ML24109A095

1 to the Updated Safety Analysis Report, Chapter 13, Plant Operations

ML24109A095
Person / Time
Site:	Monticello
Issue date:	04/17/2024
From:	Northern States Power Company, Minnesota, Xcel Energy
To:	Office of Nuclear Reactor Regulation
Shared Package
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References
L-MT-24-002
Download: ML24109A095 (1)

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MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-13 Revision 39 SECTION 13 PLANT OPERATIONS Page 1 of 19

TABLE OF CONTENTS Section Page

13.1 Summary Description................................................................................. 3

13.2 Organization, Responsibilities, and Qualifications...................................... 3

13.2.1 Organization............................................................................................... 3 13.2.2 Duties, Responsibilities and Qualification of the Operating Staff Personnel................................................................................................... 3

13.3 Personnel Experience and Training............................................................ 3

13.3.1 Experience of Initial Plant Supervisory Personnel...................................... 3 13.3.2 Experience and Training of Plant and Site Staff......................................... 4 13.3.3 Personnel Behavior.................................................................................... 4

13.4 Operational Procedures.............................................................................. 4

13.4.1 General....................................................................................................... 4 13.4.2 Procedure Development............................................................................. 5 13.4.3 Emergency Plan......................................................................................... 5 13.4.4 Security Plan.............................................................................................. 6 13.4.5 Quality Assurance Plan.............................................................................. 6 13.4.6 10CFR50.55a Inservice Inspection and Testing Programs........................ 7 13.4.7 Post-Scram Review.................................................................................... 8 13.4.8 Surveillance Frequency Control Program................................................. 10

13.5 Operational Records and Reporting Requirements.................................. 10

13.5.1 Records of Initial Tests............................................................................. 10 13.5.2 Routine Operation.................................................................................... 10 13.5.3 Abnormal Operation................................................................................. 10 13.5.4 Reporting Requirements........................................................................... 10 13.5.5 Radiographs............................................................................................. 11

13.6 Operational Review and Audits................................................................ 11

13.6.1 General..................................................................................................... 11 13.6.2 Plant Operating Review Committee.......................................................... 11 13.6.3 Management and Safety Review Committee........................................... 11

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13.7 Emergency Procedures............................................................................ 11

13.8 Technical Requirements Manual.............................................................. 12

13.9 Risk Informed Categorization and Treatment........................................... 12

13.9.1 Introduction............................................................................................... 12 13.9.2 SSC Categorization.................................................................................. 14 13.9.3 SSC Treatment......................................................................................... 14 13.9.3.1 Treatment of Component Categories....................................................... 14 13.9.3.2 Enhanced Treatment of RISC-2 SSCs..................................................... 15

13.10 References............................................................................................... 16

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13.1 Summary Description

Xcel Energy and Northern States Power Company, a Minnesota corporation have experienced plant personnel that are qualified to perform plant operations and plant maintenance that are necessary for safe operation of the plant.

Training programs are scheduled and implemented to maintain sufficient licensed operators and a competent supporting technical staff. Plant activities are conducted in accordance with Quality Assurance, Emergency, and Security Plans and written procedures implemented in response to regulatory requirements. Inspection and testing are conducted in accordance with a program which meets regulatory requirements.

13.2 Organization, Responsibilities, and Qualifications

13.2.1 Organization

The Monticello Nuclear Generating Plant site management and organization including the plant-specific titles of those personnel fulfilling the responsibilities of the positions delineated in the Technical Specifications 5.2.1.a are described in NSPM-1, Quality Assurance Topical Report.

In support of the individual responsibilities of plant personnel, an Plant Operating Review Committee (further described in Section 13.6.2) provides multi-discipline review of various plant activities.

The onsite organization includes the technically trained personnel necessary to support all aspects of plant operations.

13.2.2 Duties, Responsibilities and Qualification of the Operating Staff Personnel

The responsibilities and duties of key site and plant operating staff personnel are described in NSPM-1.

13.3 Personnel Experience and Training

13.3.1 Experience of Initial Plant Supervisory Personnel

Senior Operator licensed personnel from the Pathfinder Nuclear Plant organization were assigned to supervisory positions for the Monticello plant during initial plant operation.

All participated in the pre-operational testing, fuel loading, startup testing and operation of the Pathfinder plant.

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13.3.2 Experience and Training of Plant and Site Staff

Minimum qualifications and training requirements for plant staff (i.e., operating personnel) are contained in training programs approved by the Nuclear Regulatory Commission. An NRC approved t raining program is one that is based on the systems approach to training (SAT) and has been accredited by the National Nuclear Accrediting Board (NNAB) (Reference 13 and Reference 14).

