ML20150F809

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Rev 12 to Operational QA Plan
ML20150F809
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 05/26/1988
From: Albrecht K, Jensen R
NORTHERN STATES POWER CO.
To:
Shared Package
ML112971252 List:
References
NUDOCS 8807190068
Download: ML20150F809 (150)
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Text

O NORTHERN STATES POWER COMPANY 414 Nicollet Mall Minneapolis, MN 55401 OPERATIONAL QUALITY ASSURANCE PLAN REV 12 Reviewed By: Date: I 4 ff Kenneth J Albrecht Director Power Supply Quality Assurance Approved By: '=== -

Date: f- 1C 48 Roland J (e/fsen Senior Vice President Power Supply gs 71:= :=ghe 0QAP

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Operational Quality Assurance Plan

'Rev.12 -

r'N b Table of Contents Page 1.0 Policy Statement 1 2.0 Introduction 6 3.0 Organization 7 4.0 ' Operational. Quality Assurance Program 20 5.0 Modification Control 25 6.0 Procurement Document Control 26 7.0 Instructions, Procedures and Drawings 28 8.0 Occument Control 30 9.0 Control of Purchased Material, Equipment and Services 32 38 10.0 Identification and Control of Materials, Parts and Components 34 11.0 Control of Special Processes 35 12.0 Inspection 37 13.0 Test Control 41 14.0 Control of Measuring and Test Equipment 43 15.0- Handling, Storage end Shipping 44 38 !

16.0 Inspection, Test and Operating Status 46 17.0 Nonconforming Materials, Parts or Coraponents -49 18.0 Corrective Action 51 i

19.0 Quality Assurance Records 52  :

20.0 Audits 54 Appendix A - Monticello Structures, Systems, and Components Subject to Appendix B of 10CFR50 56 Appendix B - Prairie Island Structures, Systems, and

  • Components Subject to Appendix B of 10CFR50 61

, Appendix C - Nuclear Plant Fire Protection Program 70 Appendix 0 - Revision 12 Change Summary 84 1 OQAP ,

m Operational Quality Assurance Plan Rev 12 1.0 Policy Statement 1.1 Northern States Power Company (NSP) has established and is implementing an Operational Quality Assurance Program. This quality assurance program is applicable to NSP nuclear plants that are regulated under provisions of an NRC Operating License.

1.2 The quality assurance program, as applied to activities affecting safety related functions, shall comply with and be responsive to applicable regulatory requirements and applicable industry codes and standards including:

1. 10CFR 50, Appendix B.
2. NRC Operating Licenses.
3. The ASME Boiler and Pressure Vessel Code,Section XI, "Inservice Inspection".
4. 10CFR21 "Reporting of Defects and Noncompliance".
5. 10CFR71, Subpart H, "Quality Assurance".
6. Nuclear Plant Fire Protection Program, Operational Quality f ^'; Assurance Plan Appendix C.

() 7. NSP Plant Security Plans.

8. NSP Radiation Environmental Monitoring Program.
9. ANSI N45.2.6-1978, Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants, as modified 39 by Regulatory Guide 1.58, Revision 1.
10. ANSI N45.2.12-1977, Requirements for Auditing of Quality Assurance Programs for Nuclear Power Plants.
11. ANSI N45.2.23-1978, Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants, as modified by Regulatory Guide 1.146, August 1980, 1.3 The Operational Quality Assurance Program shall incorporate:

(1) the requirements of ANSI N18.7-1976 as modified by Table 1-1 and (2) the requirements of the following standards to the extent specified by ANSI N18.7-1976, as modified by the regulatory position of the Regulatory or Safety Guides referenced below.

1. ANS1 N18.1-1971, Selection and Training of Nuclear Power Plant Personnel (Regulatory Guide 1.8, Rev. 1).

2, ANSI N45.2-1971, Quality Assurance Program Reauirements for N'jclear Power Plants CQAP Page 1 of 88

Operational Quality Assurance Plan Rev 12

3. ANSI N45.2.1-1973, Cleaning of Fluid Systems and Associated Corrponents Ouring Construction Phase of Nuclear Power Plants (Regulatory Guide 1.37,3-16-73). i
4. ANSI N45.2.2-1972, Packaging, Shipping, Receiving, Storage and 39 !

Handling of Items for Nuclear Power Plants (During the  !

Construction Phase) l (Regulatory Guide 1.38, Rev. 2). l

5. ANSI N45.2.3-1973, Housekeeping During the Construction Phase of Nuclear Power Plants (Regulatory Guide 1.39, Rev. 1).
6. ANSI N45.2.4-1972, Installation, Inspection, and Testing Requirements for Instrumentation and Electric Equipment 39 During the Construction of Nuclear Power Generating Stations (Safety Guide 30, August 11,1972).
7. ANSI N45.2.5-1974, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Structural Concrete and Structural Steel Ouring the Construction Phase of Nuclear Power Plants (Regulatory Guide 1.94, Rev. 1).
8. ANSI N45.2.8-1975, Supplementary Quality Assurance Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Power Plants l 9. ANSI N45.2.9-1974, Requirements for Collecticn, Storage and Maintenance of Quality Assurance Records for Nuclear Power Plants (Regulatory Guide 1.88, Rev. 2).
10. ANSI N45.2.10-1973, Quality Assurance Terms and Definitions (Regulatory Guide 1.74, February, 1974). 39
11. ANSI N45.2.11-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants (Regulatory Guide 1.64, Rev. 2).
12. ANSI N45.2.13-1976, Quality Assurance Requirements for the Control of Procurement of Items and Services for Nuclear Power Plants
13. ANSI N101.4-1972, Quality Assurance for Protective Coatings 39 Applied to Nuclear Facilities (Regulatory Guide 1.54, June, 1973).

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Operationt.1 Quality Assurance Plan Rev 12 A

("' 1 1.4 Management directives and departmental instructions and procedures

.shall provide for compliance with appropriate regulatory, statutory, license and industry requirements. Specific quality assurance requirements and organizational ~ responsibilities for implementation of these requirements shall- be specified in implementing directives and instructions. ,

1.5 Compliance with this policy and the. provisions.of the Operational Quality Assurance Program is mandatory for NSP personnel with respect to nuclear plant operational activities or activities which support nuclear plant operation. Personnel shall therefore, be familiar with the requirements and responsibilities of the program that are applicable.to their individual activities and interfaces.

1.6 The Senior Vice President Power Supply, through an independent organization, shall periodically have the Operational Quality Assurance Program reviewed to assure its adequacy.

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O OQAP Page 3 of 88

Operational Quality Assurance Plan Rev 12 h'"

Table 1-1 Exception to ANSI N18.7 - 1976

1. Documentation required by ANSI 18.7-1976 may be deferred for emergency work. Emergency work is defir.ed as that work that must be completed immediately and which, if delayed, may result in an unsafe condition or significantly interfere with reliable plant operation.
2. Exceptions to Regulatory Guides and ANSI Standards are acceptable fee those principal contractors, retained by NSP, such as NSSS contractors and A/E Firms, which exceptions have been approved by the NRC.
3. Section 5.1; Delete this section. The provisions associated with identification of the Operational Quality Assurance Program scope are explicity identified in Section 4 of the Operational Quality Assurance Plan.
4. Section 5.2.2; replace the third sentence with the following "Procedure changes shall be reviewed and approved as required by the Technical Specifications." Delete the fourth sentence. l 40
5. Section 5.2.5; replace the second and third sentences with "Temporary procedures shall be reviewed and approved as required by the Technical Specifications."
6. Section 5.2.9; delete the reference to ANSI N18.17. The Plant Security Plans contain required security provisions.
7. Section 5.2.11, first sentence; change "abnormal occurrences" to "reportable events".
8. Section 5.2.13.2, fourth paragraph; change the first sentence to l 41 read "... installation or use of such items that serve a safety function." Last paragraph; change "quality" to "quantity". This change corrects an error in the standard.
9. Section 5.2.15 of ANSI N18.7-1976 shall govern review, approval, and control of required procedures except that for procedures required by the Plant Technical Specifications, the review and approval requirements stipulated in the Technical Specifications shall be utili::ed rather than those contained in Section 5.2.15.
10. Section 5.3; change the last sentence to read "Procedures shall be prepared and approved prior to implementation as required by 5.2.15."
11. Section 6; delete this section. The referenced documents are explicitly referenced in the Operational Quality Assurance Plan. NSP will evaluate new or revised ANSI Standards if appropriate for inclusion in the Operational Quality Assurance Plan.

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0QAP Page 4 of 88

Operational Quality Assurance Plan Rev 12 0 12. Sections 5.3.9 and 5.3.9.1; delete these sections.

Procedures shall be consistent with Supplement 1 to NUREG - 0737 -

Emergency Operating 2

equirements for Emergency Response Capability (Generic Letter 82-33).

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O OQAP Page 5 of 88

Operational Quality Assurance Plan Rev 12 2.0 Introduction Northern States Power Company (NSP) is involved in the construction and operation of nuclear, fossil f'-led and hydro power plants. Construction 42 of nuclear plants is conductet under a cuality assurance program on a project basis. NSP's nuclear plant operational activities are conducted under the Operational Quality Assurance Program.

The Construction Quality Assurance Program is structured to govern nuclear plant design, fabrication, construction, testing and associated procurement as required by the applicable NRC Construction Permit and pertinent regulations. The Operational Quality Assurance Program is formulated on a company-wide basis, to govern nuclear piant operational activities and associated support activities as required by NRC Operating License provisions and associated regulations. The Operational Quality Assurance Program is implemented, to the extent compatible with construction responsibilities, at least 90 days prior to initial fuel loading and is fully implemented upon satisfactory completion of the preoperational and startup test program.

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Operational Quality Assurance Plan Rev 12 C')

V 3.0 Organization 3.1 General Reouirements

1. NSP shall be responsible for the establishment and execution of the Operational Quality Assurance Program. NSP may delegate to other organizations the work of establishing and executing the Operational Quality Assurance Program, or any part thereof, but shall retain responsibility therefor.
2. The authority and duties of persons and organizations performing quality assurance functions shall be clearly established and delineated in writing. Such persons and organizations shall have sufficient authority and organizational freedom to identify quality problems; to initiate, recommend, or provide solutions; and to verify implementation of solutions.
3. Assurance of quality requires management measures which provide that the individual or group assigned the responsibility for i; checking, auditing, inspecting, or otherwise verifying that an t activity has been correctly performed is independent of the individual or group directly resporsible for performing the specific activity, f

3.2 Ot.ality Organization Summary

1. The Power Supply Quality Assurance Department is responsible for the overall administration of the Operational Quality Assurance Program. Specific responsibilities are stated in Section 3.3.2.1 and its subsections.
2. The Plant Quality Engineering Section for each nuclear plant is responsible for the overall administration of the plant quality assurance program and quality control of plant activities.

Specific responsibilities are stated in Section 3.3.2.2.1.1.1.

3. The Mt.terial & Special Processes Section of the Production Plant Maintenance Department is responsible for providing technical support for the quality control of special processes at the nuclear plants. Specific responsibilities are stated in Section 3.3.2.3.1.2.1. 3
4. The Quality Control Section of the Nuclear Engineering and Construction Department is responsible for providing quality control for projects assigned to Nuclear Engineering and Construction. Specific responsibilities are stated in Section 3.3.2.2.2.2.

O OQAP Page 7 of 88

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Operational Quality Assurance Plan Rev 12 3.3 Corocrate Oroanization with Goe-ational Quality Assurance Program Resoonsibilities 3.3.1 President & Chief Executive Officer l4 This position is responsible for all NSP activities including those associated with operating nuclear plants. This responsibility is implemented by assigning responsibility to the corporate officers of the company (See Figure 1, Page 19, for the organizational diagram of positions with quality assurance l5 responsibilities).

3.3.2 Senior Vice President Power Supply This position is designated by the President & Chief Executive Officer as responsible for the establishment and implementation of l4 an Operational Quality Assurance Program. Responsibilities include:

1. Engineering, construction and operation of all generating facilities.
2. Establishment of an Operational Quality Assurance Plan that governs activities associated with Federal Regulation (10CFR50, Appendix B).

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3. Establishment of Corporate Nuclear Administrative Control Directives that identify quality assurance requirements and positions responsible for implementing those requirements.
4. Providing status reports to management.

Positions reporting to the Senior Vice President Power Supply include: Director Power Supply Quality Assurance, Vice President Nuclear Generation, Vice President Combustion & Hydro 6 Operations, Vice President Transmission & Inter-Utility Services, Director Fuel Resources, Director Power Supply Support, and Director Power Supply Financial Operations.

3.3.2.1 Director Power Supply Quality Assurance t

' This position is responsible for the establishment, maintenance and evaluation of the Operational Quality Assurance Program.

Responsibilities include:

1. Controlling revisions to the Operational Quality Assurance Plan.
2. Stop work authority for non-conforming activities until the adverse conditions have been corrected.
3. Assisting other company organizations in implementing quality assurance program requirements.

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Operational Quality Assurance Plan Rev 12

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V 4. Providing Power Supply Quality Assurance status reports to appropriate levels of management.

Positions reporting to this Director include: Superintendent Nuclear Projects & Supplier Quality Assurance, and Superintendent Nuclear Operations Quality Assurance.

l7 3.3.2.1.1 Superintendent Nuclear Projects & Supplier Quality Assurance l7 This position is responsible for control of the supplier qualification program and for quality assurance activities associated with nuclear plant projects performed by Nuclear Engineering & Construction. Responsibilities include: l4

1. Inspections of nuclear fuel suppliers.
2. Quality assurance audits and qualification of suppliers. l8
3. Review and approval of selected A/E, vendor and contractor quality assurance programs. l9
4. Quality assurance reviews of nuclear procurement made by general office organizations. 10
5. Quality assurance reviews of project specifications and

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t procurement documents.

6. Preparation / review of internal quality assurance programs and procedures.
7. Audits /surveillances of engineering, procurement, construction and testing activities.

3.3.2.1.3 Superintendent Nuclear Operations Quality Assurance This position is responsible for quality assurance activities associated with general office organizations and internal auditing. Responsibilities include:

1. Internal audits of all levels of the Operational Quality Assurance Program. 11
2. Review of Corporate Nuclear Administrative Control Directives and Administrative Work Instructions.
3. Maintenance of Corporate Nuclear Administrative Control Directives and Administrative Work Instructions current l43 with corporate commitments and policies.
4. Program implementation monitoring and periodic trending.

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OQAP Page 9 of 88

Operational Quality Assurance Plan Rev 12 2.3.2.2 Vice President Nuclear Generation '

This position is responsible for the operation and physical control of the company's nuclear generating facilities.

Responsibilities include:

1. Operation of nuclear facilities.
2. Maintenance of nuclear facilitie
2. Modification of nuclear facilities.
4. Nuclear facility fuel utilization.
5. Operational review of new nuclear facility design.
6. Independent review and audit of nuclear plant operations and operating license ad:ninistration.
7. Corporate security. 12 P sitions reporting to this Vice President include: General Manager Nuclear Plants, General Manager Nuclear Engineering &

Construction, General Panager Headquarters Nucle " Group, Manager Corporate Securi.ty, and Manager Special Nuclear Programs.

3.3.2.2.1 General Manager Nuclear Plants This position is responsible for the overall supervision of nuclear plant management, ensuring compliance with regulatory requirements, and for providing overall direction and support to nuclear plant management in matters of staffing and employee qualifications. Responsibilities include:

1. Review of plant operating abnormalities, problems, performance, malfunctions, etc, and concurrence in corrective actions.
2. Review of quality assurance status, trend and audit reports, and follow-up of resolution to nonconformances. 13
3. hview of Safety Audit Committee reports and recommendations.
4. Performance of further actions as required to ensure safety.
5. Training support to all of Power Supply. l 14 Positions reporting to this General Manager include: Monticello and Prairie Island Plant Managers, Manager Nuclear Radiological Services and Manager Production Training. l6 l i

l 0QAP Page 10 of 88 j

Operational Quality Assurance Plan Rev 12 jq r i V 3.3.2.2.'1.-1 Monticello and Prairie Island Plut Managers These positions are responsible for ensuring that activities' and operations comply with applicable regulatory requirements.

Responsibilities include:

.1. Responsibilities as.iigned by the operating licease and the Corporate Nuclear Administrative Control Directives. l 11

2. Plant managerial control system.
3. Plant operation and maintenance.
4. Plant staffing, including qualifications, hiring, training, discipline, and administration of the labor contracts.
5. Development and implementation of the following programs:
a. Preventive maintenance
b. Surveillancr
c. Material control
d. Operating, maintenance, and testing procedural systems
e. Fire protection

/7 f. Plant quality assurance and. control

(,) g. Operating experience assessment

h. Plant physical security and guard force supervision l 15
6. Coordination of activities performed by non plant staff personnel with plant operation.
7. Nondestructive examinations not associated with inservice inspection.

The posicien reporting to the Plant Manager having specific CA responsibilities is the Superintendent Quality 6 Engineering.

3.3.2.2.1.1.1 Superintendent Quality Engineering (each plant)

This position is responsible for the administration of the Operational Quality Assurance Program requirements at the plant level. Responsibilities include:

1. Implementation of the plant quality control inspection program (except ISI). l 16
2. Review of inspection schedules (except ISI),

procedures, and results (i.e., those associated with routine maintenance and modification activities, operational activities, technical services, radioactive b material packages, emergency equipment, and fire protection).

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Operational Quality Assurance Plan Rev 12

3. Audit of selected plant level activities when determined 0 that the audit will improve plant program implementation. l17
4. Review of plant Administrative Control Directives and Instructions.
5. Plant program implementacion monitoring and periodic trending.
6. Review of procurement documents.
7. Receipt inspection performed for plant procured items.
8. Stop work authority for nonconforming activities at the plant until adverse conditions have bean corrected.
9. Providing plant quclity assurance status reports to appropriate levels of management.

3.3.2.2.1.2 Manager Nuclear Rad;ologicai Services 18 This position is responsible for providing support to the nuclear plants in the areas of radiation protection, chemistry, emergency planning ud radiation environmental monitoring. Responsibilities inci d :

1. Providing a supportive corporate radiation protection program.
2. Providing a supportive corporate nuclear chemistry program (8WR & PWR).
3. Providing emergency preparedness management.

4 Administering NRC operating licenses and technical specificcions for environmental activities.

5. Reviewing, coordinating and evaluating company emergency preparedness and environmental activities required by the NRC.
6. Conducting a radiation enviror. mental monitoring program to comply with NRC requirements.
7. Reviewing proposed and revised regulations related to emergency preparedness and nuclear environmental activities.

