure Coolant Injection System (HPCI) -Inspection and Testing Section 6.2.3.1Residual Heat Removal System (RHR) -

Design Basis Section 6.2.3.3Residual Heat Removal System (RHR) -

Performance Analysis Section 6.2.3.4Residual Heat Removal System (RHR) -Inspection and Testing Section 6.2.5.1Automatic Depressurization System (ADS) -

Design Basis Section 6.2.5.3Automatic Depressurization System (ADS) -

Performance Analysis Section 6.2.5.4Automatic Depressurization System (ADS) -

Inspection and TestingCriterion 48 - Testing of Operational Sequence of Emergency Core Cooling System (Category A)

A capability shall be provided to test under conditions as close to design as practical the full operational sequence that would bring the emergency core cooling systems into action, including the transfer to alternate power sources.Conformance 48 - Testing of Operational Sequence of Emergency Core Cooling System (Category A)

Section 6.2Emergency Core Cooling System (ECCS)

Section 8 CompletePlant Electrical Systems Section 8.2Transmission System Section 8.3Auxiliary Power System Section 8.4Plant Standby Diesel Generator Systems Section 8.5D-C Power Supply Systems Section 8.6Reactor Protection System Power Supplies Section 10.4Plant Cooling System Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 47 of 61 I/jlkCriterion 49 - Containment Design Basis (Category A)

The containment structure, including access openings and penetrations, and any necessary 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 energy release 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 Core Section 1.2.3Principal Design Criteria - Reactor Core Cooling Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 1.3.3Summary Design Description and Safety

Analysis - Plant Containment System Section 1.3.4Summary Design Description and Safety

Analysis - Plant Auxiliary and Standby Cooling

Systems Section 1.3Summary Design Description and Safety

Analysis Section 5.1Containment System - Summary Description Section 5.2.3Primary Containment System - Performance

Analysis Section 5.2.4Primary Containment System - Inspection and Testing Section 5.3.2Secondary Containment System - Design Basis Section 5.3.5Secondary Containment System - Performance

Analysis Section 5.3.6Secondary Containment System - Inspection and Testing Section 6.2Emergency Core Cooling System (ECCS)

Section 6.6Standby Liquid Control System Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 48 of 61 I/jlk Section 10.2.5Reactor Auxiliary Systems - Reactor Core Isolation Cooling System (RCIC)

Section 14.1.5Summary Description - Design Basis for Accident Analysis Criterion 50 - NDT Requirement for Containment Material (Category A)

Principal load carrying components of ferritic materials exposed to the external environment 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 of the appropriateness of features such as isolation valves and additional containment 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 Criteria Section 1.2.4Principal Design Criteria - Plant Containment Section 1.2.5Principal Design Criteria - Plant Instrumentation and Control Section 1.2.6Principal Design Criteria - Plant Electrical Power Section 1.3.2Summary Design Description and Safety

Analysis - Reactor System Section 1.3.3Summary Design Description and Safety

Analysis - Plant Containment System Section 1.3.5Summary Design Description and Safety Analysis - Plant Instrumentation Control Systems Section 1.3.8Summary Design Description and Safety Analysis - Plant Electrical Power System Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 49 of 61 I/jlkSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant Safety Section 2.2Site Description Section 5.2Primary Containment System Section 5.3Secondary Containment System Section 6.3Main Steam Line Floor Restrictions Section 7.5.2Plant Radiation Monitoring Systems - Process

Radiation Monitoring System Section 7.6.3Plant Protection System - Primary Containment

Isolation System Section 14.1.5Summary Description - Design Basis for

Accident Analysis Criterion 52 - Containment Heat Removal Systems (Category A)

Where active heat removal systems are needed under accident conditions to prevent 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 Cooling Section 1.2.4Principal Design Criteria - Plant Containment Section 1.3.2Summary Design Description and Safety Analysis - Reactor System Section 1.3.3Summary Design Description and Safety

Analysis - Plant Containment System Section 1.3.4Summary Design Description and Safety Analysis - Plant Auxiliary and Standby Cooling Systems Section 5.2Primary Containment System Section 6.2Emergency Core Cooling System (ECCS)

Section 10.2Reactor Auxiliary Systems Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 50 of 61 I/jlk Section 10.4Plant Cooling System Section 14.1.5Summary Description - Design Basis for Accident AnalysisCriterion 53 - Containment Isolation Valves (Category A)

Penetrations that require closure for the containment function shall be protected by redundant valving and associated apparatus.Conformance 53 - Containment Isolation Valves (Category A)