Each member of the plant and site staff SHALL meet or exceed the minimum qualifications of ANSI N18.1-1971 (Reference 25) for comparable positions. Exceptions to these standards are documented in the Technical Specifications.

Training enhancements required by NUREG-0737, item I.A.2.1.4, are in place as well as training in mitigating core damage required by Item II.B.4.1 (Reference 2).

13.3.3 Personnel Behavior

The Fitness for Duty Program applies to all nuclear generation personnel, including all badged contract workers and craft union personnel hired by Xcel Energy, NSPM, or its contractors. It recognizes that fatigue, stress, illness and temporary physical impairments, as well as drug and alcohol abuse, can have a negative effect on a workers fitness and jeopardize safe operations.

All personnel badged for unescorted access to the plant are subject to random drug and alcohol testing, and are trained to be observant of co-worker or visitor behavior that may indicate a fitness for duty concern. Supervisors are trained to be observant of employee behavior that might indicate excessive fatigue or unhealthy behavior patterns and to bar employees from working if they appear unfit for duty. The NSPM Fitness for Duty program meets all of the requirements of 10CFR26 (see Reference 24).

13.4 Operational Procedures

13.4.1 General

A preoperational test program was conducted to assure that all systems and equipment function properly. The initial preoperational and startup test programs are described in Appendix D. General Electric and Bechtel provided written procedures and technical direction for these programs. The plant operating staff participated in the preparation and execution of these tests.

Detailed written procedures, including the applicable check-off sheets and instructions were prepared in accordance with the Technical Specifications and ANSI N18.7-1976 (Reference 28). Currently the QATR, NSPM-1 (Reference 66) governs the detailed written procedures. Plant operations are conducted in accordance with these procedures.

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13.4.2 Procedure Development

The original operations procedures were written by members of the plant staff with the technical assistance of General Electric and were reviewed by the Operations Committee.

Procedures are periodically updated to reflect plant modifications and improvements in methods of operation as operating experience accumulates.

Special written procedures for one-of-a-kind operations are occasionally necessary.

These are prepared by qualified personnel and are reviewed by the Operations Committee. The Operations Committee may also submit these for review by the Safety Audit Committee.

Maintenance and test procedures, checklists, and other necessary records to satisfy routine inspections, preventive maintenance programs, and license requirements, have been and will continue to be developed by qualified personnel.

13.4.3 Emergency Plan

In any emergency situation at Monticello, the initial response would be made by the plant staff and, if needed, by local support agencies. It is expected that the initial response would have to extend for a period of hours, by which time the plant staff would be augmented by other segments of the overall NSPM emergency response organization. Once all centers are activated and the emergency organization is at full strength, the scope of the plant staff response will be reduced to the immediate plant site activities. The Monticello Nuclear Generating Plant Emergency Plan was submitted according to the new 10CFR50 emergency planning regulations on February 6, 1981.

Subsequent revisions to the plan are issued and reported to the NRC in accordance with 10CFR50.54(q).

The NRC has concluded that onsite and offsite emergency preparedness is adequate and that the emergency plans have been upgraded in accordance with NUREG -0737 Item III.A.2.1 (Reference 1).

The plan is directed toward the following areas:

a. Organization and actions within the plant to control and limit the consequences of an accident.

b. Organization and actions controlling site and initial offsite activities in the event of an uncontrolled release of radioactive material. This includes notification of and coordination with required offsite support agencies.

c. Identifying and evaluating the consequences of accidents that may occur and affect the public and plant personnel.

d. Describing the protective action levels and actions that are required to protect the public and plant personnel in the event of an accident.

e. Considerations necessary for the purposes of re-entry and recovery.

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f. Arrangements required for medical support in the event of injury.

g. The training necessary to assure adequate response to emergencies.

The Emergency Plan identifies the location of primary and backup Emergency Operations Facilities (EOF). The primary EOF is located in the Training Building one mile from the containment. The backup EOF is located 45 miles from the plant in conjunction with the Xcel Energy Corporate Headquarters in downtown Minneapolis.

The location of the EOFs was found acceptable by the NRC documented in a letter dated October 27, 1983 (Reference 29).

The Emergency Plan is dependent upon the Emergency Plan Implementing Procedures for implementation. Implementing Procedures were initially submitted to the NRC on February 27, 1981 (Reference 30). Revisions to procedures are issued and reported to the NRC in accordance with 10CFR50, Appendix E, Section V.