3.3.2.2.1.3 Manager Production Training This position is responsible for evaluating the training needs of Power Supply employees, developing programs to meet these needs, and providing the necessary instruction.

Responsiblilities include:

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Operational Quality Assurance Plan Rev 12

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1 1

  • / 1. Providing NRC Reactor Operator and Senior Reactor Operator license training programs. l 19
2. Providing requested training.for Power Supply personnel working at nuclear plants.
3. Providing required training to personnel temporarily working at nuclear plants.
4. Managing and operating simulator facilities.
5. Providing requested support for Pov!er Supply internal nuclear plant training.
6. Maintaining required training records.
7. Providing fire protection training.

3.3.2.2.2 General Manager Nuclear Engineering & Construction l4 This position's responsibilities include:

1. Design, procurement, manufacture, fabrication, construction, installation, quality control, preoperational testing, and startup of new nuclear generating facilities.

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2. Implementation of assigned large projects and major modifications to operating nuclear facilities.

Positions reporting to this General Manager include: Manager Monticello Plant Projects, Manager Prairie Island Plant Projects, and Project Superintendent. l20 3.3.2.2.2.1 Manager Monticello Plant Projects / Manager Prairie Island Plant Projects These positions are responsible for the execution of projects assigned to Nuclear Engineering & Construction at the plant l 4 sites, and for providing craft labor when requested by the l 23 plants. Responsibilities include:

1. Providing full management direction for all Nuclear Engineering & Construction projects assigned at the plant  ! 4 sites.
2. Providing craft labor to support projects done under plant direction and control.
3. Assuring procedure adherence by all Nuclear Engineering & l 4 Construction and contractor employees at the plant sites 7 for Nuclear Engineering & Construction projects. 4 (O Performing Request for Engineering studies as requested l

4.

by plant mangement.

0QAP Page 13 of 88

Operational Quality Assurance Plan Rev 12 3.3.2.7.2.2 Project Superintendent l 0

20 This position is responsible for quality control for projects assigned to Nuclear Engineering & Construction. l 4 Responsibilities include:

1. Developing and implementing a quality control program for all Nuclear Engineering & Construction projects. l 4
2. Performing required inspections and tests.
3. Monitoring :nntractor quality control activities.

3.3.2.2.3 General Manager Headquarters Nuclear Group This position's responsibilities include:

1. Support to nuclear plants in licensing, safety, core analysis, and related technical areas.

Positions reporting to this General Manager include: Manager NuclearSuoportServices,ManagerNuclearAnalysis,andManagerl 21 Nuclear Technical Services.

3.3.2.2.3.1 Manager Nuclear Support Services This position is responsible for providing support to the nuclear plants in the areas of licensing administration and l 22 safety audit and assessment. Responsibilitics include:

1. Independent review functions for operating nuclear plants to verify compliance with operating license requirements as required by NRC regulations.
2. License administration for nuclear plants and liaison to the NRC Office of Nuclear Reactor Regulation.
3. Engineering and technical support to nuclear plants in nuclear safety and licensing areas.

3.3.2.2.3.2 Manager Nuclear Analysis This position's responsibilities include:

1. Core and nuclear safety analysis for all nuclear plants.
2. Licensable reload core designs for nuclear plants.
3. Technical expertise, information and direction to the Nuclear Support Services Department to ensure licensability of reload core designs, and to ensure 23, adequacy of technical specifications.

0QAP Page 14 of 88

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Operational Quality issurance Plan Rev 12

'i 7 s V' '4. Technical expertise, information and direction to the Fuel Resources Department in the formulation, coordination ~and 25 implementation of reload designs,- vendor evaluations and vendor contract negotiations.

5. Technical expertise and direction to'the Power Supply-

-Quality Assurance Department to ensure that nuclear fuel.

meets company and regulatory agency requirements.

6. Feedback and technical. expertise to the Production Training Department to ensure that necessary expercise, information and computer physics models are available for training.

3.3.2.2.3 Manager Nuclear Technical Services This position's responsibilities include:

1. Providing technical support to.the' nuclear plants.
2. -Acting in the stead of the' plant organization in performing assigned responsibilities for modification or maintenance activities at the request of operating line

. management.

3. Providing operating experience assessment.

3.3.2.2.4 Manager Special Nuclear Programs This position's responsibilities ir.clude:

1. Coordinating and managing, as applicable, the storage, shipment, and disposal of spent fuel, low level waste, and high level waste for NSP.
2. Developing corporate position on legislative and regulatory issues, other than NRC actions, which may impact nuclear operations.

3.3.2.2.5 Manager Corporate Security This position's responsibilities include:

1. Implementation of the Corporation's Security Program, which 44 includes providing trained security personnel at nuclear and non-nuclear plants.
2. Nuclear Access Authorization, general employee screening, 26 investigations and other related security services.

f) 3. Development and implementation of Nuclear Generation's v Fitness for Duty Program. 44 0QAP Page 15 of 88 j

Operational Quality Assurance Plan Rev 12 3.3.2.3 Vice President Combustion & Hydro Operations l0 o

This position is responsible for the maintenance and physical control of the company's combustion and hydroelectric power plants, and support to the nuclear plants. Responsibilities include:

1. Support of nuclear plant maintenance program.
2. Certain electrical maintenance and material testing activities.

Positions reporting to this Vice President include: General Manager Maintenance & Testing, and General Manager Plant Engineering & Construction.

l4 3.3.2.3.1 General Manager Maintenance & Testing l4 This position's responsibilities include:

1. Performance of plant electrical equipment maintenance.
2. Testing laboratory services.
3. Air filter and battery capacity surveillance testing.

Positions reporting to tnis General Manager include: Manager l Production Performances & Services, and Manager Production Plant Maintenance.

l4 4 3.3.2.3.1.1 Manager Production Performances & Services l4 This position's responsibilities include:

1. Drawing and technical manual control programs.

3.3.2.3.1.2 Manager Production Plant Maintenance This position's responsibilities include:

1. Assisting in securing maintenance contracts.
2. Providing materials and special process controls.
3. Providing technical support and council in assigned areas.

Positions reporting to this Manager include: Superintendent Materials & Special Processes. 4 0

0QAP Page 16 of 88

Operational Quality Assurance Plan Rev 12 m fm.,

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V 3.3.2.3.1.2.1 Superintendent Materials & Special Processes l 4 This position is responsible for activities associated with special processes, inservice inspection and technical support in areas of quality control for special processes and materials properties. Responsibilities include:

1. Developing, preparing, and distributing welding, heat treating and nondestructive examination procedures.
2. Certifying personnel in welding and nondestructive examination and maintaining qualification records.
3. Developing, implementing, and documenting an. Inservice Inspection Examination Program in assigned areas for nuclear plants.
4. Providing technical support to plants in the areas of metallurgy, ASME Boiler Codes, welding, heat treating 28 and nondestructive examination.
5. Providing technical instruction in welding and other special processes as required.

(* 3.3.2.3.2 General Manager Plant Engineering & Construction l 4 h This position's responsibilities include:

1. Technical support, drafting, and construction electrical l29 testing services for nuclear plants.

3.3.2.3.2.1 Manager Technical Services This position's responsibilities include:

.. Orafting services for the plants. l 3.3.2.4 Vice President Transmission & Inter-Utility Services l 6 This position is responsible for electrical system operation and

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the construction, operation, and management of the transmission system.

Positions reporting to this Vice President include: Manager System Operation. 6 3.3.2.4.1 General Manager System Operation This position is responsible for operation of the company's electrical system. Responsibilities include:

1. Coordinating electrical transmission system operation and maintenance with generating facility operation.

OQAP Page 17 of 88

Operational Quality Assurance Plan Rev 12

2. Performance of plant electrical equipment maintenance. l0 6 3.3.2.5 Director Fuel Resources l 30 This position is responsible for procurement and disposition of fuel for the Company's generating plants. Responsibilities include:
1. Procuring and delivering nuclear and diesel fuels for nuclear plants.
2. Disposition of depleted nuclear fuel.
3. Coordinate between various Power Supply sections, the eview 23 of nuclear fuel cycle design, licensing, and other safety related documents provided by the nuclear fuel vendor.
4. Negotiation and administration of nuclear fuel contracts and related nuclear fuel services contracts. l 31 3.3.2.6 DirecMr Power Supply Support l 32 This position 2 :espons1bilities include:
1. Purchasing functions for material and services.

Positions reporting to this Director include:

Supply Procurement.

Manager Power l6 3.3.2.6.1 Manager Power Supply Procurement This position is responsible for Power Supply purchasing 33 functions for material and services.

3.3.2.7 Director Power Supply Financial Operations 34 This position's responsibilities include:

1. Drawing control.

6

2. Financial operations of Power Supply.

O OQAP Page 18 of 88

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7 Operational Quality Assurance Plan Rev 12 pb 4.0 Ooerational Quality Assurance program 4.1 General Requirements

1. The Operational Quality Assurance Program shall be:
a. Documented by written Directives, Instructions, or procedures.
b. Carried out throughout plant operating life in accordance with those Directives, Instructions, or procedures.
2. The Program shall include identification of:
a. The structures, systems, and components to be covered,
b. The major organizations participating in the Program, together with the designated functions of these organizations.
3. The Program shall provide control over activities affecting the quality of the identified structures, systems, and components to the extent consistent with their importance with safety.
4. Activities affecting quality shall be accomplished under

{g} suitable controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanliness; and assurance that all prerequisites for the given activity have been satisfied.

5. The Program shall take into account the need for special controls, processes, test equiptrent, tools, and skills to attain the required quality, and the need for verification of quality by inspection and test.
6. The Program shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained.
7. The adequacy and status of the Program shall be regularly reviewed.
8. Management of other organizations participating in the Program shall regularly review the status and adequacy of that part of the Program which they are executing.

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Operational Quality Assurance Plan Rev 12 4.2 General Descriotion

1. The Operational Quality Assurance Program has been established to govern the operational activities and the activities necessary to support operation of the company's nuclear plants operated under an NRC Operating License. The Operational Quality Assurance Program is thus an overall integrated company-wide program which governs all safety related, fire p.otection related and 10CFR71 related activities as they pertain to operating nuclear plants.
2. The Program has been initiated by the President of the Company l 44 issuing a single directive to the Senior Vice Dresident Power Supply establishing him as being responsible for formulating and implementing an Operational Quality Assurance Program and identifying the program objectives.
3. The Operational Quality Assurance Program shall utilize the following documents to meet the program objectives:
a. Operational Quality Assurance Plan (Plan)
b. Administrative Control Directives (Directives) at the Corporate and Plant level
c. Administrative Work Instructions (Instructions) at the Corporate and Plant level
d. Required Procedures (Procedures) at the Plant and Department level
4. The Plan shall be considered an overall document which governs the implementing dccuments (i .e. , Directives, Instructions, and Procedures).
5. For ease of administration, implementing documents shall be issued at the following program levels:
a. Coroorate: Approving Authority, Power Supply Vice Presidents and Directors,
b. Plants (Prairie Island and Monticello): Approving Authority, Plant Manager.
c. Departments Providina Nuclear Plant Sucoort: Approving Authority, Department Manager.
6. It should be noted that the plant level Directives are controlled by the Corporate Level Directives (i.e., the Corporate Level establishes the minimum requirements associated with the Plant Level).

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Operational Quality Assurance Plan Rev 12 g

(7 4.3 Operational Quality Assurance Plan

1. The Operational Quality Assurance _ Plan shall be a document which describes in general terms how compliance with the quality requirements presented in 10CFR50 Appendix 8 and 10CFR71 Subpart H is accomplished with respect to company nuclear plants regulated by an NRC Operating License.
2. The Operational Quality Assurance Plan shall be issued under the authority of the Senior Vice President Power Supply and shall be reviewed periodically.
3. The Operational Quality Assurance Plan shall be controlled to a

assure current copies are made available to each Approving Authority of the two Directive levels of the program, to those personnel responsible for administration of the program, and to those individuals or organizations responsible for reviewing the program.

4. All changes to the Operational Quality Assurance Plan shall be approved by the Senior Vice President Power Supply or equivalent management position.

4.4 Administrative Control Directives

p d 1. Administrative. Control Of rectives (Of rectives) shall be documents which establish responsibility and requirements governing activities associated with plant operation. Directives shall be first tier implementing documents and shall receive a quality review prior to issuance. The quality review shall assure compliance with the Operational Quality Assurar.ce Program objectives. Required Directives shall be controlled and reviewed periodically.
2. Administrative Control Directives shall be issued as necessary.

It is mandatory that the Directives at the Corporate level

. assure compliance with all applicable requirements of 10CFR50 Appendix B and 10CFR71 Subpart H. The Directives issued at the plant levels are not expected to satisfy all 10CFR50 Appendix B and 10CFR71 Subpart H requirements but shall implement responsibilities assigned by the higher level Directives.

4.5 Administrative Work Instructions

1. Administrative Work Instructions (Instructions) shall be documents which provide guidelines or instructions for the implementation of the requirements of Administrative Control Directives. Instructions shall be second tier implementing documents and shall receive a quality review prior to issuance.

The quality review shall assure compliance with pertinent p Directive requirements and assigned responsibilities.

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Operational Quality Assurance Plan Rev 12 l 1

2. Administrative Work Instructions may be issued at the Corporate and Plant level. Instructions shall generally be utilized for department interfacing. Required Instructions shall be controlled and reviewed periodically.

4.6 r P_racedures

1. Procedures shall be documents which provide specific instructions for performing an activity. Procedures shall be second or third tier documents utilized to perform safety related, fire protection, and 10CFR71 related activities as required by the applicable NRC Operating License Technical Specifications.
2. Procedures shall be provided where applicable, to assure that activities important to safety are performed in the required manner. Required procedures shall be reviewed and approved as required by the applicable Technical Specifications. Approval of procedures not required by the Technical Specifications shall be by a memb1r of the responsible area management. Review of procedures not required by Technical Specifications shall be by an independent knowledgeable person. Required procedures shall be controlled and reviewed periodically.

4.7 Program Administration

1. Administration of the Corporate level of the Operational Quality Assurance Program shall be performed by the Director Power Supply Quality Assurance.
2. Administration of the Plant level of the Program shall be l44 performed by the Superintendent Quality Engineering.
3. Disputes between Quality Assurance persennel and other organizations relative to Program requirements shall be referred l 44 to the Approving Authority (as identified in section 4.2 of this Plan) responsible for establishing the pertinent requirement.
4. Program administration shall include the following activities:
a. Quality review of Directives.
b. Quality review of Instructions.

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c. Procurement review.
d. Performance of required audits. (
e. Reporting to management concerning:
1. Program status.
2. Program discrepancies including quality trends.

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Operational Quality Assurance Plan Rev 12 g

C 4.8 Program Boundary

1. The structure, systems, components, and other items requiring quality assurance are listed in Appendices A and B. The Program 44 shall also include shipment of radioactive materials as required by 10CFR71 and systems and activities associated with fire protection as identified in Appendix C.
2. An index shall be established and maintained by the Director Power Supply Quality Assurance which identifies the Directives and Instructions that are utilized to implement the requirements of ANSI N18.7-1976 that are committed to in Section 1 of this plan and the requirements identified in the remaining sections of this plan.

4.9 Quality Assurance Training Training programs shall be established for those personnel performing quality-affecting activities such that they are knowledgeable in the quality assurance documents and their requirements and proficient in implementing these requirements.

These training programs shall assure that:

1. Personnel responsible for performing quality-affecting p

sj activities are instructed as to the purpose, scope, and implementation of the quality-related Directives, Instructions, and Procedures.

2. Personnel performing quality-affecting activities are trained and qualified, as appropriate, in principles and techniques of the activity being performed.
3. The scope, the objective, and the method of implementing the training programs are documented.
4. Proficiency of personnel performing quality-affecting activities is maintained by retraining, re-examination, and/or recertification as appropriate.
5. Methods are provided for documenting training sessions describing content, attendance, date of attendance, and the results of the training session, as appropriate.
6. Fire protection training is accomplished in accordance with Appendix C.

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OQAP Page 24 of 88

Operational Quality Assurance Plan Rev 12 5.0 Modification Control 5.1 General Reautrements Modifications shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the company designates another responsible organization.

5.2 Uniform Modification Process A uniform process for controlling modifications to nuclear plants shall be provided in the Operational Quality Assurance Program.

Measures shall be established to assure that:

1. The requirements of ANSI N45.2.11-1974 are implemented.
2. Reviews and' approvals are performed.
3. Plant documentation is updated.
4. Appropriate installation procedures are prepared and utilized.
5. Tests and inspections are performed.as necessary.
6. Plant procedures are reviewed and revised as appropriate.
7. 10CFR50.59 is complied with.
8. Fire protection reviews are performed as required by Appendix C.

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M Operational Quality Assurance Plan Rev 12

.m 6.0 Procurement Document Control 6.1 General Requirements Measures shall be established to assure that applicable regulatory.

requirements, design bases, and.other requirements which~are

.necessary.to assure adequate quality are suitably included or referenced in the documents for procurement of material, equipment, and services, whether purchased by NSP or by its contractors or subcontractors. To the extent necessary, procurement documents shall require contractors or subcontractors to provide a quality assurance program consistent with the pertinent provisions of 10CFR50 Appendix 8 or 10CFR71 Subpart H.

6.2 Technical and Quality Reouirements

1. The Operational Quality Assurance Program shall contain provisions for controlling procurement of material, equipment, components, and services.that are safety, fire protection, or 10CFR71 related and utilized at or for an operating nuclear -

plant.

2. Procurement documents shall contain specific technical and quality requirements. Renewal, spare, and replacement parts g shall be required to meet the original specification (or

~ {Q properly reviewed and approved-revision) or construction code, quality assurance documentation requirements, and vendor quality

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assurance program requirements. .

3. Quality assurance requirements that are required of the Vendor shall be included. Quality assurance requirements shall be based on_ ANSI N45.2-1971 (or equivalent standard).

Documentation requirements shall include, as applicable, chemical analysis reports, material certification, testing results, and testing reports. Time and frequency of submittals should be included.

4. Procurement documents shall contain provisions which establish
, the right of access to vendor-facilities and records for source inspection and audits as appropriate.
5. Procurement documents for contracting packages for transport of radioactive materials shall require a copy of the package license, certificate, or other NRC approval authorizing use of l

the package. The procurement documents shall also require copies of all documents referred to in the license, certificates, or other NRC approval as applicable.