Section 5.2.1.3Primary Containment System - Containment Penetrations Section 5.2.2.5.3Primary Containment System - Isolation System Section 5.2.3.7Primary Containment System - Penetrations Section 5.2.3.6.2Primary Containment System - Isolation System Section 5.2.4Primary Containment System - Inspection and Testing Section 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 Containment Section 5.2.1Primary Containment System - Design Criteria Section 5.2.3Primary Containment System - Performance Analysis Section 5.2.4Primary Containment System - Inspection and Testing Section 5.3.2Secondary Containment System - Design Basis Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 51 of 61 I/jlk Section 5.3.5Secondary Containment System - Performance Analysis Section 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 Containment Section 5.2.1Primary Containment System - Design Criteria Section 5.2.3Primary Containment System - Performance Analysis Section 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 Criteria Section 5.2.3Primary Containment System - Performance Analysis Section 5.2.4Primary Containment System - Inspection andTesting Section 5.3.5Secondary Containment System - Performance Analysis Section 5.3.6Secondary Containment System - Inspection and Testing Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 52 of 61 I/jlkCriteria 57 - Provisions for Testing of Isolation Valves (Category A)

Capability shall be provided for testing functional operability of valves and associated apparatus essential to the containment function for establishing that no failure has occurred and for determining that valve leakage does not exceed acceptable limits.

Conformance 57 - Provisions for Testing of Isolation Valves (Category A)

Section 7.6.3.1Plant Protection System - Design Basis Section 7.6.3.3Plant Protection System - Performance Analysis Section 7.6.3.4Plant Protection System - Inspection and Testing Section 7.5.2Plant Radiation Monitoring Systems - Process Radiation Monitoring System Criterion 58 - Inspection of Containment Pressure-Reducing System (Category A)

Design provisions shall be made to facilitate the periodic physical inspection of all important components of the containment pressure-reducing systems, such as, pumps, valves, spray nozzles, torus, and sumps.

Conformance 58 - Inspection of Containment Pressure-Reducing System (Category A)

Section 5.2.4Primary Containment System - Inspection andTesting Section 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 Systems Components (Category A)

Section 6.2.1.1Emergency Core Cooling System (ECCS) -

Design Basis Section 6.2Emergency Core Cooling System (ECCS)

Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 53 of 61 I/jlkCriterion 60 - Testing of Containment Spray Systems (Category A)

A capability shall be provided to test periodically the delivery capability of the containment spray system at a position as close to the spray nozzle as is practical.Conformance 60 - Testing of Containment Spray Systems (Category A)

Section 6.2.1.1Emergency Core Cooling System (ECCS) -

Design Basis Section 6.2Emergency Core Cooling System (ECCS)Criterion 61 - Testing of Operational Sequence of Containment Pressure-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 containment pressure-reducing systems into action, including the transfer to alternate power sources.Conformance 61 - Testing of Operational Sequence of Containment Pressure-Reducing Systems (Category A)

Section 5.2 CompletePrimary Containment System Section 7.6.3.3Plant Protection System - Performance Analysis Section 7.6.3.4Plant Protection System - Inspection and Testing Section 6.2.1.1Emergency Core Cooling System (ECCS) -

Design Basis Section 6.2Emergency Core Cooling System (ECCS)

Section 8 CompletePlant Electrical Systems Criterion 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 54 of 61 I/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

Analysis Section 5.3.6Secondary Containment System - Inspection and Testing Section 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 - Performance Analysis Section 5.3.6Secondary Containment System - Inspection and Testing Section 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 air cleanup 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 - Performance Analysis Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 55 of 61 I/jlk Section 5.3.6Secondary Containment System - Inspection andTesting Section 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 systems to 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

Analysis Section 5.3.6Secondary Containment System - Inspection andTesting Section 7.5.2Plant Radiation Monitoring Systems - Process Radiation Monitoring System Section 7.6.1Plant Protection System - Reactor Protection

System Section 8.4Plant Standby Diesel Generator Systems Section 8.5D-C Power Supply Systems Section 8.6Reactor Protection System Power Supplies Section 10.3.2Plant Service Systems - Plant Heating, Ventilating and Air Conditioning SystemsE.2.8Group VIII - Fuel and Waste Storage Systems The 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 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 56 of 61 I/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 provides containment), 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 and arranged 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 individual criteria 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 or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls.

Conformance 66 - Prevention of Fuel Storage Critically (Category B)

Section 1.2.9Principal Design Criteria - Plant Fuel Handling and Storage Section 1.3.6Summary Design Description and Safety

Analysis - Plant Fuel Storage and Handling

Systems Section 6.6.3Standby Liquid Control System - Performance

Analysis Section 10.2.1.1Reactor Auxiliary Systems - Design Basis Section 10.2.1.2Reactor Auxiliary Systems - Description01081199 Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 57 of 61 I/jlkCriterion 67 - Fuel and Waste Storage Decay Heat (Category B)

Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public environs.Conformance 67 - Fuel and Waste Storage Decay Heat (Category B)

Section 1.2.7Principal Design Criteria - Plant RadioactiveWaste Disposal Section 1.2.9Principal Design Criteria - Plant Fuel Handling

and Storage Section 1.3.4Summary Design Description and Safety

Analysis - Plant Auxiliary and Standby Cooling

system Section 1.3Summary Design Description and Safety

Analysis Section 6.2.1.2Emergency Core Cooling System (ECCS) -

Description and Function of ECCS Section 10.2.1Reactor Auxiliary Systems - Fuel Storage and Fuel Handling Systems Section 10.2.2Reactor Auxiliary Systems - Spent Fuel Pool Cooling and Demineralizer System Section 10.2.3Reactor Auxiliary Systems - Reactor Cleanup

Demineralizer System Section 10.2.4Reactor Auxiliary Systems - Reactor Shutdown

Cooling System Section 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 and waste storage facilities as required to meet requirements of 10 CFR 20.

Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 58 of 61 I/jlkConformance 68 - Fuel and Waste Storage Radiation Shielding (Category B)

Section 1.2.8Principal Design Criteria - Plant Shielding and Access Control Section 1.3.6Summary Design Description and Safety

Analysis - Plant Fuel Storage and Handling Systems Section 1.3.9Summary Design Description and Safety Analysis - Plant Shielding, Access Control, and

Radiation Protection Procedures Section 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 Section 12.3Shielding And Radiation Protection Section 9.2.1Liquid Radwaste System - Design Basis Section 9.2.3Liquid Radwaste System - Performance Analysis Section 9.3.1Gaseous Radwaste System - Design Basis Section 9.3.3Gaseous Radwaste System - Performance

Analysis Section 9.4.1Solid Radwaste System - Design Basis Section 9.4.3Solid Radwaste System - Performance Analysis Section 10.2.1.1Reactor Auxiliary Systems - Design Basis Section 10.2.1.2Reactor Auxiliary Systems - Description Section 10.2.1.3Reactor Auxiliary Systems - Performance Analysis Criterion 69 - Protection Against Radioactivity Release from Spent Fuel andWaste Storage (Category B)

Containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environs.

Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 59 of 61 I/jlkConformance 69 - Protection Against Radioactivity Release from Spent Fuel andWaste Storage (Category B)

Section 1.2.4Principal Design Criteria - Plant Containment Section 1.2.8Principal Design Criteria - Plant Shielding and Access Control Section 1.3.6Summary Design Description and Safety

Analysis - Plant Fuel Storage and Handling

Systems Section 1.3.9Summary Design Description and Safety Analysis - Plant Shielding, Access Control, and Radiation Protection Procedures Section 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 Section 5.1Containment System - Summary Description Section 5.3Secondary Containment System Section 9 CompletePlant Radioactive Waste Control Systems Section 10.2.1Reactor Auxiliary Systems - Fuel Storage and Fuel Handling Systems Section 10.2.2Reactor Auxiliary Systems - Spent Fuel Pool Cooling and Demineralizer System Section 1.2.7Principal Design Criteria - Plant RadioactiveWaste Disposal Section 1.2.8Principal Design Criteria - Plant Shielding and

Access Control Section 14.7.6.4.2Refueling Accident Analysis - Radiological Consequences Section 14.7.4Accident Evaluation Methodology - Fuel Loading Error Accident Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 60 of 61 I/jlkE.2.9Group IX - Plant Effluents The 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 engineered safety 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 the established release limits at the elevated plant stack. (Criterion 70)

References to applicable sections of the USAR are given for the individual criteria 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. Appropriate holdup 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 normal operations and for any transient situation that might reasonably be anticipated to occur 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 Containment Section 1.2.7Principal Design Criteria - Plant RadioactiveWaste Disposal Section 1.2.8Principal Design Criteria - Plant Shielding and

Access Control Revision 25 USAR E.2MONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 61 of 61 I/jlk Section 1.3.9Summary Design Description and Safety Analysis - Plant Shielding, Access Control, and Radiation Protection Procedures Section 1.3.10Summary Design Description and SafetyAnalysis - Plant Radioactive Waste Control

SystemsSection 1.3.11Summary Design Description and Safety Analysis - Summary Evaluation of Plant Safety Section 2.2Site Description Section 5 CompleteContainment System Section 12 CompletePlant Structures and Shielding Section 7.5Plant Radiation Monitoring Systems Section 8 CompletePlant Electrical Systems Section 9 CompletePlant Radioactive Waste Control Systems Section 10.3.6Plant Service Systems - Plant Equipment and Floor Drainage Systems Section 10.3.7Plant Service Systems - Plant Process Sampling SystemSection 11.3.2Main Condenser System - Main Condenser Gas Removal System Section 13 CompletePlant Operations Section 14 CompletePlant Safety Analysis Revision 22 USAR APPENDIX EMONTICELLO UPDATED SAFETY ANALYSIS REPORT Page 1 of 1APPENDIX EPLANT COMPARATIVE EVALUATION WITH THE PROPOSED AEC 70 DESIGN CRITERIA I/mabTABLE OF CONTENTS Section Page E.1 Summary Description