13.4.4 Security Plan

The security plan consists of documents referred to as the Monticello Nuclear Generating Plant Physical Security Plan as approved by the NRC via Amendment 58 to Facility Operating License DPR-22, dated December 13, 1988 and Cyber Security Plan for NSPM as approved by the NRC via Amendments 166 and 186 to the Renewed Facility Operating License DPR-22. The security plans are periodically revised to meet changing requirements. Revisions to the security plans, not requiring prior NRC approval, are issued and reported to the NRC in accordance with 10CFR50.54(p).

13.4.5 Quality Assurance Plan

Xcel Energys nuclear plant operational activities were conducted under the Operational Quality Assurance (QA) Plan (Reference 31). Now the operational activities are conducted under the current revision of the Northern States Power Company-Minnesota QATR, NSPM-1. The QATR describes the quality assurance program and how it satisfies the applicable regulations and guidelines. It also contains the duties, responsibilities and authority of those individuals and groups involved in carrying out activities required by the QA program.

The Northern States Power Company-Minnesota QATR, NSPM-1 (Reference 66) is responsive to the requirements of Appendix B to 10CFR50. The Quality Assurance Plan is periodically revised to meet changing requirements and the current revision is maintained on file at the plant and corporate headquarters.

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13.4.6 10CFR50.55a Inservice Inspection and Testing Programs

Inservice Inspection (ISI) of components and their supports equivalent to ASME Code Classes 1, 2, 3 are performed in accordance with the Monticello Nuclear Generating Plant ASME Code Section XI Inservice Inspection Program. Inservice Inspection (ISI) of the metal containment (MC) components are performed in accordance with the Monticello Nuclear Generating Plant ASME Code Section XI Containment Inspection Program. Testing and examinations are performed in accordance with formal administrative work instructions (AWIs). The ISI Program is composed of two inspection plans:

a. The ISI Plan is developed and maintained at the site. (Reference 60). When practical, the tests and examinations conform to ASME Code,Section XI, as specified in 10CFR50.55a(g).

b. The Metal Containment Examination Plan (IWE) is maintained on site available for review. When practical, the tests and examinations conform to ASME Code,Section XI, as specified in 10CFR50.55a(g).

Inservice Testing (IST) of program pumps and valves are performed in accordance with the Monticello Pump and Valve IST Program for (Reference 61) in accordance with the Plans ASME Operation and Maintenance (OM) Codes of Record (Reference 63).

Testing of IST Program pumps and valves is performed in accordance with formal procedures.

Preservice and inservice examination and testing of dynamic restraints (snubbers) required to perform a specific function in shutting down the reactor to safe shutdown condition, in maintaining the safe shutdown conditions, or in mitigating the consequences of an accident are performed in accordance with the Monticello Snubber IST Program Plan in accordance with the Plans ASME Operation and Maintenance (OM) Codes of Record (Reference 63). Testing and examinations are performed in accordance with formal procedures.

The IST Program is composed of two inspection plans:

a. The IST Plan is developed and maintained at the site. Tests conform to ASME OM Code as specified in 10CFR50.55a(f) except where relief from such requirements has been granted.

b. The Snubber Plan is developed and maintained at the site. Tests conform to the ASME OM Code as specified in 10CFR50.55a(f).

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13.4.7 Post-Scram Review

As a result of a failure of scram circuit breakers at the Salem Nuclear Power Plant, the NRC requested all operating plants to initiate an in-depth review into four areas: (1)

Post-Scram Review, (2) Equipment Classification and Vendor Interface, (3)

Post-Maintenance Testing, and (4) Reactor Trip System Relia bility. The NRC requirements were described in Generic Letter 83 -28 (Reference 44).

Northern States Power Company completed a review of the generic implications of the event at the Salem Plant, as required by the Generic Letter and submitted the results of this review to the NRC.

One of the key lessons learned from the Salem event was the need for a comprehensive post-scram review prior to returning the unit to service. The Monticello post-scram review program specifies:

a. Criteria for determining acceptability of restart.

b. Qualifications, responsibilities, and authorities of personnel who perform the review and analysis.

c. Methods and criteria for comparing event information with known or expected plant behavior.

d. Criteria for determining need for independent assessment of an event.

e. Procedures to ensure that all physical evidence necessary for an independent assessment is preserved.

f. Systematic safety assessment program to assess unscheduled plant scrams.

The NRCs review and acceptance of the Monticello post-scram review program is reflected in their July 3, 1985 Safety Evaluation Report (Reference 7) of NSPs November 14, 1983 response to Generic Letter 83-28 Item 1.1 (Reference 6).