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Operational Quality Assurance Plan Rev 12 6.3 Review and Approval Documents, and changes thereto initiating procurement of safety related, fire protection related, 10CFR71 related material, equipment, components or services shall be approved by appropriate management personnel and shall be subject to a quality review to insure applicable regulatory requirements, design bases, quality assurance, and other requirements are adequately satisfied prior to release.

6.4 Fire Protection Procurement Control The additional procurement controls identified in Appendix C shall be applied to purchasing fire protection systems and equipment.

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-Operational Quality. Assurance P1an

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1 Rev 12 l l

(,/ . 7.0 Instructions, Procedures and-Orawings 7.1. General Requirements

1. Directives,' Instructions, procedures', and drawings'of a type appropriate to the circumstances shall .be provided for the control and performance of activities which affect quality.
2. Directives, Instructions, procedures, and drawings shall include aporopriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

7.2 Directives and Instructions Directives and Instructions shall be issued which establish procedural requirements for appropriate functional areas. Such procedural requirements shall include the following as appropriate:

1. Procedure review and approval requirements.
2. Procedure control requirements.
3. Procedure content requirements.

7.3 Procedures

1. Procedures of a type appropriate to the circumstances shall be provided for the performance of activities which affect the quality of safety related, fire protection related, or 10CFR71 related structures, systems, or components.
2. The.following procedures shall be provided:
a. Operating procedures
b. Emergency procecures
c. Surveillance test procedures
d. Routine or preventive maintenance procedures
e. Calibration procedures
f. Plant chemistry and count room procedures
g. Radiation protection orocedures
h. Emergency plan procedures i
i. Special process procedures a

0QAP Page 28 of 88

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Operational Quality Assurance Plan Rev 12

j. Preoperational & operational test procedures
k. Audit procedures
1. Fire fighting procedures
m. Document control procedures
n. Radioactive material shipment procedures
o. Inspection procedures 7.4 Drawinas and Technical Manuals Drawings and technical manuals of a type appropriate to the circumstances may be used as procedural documents for conducting activities that affect the quality of safety, fire protection, or 10CFR71 related structures, systems, or components.

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i 0QAP Page 29 of 88

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1 Operational Quality Assurance Plan Rev 12

.h,, 8.0 Document Control-8.1 General Requirements

1. Measures.shall be established to control the issuance of documents,.such as Directives, Instructions, procedures, and drawings, including changes thereto,~which prescribe activities affecting quality.
2. These measures shall assure that documents, including changes, are:
a. Reviewed for adequacy and approved for release by authorized ,

personnel, and

b. Are distributed to and used at the location where the prescribed activity is performed.
3. Changes to documents shall be reviewed and approved by the same organization that performed the original review and approval or another designated responsible organization.

8.2 Directive Control

1. Directives issued to implement the Operational Quality Assurance Program shall be controlled to assure that current copies and appropriate indexes are made available to personnel performing the prescribed activities. Directives shall be reviewed by quality assurance personnel to assure their compatibility with the Operational Quality Assurance Program objectives and shall be approved oy the designated management.
2. Changes to Directives shall be reviewed and approved in the same manner as the original.

8.3 Instruction Control

1. Instructions issued to implement provisions of Directives shall be controllea to assure that current copies and appropriate indexes are made available to personnel performing the prescribed activities. Instructions shall be reviewed by quality assurance personnel to assure that they are compatible with pertinent Directive provisions and shall be approved by designated management.
2. Changes to Instructions shall be reviewed and approved in the same manner as the original.

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Operational Quality Assurance Plan Rev 12 8.4 Procedure Control

1. Required procedures shall be controlled to assure that current copies are made available to personnel performing the prescribed activities. Required procedures shall be reviewed by a knowledgeable individual and shall be approved by a management member of the organization responsible for the prescribed activity. Required procedures shall be reviewed and approved as required by the Technical Specifications. Appropriate indexes of standing procedures shall be formulated and made available to personnel responsible for performing the prescribed activities.
2. Significant changes to required procedures shall be reviewed and approved in the same manner as the original and shall comply with the Technical Specifications.

8.5 Orawing Control

1. Drawings which represent the physical and functional aspects of the operating nuclear plants and which are critical to safe plant operation or safety of personnel shall be maintained in a current status. Appropriate indexes shall be formulated and made available to personnel responsible for plant operation, maintenance, and modification.
2. Measures shall be established for revising plant drawings and for distributing revised drawings. Proposed revisions to drawings shall be reviewed by a kr.owledgeable individual to determine the safety significance and appropriateness of the change.

8.6 Specifications Plant design specifications shall be controlled to assure that current copies and appropriate indexes are made available to personnel responsible for plant operation, maintenance, and modification.

8.7 Rjaicactive Shipment Package Documents All documents related to a specific shipping package for radioactive material shall be controlled by appropriate instructions; all significant changes to such documents shall be similarly controlled.

8.8 Updated Safety Analysis Reoorts Updated Safety Analysis Reports shall be updated in accordance with the applicable provisions of 10CFR50.

8.9 Technical Manuals Technical manuals that are used as procedural documents shall be controlled.

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Operational Quality Assurance Plan g Rev 12'

. ,a 9.0 Control of Purchased Material, Equipment and Services

.Q 9.1 General Reouirements

1. Measures shall be established to assure that purchased material, equipment and services conform to the procurement' documents.

These measures shall include provisions, as appropriate, for vendor evaluation and selection, objective evidence of quality furnished by the vendor, inspection at the vendor source, and examination of products upon delivery.

2. Documentary evidence that material and equipment conform to the procurement requ!rements shall be available at the plant site prior to ir.stallation or use of such material and equipment.

. This documentary evidence shall be retained at the plant site and shall be sufficiant to indicate that the purchased material and equipment meet the specific requirements of the codes, standards, or specificctions.

3. The effectiveness of the control of quality by vendors shall be assessed at intervals consistent with the importance, complexity and quantity of the product or service.

9.2 Quality Review IT 1. Documents initiating procurement of safety related, fire b protection related, and 10CFR71 related material, equipment and services shall be subject to a quality review to en<,ure applicable regulatory requirements, design bases, quality assurance, and other requirements are adequately satisified.

2. Quality assurance requirements shall include identification of applicable elements of ANSI N45.2-1971 (or equivalent) that are required to be included in the vendor's quality assurance program.

9.3 Vendor Evaluation and Verf fication

1. The adequacy of vt dor's quality assurance orogram specified in procurement documentation shall be verified prior to use of the procured material, equipment, or service. Vendor's adherence to their quality assurance program to the extent appropriate for the procured material, equipment or service shall be verified.
2. Vendor evaluations shall include inspections, audits, or monitoring as appropriate. These activities shall be planned and performed in accordance with written procedures based upon procurement document requirements.
3. Material and equipment may be procured and used based on appropriate certificates of conformance, provided the validity of such certificates are periodically evaluated by audits,

[k independent inspection or tests and that such certificates comply with applicable code provisions.

OQAP Page 32 of 88

Operational Quality Assurance Plan RW 12 9.4 Receiot Insoection

1. Material and equipment shall be inspected epon receipt at the plant site prior to use or storage to determine that procurement requirements are satisfied. This inspection shall include verificatic, that required documentation is complete.
2. Nonconforming material and equipnient shall be controlled to assure such material or equipment is not utilized to fulfill a safety related, fire protection related or 10CFR71 related function prior to an acceptable resolution of the discrepancies.

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o Operational Quality Assurance Plan b Rev 12

-(/ 10.0 Identification and Control of Materials, Parts and Comoonents 10.1 General Reauirements

1. Measures shall be established-for the identification and control of materials, parts and components, including partially

. fabricated assemblies. These measures shall assure that identification'of the item is maintained by heat number, part number, serial number or'other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, erection, installation and use of the item.

2. These identification and control measures shall be designed to prevent the use of incorrect or defective material, parts and components.

10.2 Soare Parts Control

1. Spare parts held for future use on safety related, fire protection related, and 10CFR71 related components shall be controlled in such a manner that assures they will perform their safety function when utilized.
2. Measures shall be taken which assures these items are in an

, appropriate condition for use or will be placed in such a y condition prior to use.

10.3 Material Control Material held in storage for use on safety related, fire protection related, and 10CFR71 related systems, structures or equipment shall be controlled in such a manner as to prevent its degradation and to assure the rejection of incorrect or defective material. This material shall be identified by heat number or other appropriate means, either on the item or on records tractable to the item. The method utilized in identification shall not significantly affect the fit, function, or quality of the item being identified.

10.4 Receipt Insoection Material, parts and components that are to be utilized to fulfill a safety related, fire protection related, and 10CFR71 related function or used for shipment of rt.dioactive materials shall be inspected upon receipt to assure that associated procurement document provisions have been satisfied. Measures shall be established for identifying nonconforming material, parts and components.

10.5 Nuclear Fuel Control

^ Measures shall be established te protect special nuclear material

against theft or diversion in accordance with applicable NRC regulations.

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Operational Quality Assurance Plan Rev 12 11.0 Control of Special Processes 11.1 General Requirements Measures shall be established to assure that special processes, including welding, heat treating, and non-destructive examination are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria and other special requirements.

11.2 Welding Procedures

1. Safety related welding and brazing shall be performed in accordance with qualified procedures. Safety related welding and brazing procedures shall be qualified in accordance with applicable codes and standards and shall be reviewed to assure their technical adequacy and approved by management.
2. Measures shall be established for controlling welding and brazing procedures that assure such procedures are qualified, reviewed and approved, as required, prior to use.

11.3 Welder Qualification

1. Measures shall be established that assure safety related welding and brazing is performed by qualified personnel. Welders and brazers shall be qualified, and requalified, in accordance with applicable codes and standards.
2. Measures shall be established for controlling welder and brazer qualification and requalification that assure qualified personnel are utilized to perform safety related welding and brazing.

11.4 Heat Treating Procedures

1. Heat treating shall be performed in accordance with procedures formulated and approved in accordance with applicable codes and standards.
2. Measures shall be established for controlling heat treating procedures that assure such procedures are qualified, reviewed, j and approved, as required, prior to use.

l 11.5 NDE Procedures

1. Safety related non-destructive examinations (NDE) shall be performed in accordance with procedures formulated in accordance with applicable codes and standards end shall be reviewed to assure their technical adequacy and approved by management.
2. Measures shall be established for controlling NOE procedures that assure such procedures are reviewed and approved, as required, prior to use.

OQAD Page 15 of 88

.g .. .. . - . . . . _ _

Operational. Quality Assurance Plan Rev 12 11.6 NDE Personnel Qualification

1. Measures shall be established that assure safety related non-destructive examinations (NDE) are performed by personneI qualified and requalified in accordance with applicable codes and standards.
2. Measures shall. be established for controlling NDE personnel qualification and requalification that assure qualified personnel.are utilized to perform safety related non-destructive examinations.

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Operational Quality Assurance Plan Rev 12 12.0 Irspection 12.1 General Reouirements

1. Measures shall be established for inspection of activities affecting quality to verify confermance with the documented instructions, procedures and drawings for accomplishing the activity. Such inspections shall be performed by individuals other than those who performed the activity being inspected or directly supervised the activity being inspected.
2. Examinations, measurements or tests of material or products shall be performed for each work operation where necessary to assure quality. If inspection of processed material or products is impossible or disadvantageous, indirect control by menitoring processing methods, equipment and personnel shall be provided.
3. Both inspection and process monitoring shall be provided when control is inadequate without both.
4. If mandatory inspection hold points, which require witnessing or inspection and beyond which work shall not proceed without prior consent are required, the specific hold points shall be indicated in appropriate documents.

12.2 Plant Ooeration

1. Measures shall be established that assure periodic inspection of safety related and fire protection systems, components, and structures. Such inspection of plant systems and equipment shall be performed to assure that such systems and equipment are in the requ1 red status and configuration. Routine general inspections of the accessible plant facilities to vreify appropriate safety measures are maintained including fire protection shall also be performed.
2. In addition, an inspection of the core shall be performed prior to startup following initial fuel leading nr refueling to assure specified fuel and reactor internal configuration.

12.3 Inservice Insoection Measures shall be established that assure inservice inspection examinations are performed in accordance with applicable provisions of the ASME Boiler and Pressure Vessel C;de,Section XI as required by 10CFR50.55 (see Sectica 13.2 relative to Inservice Inspection functional testing).

O OQAP Page 37 of 88

Operational Quality Assurance Plan Rev 12

p)

( 12.4 Inspection of Maintenance and Modifications Measures shall be established which assure that activities associated with plant maintenance and modifications are inspected when datermined appropriate by quality or other qualified personnel. Such inspections shall include verification that:

1. Appropriate procedures are available,
2. Plant equipment control exists,-
3. Applicable procedures are adhered to,
4. Qualified personnel are utilized,
5. Fire protection measures are established,
6. Radiation protection measures are established,
7. Appropriate materials and replacement parts are utilized,
8. Work is completed as required,
9. P! ant equipment is returned to service as required,

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10. Actietties'are appropriately documented, and
11. Redu, dant equipment is available.

12.5 Modifications and Non Routine Maintenance

1. Measures shall be established which assure that non routine maintenance and modification receive prior review by a qualified individual to identify applicable inspectiens. Such reviews shall include considering, (1) required meche.*ical inspections, electrical inspections, instrumentation & control inspections, structural inspections, and inspection of non-NDE special processes, (2) appropriate inspection procedures, and (3) appropriate qualification of inspection personnel.
2. Measures shall also Se established which assure that the results of identified inspections are evaluated by a qualified individual to verify their adequacy.

12.6 Technical Services Measures shall be established which assure that activities associated with technical services (such as surveillance testing, instrument calibration, laboratory services, etc.) are inspected by qualified personnel when determined appropriate by quality or m other qualified personnel.

U OQA? Page 38 of 88

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Operational Quality Assurance Plan Rev 12 12.7 Receiot Insoection Measures shall be established which assure that received items are inspected by qualified personnel (See Section 9.4).

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12.8 Vendor Insoection 1 Measures shall be established which assure that inspections and process monitoring specified in apprcpriate procurement documents for materials, components, ard equipment are performed by qualified personnel.

12.9 Fire Protection Insoections Measures shall be established which assure that fire protection '

inspections required by applic.ble Technical Specifications are performed by qualified personnel.

12.10 Radioactive Material Packages Measures shali be established which assure that packages utilized to ship licensed radioactive material off site are inspected in accordance with the applicable provisions of 10CFR71.

12.11 Emergency Eauipment Measures shall be established which assure that emergency equipment required to implement emergency plans is inspected when determined appropriate by qualified personnel.

12.12 Handlino Equioment Measures shall be established which assure that plant handling equipment (such as cranes, lift trucks, fuel handling tools) is inspected by qualified personnel and whe' determined appropriate by quality or other qualified personnel.

12.13 Insoection Procedures

1. Required inspections shall be performed in accordance with appropriate instructions, procedures, and checklists. Such instructions, recedures, and checklists shall contain a description _' objectives; acceptance criteria and prerequisites for perform'ng the inspections. These procedures shall also specify any special equipment or calibrations required to conduct the inspection. Inspection results shall be documented and evaluated by responsible authority to assure that inspection requirements have been satisfied.

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, Operational Quality Assurance Plan Rev 12

2. Where activities are to be inspected, the activity procedure shall identify hold points in the activity sequence to permit inspection. The activity procedure shall require appropriate approval for the work to continue beyond the designated hold

- point and identification of those performing the inspection.

The inspection procedure or checklist shall requirc recording the data, identification of those performing the insp aticn',

and as-found condition.

12.14 Personnel Qualif* cat.fon

1. Personnei performing required inspections shall be qualified in accordance with applicable codes, standards and training programs. Required inspections shall not be performed by individuals who performed the inspected activity or-directly supervised the inspected activity.
2. Personnel performing inspections required by sections 12.2, 12.4,12.6,12.10,12.11, and 12.12 shall be qualified based upon experience and training applicable to area of inspected activity and upon training in inspection methods.
3. Personnel' performing inspections required by sections 12.3, 12.5,12.7, and 12.8 shall be qualified in accordance with ANSI N45.2.6-1978 as modified by Regulatory Guide 1.58, -

n)

E Revision 1.

4. Personnel performing inspections required by section 12.9 shali be qualified in accordance with section 14.0 of Appendix C.

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OperaSional Quality Assurance Plan Rev 12 13.0 Test Control g 13.1 General Reouirements

1. Measures shall be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service is. identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable documents. Proof tests prior to installation, preoperational tests, and operational tests during nuclear power plant operation, of structures, systems and components shall be included as appropriate.
2. Test procedures shall include provtsions for assuring that all prerequisites for the given test have been met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmen*31 conditions. Test results shall be documented and evaluated to assure that test requirements have been satisfied.

13.2 Surveillance Tests

1. A surveillance test program shall be established to assure that testing required to demonstrate that safety related and fire protection structures, systems, and components will perform satisfactorily in service. Surveillance tests shall be identified and performed in accordance with written test procedures which incorporate the requiremer+.s and acceptance limits contained in applicable documents. The surveillance test program shall include, as a minimum, those surveillance tests specified in applicable Technical Specifications and functional Inservice Inspection testing of pumps and valves. Surveillance requirements for fire detection and protection systems in o'.her areas of the plants shall be developed using appropriate NFPA for guidance.
2. Surveillance test results shall be documented and evaluated to

&ssure that test requirements have been satisfied or deficient items satisfactorily resolved. Functional Inservice Inspection tests shall be performed by personnel qualified in accordance with applicable requirements.

13.3 Preoperational and Operational Tests Measures shall be established to assure that appropriate preoperational and operational tests are performed on safety related and fire protection related structures, systems and components that have been subject to modification or significant maintenance. Such tests shall be performed in accordance with the original design and testing requirements or acceptable alternatives. Test results shall be documented and evaluated to assure that test requirements have been satisfied or deficient items satisfacterily resolved.

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Operational QualityiAssurance Plan Rev 12 If~h

. sj 13.4 Proof Tests Measures shall be established to assure that appropriate proof tests are specified in procurement documents for safety related end . fire protection related replacement material and equipment and that such tests are performed and documented prior to installation.

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13.5 Soecial Tests Measures shall be established that assure safety related tests are reviewed and approved as required by 10CFR50.59 and applicable Technical Specifications. Such tests shall be performed in accordance with appropriate procedures. Test results shall be documented and evaluated to assure test requirements have been satisfied.

OQAP Page 42 of 88

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Operational Quality Assurance Plan Rev 12 14.0 Control of Measuring and Test Equioment 14.1 General Requirements Measures shall be established to assure that tools, gauges, instrumer.ts and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated and adjusted at specified periods to maintain accuracy within necessary limits.