The Monticello post-scram review program includes the following data and information capability:

a. Equipment to record the sequence of events and time history data needed for post-scram review.

b. Established and identified parameters to be monitored and recorded for post-scram review.

c. Means for storage and retrieval of information gathered by the sequence of events and time history recorders, and for the presentation of this information for post-scram review and analysis.

d. Data and information used during post-scram review retention for the life of the plant.

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The NRCs review and acceptance of the Monticello post -scram data and availability is reflected in their June 2, 1986 Safety Evaluation Report (Reference 10) of NSPs November 14, 1983 (Reference 6) and May 5, 1986 (Reference 9) responses to Generic Letter 83-28 Item 1.2.

Generic Letter 83-28, Item 2.1 required NSP to confirm that all components required to trip the reactor are properly identified as safety related in all plant documentation and to confirm that an interface is established and maintained with the vendors of these components through a program of periodic communications. The NRCs review and acceptance of NSPs November 14, 1983 (Reference 6) and June 9, 1988 (Reference 56) submittals is reflected in their February 13, 1989 Safety Evaluation Report (Reference 16).

Generic Letter 83-28, Item 2.2, Part 1, required NSP to describe the Monticello program for classifying and identifying safety related components other than those in the Reactor Trip System. The NRCs review and acceptance of NSPs November 14, 1983 (Reference 6), March 31, 1987 (Reference 55) and June 9, 1988 (Reference 56) responses is reflected in their September 15, 1989 Safety Evaluation Report, (Reference 17). Generic Letter 90-03 (Reference 57) relaxed the NRC position stated in Generic Letter 83-28, Item 2.2, Part 2, relative to the interface with vendors of these components. The NRC acknowledged NSPs September 25, 1990 (Reference 22) commitment to implement the guidance in Generic Letter 90 -03 in an October 10, 1990 letter (Reference 23).

Generic Letter 83-28, Item 3.1 and 3.2 required NSP to assure that post maintenance operability testing of reactor trip system and other associated components is performed in accordance with vendor and engineering recommendations to demonstrate that all safety related equipment is capable of performing its safety functions before being returned to service. The NRCs review and acceptance of NSPs November 14, 1983 (Reference 6), December 30, 1983 (Reference 52), December 26, 1984 (Reference 8) and February 28, 1986 (Reference 53) submittals is reflected in their May 2, 1986 (Reference 54) and March 21, 1986 (Reference 11) Safety Evaluation Reports.

Generic Letter 83-28, Item 4.5 requires on-line functional testing of the reactor trip system, including the scram pilot solenoid valves and initiating circuitry. In its November 14, 1983 submittal (Reference 6), NSP affirmed that such testing was being performed with the exception of the backup scram valves which are tested as part of the plant restart testing for each refueling outage. The NRCs March 21, 1986 Safety Evaluation Report (Reference 11) found Monticellos review of part 4.5.1 acceptable. The NRC found Monticellos review part 4.5.2 acceptable in its January 27, 1989 Safety Evaluation Report (Reference 18). Monticello endorsed (Reference 19) the BWR Owners Group resolution of part 4.5.1 as documented in General Electric Topical Report No. NEDC-30844 (Reference 45). The NRC staff concluded that the intervals for on-line functional testing at Monticello are consistent with achieving high reactor trip system availability (Reference 20).

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13.4.8 Surveillance Frequency Control Program

Surveillance frequencies are based primarily upon deterministic methods such as engineering judgment, operating experience, and manufacturer's recommendations.

The Surveillance Frequency Control Program (SFCP) and Technical Specifications (TS)

Section 5.0, Subsection 5.5.15, references NEI 04-10, Revision 1, which provides a risk-informed methodology using plant-specific risk insights and performance data to revise surveillance frequencies within the SFCP. TS define which surveillance requirements are within the SFCP. This methodology supports relocating surveillance frequencies from TSs to a licensee-controlled document, provided those frequencies are changed in accordance with the NRC-approved NEI 04-10, Revision 1. The SFCP complements the deterministic approach and supports a defense-in-depth philosophy.

13.5 Operational Records and Reporting Requirements

13.5.1 Records of Initial Tests

All preoperational procedures, test data, and reports are kept on file at the plant site.

Complete records of the plant startup tests are kept at the plant site in the test file.

These records include:

a. Startup test procedures. This is the final, as run, test procedure, including approvals and data sheets.

b. Pertinent recorder charts and log sheets.

c. Test reports - This includes any reports prepared by NSP, GE or Bechtel.

13.5.2 Routine Operation

Operating, maintenance and testing records and logs are kept on file in accordance with the Technical Specifications, Federal Regulations and NSP policy.