14.2 Installed Plant Instrumentation Measures shall be established to assure that installed safety related plant instrumentation is maintained and calibrated at specified periods to maintain accuracy within necessary limits.

Maintenance and calibration of safety related instrumentation shall be performed in accordance with appropriate procedures and shall be controlled and documented.

14.3 Measuring and Test Instrumentation Measures shall be established to assure that tools (micrometer, calipers, etc.), gauges, instruments and other inspection, measuring, test equipment and devices used to verify conformance to established requirements are maintained and calibrated at specified periods to maintain accuracy within necessary limits. Calibration of such measuring and test equipment shall be controlled and shall W

be traceable to tiie National Bureau of Standards or where national standards are not available, the basis of calibration shall be documented.

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Operational Quality Assurance Plan Rev 12 C./ 15.0 Handling, Storage and Shipping 15.1 General Requirements

1. Measures shall be established to control the handling, storage, shipping, cleaning and preservation of safety related material and equipment in accordance with work and inspection instructions to prevent damage or deterioration.
2. When necessary for particular products, special protective environments such as inert gas atmosphere, humidity levels, and temperature levels, shall be specified and identified.

15.2 Storage Facilities Storage facilities shall be provided at each operating nuclear plant for storage of safety related and fire protection related operating and maintenance supplies, spare parts, replacement parts, replacement equipment, materials and tools. These storage facilities sSall assure physical protection and protection from environmental conditions including temperature and moisture as appropriate. Storage facilities shall be arranged and equipped to facilitate control of the stored safety related items.

p 15.3 Nuclear Fuel Storage Areas shall be provided for :torage of nuclear fuel which assure physical protection, subcritical arrangement, adequate cooling, adequate radiation shielding and containment of radioactive material as appropriate for the condition of the stored fuel.

15.4 Radioactive Material Storage

1. Areas shall be provided for storage of radioactive material which assure physical protection, as low as reasonably achievable radiation exposure to personnel, control of the s u red material, and containment of radioactive material as a m opriate.
2. Handling, storage, and shipment of radioactive material shall be controlled based uoon the following criteria:
a. Established safety restrictions concerning the handlino.

'torage, and shipping of packages for radioactive trateria shall be followed.

b. Shipments shall not be made unless all tests, certifications, acciptances, and final inspections have been completed.
c. Work vstructions shall be provided for handling, storage, r3 and shipping operations, j

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Operational Quality Assurance Plan Rev 12 15.5 Storage Control Stored material, parts and equipment shall be controlled in a manner that assures safe plant operation when and if the items are utilized. Stored safety related and fire protection related items shall be controlled to assure that the item will perform its safety function when utilized.

15.6 Material Handling Safety related material, supplies, equipment and parts shall be handled in accordance with procurement documentation and in accordance with appropriate material handling practices. Material handling equipment shall be subject to periodic testing and preventive maintenance which assures its operability. Appropriate operating instructions and procedures shall be provided for handling equipment.

15.7 Shipping and Pack: 4 ing

1. Shipping and packaging requirements shall be prepared for material, equipment, and components that are to be shipped off site and returned for use at a nuclear plant to perform a safety related function. Such requirements shall assure that the item's safety related function is not significantly degraded while in transit.
2. Shipping and packaging documents for radioactive material shall be consistent with pertinent requirements of 10CFR71.

O OQAP Page 45 of 88

' Operational Quality Assurance Plan Rev 12

()

,a 16.0 Inspection, Test and Ooerating Status

-16.1 General Requirements

1. Measures shall be established to indicate, by the use of markings such as stamps, tags, labels, routing cards, or other suitable means, the status of' inspections and tests performed upon individual items of the nuclear plant. These measures shall provide for the identification of items which have satisfactorily passed required inspections and tests, where necessary to preclude inadvertent bypassing of such inspections-and tests.
2. Measures shall also be established for indicating the operating status of structures, systems and components of the nuclear power plant, such as by tagging valves and switches to prevent inadvertent operation.

16.2 Maintenance Control

1. Measures shall be established for the control of maintenance to safety related and fire protection related structures, systems and components that assure that:

'q a. Affected structures, systems and components are removed from service and secured in a manner consistent with operability and isolation requirements of the Technical Specifications.

b. Repair and modification activities are performed in a manner consistent with its importance to safety.
c. Upon completion of repairs and modifications the affected structures, systems and components are inspected and tested to determine that the required work was performed satisfactorily and that they will perform their safety function in the required manner.
2. In addition, measures shall be established to control maintenance activities that assure resulting radiation exposure to personnel is maintained as low as reasonably achievable (ALARA) and consistent with pertinent NRC regulations.
3. The above measures shall be implemented by utilizing appropriate work authorization processes, work procedures, safety tagging, bypass control, key control and area posting, as appropriate, for the involved activity.

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Operational Quality Assurance Plan Rev 12 16.3 Test Control

1. Measures shall be established for the control of tests to safety related and fire protection related structures, systems and components that assures that:
a. Proposed tests are reviewed and approved as required prior to performance,
b. The plant is placed in an acceptable status prior to the test, maintained in an acceptable status during the test, and returned to its normal status upon completion of the test,
c. Test results are reviewed and approved as appropriate.
2. The above measures shall be implemented by utilizing appropriate work authorization processes, test procedures, safety tagging, bypass control, and key control, as appropriate, for the involved test.

16.4 Safety Tagging A safety tagging program shall be developed and utilized for control of nuclear plant equipment. This program shall contain provisions for uniquely identifying components whose operation is restricted or prohibited based upon safety considerations. Provisions shall be made for review, application, independent verification, removal, and documentation of such tagging.

16.5 Key Control Measures shall be established for controlling keys for safety related and fire protection related switches or key devices important to plant security. These measures shall include restricted distribution and periodic inventory of such keys or key devices.

16.6 Bypass Control

1. Measures shall be established for controlling the application of devices utilized to bypass component functions that are important to safety. Such measures shall assure that:
a. Proposed bypasses to safety related and fire protection related items are reviewed to determine that the plant will be placed in an acceptable status when the bypass is applied.
b. Applied bypasses are independently verified.
c. Removal of bypasses from safety related and fire protection relateil items are reviewed prior to removal .

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k;U Operational' Quality Assurance Plan Rev 12 o

'Q' d. Application of bypasses to safety related and fire protection related items is authorized by responsible personnel.

2. -The application of safety and fire protection related bypasses shall be considered a temporary measure and shall be reviewed periodically.
3. All_ required activities associated with the application, review, approval and removal shall be documented.

16.7 Radioactive Material Control Inspection, test, and operating status of equipment and components associated with shipment of radioactive material shall be established based upon the following criteria:

1. Inspection, test, and operating status of packages for radioactive material shall be indicated and controlled by established procedures.
2. Status shall be indicated by tag, label, marking.or log entry.
3. Status of non-conforming parts or packages shall be O positively maintained by established procedures.

. 16.8 Reactor Startuo and Restart Control Measures shall be established for controlling reactor startups and restarts. Such measures shall assure that safety related systems, components, and structures have been placed in the required status and reviews have been completed to assure that the cause of any reactor trips (scram) has been investigated and satisfactorily resolved.

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Operational Quality Assurance Plan Rev 12 17.0 Nonconforming Materials, Parts or Comoonents 17.1 General Requirements

1. Measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. These measures shall include, as appropriate, procedures for identification, documentation, segregation, disposition and notification to I affected organizations.
2. Nonconformance items shall be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures.

17.2 Receiot Insoection

1. Measures shall be established which assure that safety < elated and fire protection related material, supplies, equipment, and components are inspected to determine that they conform to specified requirements of pertinent procurement documents upon raceipt at the plant site. The absence of required documention or discrepant documentation shall constitute nonconformance.
2. Provisions shall be made for identifying nonconforming items and for segregation of nonconforming items. Nonconforming items shall not be used to fulfill a safety related and fire protection related function until the discrepancy is satisfactorily resolved.

17.3 Maintenance Insoection Equipment, components or parts found nonconforming in a manner that could significantly affect its ability to fulfill its safety and fire protaction related function shall be identified as a nonconforming item and shall be segregated. Nonconforming items shall not be used until the discrepancy is satisfactorily resolved.

17.4 Disposition of Nonconforming Items

1. Measures shall be established which assure that nonconforming items are disposed of in a manner which prohibits their inadvertent use or installation. Provisions shall be made for reviewing the nonconformance and correcting discrepancies by repair or rework if appropriate.
2. The acceptability of such rework or repair of materials, parts, components, systems, and structures shall be verified by reinspection and retesting the item as originally inspected and tested or by a method which is equivalent to the original inspection and testing method. Inspection, testing, rework, and repair procedures shall be documented.

OQAP Page 49 of 88

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_ Operational Quality Assurance' Plan Rev 12

/T t, ,/ 3. Normally, nonconforming safety and fire protection related items shall not be installed prior to satisfactory resolution of outstanding discrepancies. In exceptional cases nonconforming items may be installed provided specific action is taken, which assures the item is not utilized to fulfill a safety function, prior to resolution of the discrepancy.

17.5 Nanconformance Documentation

1. Nonconformance reports shall tre initiated for significant deviations from specified requirements. Such reports shall identify the nonconforming item, describe the nonconformance, the disposition of the nonconformance, and the inspection requirements. Nonconformances shall be reviewed and approved by appropriate quality personnel.
2. Nonconformance reports shall be periodically analyzed to show1 quality trends and_the results of this review shall be reported to the appropriate level _of management for review and assessment.

17.6 Reporting Measures shall be established which assure that defects as defined in 10CFR21 and failures to comply with the Atomic Energy Act of

(,)

fs 1954, as amended, or any applicable rule, regulation, order or license of the NRC relating to a substantial safety hazard are reported in accordance with the applicable requireme.'ts of 10CFR21.

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, Operational Quality Atsurance Plan Rev 12 18.0 Corrective Action 18.1 General Requirements

1. Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, discrepancies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition.
2. The identification of the condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.

18.2 Operating Occurrences and Events Measures shall be established which assure that operating occurrences and events that could have a significant safety effect are investigated, reviewed, and reported as required by the Technical Specifications. Such measures shall assure that appropriate corrective action is taken and that the event or occurrence is reported to responsible levels of management.

Corrective action includes provisions which preclude recurrence.

18.3 Administrative Control Discrecancies Measures shall be established which assure that significant discrepancies identified during quality assurance program audits are reported to those responsible for the activity and to appropriate levels of management. These measures shall include corrective action designed to preclude repetition of the discrepancies identified and verification of implementation.

O 0QAP Page 51 of 88 L

Operational Quality Assurance Plan Rev 12 19.0 Quality Assurance Records 19,1 General Recuirements

1. Sufficient records shall be maintained to furnish evidence of activities affecting quality. These records shall include at least the following:
a. Operating logs and the results of reviews, inspections, tests, audits, monitoring of work performance and material analysis,
b. The records shall also include closely related data such as qualifications of personnel, procedures, and equipment.
c. Inspection and test records shall, as c. minimum, identify the inspector or data recorder, the type of observation, the results, the acceptability, and the action taken in connection with any deficiencies noted.
2. Records shall be identifiable and retrievable.
3. Requirements shall be established concerning record retention, such as duration, location, and assigned responsibility which are consistent with applicable regulatory requirements.

,9

) 19.2 Ooerating Records Measures shall be established which assure that records as they apply to plant operation are generated and retained a; required by the Technical Specifications or other regulatory requirements.

19.3 Plant Modification Records Measures shall be established which assure that adequate records are generated and retained to reconstruct plant modifications that are safety related or fire protection related.

19.4 Plant Maintenance Records Measures shall be established which assure that records pertaining to maintenance of plant safety related and fire protection related structures, equipment and components are generated and retained.

19.5 Personnel Qualification Records Measures shall be established which assure personnel qualification records are generated and retained.

19.6 Procurement Records (g

f

)

Measures shall be established which assure that safety, fire protection, or 10CFR71 related procurement documents and associated documents are generated and retained.

OQAP Page 52 of 88

Operational Quality Assurance Plan Rev 12 19.7 Surveillance Test Records Measures shall be established which assure that records associated with Surveillance Testing including Inservice Inspections, are generated and retained.

19.8 Audit Reoorts Measures shall be established which assure records pertaining to audits of quality activities are generated and retained.

19.9 Radioactive Material Control Measures shall be established which assure that records associated with radioactive material control are generated and maintained.

19.10 Drawings Measures shall be established which assure that records of drawing changes made to plant safety related and fire protection related structures, equipment and components are generated and retained.

19.11 Records Management

1. Records management systems shall be established which assure that the required records are collected, stored, and maintained in accordance with ANSI N45.2.9-1974. Such records shall at least be stored in insulated filing devices (rated 350-1 hour by UL as to fire resistance only) located in an areas having combustible loaaing of less than 5 lb/sq ft, or duplicate records shall be maintained in remote locations.

Specific records shall be identified in implementing or source documents. Identification shall indicate records required by Technical Specifications, committed to standards, and other regulatory documents.

2. Records management systems shall be established which assure that those records used to demonstrate program implementation are collected, stored, and maintained in accordance with good records management practices. Such systems shall assure that these records are made available to auditors and inspectors in a timely manner.

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' Operational Quality Assurance Plan

-Rev 12 20.0 Audits 20.1 General Requirements A comprehensive system of planned and periodic audits shall be carried out to verify compliance with all a.c.pects of the Operational Quality Assurance Program and to determine the effectiveness of the program. The audits shall be performed in accordance with written procedures or checklists by appropriately trained personnel not having direct responsibility in the areas being audited. Audit results shall be documented and reviewed by management having responsibility in the area audited. Followup action, including re-audit of discrepant areas, shall be taken where indicated.

20.2 Required Audits Measures shall be established which assure that the provisions of the Operational Quality Assurance Program are audited periodically.

In addition, an overall audit shall be performed periodically which determines the adequacy of the program with respect to requirements contained in the Operational Quality Assuranco Plan. This overall audit shall be performed by an organization other than that responsible for administration or implementation of the program.

20.3 Audit Schedules

1. Required audits shall be performed each year except that this time period may be extended to not more than two years provided such extensions are justified based upon past experience.

Special audits may be scheduled on the initiative of quality assurance personnel based upon suspected or known discrepancies or as directed by, management.

2. Appropriate audit schedules shall be prepared each year.

20.4 Audit Procedures Required audits shall be performed in accordance with appropriate audit procedures. Checklists may be used as audit procedures or in conjunction with audit procedures. Procedures shall include auditing requirements at various levels of the Operational Quality Assurance Program.

20.5 Audit Reports

1. Reports of the results of each audit shall be prepared. These reports shall include a description of the area audited, identification of individuals responsible for implementation of the audited provisions and for performance of the audit, identification of discrepant areas, and recommended corrective

- action as appropriate, a

0QAP Page 54 of 88

Operational Quality Assurance Plan Rev 12

2. Audit reports shall be distributed to the appropriate management level and to those individuals resoonsible for implementation of audited provisions.
3. Audit reports and associated nonconformance reports shall be periodically analyzed for quality trends and the results reported to the appropriate level of management for review and assessment.

20.6 Corrective Action Measures shall be established which assure that discrepancies identified by audits or other means are resolved. These measures shall include notification of the manager responsible for the discrepancy, recommended corrective action, and verification of satisfactory resolution. Discrepancies shall be resolved by the manager responsible for the discrepancy. Line management shall resolve disputed discrepancies.

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OQAP Page 55 of 88 l

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' Operational Quality-Assurance Plan-Rev 12 ' Appendix A

' APPENDIX A-

- 'Monticello Structures, Systems, and Comoonents Subject to Appendix B of 10CFR50

. 1. STRUCTURES w.

Reactor Building Plant Control'and Cable Spreading Structure Off Gas Stack

Intake Structure (Service Water pump' area)

Ofesel Generator Building '

Diesel Fuel 011 Day Tank Rooms.

Turbine Building (parts which house, support and/or protect safety related equipment).

Off Gas Compressor and Storage Building (parts which house, support and/or protect safety related equipment)-

. 2. MECHANICAL SYSTEMS AND COMP 0NENTS COMPONENTS I

1- Reactor Coolant System -

Reactor Vessel Reactor Vessel Support Skirt Reactor Vessel Stabilizer Recirculation System Piping Recirculation System Pumps and Valves Main Steam Piping (to and including outermost '

containment isolation valve)

Main Steam Safety Relief Valves Main Steam Safety Relief Valve Otscharge Piping Feedwater Piping (to and including outermost containment isolation valve)

Control Rod Drive Housing Supports Reactor Vessel Internals +

Fuel Assemblies i Core Support Structure i Jet Pumps Control Rods L Liquid Poison Pipe

Core Spray Sparger ,

Control Rod Drive System Control Rod Drives g Control Rod Drives Accumulators s Scram Discharge Volume Scram Piping OQAP Page 56 of 88

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, Operational Quality Assurance Plan Rev 12 Appendix A COMPONENTS Standby Liould Control System SLC Tank SLC Pumps SLC Explosive Valves SLC Piping Primary Containment Drywell Torus Drywell Vent Piping / Vacuum Breakers Torus Ring Header and Downcomers Containment Penetrations Containment Piping and Valves (to and incitding outermost isolation valve)

Secondary Containment RB Ventilation Isolation Dampers Standby Gas Treatment Filters and Fans Residual Heat Removal System RHR Piping, Pumps and Valves RHR Heat Exchangers (shell side)

Core Soray System Core Spray Piping, Pumps, and Valves High pressure Coolant Injection System HPCI Steam Piping and Valves Inside Containment (to and including outermost isolation valve)

HPCI Steam Supply and Exhaust Piping and Valves (outside containment)

HPCI Pump-Turbine HPCI Injection Piping and Valves HPCI Suction Piping Reactor Core Isolation Cooling System RCIC Steam Piping and Valves Inside Containment (to and including outermost isolation valve)

RCIC Steam Supply and Exhaust Piping and Valves (outside containment)

R01C Pump-Turbine RCIC Injection Piping and Valves RCIC Suction Piping 0QAP Page 57 of 88

N n Operational Quality Assurance Plan Rev 12 Appendix-A

,m.

( ) Service Water System Emergency Service Water Pumps Emergency Service Water Piping and Valves RHR Service Water Pumps

-RHR Heat Exchanger (tube side)

RHR Service Water Piping & Valves ,

EFT Emergency Service Water Pumps, Piping, and Valves Reactor Water Cleanuo System RWCU Piping and Valves (to and including outermost isolation valve)

Spent Fuel Storace Systems Spent Fuel Pool Otesel Generator Sucoort System Air Start System from Receivers to Air Start Solenoids Fuel Oil System from Day Tank to Injectors Diesel Coolers and Associated Piping and Valves (water side)

Diesel Fuel Oil Heating and Ventilatino System Emergency Filtration Train (EFT) for Control Room and EFT Builaing Combustible Gas Control System SECTION 2 NOTES:

1. Mechanical components included within each mechanical system include hangers (up to and including the first anchor supporting a safety related section of piping), fittings, flanges, vessels, tanks, etc.

as necessary to perform the system safety functions.

3. ELECTRICAL SYSTEMS AND COMPONENTS COMPONENTS 4160 Volt Bus 15 Breaker 152-602, 601, 609, 610, Feed Breaker 408 RHR Service Water Pump B Motor RHR Service Water Pump D Motor -

RHR Pump B Motor RHR Pump D Motor Core Spray Pump B Motor 480 Volt Load Center 104

, No. 12 CR0 Pump Feed Breaker 152-606 0QAP Page 58 of 88

Operational Quality Assurance Plan Rev 12 Appendix A 4160 Volt Bus 15 Breaker 152-502, 501, 509, 511, Feed Breaker 308 RHR Service Water Pumo A Motor RHR Service Water Pump C Motor RHR Pump A Motor RHR Pump C Motor Core Spray Pump A Motor 480 Volt Load Center 103 No. 11 CR0 Pump Feed Breaker 152-506 480 Volt Switchaear Load Center 104 480 V MCC 142 (1) (Essential) 480 V MCC 143A (1) 480 V MCC 1438 (1) 480 V MCC 144 (1) 480 Volt Switchgear Load Center 103 480 V MCC 133A (1) 480 V MCC 133B (1) 480 V MCC 134 (1)

Diesel Generator No. 11 (1)

Diesel Generato- No. 12 (1) 250 V DC (Division 1) Distribution _ Panel 031 (1) 250 V OC (Division 2) Distribution Panel 0100 (1) 125 V DC Distribution Panel 0-11 (1) 125 V DC Distribution Panel 0-21 (1) 120/240 Volt AC Instrumentation Distribution Panel (1)

SECTION 3 NOTES:

1. For those electrical systems or components designated with the Note (1) above, Quality Assurance electrical program requirements are applicable only to those portions of systems as defined in Section 2 as necessary to perform the system safety function.
2. Electrical components included within each electrical system include power source, breaker, control circuit, cable, relaying and operating device (motor, solenoid, heater, relay, etc.) as necessary to perform the system safety function.
3. Certain components are excluded from the QA program requirements if they nieet the criteria described in Section 5.

00AP Page 59 of 88

Operational Quality Assurance Plan Rev-12 Appendix A

( 4. INSTRUMENTATION SYSTEMS AND COMP 0NENTS Reactor Protection System Primary Containment Isolation System High Pressure Coolant Injection, System Initiation and Isolation Reactor Core Isolation Coolino System Initiation and Isolation Core Spray System Initiation Low Pressure Coolant Injection System Initiation Automatic Blowdown System Neutron Monitoring System (IRM and APRM)

Standby Gas Treatment System Initiation SJAE Off Gas Radiation Monitor EFT System Initiation and Operation Combustible Gas Control System SECTION 4 NOTES:

1. For those instrumentation systems designated above, Quality Assurance instrumentation program requirements are applicable only to those portions of systems defined in Section 2 as necessary to perform the system safety function.
2. Instrumentation components included within each instrumentation system include power supply, sensors, relays, wiring and final operating device (solenoid, relay, etc.) as necessary to perform the -

system safety function.

3. Certain components are excluded from the QA program requirements if they meet the criteria described in Section 5.
5. ELECTRICAL AND INSTRUMENTATION SYSTEM COMPONENT EXCLUSION CRITERIA
1. Any component of an electrical system Section 3 or instrumentation system Section 4 is excluded from the QA Program requirements if it meets the following criteria:
a. A failure of the component by electrical shorting, open circuiting, grounding or mechnaical failure would not render the system incapable of performing its intended safety function.
b. A failure of the fluid pressure boundary of the component would not render the system incapable of performing its intended safety function.
2. Small spare parts having no traceability, such as commercial off-the-shelf items, may be purchased as nonsafety-related and then qualified for use in equipment requiring Quality Assurance.

Examples of such items are resistors, capacitors, switches,

, indicators, coils, wire, connectors, solid state devices and

[-

x miscellaneous hardware.

OQAP Page 60 of 88

Operational Quality Assurance Plan Rev 12 Appendix B APPENDIX C Prairie Island Structures, Systems, and Components Subject to Accendix B of 10CFR50

1. REACTOR SYSTEM AND FUEL A. Reactor Vessel and Coolant System Reactor vessel Reactor vessel support Reactor vessel internals Full length control rod drive mechanism housing Part length control rod drive mechanism housing Steam generator (tube side & shell side)

Pressurizer, including instrumentation, piping, and components Reactor coolant hot and cold leg piping, fittings Surge pipe, fittings Loor bypass line Te .-ature detector bypass manifold Rei .)r coolant thermowell Rea nor coolant thermowell boss Safety valves Relief valves Reactor coolant system boundary valves Control rod drive mechanism head adapter plugs Reactor coolant pump Pump casing Main flanges Thermal barrier Seal housing Pressure retaining bolting Reactor coolant pump motor Shaft coupling Flywheel Reactor coolant pump internals RCC thimble plug (rod control clusters)

Primary and secondary sources Electric modules with safety function Cable with safety function B. Fuel Assemblies Fuel assemblies, sub-assemblies, components and materials, including fuel material O

00AP Paga 61 of 88

E' Operational Quality Assurance Plan

n Rev 12 Appendix B v5 2.

)] REACTIVITY CONTROL SYSTEMS Orive mechanisms including:

Control rod cluster drive shaft assembly, including latch assembly Reactor trip breakers Control rods and rod cluster assemblies Control rod guide tube Centrol rod drive housing' Electric modules with safety function Cable with safety function

3. CHEMICAL AND VOLUME CONTROL SYSTEM Regenerative heat exchanger-Letdown heat exchanger Reactor coolant filter 2 Volume control tank Dositive displacement charging pump and motor Seal water filter-Letdown orifices and letdown valves Excess letdown heat exchanger Seal water heat exchanger Boric acid tanks 4 -

Boric acid transfer pump Boric acid filter Reactor coolant pump seal and bypass orifice Piping, inboard of isolation valves Electric modules with safety function Cable with~ safety function Heat tracing

4. INCORE INSTRUMENTATION Thimble guide tubes Seal table
5. BORON RECYCLE SYSTEM Recycle holdup tanks, piping and valves associated with gaseous radioactive waste
6. EMERGENCY CORE COOLING SYSTEM Accumulators High head safety injection pumps Piping, inboard of isolation valves Motors, electric modules, with safety function Cable with safety function O

0QAP Page 62 of 88

Operational Quality Assurance Plan Rev 12 Appendix B

7. CONTAINMENT SPRAY SYSTEM Refueling water storage tank Spray additive tank Spray pumps Spray rings and nozzles Piping and valves Pump motors Electric modules with safety function Cable with safety function
8. RESIDUAL HEAT REMOVAL SYSTEM Pumps and motors Heat exchanger Piping and valves with safety functian Electric modules with safety function Cable with safety function
9. SPENT FUEL POOL COOLING SYSTEM Piping and valves whose failure could result in significant release of pool water
10. CONTAINMENT FAH COOLER SYSTEM Ductwork Fans Dampers Fan coolers Electric modules with safety function Cable with safety function
11. WASTE PROCESSING SYSTEM Gaseous and liquid waste piping and valves forming part of cont.ainment boundary l Systems handling gaseous radioactive materials Electric modules 'c'th safety function Cable with saf.'.y function The Waste Gas Disposal System shall be maintained in accordance with the guidance established in Regulatory Guide 1.143 Revision 1 October 1979.
12. SAMPLING SYSTEMS Valves and piping from the reactor coolant system up to the second isolation valve outside containment Valves and piping to the first isolatian valve from other safety related systems OQAP Page 63 of 88

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',- ;0perational Quality' Assurance Plan Rev 12 Appendix B .

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(A 13. STEAM GENERATOR BLOWOOWN Piping from iteam generator to containment isolation valves

14. REACTOR PROTECTION SYSTEM l Electrical modules Cable a
15. PROCESS RADIATION MONITORS Radiation Monitors, including electric modules and cable with a safety

~3 function, associated with the Shield Building, Auxiliary Building, Spent Fuel Pool and Control Room Ventilation Systems

16. CONTAINMENT HYDROGEN CONTROL SYSTEM Piping and valves with safety function Electric modules and cable -
17. REACTOR VESSEL SERVICE EQUIPMENT Containment polar crane Vessel head handling equipment O Crane structural supports V

~

Crane electrical, cable, controls and instrumentation with safety function

18. REFUELING EQUIPMENT Spent fuel cask '

Auxiliary Building crane Auxiliary Building crane structural supports Crane electrical, cable contrcls and instrumentation with safety function Fuel transfer tube Spent Fuel Bridge Crane Manipulator Crane

19. FUEL STORAGE New fuel racks  ;

Spent fuel racks Spent fuel pool struc w ? and enclosure

20. CONTROL li PANELS

~

Electric modules, with safety function Cable with safety function O

V 00AP Page 64 of 88

Operational Quality Assurance Plan Rev 12 Appendix 8

21. LOCAL PANELS AND RACKS Electric modules with safety function Cable with safety function
22. MAIN STEAM SYSTEM Main steam piping and valves from steam genera':oes up to and including piping restraints downstream of the main steam isolation valves Main steam piping and valves from main steam lines to auxiliary feedwater pump turbine Steam line flow restrictor Safety and relief valves Piping to first isolation valves and safety and relief valve J scharge Electric modules with safety function Cable with safety function
23. FEE 0 WATER SYSTEM Feedwater piping and valves inside containment structure up to and including first isolation valve outside containment structure Electric modules with safety function Cable with safety function
24. AUXILIARY FEE 0 WATER SYSTEM Piping and valves supplying auxiliary feedwater from and including containment isolation valves to connections with feedwater lines Auxiliary feedwater pumps (turbine and motor-driven)

Piping and valves supplying auxiliary feedwater from the cooling water system Electric modules with safety function Cable with safety function

25. COOLING WATER SYSTEMS Component cooling water systems (essential)

Piping (except to turbine building and non-essential equipment)

Heat exchangers, with safety function Pumps Pump motors Surge tank Valves, isolation 1

Valves, other, with safety function Electric modules with safety function l Cable with safety function Cooling water systems (essential)

Piping (except to tur'ine o building and non-essential equipment)

Diesel engine pumps Strainers, with safety function Valves, isolation O ;

l 00AP Page 65 of 88 l

l

.- - - .. = . . - _ . . . - . -... . . . . - . _.

Operational. Quality Assurance Plan Rev 12 Appendix B

~Q b Valves,'other, with safety function 1 Screen wash systems, with safety function Trave _ ling screens, with safety function

- Electric modules with safety function Cable with safety function Diesel engine pump auxiliaries as follows:

Diesel oil storage tanks Day tanks Fuel oil transfer. pumps and motors Fuel oil piping and valves with a safety function Starting air compressors Air receivers ~

Starting air piping and valves with a safety function Cooling water piping and valves with a safety function Electric modules with safety furetion Cable with safety function Diesel engine, lubricating oil and jacket cooling systems Diesel-fuel oil 26 INSTRUMENT AIR SYSTEM Piping and valves associated with containment penetration

27. DIESEL GENERATOR Diesel oil storage tanks Day tanks-Pumps and motors, fuel oil transfer Diesel filter Valves, with safety functicn Piping except vent and fill piping downstream of last valve Cooling water system pipe and valves Diesel generator jacket cooling water system Diesel generator lubricating oil system Air intake Electric modules with safety function Cable with safety function Diesel fuel oil
28. DIESEL GENERATOR AIR STARTING SYSTEM Compressor Air receivers Piping and valves from receiver to diesel generator Piping between compressor and receiver OQAP Page 66 of 88

Operational Quality Assurance Plan Rev 12 Appendix B

29. ELECTRICAL, CLASS 1E SYSTEMS Switchgear, transformers, motor control centers, load centers, batteries and chargers, and associated equipment with safety function NOTE: Point of interface with onsite electric power systems, i.e., at l point of interface with Class 1E breakers which isolate main Class 1 1E cnsite buses from the offsite power system; and including i components and circuitcy interfaces that affect the proper performance of such interfacing breaker.

4,160 - 480 V switct: gear from engineered safety systems (ESF), l includtr.g ESF buses l 4,160 - 480 V transformers (ESF load centers) l 480 - 120/208 V transformers (control room and ESF area emergency 1 lighting) 480 V switchgear (ESF load centers) 480 V motor control and motor control centers 125 V station batteries and racks (control and vital instrumentation power supplies) 125 V de panels and switchgear (vital de power distribution) 120 V ac instrument bus panels (vital instrumentation ac power distribution)

Containment penetration assemblies Main control board Radiation monitor panel Hot shutdown panel Control room air conditioning control panel Post LOCA Hydrngen control panel Emergency Itghting Emergency communications Diesel generator and accessories Diesel generator control panels Relay boards and racks Wire and cable raceway system Underground electrical duct bank system Cable system (power, control and instrumentation)

Instrument racks Electrical supports Heat tracing / free:e protection

30. INSTRUMENTATION AND CONTROL SYSTEM COMPONENTS _

Reactor trip system Engineered safety features (ESF) actuation ,ystem Systems required for safe shutdown Safety related instruments, tubing and fittings l

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Operational Quality Assurance Plan

'Rev 12 Appendix B A $

31. HEATING, VENTILATION AND AIR CONDITIONING SYSTEMS (HVAC)

Control and Relay Room HVAC System Air handling units Fans, ductwork and dampers Filters Chillers and chilled water pumps Auxiliary Building Special Ventilation System Fans, ductwork and dampers

. Filters Screenhouse Ventilation System Fans and dampers associated with diesel engine ventilation Shield Building Ventilation System Fals, ductwork and dampers F;iters Battery Room Special Ventilation System Fans, ductwork and dampers Spent Fuel Pool Special Ventilation System Exhaust fans, ductwork, dampers Exhaust filters Diesel Generator Rooms Cooling System Fans, ductwork'and dampers Auxiliary Building Normal Ventilation System

()

k/

Ductwork and dampers associated with steam exclusion Turbine Building Ventilation System Ductwork and dampers associated with steam exclusion Electric modules with safety function Cable with safety function HVAC sensors and monitors having safety functic,

32. CIVIL STRUCTURES AND FOUNDATIONS Containment and structures Containment airlocks Containment isolation (valves, piping, canisters)

Containment penetrations Shield building Auxiliary building Control room Diesel generator room Radwaste building Cooling water intake structure Electrical tunnels, with safety function Pipe tunnels, with safety function Shielding structures Turbine Building (housing emergency diesel generator, cooling water pipes, batteries, safeguards switchgear, auxiliary feedwater pumps) b a

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Operational Quality Assurance Plan Rev 12 Appendix B

33. OTHER
1. Fire protection system piping associated with the safeguards ventilation exhaust filters and containment penetration
2. Turbine building crane Crane structural support Crane electrical, cable, controls and instrumentation with safety function O

t 0

00AP Page 69 of 88

(

Operational Quality Assurance Plan Rev 12 Appendix C f

()/ - APPENDIX C-Nuclear Plant Fire Protection Proaram 1.0 Policy Statement Northern States Power Company (NSP) has e:,tablished a system of Administrative Control Directives (ACOs) that implement the Operational Quality Assurance Plan. This system shall be used to implement the requirements of the operating nuclear power plant fire protection program. The basic requirements of the fire protection program are

, specified in this appendix to the Operational Quality Assurance Plan.

2.0 Organization 2.1 General Requirements

1. NSP'shall be responsible for the establishment and implementation of the fire protection program. NSP may delegate to other organizations the work of establisding and implementing the fire protection program, or any part thereof, but shall retain responsibility for the program.

O V 2. The authority and duties of persons and organizations. involved in the fire protection program shall be clearly established and delineated in writing.

3. To assure adherence to the fire protection program, management measures shall be established which provide that the individual or group assigned the responsibility for checking, auditing, inspecting, or otherwise verifying that an activity has been correctly performed is independent of the individual or group directly responsible for performing the specific activity.

2.2 Fire Protection Organization Summary The NSP organization is summarized in Section 3.0 of the Operational Quality Assurance Plan. In addition to that summary, the following additional responsibilities shall pertain to the fire protection program.

1. Director Power Supply Quality Assurance
a. Scheduling and assuring completion of independent off-site fire protection inspections and audits,
b. Reviewing non plant (other than Nuclear Engineering and Construction) purchase requisitions related to fire protection.

' (D U 2. Plant Managers 0QAP Page 70 of 88

k Operational Quality Assurance Plan Rev 12 Appendix C

a. Routine inspection of the plant for fire hazards,
b. Establishing plant fire brigades,
c. Procurement of equipment for the fire brigades,
d. Ensuring that fire brigade membert receive required training and physical evaluations.
e. Coordinating fire dr;11s and determining their effectiveness.
f. Establishing cooperation with the local fire department, including joint drill. and training sessions to familiarize fire department personnel with plant access routes, layout, equipment, and special hazards.
g. Establishing storage requirements to insure no additional fire hazards are created.
h. Establishing a surveillance program for fire protection systems and fire fighting equipment.
1. Establishing a system to control nonconforming items.

J. Reviewing require'd work processes for fire hazards and possible reduction of fire protection system effectiveness.

k. Reviewing modifications to determine if they would cause an unreviewed fire hazard or reduce the effectiveness of the fire protection systems.
1. Establishing a fire salvage program (when required).
m. Reviewing purchase requisitions initiated by the plant and Nuclear Engineering and Construction that are related to fire protection.
n. Developing instructions for fighting fires in specific areas and identifying effects of fires in specific areas.
o. Establishing a policy for the security actions to be taken by the guard force during a fire.
p. Preparing news release information for NSP's Communications Department.
3. Manager Production Training
a. Establishing a training program for the fire brigades.