13.5.3 Abnormal Operation

In the event of any unusual, unexplained, or potentially unsafe occurrence, appropriate members of the plant staff will be assigned to conduct an investigation and prepare a report. Instructions for conducting investigation and the report format are outlined in plant administrative procedures. A complete file of investigation reports is maintained.

13.5.4 Reporting Requirements

Reports will be submitted to the Commission to satisfy the requirements of Title 10, Code of Federal Regulations, and the Monticello Technical Specifications.

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13.5.5 Radiographs

Microfilmed Radiographs of piping system welds meet the requirements of ASME Section III, Paragraph NCA-4134.17, Quality Assurance Records (Reference 46) and ASME Section XI, Paragraph IWA-6320, Reproduction and Microfilming (Reference 47 and 15).

13.6 Operational Review and Audits

13.6.1 General

Review of facility operations is performed by the Management and Safety Review Committee (MSRC) and/or the Plant Operating Review Committee (PORC).

13.6.2 Plant Operating Review Committee

The function of the committee is to review and evaluate proposed tests, modifications to plant systems or equipment, changes in plant normal or emergency procedures, certain plant events and other activities having nuclear safety significance. Detailed discussions of committee membership, frequency of meetings, authority, responsibilities, procedural requirements and record management are defined in the applicable section of the QATR.

13.6.3 Management and Safety Review Committee

The Management and Safety Review Committee (MSRC) is an independent review group whose basic responsibility is to advise NSPM management on the nuclear safety of plant operations. This Committee continuously reviews information on plant activities and periodically meets as a group to discuss this information. Detailed discussions of committee membership, qualifications, frequency of meetings, authority, responsibilities, audits, procedural requirements and record management were originally defined in the applicable section of the OQAP. Current practices are defined in the Management and Safety Review Committee (MSRC) Procedure.

13.7 Emergency Procedures

Monticello Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMGs) have been developed to satisfy the guidance contained in:

• Supplement 1 to NUREG-0737 (Reference 58)

• Section 5 of Nuclear Energy Institute (NEI) Report 91-04, Revision 1 (Reference 59)

Revisions to the EOPs and SAMGs are processed in accordance with the guidance provided in the NRCs April 17, 1990 Safety Evaluation Report (Reference 21) regarding the Monticello Procedures Generation Package submitted in response to Generic Letter 82-33 (Reference 49).

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The Monticello EOPs allow operators to take actions immediately. Early operator actions taken in accordance with the EOPs enhance the ability to mitigate the consequences of events. Should the operator not take actions immediately, the plant will remain within the margins established by the design basis analysis because the automatic plant systems will still respond as designed.

13.8 Technical Requirements Manual

The Technical Requirements Manual (TRM) is a licensee-controlled document that provides a location for items removed from the Technical Specifications that do not meet the criteria of 10 CFR 50.36 (Reference 68).

For purposes of making changes to the TRM: (1) Changes to the TRM will be controlled by the provisions of 10 CFR 50.59, (2) summaries of 50.59 evaluations will be submitted to the NRC, and (3) the TRM is a general reference in the USAR, and as such, changed pages will not be submitted to the NRC.

13.9 Risk Informed Categorization and Treatment

13.9.1 Introduction

On November 22, 2004, the NRC issued 10 CFR 50.69 that presented nuclear management with an opportunity to further enhance equipment reliability and plant safety by focusing on those critical structures, systems, and components (SSCs) with the highest safety significance. The rule broadly adjusts the scope of safety-related components that are subject to the existing NRC regulations. The implementation of this new rule is strictly voluntary on the part of each licensee.

For those safety related components that are categorized as Low Safety Significant, 10 CFR 50.69(b)(1) allows compliance with alternative requirements in lieu of the following special treatment requirements.

a. 10 CFR Part 21

b. The portion of 10 CFR 50.46a(b) that imposes requirements to conform to Appendix B to 10 CFR Part 50.

c. 10 CFR 50.49

d. 10 CFR 50.55(e)

e. The in-service testing requirements in 10 CFR 50.55a(f): the in-service inspection and repair, and replacement (with the exception of fracture toughness),

requirements for ASME Class 2 and Class 3 SSCs in 10 CFR 50.55a(g); and the electrical component quality and qualification requirements in Section 4.3 and 4.4 of IEEE 279, and Sections 5.3 and 5.4 of IEEE 603-1991, as incorporated by reference in 10 CFR 50.55a(h).

f. 10 CFR 50.65, except for paragraph (a)(4)

g. 10 CFR 50.72

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h. 10 CFR 50.73

i. Appendix B to 10 CFR Part 50

j. The Type B and Type C leakage testing requirements in both Options A and B of Appendix J to 10 CFR Part 50, for penetrations and valves meeting the following criteria:

1. Containment penetrations that are either 1-in. nominal size or less, or continuously pressurized.

2. Containment isolation valves that meet one or more of the following criteria:

a) The valve is required to be open under accident conditions to prevent or mitigate core damage events;

b) The valve is normally closed and in a physically closed, water-filled system;

c) The valve is in a physically closed system whose piping pressure rating exceeds the containment design pressure rating and is not connected to the reactor coolant pressure boundary; or

d) The valve is 1-in. nominal size or less.