O 0QAP Page 71 of 88

~

- -0perational. Quality Assurance Plan Rev.12 Appendix _C 2.3 Fire Protection Engineer

1. A. fire protection engineer (or engineering consultant) shall be used to. provide the following types of services:
a. Review of design for a .significant modification to a_ fire

. protection' system.

b. Review of proposed plant modifications which would introduce major hazards--not analyzed in the Fire Hazards Analysis.
c. Tr.iennial independent fire protection inspections (see section 14.2).
2. The fire protection engineer (or engineering consultant) shall meet the following qualifications:
a. A graduate of an engineering curriculum of accepted standing.

who has completed not less than six years of engineering.

attainment indicative of growth in engineering competency and achievement, three of which shall have been in responsible charge of fire protection engineering work, or

b. A member in ht e Society of Fire' Protection Engineers.

3.0 Nuclear Plant Fire Brigades

.3.1 Monticello

1. A fire brigade shall be established in accordance with the requirements of 10CFR50 Appendix R Section III.H, and the l35~-

requirements listed below:

a. Fire. brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate the unexpected absence of fire brigade members. Under this circumstance, immediate action shall be

,. taken to restore the fire brigade to within the minimum requirements.

3.2 Prairie Island

1. A fire brigade of five persons shall be on-site at all times in addition to the minimum shift crew complement nee:bd to safely t

shut down the unit (s).

a. Fire brigade composition may be less than the minimu.:

requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate the unexpected absence of fire brigade members. Under this circumstance immediate action shall be taken to restore the fire brigade to within the minimum O requirements.

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Operational Quality Assurance Plan Rev 12 Appendix C

2. Each fire brigade shall have an appointed leader. This leader shall not be the Shift Supervisor (the Unit No.1 Shif t Supervisor at Prairie Island).
3. All new members of the fire brigades shall have an initial physical examination for strenuous physical activity as experienced in fire fighting. Annual followup physical examinations shall include respiratory protection qualification testing which screens all respirator users (including fire brigade members) for cardiopulmonary deficiencies. Physical examinations shall be conducted by a physician. A program shall be established by NSP's corporate physician to ensure that all respirator users, when subject to even the most severe working conditions, are physically fit to wear a respirator. The program shall include pulse, blood pressure, and spirometry testing, and a medical history review in which the possibility of past or present heart disease is determined. The program shall be administered by nursing personnel who will perform the necessary cardiopulmonary screening function.

4.0 Fire Protection Training l 4.1 Monticello and Prairie Island

1. Level I fire protection training shall be general training given to operations and maintenance personnel assigned to nuclear power plants. Following initial training, these topics shall be reviewed at least annually with required personnel. Level I shall cover, as a minimum, the following areas:
a. Basic principles of fire chemistry and physics.
b. Fire hazards.
1) Common fire hazards
2) Combustibles, general
3) Flammable liquids
4) Flammable gases
c. Fire detection systems
d. Types of extinguishing systems
e. Special fire hazards associated with nuclear power plants
f. Emergency Plan w"th emphasis on fire emergency
2. Basic instruction in fire protection shall be given to contractor personnel before granting them unescorted access to l safety related areas of the plant.

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Operational Quality Assurance Plan Rev 12 Appendix C

.j \.

V 4.2 Monticello

1. Fire brigade training shall be conducted in accordance with the requirements of Appendix R Section I.

4.3 Prairie Island

1. Level II fire protection training shall be given to all fire brigade members. An initial training program with annual retraining shall be conducted. Retraining shall repeat all level II subject material over a period of approximately two years.

Level II shall include a detailed treatment of the subject matter in Level I. In addition, the following items shall be i covered:

a. The identification and location of fire hazards and associated types of fires that could occur in the plant.
b. The identification and location of fire fighting equipment in each fire area.
c. Familiarization with layout of the plant including access and egress routes in each area.
d. The proper use of fire fighting equipment.
e. Methods of fighting each type of fire.
f. Review of the plant fire fighting strategies with specific coverage of each individual's responsibilities.
g. Proper use of communication, lighting, ventilation and emergency breathing equipment.
h. Considerations of radiation and contamination in fire areas.
2. Level III training shall be presented to the fire brigade leaders. Initial training with annual retraining shall be provided. Included will be a detailed review of Level I and II training and the following additional material:

I

a. The direction and coordination of fire fighting activities.
b. The proper method of fighting fire inside buildings and confined areas.
c. Evaluation of fire hazards.

1

'. O.

V l 00AP Page 74 of 88

Operational Quality Assurance Plan Rev 12 Appendix C

3. Training Documentation Classroom training sessions, practice sessions, and drill'; for the fire brigade shall be documented. The following should be included in the documentation for persons participating:
a. Name
b. Date
c. Summary of what was done
d. Evaluation by observer 5.0 Orills and Practice 5.1 Monticello
1. Fire protection drills and practices shall be conducted in accordance with the requirements of Appendix R Section I and the requirements listed below:
a. A meeting shall be held after each drill to discuss the drill and the need to repeat portions of the training program that are applicable to the type of drill performed.
b. Preplanned strategies, and the proficiency of brigade members and control room operators in their use shall be tested during drills.

l 36

c. Each year, one drill, conducted with any fire brigade, shall involve the local on-duty fire department.

5.2 Prairie Island

1. Orills
a. Orills shall be scheduled so that each fire brigade will participate in at least four drills per year. The following types of drills shall be scheduled:
1) Involve local on-duty fire department. This shall be done at least once a year, with any fire brigade.
2) At three year intervals, a randomly selected, announced drill shall be critiqued by qualified individuals 37 independent of NSP's staff.
3) Back shif t, conducted by the fire brigade leader on duty at the time. This shall be scheduled at least once per year f:- each brigade.

O OQAP Page 75 of 88

Operational Quality Assurance' Plan Rev 12 Appendix C y) b. Except as required above, drills may be. announced and may involve only the shift fire brigade members (to preclude the disruption of essential plant activities).

c. All drills shall be preplanned and critiqued. A meeting shall be held after each drill to discuss the drill ~and the.

need to repeat portions of the training program that are applicable to the type of drill performed.

d. To-the extent practical, fire brigade members shall use protective equipment, suppression systems, and other equipment used to fight an actual fire during all drills.

Preplanned strategies shall be tested during drills as well as the proficiency of brigade members and Control Room operators in their use.

2. Practice
a. Practice sessions shall be held at least once every year.

These sessions shall involve actually fighting fires which are similar to those which might be encountered in the plant. These sessions shall include:

1) Use of fire fighting equipment.
2) Use of breathing equiptrent under strenuous conditions.
3) Extinguishment of fire.
4) Best method by which to approach each type of fire, to the extent possible.
b. Brigade members missing a practice session shall be rescheduled to attend a later session with another brigade.

If this is not possible, they.shall be required to review the training material covered during the practice session.

6.0 Control of Combustibles and Ignition Sources 6.1 Monticello

1. Control of combustibles and ignition sources shall be in accordance with the requirements of Appendix R Section K, and the requirements listed below:
a. All areas containing safety related equipment or cables shall be surveyed once each working day for fire hazards by a member of the p,lant staff.
b. Storage of combustible materials shall be permitted only in p posted areas or in approved cabinets and containers.

.V 00AP Page 76 of 88

Operational Quality Assurance Plan Rev 12 Appendix C

c. Transient combustibles in any safety related area, or area containing safe shutdown equipment, shall be limited to the equivalent of 2 gallons of combustible liquid. Use of larger amounts of combustible material shall be governed by written procedures which specify augmented fire protection measures,
d. A person designated as a fire watch and equipped to prevent and combat fire shall be assigned to safety related areas where cutting, welding, grinding and open flame work is involved.
e. The fire watch shall remain in the assigned area for 30 minutes after work involving the cutting, welding, grinding, or open flame is completed,
f. Where feasible, all movable combustible material below or within 35 feet of cutting, welding, grinding, or open flame work shall be removed, and all immovable combustibles below or within 35 feet shall be protected.
g. Smoking shall be prohibited in all safety related areas, except those specifically designated by the plant management.

6.2 Prairie Island

1. Permanent and Temporary Storage
a. Measures shall be established to minimize fire hazards in areas containing safety related equipment or equipment

. required to safely shut down the reactor (s) which:

1) Govern the handling and limitation of the use of ordinary combustible materials, combustible and flammable gases and liquids, high efficiency particulate air and charcoal filters, dry ion exchange resins, or other combustible supplies in safety related areas.
2) Govern the removal from the area of all waste, debris, scrap, oil spills, or other combustibles resulting from the work activity immediately following completion of the activity, or at the end of each work shift, whichever comes first.
3) Govern the handling of transient fire loads such as combustible and flammable materials during maintenance, modification, or refueling operations.

O OQAP Page 77 of 88

Operational Quality Assurance Plan Rev 12 Appendix C 7

('j ' 4) Govern the use of specific compustibles in safety related areas.

a) All wood used in safety related areas during maintenance, modification, or refueling operations (such as laydown blocks or scaffolding) snall be treated with a flame retardant.

b) All untreated wood in safety related areas (during operations other than maintenance, modification or refueling) shall be limited to less than 2 cubic feet per area, c) Equipment or supplies (such as new fuel) shipped in untreated combustible packing containers may be unpacked in safety related areas if required for valid operating reasons. However, all combustible materials shall be removed from the area immediately following the unpacking.

d) Large amounts of combustible material shall not be left unattended during lunch breaks, shift changes, or other similar periods.

'O e) Loose combustible packing material such as wood V or paper and excelsior shall be placed in metal containers with tight-fitting self-closing metal covers.

b. All areas containing safety related equipment or cables shall be surveyed once each working day for fire' hazards by a member of the plant staff. Storage of combustible materials shall be permitted only in posted areas or in approved cabinets and containers. Unnecessary transient combustibles shall not be stored in areas containing safety related equipment or areas containing safe shutdown equipment or other essential auxiliary equipment area (e.g., HVAC equipment room).
c. Transient combustibles in any safety related area or area containing safe shutdown equipment shall be limited to the equivalent of 2 gallons of combustible liquid. Use of larger amounts of combustible material shall be governed by written procedures which specify augmented fire protection measures.
O OQAP Page 78 of 88

Operational Quality Assurance Plan Rev 12 Appendix C

)

2. Cutting, Welding, Grinding and Open Flame
a. Cutting, welding, grinding and open flame work in safety related areas shall be administratively controlled. A person designated as fire watch and equipped to prevent and combat fire shall be assigned to safety related areas where cutting, welding, grinding and open flame work is involved.

The fire watch shall remain in these assigned areas for 30 minutes after the work involving the cutting, grinding or open flame is completed.

b. Where feasible, all movable combustible material below or within 35 feet of cutting, welding, grinding, or open flame work shall be removed and all immovable combustibles below or within 35 feet shall be protected.
c. Smoking shall be prohibited in all safety related areas, except those specifically designated by the plant maragement,
d. Fire barrier penetration leak testing shall be done with approved and reviewed procedures. Permission to do this leak testing shall be obtained from the shift supervisor.

7.0 Fire Fighting Procedures 7.1 Monticello O Fire fighting procedures shall be established in accordance with the reouirements of Appendix R Section K.

7.2 Prairie Island

1. Fire fighting procedures or instructions shall be developed to cover the following areas:
a. Discovery of fire including:
1) Notification
2) Attempts to extinguish fire
b. Action of Control Room operator including:
1) Announcement
2) Sounding of fire alarm
3) Who to notify
c. Selection and delineation of responsibilities of fire brigade members.

00AP Page 79 of 88 L - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - . _ _ _ _

Operational Quality Assurance Plan Rev 12. Appendix C 7

(f d. Coordination of off-site fire department activities.

e. Actions of sec;rity guards during a fire emergency.
f. Delineation of recoonsibilities of other plant personnel.
g. Instructions and preplanned strategies for fighting fires in specific. areas of the plant when the general instructions are not adequate. These instructions shall include:
1) Identification of combustibles in area.
2) Identification of safe shutdown equipment in area and alternate equipment available for performing that

, function.

3) Fire suppre.;sion equipment available in the area.
4) Information showing ventilation control (power sources),

access hallways, stairs, and doors.

5) Identification of plant systems that should be managed to. reduce the damage potential from a fire in the area.

.m

( 6) Identification of radiological and toxic hazards in the area.

7) Ventilation system lineups to minimize spread of smoke and to remove smoke from the area.
8) Identification of actions which must be coordinated with operations personnel.
2. Instructions and preplanned strategies shall be tested during drills.

8.0 Modification Control (Monticello & Prairie Island) 8.1 Review of modifications for possible impact on plant fire protection provisions shall be performed if determined required by a designated member of the plant technical staff. The following guidelines shall be used in re.aking this determination:

1. Could the modification present a hazard not considered in the Fire Hazards Analysis? Will additional analysis be required?
2. Could the modification have the potential to interfere with installed fire protection equipment or does it modify existing fire protection equipment?
3. Could the fire protection system require modification because of the change?

0QAP Page 80 of 88

Operational Quality Assurance Plan Rev 12 Appendix C 8.2 If a fire protection review is required, the individual assigned to perform the review shall use tne following as a guide:

1. Does the modification reduce the fire protection provisions for safety related or "safe shutdown" equipment?
2. Will it be necessary to do a fire hazards analysis?
3. Does the design present an obstruction to installed fire protection equipment?
4. Will the installation of the equipment temporarily remove a fire protection system from service?
5. Does the modification involve thermal stress relieving and, if so, have precautions been taken?
6. Will any fire barriers be affected by the modification?

8.3 A modification shall be allowed to proceed only after satisfactory resolution of these concerns.

9.0 Procurement Control (Monticello & Prairie Island) 9.1 Underwriters Laboratories (UL) and Factory Mutual (FM) directories shall be reviewed to determine if the item is listed as being UL or .

FM approved. If the item is listed, a manufacturer shall be identified and the item procured in accordance with NSP's procurement process for nuclear plants.

The one exception for not buying an item that is UL or FM listed is if it is a replacement of original equipment or NSP standard type, then it shall be identified as such and procured from the origirial supplier or manufacturer. As a minimum, the s tem or equpment shall, by appropriate testing, meet NFPA standards.

9.2 If the item is not listed by UL or FM, the following process shall be used:

1. An evaluation shall be made to determine the compatibility of the item to the existing system or component, or
2. If the item has been manufactured for a long period of time, and
a. The item is standardized, and
b. The item has a satisfactory performance history, and
c. Appropriate receipt inspection is identified in the procurement dccuments, then an evaluation is unnecessary.

The fact that the supplier and item meets these requirements shall be documented in the procurement files.

0QAP Page 81 of 88

Operational Quality Assurance Plan Rev 12 Appendix C 9.3 Parts of components and equipment that have UL or FM approval as a unit shall be procured as follows:

~1. The part shall be manufactured by the original manufacturer of the component or equipment whenever possible.

s

2. The model number of the component or equipment shall be identified.
3. The specific part number shall be identified.
4. Documentation from the supplier shall be requested that indicates the part delivered meets the specification of the part s used in the original component or equipment. If the part has been changed, the manufacturer shall be a'sked to indicate any-changes in the operation of the component or equipment. In lieu of this documentation, the acceptance of the part shall be based on inspection or testing.

9.4 All purchase requisitions pertaining to fire protection systems and equipment shall be reviewed. Plant requisitions and Nuclear Engineering and Construction requisitions shall be reviewed by an individual designated by the Plant Manager. Non plant requisitions shall be reviewed by an individual designated by the

, Director Power Supply Quality Assurance.

d 10.0 Instructions, Procedures, and Drawings (Monticello & Prairie Island) 10.1 The system of Administrative Control Directives (ACDs) shall be used-to delineate responsibilities and requirements for the fire protection program.

10.2 Departmental instructions and procedures shall be revised or fssued to implement the fire protection program responsibilities and requirements contained in the ACDs.

10.3 Fire protection maintenance, modifications, inspections, tests, administrative controls, drills, and training shall be prescribed by written instructions, procedures, and drawings.

11.0 Surveillance and Inspection (Monticello & Prairie Island) 11.1 The Technical Specifications specify the surveillance and inspection requirements for the fire protection system. Surveillance shall be scheduled, performed, and documented in accordance with standard directives governing the surveillance testing program.

11.2 Procedures shall be developed to assure adequate preventive maintenance of fire protection equipment, including fire suppression water system pumps and hydrants.

O OQAP Page 82 of 88

Operational Quality Assuranco Plan Rev 12 Appendix C 12.0 Conditions Adverse to Fire Protection (Monticello & Prairie Island) 12.1 Administrative Control Directives shall establish criteria for housekeeping.

12.2 Work control process procedures shall be used to correct equipment failures, malfunctions, deficiencies, and defective components of fire protection systems.

12.3 As part of the training process, plant personnel shall be instructed on how to identify fire hazards and report them to their supervisor.

13.0 Records (Monticello & Prairie Island)

Plant and General Office directives establish nuclear plant records, creation, and retention requirements. Fire protection records requirements shall be included in the scope of these directives.

14.0 Audits (Monticello & Prairie Island) 14.1 In addition to normal quality assurance audits (at least biennial),

an independent fire protection and loss prevention inspection and audit shall be performed annually at each plant utilizing either qualificd off-stte NSP personnel or an outside fire protection engineer (or engineering consultant)(annual independent inspection).

14.2 An inspection and audit by an outside qualified fire protection engineer (or engineering consultant) (see section 2.3) shall be performed at each plant at least every three years (triennial independentinspection).

14.3 Inspection and audit results shall be reported to levels of management having fire protection program responsibilities in those areas audited or inspected.

O 00AP Page 83 of 88

ig Operational Quality Assurance Plan Rev 12 Appendix 0

.-Q. /') APPENDIX 0 Revision 12 Change Summary This appendix summarizes the changes made in Revision 12 to the Operational Quality-Assurance Plan. The intent of this appendix is to _ fulfill the requirements for identifying changes in accordance with 10CFR50.54(a)(3), Conditions of Licenses.

This appendix'is not a part of the Operational Quality Assurance Plan.

Change Number identifies the change number next to the sideline on the affected pages.

Page identifies the page numbers containing the change.

Reason (R) identifies the reason for the change.

Basis (8) identifies the basis for concluding that the revised program incorporating the change continues to satisfy 10CFR50, Appendix B and the quality assurance program description commitments previously accepted by the NRC.

Change Reason Numby Page(s) Basis p 1 i R: Revision number updated to 12.

Not required; editorial item.

t] _E .

2 5 R2 Exception was taken to ANSI N18.7-1976, Section 5.3.9, Emergency' Procedures, and Section 5.3.9.1, Emergency Procedure Format and Content, because requirements were not consistent with Generic Letter 82-33.

B:

NSP has committed to Generic Letter 82-33, Supplement 1 to NUREG - 0737, Requirements for Emergency Response Capability. This is

, an upgrade to previous requirements. It does not reduce previous commitments.