• Appendix A to Part 100, Sections Vl(a)(1) and Vl(a)(2), to the extent that these regulations require qualification testing and specific engineering methods to demonstrate that SSCs are designed to withstand the Safe Shutdown Earthquake and Operating Basis Earthquake.

It should be noted that 10 CFR 50.69 does not replace the existing safety-related and non safety-related classification. Instead, 10 CFR 50.69 divides these classifications into two subcategories based on high or low safety significance, such that there are four categories of risk-informed safety class (RISC), as shown below:

• RISC-1: safety-related SSCs that perform (high) safety significant functions.

• RISC-2: non safety-related SSCs that perform (high) safety significant functions.

• RISC-3: safety-related SSCs that perform low safety-significant functions.

• RISC-4: non safety-related SSCs that perform low safety-significant functions.

When applying alternative treatment, 10 CFR 59.69(d) requires that the licensee shall ensure, with reasonable confidence, that RISC-3 SSCs remain capable of performing their safety related functions under design basis conditions, including seismic conditions and environmental conditions and effects throughout their service life. Periodic inspection and testing activities will be conducted to ensure RISC-3 SSCs remain capable of performing their safety-related functions. The corrective action program will be used to document and correct in a timely manner any conditions that would prevent a RISC-3 SSC from performing its safety-related functions.

Monticello received approval from the NRC to implement 10 CFR 50.69 as outlined in Reference 69.

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13.9.2 SSC Categorization

As outlined in the Safety Evaluation Report (SER) for License Amendment 203 to Renewed Facility Operating License No. DPR-22, Monticello will use the methodology outlined in NEI 00-04, Reference 70.

10 CFR 50.69(f)(2) requires updating the UFSAR to reflect which systems have been categorized. The following table is revised as part of the periodic UFSAR update to reflect systems that have been categorized.

System Description CRV/EFT Control Room Heating & Ventilation -

Emergency Filtration Trains RHR Residual Heat Removal FSW EFT Emergency Service Water ARM Area Radiation Monitoring HPC High Pressure Coolant Injection CSP Core Spray Cooling RSW RHR Service Water

13.9.3 SSC Treatment

13.9.3.1 Treatment of Component Categories

The programs or processes that implement the special treatment requirements are revised to recognize that the special treatments no longer apply to RISC-3 SSCs. The programs or processes either allow continued application of the special treatments or acceptable alternative treatments, as applicable, to provide reasonable confidence that these SSCs would perform their safety-related function under design basis conditions.

The following information provides the general approach for applying treatment for the component categories:

a. RISC-1 Components RISC-1 SSC should continue to satisfy all of the existing regulatory requirements that are applicable, including those insights that were considered during the SSCs categorization.

b. RISC-2 Components The purpose of treatment applied to RISC-2 SSCs is to maintain their ability to perform risk-significant functions consistent with the categorization process.

These components will continue to receive any existing special treatment required by NRC regulations. Additionally, the risk significant functions of these components will receive consideration for enhanced treatment. This consideration is described in paragraph 13.9.3.2.

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c. RISC-3 Components These components may receive alternative treatments, in lieu of special treatments, as described in 13.9.1

d. RISC-4 Components The treatment of these components are not subject to regulatory control.

e. Uncategorized Components Until a component is categorized, it continues to receive the special treatment required by NRC regulations and associated Monticello implementing programs, as applicable.

13.9.3.2 Enhanced Treatment of RISC-2 SSCs

The 10 CFR 50.69 procedures and 10 CFR 50.69(d)(1) require that RISC-2 SSCs perform their functions consistent with the categorization process assumptions by evaluating treatment being applied to these SSCs to ensure that it supports the key assumptions in the categorization process that relate to their assumed performance.

Non safety-related HSS components may perform risk-significant functions that are not addressed by the special treatment requirements in NRC regulations or current Monticello programs.