3 7 R:

Section reference renumbered to reflect the reorganization of top level management.

82 This is a reference correction only. It does not reduce previous commitments.

4 8,9,13,14,16,17 R2 The wo-d "and ' changed to &,

83 Not required; editorial item.

5 8 R1 The reference to the organizational chart figure number and location was corrected.

82 Hot required; editorial item.

f"'

(

00AP Page 84 of 88

-- w--,y- ---,,-e,- - , , , wm,-w- =- , , v -

w w e 9

Operational Quality Assurance Plan Rev 12 Appendix 0 Change Reason Number Page(s) Basis 6 8,10,11,17,18 L These changes reflect reorganization of top level management responsibilities.

B_:, This is a responsibility description only.

It does not reduce previous commitments.

7 9 h The Superintendent Suppliar Quality Assurance and the Superintendent Nucelar Projects Quality Assurance were combined into the Superintendent Nuclear Projects & Supplier Quality Assurance.

B; This is a responsibility description only.

It does not reduce previous commitments.

8 9 R; The word "inspections" replaced with the word "qualification" for consistency with implementing documents.

82 This is an improvement of the QA Program description. It does not reduce previous commitments.

9 9 R: The word "selected" was removed.

EI This is an improvement of the QA Program description. It does not reduce previous commitments.

10 9 h The words "except those for Nuclear Engineering and Construction" were removed.

B; The new position of Superintendent Nuclear Projects & Supplier Quality Assurance has incorporated this exception.

11 9,11 h The word "the" was added.

h Not required; editorial item.

12 10 R2 Corporate Security added to the responsibilities for the Vice President Nuclear Generation.

B; This is a responsibility description only.

It does not reduce previous commitments.

13 10 RJ The lower case used for the words "quality assurance".

82 Not required; editorial item.

14 10 h The responsibility for Power Supply training added to the General Manager Nuclear Plants.

82 This is a responsibility description only.

It does not reduce previous commitments.

l OQAP Page 85 of 88

r

(.

Operational Quality Assurance Plan Rev 12 Appendix 0 fx

(,) Change Reason

' Number Pajds] Basis

' 15 11 RJ Plant security added to the responsibilities of the Plant Managers.

h This is a responsibility description only.

3- It does not reduce previous commitm nts.

The words except ISI" moved to the end of

~

16 11 RJ -

the sentence, h Not required; editorial item. <

. 17 12 R: The word "that' added.

, E Not required; editorial item.

18 12 RJ Title change from General Superintendent Radiation Protection and Chemistry to Manager Nuclear Radiological Services.

82 This is a responsibility descripi.fon only.

It does not reduce previous commitments.

19 13 R2 The word "program" made plural.

82 Not required; editorial item.

l r^)

t g

20 13 h Title change from Superintendent Quality Control.

83 This is a title change only. It does not reduce previous commitments.

21 14 RJ The Manager Production Training now reports to the General Manager Nuclear Plants. This reflects reorganization of top level management responsibilities.

82 This is a responsibility description only.

It does not reduce previous commitments.

22 14 h lower case used for the words "nuclear plants".

82 Not required; editorial item.

23 8,13,14,18 RJ A comma added.

BJ Not required; editorial item.

24 14 R2 The word "insure" changed to "ensure".

82 Not required; editorial item.

25 15 RJ Title change from Fuel Supply Department.

h This is a title change only. It does not reduce previous commitments.

O b

l l

00AP Page 86 of 88

Operational Quality Assurance Plan Rev 12 Appendix 0 Change Reason Number Page(s) Basis 25 15 R2 Additional detail is incorporated in the responsibilities for Manager Corporate Security.

82 This is a responsibility description only.

It does not reduce previous commitments.

27 16 h Title change from Vice President System Operations & Maintenance.

82 This is a title change only. The responsibility for Electrical System Operations are now under the Vice President Transmission & Inter-Utility Services.

28 17 R: The word "and" was deleted.

_5~ Not required; editorial item.

29 17 R2 The responsibility for providing drafting services to the plant added to the Manager Technical Services under the General Manager Plant Engineering & Construction.

82 This responsibility description only. It does not reduce previous cormitments.

30 18 RJ Title change from Director Fuel Supply.

82 This is a title cb.oge only. It does not reduce previous .ommitments.

31 18 R; The plural of the word "service" was deleted.

B; Not required; editorial only.

32 18 RJ Title change from the Director Administrative Services.

82 This is a title change only. It does not reduce previous commitments.

33 18 RJ Title change from the Manager Purchasing.

h This is a title change only. It does not reduce previous commitments.

34 18 R: The resoonsibility for the control of drawings is now under the Director Power Supply Financial Operations.

82 This responsibility description only. It does not reduce previous commitments.

35 72 h Added Section III to the Appendix R reference for clarification.

h This is an improvement in the QA program description. It does not reduce previous commitments.

OQAP Page 87 of 88

Operational Quality Assurance Plan

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Rev 12 p/

C Change . Reason

-Number Page(s) Basis 36 '75 R_1 This sentence was deleted due to the NRC acceptance of the NSP January 22, 1988 submittal. The proposal was to alter NSP commitments to comply with 10CFR50 Appendix R, Section III.I. The change was from a yearly, unannounced drill, to a three year interval for independent critiques of fire drills which.is consistent with Appendix R.

82 This change results'in a minor reduction in the commitments contained in the QA Program relating to unannounced fire drills. The change received aporoval from the NRC.

37 75 h This sentence was reworded due to the NRC acceptance of the NSP January 22, 1988 submittal. The proposal was to alter NSP commitments to comply with 10CFR50 Appendix R, Section III.I, regarding yearly unannounced drills, to three year intervals for independent critiques of fire drills only.

82 This change results in a minor reduction in s the commitments contained in the QA Program

-(Q relating to unannounced fire drills. The change received approval from the NRC.

38 i R2 The "&" changed to the word "and."

h Not required; editorial item.

39 1,2 RJ Corrections made to reflect actual document titles and punctuation.

h Not required; editorial item.

40 4 R2 The plural of th' e"Specifi.:ation" was added.

B_1 Not required; editorial item.

41 4 -

R: The word "Ist" was replaced with the word "first."

h Not required; editorial item.

42 6 R; The words "and hydro" added to the types of NSP power plants.

82 Change reflects the actual types of plants within NSP. It does not reduce previous commitments. i 43 9 -

R: The word "Administrative" added to reflect actual title of document type. l q BJ Not required; editorial item. )

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44 15,21,23,24 R2 The word "Program" capitalized.

82 Not required; editorial item.

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APPENDIX I - EVALUATION OF HIGH ENERGY LINE BREAKS OUTSIDE CONTAINMENT O

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~' REV 5 12/86

MONTICELLO

, EXECUTIVE

SUMMARY

This report providcs a review of tnose outside containme nt systems identified as having high energy lines and evaluates the e f fect s of pos tula ted piping ruptures of these lines on the equipment required for safe shutdown of the unit. The report is divided into seve ral se ct io ns , wh ich ide nt if y the evalua tion criteria used, the safe shutdown equipme n t , a nd the high ene rgy l i r.es . The repo rt reviews the interact io n of pos tula ted high ene rgy line breaks (H ELB 's ) , a nd assesses the capabili ty to safely shutdown the unit following a pos tula ted HELB .

The definition of high energy piping used in this report is pip-ing with normal ope ra ting tempe ra tu res eq ual to or exceedi ng 200* F a nd des ign pressures equal to or exceeding 27 5 ps ig . Using this criterion piping sys tems were selected for cons ide ration as h igh ene rgy pipi ng and a c omprehe ns ive list of the high energy lines was produced. The systems identified as having high energy piping, which could adve rsely af fect safe shutdoun equipment, include:

o Primary Steam o Fe edwate r o Conde ns a te o Rea cto r Wa te r Clea nu p o High Pressure Coolant Injection (Steam) o Re acto r Core Iso lat io n Coo li ng (Steam)

Addi tional sys tems. we re evaluated fo r their ef fect on safe shut down eq uipme nt. Howe ve r , none of these systems had pos tula ted H ELB ' s , wh ich could af fect safe shutdown equipme nt or could create a condition within a s pe cif ic c ompa r tme nt that was not e nveloped by the condi tions created by one or more of the above listed systems.

The targe ts of the pos tula ted H8LB's were selected as those systems, c ompone nt s, a nd structures r equir ed to mit igate the consequences of pos tula ted HELB's a'nd bring the reactor to a cold shutdown condition without fuel damage or breach of Prima ry Containment. Since a coincident loss of of f-site power was also r equired to be postulated, the sys tems on which credit fo r s hutdown could be take n , were limi ted to the sys tems capable of being powered from the diesel ge ne ra to r. As such, pos tula ted paths to safe shutdown were de termined using only these systems.

The results of the HELB evaluat io ns revealed that there werea three compa rtme nt s where pos tula ted HELB's would have s ignif icant impact on safe shutdown. These compa rtments included the Primary Steam Chase, Conde nser Bay, and the Feedwater Pump Area. For pos tulat ed H ELB 's in each of these areas, ad dit ional me asures in the fo rm of equipme nt modi f ica tio ns , procedural revisions, and/or jet impingement shields were taken to protect I il REV 6 12/87

MONTICELLO paths to cafe shutdown. For all other compartments, safe shutdown paths were identified.

The conclusion of this evaluation is that a path to safe shutdown exists for any postulated HELB. This-conclusion has been reached with the assumption of a coincidental loss of of f-site power and

a. single active failure in any of the safe shutdown systems.

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MONTICELLO TABLE OF CONTENTS Page EXECUTIVE SUFMARY I-iii LIST OF TABLES I-ix LIST OF FIGURES I_x I.1 INTRODUCTION I.1-1 I.1-1 Purpose and Scope I.1-1 I.1-2 Definition of Terms I.1-1 I.1-3 Tec).nical Ove rview I.1-3 I.2 EVALUATION CRITERI A I.2-1 I.2-1 General Criteria I.2-1 I.2-2 As sump tio ns I.2-1 I.2-3 Identification of High Energy Piping I.2-2 I.2-4 Selection of Break Locations I.2-3

. I.2-5 Break Types I . 2- 4' ..

I.2-6 Hazards Considered I.2-4 I . 2- 6.1 Pipe Whip Ef fects I . 2- 5 I.2-6.2 Jet Impingement Ef fects I.2-6 I.2-6.3 Flooding Ef fects I . 2- 7 I.2-6.4 Environmental Ef fects and I.2-8 Compa rtme nt Pressurizat ion I.2-7 Basis for Ide nt i. cation of Targe ts I.2-8 I.2-8 Comparison of Applied Criteria with I . 2- 8 the 1973 Licensing Basis I.3 EVALUATION METHODOLOGY I.3-1 I.3-1 Ider2ification of Plant Areas, I.3-1 Compartments and Rooms I.3-2 Identification of High Energy Piping I.3-4 I.3-3 Identification of Potential Targets I . 3- 5 I.3-4 Site Survey Methodology I.3-5 I.3-5 Compliance Evaluation I.3-6 SAFE SHUTDOWN SYSTEMS REQUIREMENTS I.4 l 2.4 I.4-1 Safe Shutdown Systems Summary I.4-1 I.4-1.1 Reactor Protect ion Sys tem, Control I.4-2 Rod Drive System and Control Rods l

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MONTICELLO O

(,) TABLE OF CONTENTS (Continued)

Page I.4-1.2 High Pressure Coolant Injection I.4-2 (HPCI)

I.4-1.3 Safety / Relief Valves (S/RV's) I.4-3 I.4-1.4 Reactor Core Isolation Cooling I.4-3 (RCIC)

I.4-1.5 Residual Heat Removal (RHR) I.4-4 I.4-1.6 Core Spray (CS) I.4-4 I.4-1.7 RHR Service Wa ter (RHRSW) I.4-4 I.4-1.8 Shutdown Instrumentation I.4-5 I.4-1.9 Auxiliary Support Systems I.4-5 1.4-2 Description of Safe Shutdown Paths I.4-5 I.4-3 Safe Shutdown System Components I.4-6 Locat ion I.4-3.1 Reactor Protection System I.4-8 I.4-3.2 Control Rod Drive System I . 4- 8 I.4-3.3 High Pressure Coolant Injection I.4-8 (HPCI) System l(,r ;g I.4-3.4 Safety / Relief Valves (S/RV's) I.4-8 I.4-3.5 Reactor Core Isolation Cooling I.4-9 (RCIC) System I.4-3.6 Residual Heat Removal (RHR) Sys tem I . 4-9 I.4-3.7 Core Spray (CS) System I.4-10 I.4-3.8 RHR Service Water System I . 4-l l I.4-3.9 Shutdown Instrumentation I.4-12 1.4-3.10 Emergency Service Wa ter I.4-13 (ESH) System I.4-3.11 Diesel Generators and Auxiliaries I.4-15 I.4-3.12 Auxiliary Power Distribution I.4-15 Systems I.4-3.13 DC Power Systems I.4-16 I.4-3.14 HVAC Systems I.4-16 I.5 HIGH ENERGY SYSTEMS I.5-1 I.5-1 Description of High Energy Systems I.5-1 and Boundaries I.5-1.1 Primary Steam System I.5-1 I.5-1.2 Feedwater System I.5-2 I.5-1.3 Condensate System I.5-2 I.5-1.4 HPCI (Steam) System I.5-2 I.5-1.5 RCIC (Steam) System I.5-3 I.5-1.5 Reacto r Wate r Clea nup ( RWCU) I.5-3 O. System I-v REV 6 12/87 s

MONTICELLO TABLE OF CONTFNTS (Continued)

Page I.5-1.7 Instrument and Sample Lines I.5-3 I.5-1.8 Core Spray System I.5-4 f.5-1.9 Residual Heat Removal (RHR) I.5-4 System I.5-1.10 HPCI (Water) System I.5-4 I.5-1.11 RCIC (Water) System I.5-5 I.5-1.12 Standby Liquid Control I.5-S I.5-1.13 Offgas System I.5-5 I.5-1.14 Control Rod Drive System I.5-5 I.5-1.15 Extraction Steam System I.5-6 I.5-2 Description of High Energy Piping I.5-6 I.5-3 Description of Break Locations I.5-16 I.5-3.1 Primary Steam Break Locations I.5-16 I.5-3.2 High Pressure Coolant Injection I.5-20 Break Locations I.5-3,3 Reactor Core Isolation Cooling I.5-21 System Break Locations I.5-3.4 Feedwater Break Locations I.5-21 I.5-3.5 Condensate System Break Locations I.5-23 I.5-3.6 Reactor Water Cleanup Break I.5-23 Break Locations I.5-3.7 Other Systems I.5-23 I.6 EVALUATION RESULTS I.6-1 I.6-1 Single Active Failure Evaluation I.6-1 I.6-2 HELB Evaluations by System I.6-4 I.6-2.1 Main Steam fystem I.6-5 I.6-2.2 Feedwater System I.6-9 I.6-2.3 Condensate System I.6-13 I.6-2.4 HPCI (Steam) I.6-15 I.6-2.5 RCIC (Steam) I.6-16 I.6-2.6 Reactor Water Cleanup System I.6-18 I.1-2.7 Core Spray System I.6-21 I.6-2.8 Residual Heat Removal System I . 6 - 2 '2 I.6-2.9 HPCI (Water) System I.6-23 I.6-2.10 RCIC (Water) System I.6-23 I.6-2.ll Standby Liquid Control System I.6-23 I.6-2.12 Offgas System I.6-23 I.6-3 Table of System Effects I.6-234 I.6-4 Use of RCIC in Safe Shutdown Sequence I.6-27 O

I- v i REV 6 12/87

MONTICELLO

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I.1-0 INTRODUCTION I.1-1 Purpose and Scope The purpose of this repo rt is to document the results of the evaluat ion for the ef fects of pos tulat ed High Energy -Line Breaks ( H ELB 's ) on the Monticello Nucle ar Generating Pla n t 's equipment that is es se ntial to saf e shutdown of the unit. The objective of this evaluation was to review the ef fects of HELB's, pr ovide deta iled inf o rmation rega rding the me thod of compliance, and identify a ny interact io nr wh ich would result in the minimum safe shut down eq .pme nt be ing unava ilable to cafely shutdown the unit. The evaluation inc luded a comprehe ns ive review of all high energy inte ract ions and reflects the present phys ical co nf iguration of the unit. This report supe rcedes the 1973 high energy line break report (Reference I.8-1) written fo r Monticello.

As required by the criteria provided in the December 18, 1972 Giambus so lette r (Reference I.8-2), this report ide nt if ies all of the high energy lines outside of con-tainment associated with the Monticello Nuclear Generat-ing Pla nt and evalua tes the ef fects of H ELB 's on safe shu tdown (SSD) equipme nt. Each system with high energy N piping was evaluated individually fo r the ef fects of pipe wh ip , jet impi ngeme n t, c ompa rtmental pres s uriza-tion, flooding, and equipment quali f ica tio n. This report also provides background i nfo rmat ion on the systems, components, and structures used for safe shut-down of the unit, the project evaluation approach and me thodology, the results of the evalua tions , and the conc lus io ns.

I.1-2 Def inition of Te rms_

The following list of terms are defined as they pe rtain to this report:

Circumf e rent ial Break - A pipe break pe rpe ndicula r to the pipe axis with a bre ak area eq uivalent to 'the internal cross-sectional area of the ruptured pipe.

Compa rtme nt -

The collect ive te rm used to describe rooms, bays, hallways , e tc . , that makeup an enclosure in i

wh ich a high ene rgy pipe is routed or pipe break may take place .

Compa rtme ntal Pressurization - The condition of pres-

' s urizat ion of a compa rtment above its normal pressure due to a high energy line break within the compartment.

l I.1-1 REV 6 12/87

M0Na'ICELLO Critical Crack - The s ing le pos tula ted crack on a high ene rgy line that produces the most ad ve rse ef fect {

on safe shutdcrn equipment within a compa rtme nt. '

Dime ns icn ally the critical crack s ize is pos tulat ed to be one-half the pipe ins ide di ame ter in le ngth by one- .

half the pipe thickness in width. (

Equipmant Qualifica tion - The term used to describe the l abili ty of equipme nt to remain f unct ional in the abnor-mal tempe ra tur e , pressure, humidi ty , and radiation e nviro nme nts generated by a pos tulat ed H ELB within a c cupa rtme nt .

Fire Area /Zono - The subdivision of the pla nt 's build-ingc (using Ap pe ndix R nomenc lature ) chosen to provide cons istent ide nt if ica tion of the compa rtme nt s evaluated in this report.