When a non safety-related component is categorized as HSS, determine whether enhanced treatment is warranted to enhance the reliability and availability of the component in support of its HSS function(s). In particular, evaluate the treatment applied to the component to ensure that the existing controls are sufficient to maintain the reliability and availability of the component in a manner that is consistent with its categorization. This process evaluates the reliability of the component, the adequacy of the existing controls, and the need for any changes. If changes are needed, additional controls are applied to the component. In addition, the component is placed under the Maintenance Rule monitoring program, if not already scoped in the program. Components under these controls will remain non safety-related, but the enhanced treatments will be appropriately applied to give additional confidence that the component will be able to perform its HSS function(s) when demanded.

These identified processes provide reasonable confidence that HSS components will be able to perform their risk significant functions.

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13.10 References

1. NRC (D B Vassallo) letter to NSP (D M Musolf), NUREG-0737 Item III.A.2.1 Emergency Plan Upgrade to Meet Rule, dated May 18, 1983.

2. NRC (D B Vassallo) letter to NSP (D M Musolf), NUREG-0737 Item I.A.2.1.4 Upgrading of RO and SRO Training, Item II.B.4 Training for Mitigating Core Damage, dated June 3, 1983.

3. Deleted.

4. Deleted.

5. Deleted.

6. NSP (D M Musolf) letter to the NRC, Generic Implications of Salem ATWS Events (Generic Letter 83-28), dated November 14, 1983.

7. NRC (D B Vassallo) letter to NSP (D M Musolf), Generic Letter 83-28, Item 1.1 -

Post-Trip Review, dated July 3, 1985.

8. NSP (D M Musolf) letter to the NRC, Generic Implications of Salem ATWS Events (Generic Letter 83-28), dated December 26, 1984.

9. NSP (D M Musolf) letter to the NRC, Additional Information Related to Generic Letter 83-28, Item 1.2., dated May 5, 1986.

10. NRC (R Auluck) letter to NSP (D M Musolf), Generic Letter 83-28, Item 1.2, Post-Trip Review (TAC 53608), dated June 2, 1986.

11. NRC (J A Zwolinski) letter to NSP (D M Musolf), Safety Evaluation - Generic Letter 83-28, Items 3.1.1, 3.1.2 and 4.5.1 (TAC 52937, 54084), dated March 21, 1986.

12. Deleted.

13. NSP (D M Musolf) letter to the NRC, NRC License Operator Training Program and Licensed Operator Requalification Program, dated March 21, 1988.

14. NSP (D M Musolf) letter to the NRC, Shift Technical Advisor Training, dated April 8, 1988.

15. NRC (J J Harrison) letter to NSP (C E Larson), Special Safety Inspection Report No.

263/88022(DRS), dated November 28, 1988.

16. NRC (J J Stefano) letter to NSP (D M Musolf), Acceptability of Monticello Nuclear Generating Plant Programs Required by NRC Generic Letter 83-28, Item 2.1 (Parts 1 and 2) TAC No. 52856, dated February 13, 1989.

MONTICELLO UPDATED SAFETY ANALYSIS REPORT USAR-13 Revision 39 SECTION 13 PLANT OPERATIONS Page 17 of 19

17. NRC (W O Long) letter to NSP (T M Parker), Safety Evaluation for Item 2.2 (Part 1) of Generic Letter 83-28, Equipment Classification (Programs for all Safety-Related Components) TAC No. 53691, dated September 15, 1989.

18. NRC (J J Stefano) letter to NSP (D M Musolf), Safety Evaluation for Generic Letter 83-28, Item 4.5.2 (Reactor Trip System Reliability on-line Testing) TAC No. 54001, dated January 27, 1988.

19. NSP (D M Musolf) letter to the NRC, Additional Information Related to Generic Letter 83-28 Item 4.5.3 Applicability of General Electric Topical Report NEDC-30844, dated June 17, 1988.

20. NRC (J J Stefano) letter to NSP (T M Parker), Safety Evaluation for Generic Letter 83-28, Item 4.5.3, Reactor Trip Reliability-On-Line Functional Testing of the Reactor Trip System (TAC No. 54001), dated June 7, 1989.

21. NRC (W O Long) letter to NSP (T M Parker), Procedures Generation Package -

Safety Evaluation Report (TAC No. 44317), dated April 17, 1990.

22. NSP (T M Parker) letter to the NRC, Response to Generic Letter 90-03, Relaxation of Staff Position in Generic Letter 83-28, Item 2.2 Part 2 - Vendor Interface for Safety-Related Components, dated September 25, 1990.

23. NRC (W O Long) letter to NSP (T M Parker), Response to Generic Letter 90-03 for Monticello Nuclear Generating Plant (TAC No. 76260) dated October 10, 1990.

24. NSP (C E Larson) letter to the NRC, Certification of Compliance to 10 CFR 26 Fitness for Duty Program, dated January 3, 1990.

25. American National Standard, ANSI N18.1 - 1971, Selection and Training of Nuclear Plant Personnel.

26. Deleted.

27. Deleted.

28. American National Standard ANSI N18.7 -1976, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants, approved February 19, 1976.

29. NRC (D G Eisenhut) letter to NSP (D M Musolf), Primary and Backup Emergency Operations Facilities, dated October 27, 1983.

30. NSP (L O Mayer) letter to the NRC, Emergency Response Plan Implementing Procedures, dated February 27, 1981.

31. NSP Operational Quality Assurance Plan, Revision 2, November 15, 1977.

32. NRC (D K Davis) letter to NSP (L O Mayer), Operation Quality Assurance Program, Revision 2, dated December 29, 1977.

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33. Deleted.

34. Deleted.

35. Deleted.

36. Deleted.

37. Deleted.

38. Deleted.

39. Deleted.

40. Deleted.

41. Deleted.

42. Deleted

43. Deleted.

44. NRC (D G Eisenhut) Generic Letter 83-28, Required Actions Based on Generic Implications of Salem ATWS Events, dated July 8, 1983.

45. General Electric report NEDC-30844A, BWR Owners Group Responses to NRC Generic Letter 83-28, Item 4.5.3, March 1988.

46. ASME Boiler & Pressure Vessel Code,Section III, Paragraph NCA-4134.17, Quality Assurance Records.

47. ASME Boiler & Pressure Vessel Code,Section XI, Paragraph IWA-6320, Reproduction, Digitization, and Microfilming.

48. Deleted.

49. NSP (D M Musolf) letter to the NRC, Emergency Operating Procedures Generation Package Submittal, dated July 31, 1984.

50. Deleted.

51. Deleted.

52. NSP (D M Musolf) letter to the NRC, Generic Implications of Salem ATWS Events (Generic Letter 83-28), dated December 30, 1983.

53. NSP (D M. Musolf) letter to the NRC, Generic Letter 83-28, Salem Action Plant, Item 3.2, Check of Vendor and Engineering Recommendations, dated February 28, 1986.

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54. NRC (J A Zwolinski) letter to NSP (D. M. Musolf), Safety Evaluation - Generic Letter 83-28, Items 3.2.1 and 3.2.2 (TAC 53774), dated May 2, 1986.

55. NSP (D M. Musolf) letter to the NRC, Response to NRC Request for Further Information on NSP Response to Generic Letter 83-28, Item 2.2.1, dated March 31, 1987.

56. NSP (D M Musolf) letter to the NRC, Additional Information Related to Generic Letter 83-28, dated June 9, 1988.

57. NRC (J G Partlow) Generic Letter 90-03, Relaxation of Staff Position in Generic Letter 83-28, Item 2.3 Part 2 - Vendor Interface for Safety Related Components, Generic Letter 83-28, dated March 20, 1990.

58. NRC (D G Eisenhutt) Generic Letter 82-33, Supplement 1 to NUREG-0737 -

Requirements for Emergency Response Capability, December 17, 1982.

59. Nuclear Energy Institute (NEI) Report 91-04, Revision 1, Severe Accident Issue Closure Guidelines, dated December 1994.

60. Monticello ISI Plan.

61. NSPM letter to the NRC, "Pump & Valve In-Service Testing Program Plan, Fifth Interval-Revision 1", dated December 12, 2012.

62. NSPM letter to the NRC, "In-service Testing (IST) Program Plan for Safety Related Seismic Restraints (Snubbers) for the Fifth 10-Year Interval", dated November 30, 2012.

63. ASME Operation and Maintenance Code, 2004 Edition with Addenda through 2006.

64. Deleted.

65. Deleted.

66. Quality Assurance Topical Report (QATR), NSPM-1.

67. Transmittal of Monticello Emergency Plan, Revision 21, dated August 15, 2002.

68. NRC letter to NSPM, "Monticello Nuclear Generating Plant (MNGP) - Issuance of Amendment for the Conversion to the Improved Technical Specifications with Beyond-Scope Issues (TAC Nos. MC7505, MC7597 through MC7611, and MC8887)", dated June 5, 2006.

69. Safety Evaluation Report by the Office of Nuclear Reactor Regulation related to Amendment No. 203 to Renewed Facility Operating License No. DPR-22, Monticello.

70. NEI 00-04, 10 CFR 59.69 SSC Categorization Guidance, Rev. 0