Floodi ng - The condi tion encountered within a compart-mont or ed joini ng c ompa rtments when liquid wa te r or condo ns ing steam escapes from a ru ptured high energy line or other component at a f aster rate than the fluid can be drained away.

High Ene rg y System - Anf mechanical sys tem with piping containing high ene rgy fluid.

High "n e rg y Line - Any pipe which ca rries fluid with a des ign press ure greater than or equal to 27 5 ps ig a nd a normal ope ra ting teuperature greater than or equal to 200'F.

Jet Impingeme nt - Tho. forces exerted on targe ts by high pressure f luid exiting a break in the pipe.

i Lo ngi tudi nal Break -A pipe break pa ralle l to the pipe I axis and oriented at a ny point around the pipe circom-f orence with a break area equal to the ef fective cross-se ct ional flow area immedia t e1.y upstream of the break loca t io n .

Normal Ope ra ting Condi tions - Plant ope ra t ing condi tlans during reactor startup, operation at powe r , hot standby ,

or reactor cooldown to the cold shItdown condition.

Pipe Whip - The mechan ical reaction c aused by h igh energy fluid exiting a ruptured pipe.

Safe Shu tdown - Shutdown of tha unit from powe r to a s ub e ri * '.c a l condition with a tempe ra tur e in the reactor of less than or equal to 212*F a nd no resulting fuel damage or breach of the Primary Containment.

I.1-2 REV 6 12/87

MONTICELLO I.2 EVALUATION CRITERIA

{v- The criteria used for the determination .C the high energy lines and the effects of the postulated breaks on these lines on safe shutdown equipment are identified in the December 18, 1972 Giambusso letter as clarified by Star.dard Review Plan (SRP) 3.6.1 (Reference I.8-3).

These criteria are utilized as the basis for the determination of the high energy lines, break locations, and the evaluation of-effects on SSD equipment.

This section describes the cr'iteria and assumptions used to select the high energy lines, - determine the break locations, and evaluate the effects of the high energy breaks on SSD capability. In addition, a comparison to-the original licensing basis is presented.

1.2-1 General Criteria The evaluation requirements for high energy line breaks outside of the containment are contained in Criteria No.

4 of the General Design Criteria listed in Appendix A of 10 CFR Part 50 and in the 1972 Giambusso letter (Reference I.8-2). These criteria require that systems and components necessary to ensure a cold shutdown of the plant be capable of withstanding all expected con-ditions resulting from high energy line breaks outside O

5 of the containment including pipe whig, jet impingement and environmental effects. The application of thes.

criteria is restricted to those systems, structures, anc components required to bring the plant to, and maintain a cold shutdown condition. Therefore, for any postu-lated HELB within a compartment, a path 'for achieving and maintaining a cold shutdown condition of the unit must be demonstrated.

I.2-2 pssumptions The following asrumptions were used during the HELS evaluation:

I.2-2.1 A postulated break was assumed to occur during normal steady state operating conditions at 100% of rated power (Criterion B.3.a - Reference I.8-3).

I.2-2.2 Loss of off-site power concurrent with the line break was assumed, unless it was more conservative to assume the availability of of fsite power, as was the case for a break in a Feedwater line (Criterion B.3.b.1 - Reference I.8-3).

I.2-1 REV 5 12/86

MONTICELLO I . 2-2. 3 No fires or other simultaneous line breaks or accidents were considered in the evaluation of the data (Criterion B .3.a - Ref erence I .8-3 ) .

I.2-2.4 A single active component failure was assumed in systems used to mitigate the consequences of the pos tula ted piping failure and to shut down the reactor. The single act ive component f a ilure was assumed to occur in addition to the postulated piping failure and any direct cons eq ue nces of the piping failure, such as unit trip and loss of off site powe r (Crite rion 20 -

Reference I.8-2 and as clarified in Criterion B.3.b.2 of Reference I.8-3)

I.2-2.5 Circumfe rent ial breaks were assumed to result in complete pipe severance with full separation of the two seve red pipe ends ( i .e . , guillotine). The break was assumed to be pe rpe ndicular to the lo ngi tudin al axis of the pipe. Circumfe rential breaks pos tula ted at fittings were assumed to La at the fi t t i ng- to-pi pe weld (s)

(Footnote 9 to Criterion 3 - Reference I.8-2).

I.2-2.6 Iongitudinal bre aks were assumed to be oriented pt.rallel to the long i tudi nal axin of the pipe oriented at any point around the pipe circumference with' a braak area equal to the ef fect ive cross-sectional flow area immediately upstream of the break lo ca t ion. Lo ngi t u-dinal breaks pos tula ted at fittings we re assumed to be at the conter of the fitting (Footnote 8 to Criterion 3

- Reference I.8-2).

I.2-2.7 Seismic Ca cego ry I Systems, for which stress data was not ava ilab le to determine break locations, were treated as seismic Category II Systems (Criterion B.3.a.2 Re ference I .8-4 ) .

I.2-2.6 A normally closed check valve in a piping line utilized as an inboard containment isolation valve was assumed to leak. The Icakage was assumed to result in pressuriza-tion of that se ct ion of pipe out to the fi rs t no rmally closed valve outside the Primary Containment.

I.2-2.9 A normally closed check valve was assumed to remain closed in the event of a pos tula ted line break between the check valve a nd the first no rmally closed valve outside the primary containment. For this condi tio n ,

the check valve was assumed to be a passive component.

I.2-3 Ide ntif ica tion of High Energy Piping High energy (HE) r ipi ng was assumed to be all piping with a no rmal ope ra ting tempe ra ture equal to or l

l

! I.2-2 REV 6 12/87 l

,- 4 MONTICELLO e xceeding 200'F and a des ign pres sur e equal to or exceeding 27 5 ps ig ( Re ference I.8-9). If the actue' ope ra ting temperature was not known, the des ign tempera ture was co nse rvat ively s ubs ti tut ed. These criteria were applied to the Primary Steam, Reactor Core Iso lat io n Cooling (RCIC) s te am, High Pressure Coolan t Injection (HPCI) s team , Reactor Water Cle anu p (RUCU),

Co nde nsa te , Fe edwa te r, high energy sampling and ins trume nt se ns ing li nes , Core Spray, Res idual Heat Removal (RHR), HPCI (injection), RCIC (i nje ct io n ) ,

Standby Liquid Control (SBLC), Of fgan, Control Rod Drive (C RD) , and Extract ion Steam Systems. The Pipi ng and Instrumentation Diagrams and the Piping Line Designation Tables were utilized to generate the list of high ene rgy lines.

For the Of fgas System, the normal operating temperatures and pressures were used to de te rmine the h igh ene rgy lines on the sys tem . However, no lines of the Of fgas System were excluded from the list of nigh energy piping based upon an operating pressure less than 275 psig.

I . 2-4 Selection of Break tocations O Break loca tio ns were pos tula ted for all high energy U piping with a nominal diame te r greater (Criterion 3 - Reference I.8-2), and a normal operation than 1 inch time exceeding 2% of the total unit ope rat ion time (Criterion B .2.e -

Re f e rence I.8-4) in accordance witn the following criteria:

I.2-4.1 Breaks we re pos tula ted for each Seismic Category I HE line as follows:

(1) At the terminal ends of the pressurized sections of the run (Criterion 2.b.1 - Reference I.8-2);

(2) At any intermediate location where normal operating and seismic stresses exceeds 0.8(Sh + SA) or .8(SA)

(Criterion 2.b.2 - Reference I.8-2);

(3) Not less than two (2) tot al inte rmedia te lo ca t ions we re selected on the basis of highes t relative s tress if line s tres ses were below the criteria s pe cified in Item 2. If the stressee dif fered by less than ten perce nt at the two h ighes t s tressed locations and their locations were not separated by a cha nge in the direct ion of the pipi ng run, the n additional points were selected. These addi tional points were chosen until a minimum of two locations were identified which were separated by a change in I.2-3 REV 6 12/87

MONTICELLO d irect ion of the piping run (Section 5.2.c -

Re ference I .8-10 ) .

(4) If the piping run had calcula ted stresses lower than the limits es t ab lished by Item 2 eve rywhe re between its terminal ends and had only one change of dire ct io n, only one inte rmedia te rup ture loca tio n was pos tula ted . Intermediate ru ptu re locations were not pos tulated on s hort pipi ng runs (less than 20 pipe diame ters in lo ng th) with calculated s tresses lower than the limit of Item 2 eve rywhere be twe en te rmi nal e nds and only one change of direction (Section 5.2.c - Reference I.8-10).

I.2-4.2 Break loca tions pos tula ted for each seismic Category II (i.e., non-seismic) h igh e ne rgy line were sele ct ed as fol1ows (Criterion B.l.d.2 - Reference I.8-4):

(1) At the terminal ends of the pressurized sections of the run; (2) At each intermediate location of pot ent ial high s tress or fat igue such as pipe fittings, valves, flanges and welded on attachments.

I.2-4.3 Critical cracks were pos tula t ed to occur in piping ca rrying high ene rg y fluids routed in the vicinity of systems req uir ed for sa fe shutdown of the unit. The critical crack size was take n to be 1/2 the pipe diameter in le ngth a nd 1/2 the wall thickness in width (Criterion 2 - Reference I.8-2).

I.2-5 Break Types Cirwmferential breaks were pos tulated and evalua ted at te rminal e nds. In addition, circumferential breaks were pos tulated at intermediate break loca tio ns on HE piping with a nominal diameter exceeding 1 inch. Lo ngi tudinal breaks we re pos tula ted at all inte rmedi ate loca tions on HE piping with a nominal diameter equal to or exceeding 4 inches. Longi tudi nal brecks were pos tula ted with an orientation of a ny point ar ound the pipe circumference and a break area equal to the ef fective cross sectional flow area immediately ups tre am of the break lo ca t io n.

(Criterion 3 - Reference I.8-2).

I.2-6 Hazards Cons idered The ef fects on Safe Shutdown Equipment from high energy lines can be divided in to five (5) separate types of int e rac t io ns . These interactions include:

I.2-4 REV 6 12/87

p MONTICELLO i

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Pipe Whip Jet Impingeme nt o Flooding o Environmental Ef fects o Compartment Over pressurization The evaluation criteria for each of these interactions i (hazards) identified was pe rfo rmed by applyi ng the following criteria I.2-6.1 Pipe Whip Ef fects For each pos tula ted break, the ef fects of pipe whip were evaluated. Pipe whip movement was assumed to occur about a plastic hinge located at a point or points selected using the following guidelines ( Re ference I.8-3):

(1) The pipe run terminal point; (2) The nearest pipe fitting which will experience high ,

be ndi ng mome nt . Generally this will occur at the second elbow from the break point.

n (3) A pipe whip res traint; U (4) Any structure or equipment wh ich can reasonably be expe ct ed to restrain the moveme nt of the pipe (e.g. , concrete wall or pressure vessel).

(5) Ordinary pipe supports were considered inef fective restraints during whip.

Circumferential breaks were assumed to cau se pipe whip about the plastic hinge in a plane defined by the pipe geometry and in the direction of jet reaction (Criterion B.3.a.5 -

Re ference I.8-4). Iongitudinal bre aks were assumed to cau se pipe wh ip movement in the direct ion op posite of the fluid flow no rmal to the axis of the pipe at the point of the break (Footnote 8 to Criterion 3 - Reference I .8 -2. Ad di t ion al ly, it was assumed that the geometry of the pipe segme nt be twe en the selected hinges remc.ined unchanged throughout the pipe whip path.

The area of influence of pipe whip was considered to be the wo rs t case area de te rmined applyi ng the criteria ide nt if ied above. In geaeral, circumf e rential breaks cau se the wo rst case pipe whip conditions.

In evalua ting the ef fect of pipe whip on safe shutdown components a nd structures, the following c rite ria was O u sed :

I.2-5 REV 6 12/87

E MONTICELLO (1) The energy level in a whipping pipe was considered 9

to be insuf ficient to rupture an impacted pipe of:

A) Ec ;al or greater nominal pipe size and P? T 4ual er heavie r wall thickness (Footnote 1 to Criterion 1.d - Reference I.8-2).

(2) Impacted pipe of lesse r nominal pipe s ize or thinner wall thickness was assumed to ru pture or require f urther evaluation (Footnote 1 to Criterion 1.d - Reference I.8-2).

(3) All electrical cables in cable trays, condu i ts , or other raceways impact ed by a wh ip pi ng pipe were assumed to be severed or require further evalua-tion. Other electrical and instrument and control (I & C) canponents were assumed to fa.il or require f urther evaluation (Reference I .8-2).

~~

(4) Structural components impact ed by a wh ipping pipe were assumed to fail or require f urther evaluation (Reference I.8-2).

I.2-6.2 Jet Impingement Ef fects For each pos tula ted breake the ef fects of jet impi nge-ment were evaluated. Tne criteria used to evalua te these ef fects were as follows:

(1) All jets were assumed to be i nfluenc ed by gravity (Sect ion 7.1 - Re f e rence I . 8-10 ) . Howeve r, because of the jet f luid - velo cities a nd the relat ive proximity of the t a rg e t s , s traigh t line jet travel was assumed.

(2) Jets frcra a circumferential break we re assumed to sweep the arc traveled during the whip (Section 8-1

- Reference I.8-10).

(3) A lorg i tudi nal break was assumed to occur at any azimuth locat ion on the pipe circumference with a break area equal to the ef fective cross sectional flow area immediat ely ups tre am of the break loca tio n. (Footnote 8 to Criterion 3 -

Reference I.8-2).

(4) A jet discharging saturated steam, or a mixture of steam and water, with a fluid tempe ra ture grea te r than the saturation tempe rature at the surroundi ng e nviro nment al pressure, was as sumed to expa nd in a 10' half-angle cone. (Section 7.2 - Reference I.8-I.2-6 REV 6 12/87

MONTICELLO 10 and page 3.6.2-8 Re f erence I . 8-4) . (Note: Th is (O/ criterion is consistent Monticello HELB evaluation.

with the previou s Re f erences I.8-1 and I . 8 - 9. ) .

(5) A je t discharging saturated liquid water or s ubcooled wate r was characterized by a constant d lame ter je t for purposes of determining the jet force on targets. For the purpose of-determining potential targets saturated liquid water or subcooled wate r jets were assumeg to e xpa n d, at least at a 10' half angle cone (Section 7.2, Reference I.8-10 and page 3.6.2 Peference I .8 -

4). (Note: Th is criterion is more conservative than that used in the previcus Monticello HELB evaluation - Re f erence I.8-1 a nd I.8-9) .

(6) The magnitude of jet force was determined using th e criteria of It em C.(4 ) on Page 3.6.2-7 of SRP 3.6.2

( Re ference I.8-4). The criteria for taking into account the geometric considerations of a fluid jet was provided in Appendix D of Reference I.8-10.

(7) A postulated jet was considered ef fective until it struck a solid barrier upon which tne jet momentum rs is reduced to zero. All safe shutdown components, pipes, and structures impacted by the jet were con-sidered incapable of performing their safe shutdown function or required further evaluation ( Sectio n

7. 2 - Re f erence I.8-10) .

(8) Where the jet must travel a significant distance to impact safe shutdown components or structures, s impli fied calculations we re used to demonstrate that the jet impingement forces were negligible.

I.2-6.3 Flooding Ef fect s In each area or compartment whera breaks were pos t-ulated, the potential adverse ef fects from flooding were identified using the following criteria (Section 10 -

Reference I.8-10):

1. Because of the uncertainties associated with break orientation and jet geometry, an additional degree o f conservatism was established to include potential targets beyond the 10' half angle cone. In some cases, targets up to 4 5' of f the axis of the jet were considered hit and incapable of performing their saf e

, shutdown function.

]

I.2-7 REV 6 12/87 i

MONTICELLO (1) Vulnerability of safe shutdown equipment due to flooding because of location a nd configuration.

(2) Potential rupture sizes and the available qua nt ity cf water.

(3) Absence of floor drains or doorways which could provide drainage.

(4) Existence of flood pr ot ect ion in the form of dams or sumps with pumps.

I.2-6.4 Environmental Ef fects and Compartment Pressurization A review was conducted to dete rmine if the safe shutdown ecmponents in compartments or areas in which breaks have been pos tulated are qualified for worst case e nviron-me nt al condi tions including the of fects of ccupa rtme nt press ur iza t ion. Attention was focused on closed compa rtme nts wh ich contained high ene rgy lines. The bounding or largest available high energy line break for each compa rtment was compared with that assumed in the Monticello Nuc lea r Gene rati ng Pla nt Enviro nme ntal Effects Report (Reference I.8-ll) to es tablish limiting co ndit ion for the compa rtme nt. If other lines appeared to produce a more severe condition, this information was noted and evaluated in Section 6, of this report.

I.2-7 Bas is for Identification of Targets The following criteria were used to ide nt ify the potential ta rge ts of the pos tulat ed high ene rgy line b reaks :

7.2-7.1 The potential targets of the pos tulated high energy line bre aks inc luded those s ys t ems , components, a nd struc-tures requi red to mitigate the consequences of the HELS and safely shut down the unit with a coincident loss of of f-site power (Reference I.8-3 & I.8-4).

I.2-7.2 Spe cif ic systems a nd compone nts chosen as ta rge ts were those required to ach ieve and maintain a safe shutdown condition as identified in the USAR (Re ference I .8-5 ) ,

the original H ELB evaluation repo rt (Reference I . 8 -1 ) ,

and the Safe Shutdown Analysis (Reference I.8-6).

I . 2- 8 Compa riso n of Ap pli ed Criteria with the 1973 Lice ns ing Basis In orde r to demons trate conf o rma nce with the original HELB lice nsi ng basis (Reference I.8-2), a comparison of the original criteria was made with the criteria used in I.2-8 REV 6 12/87

/

MONTICELLO this evalua tio n. The - comparison was pe rf o rmed on a point by.' point basis ad dressing each of 21 criteria

-ide nt if ied in the Giambusso *1e tter (Reference . I .8-2 ) .

Each criterion has been evaluated to de te rmine - the degree of compliance be tween this HELB evaluation and

- the - re ferenced c ri te ri a . - The Gi ambus so criteria are

- provided in Appendix C of this report.- -

Lompliance with criterion 1:

The criteria used in this evaluation is consistent with the reference criterion except for two items. The high e nergy piping is def ined in this evaluation as piping having a normal operating tempera ture 1(a) equal to ' o r above 200*F and a design pressure equal to or above 275 psig. The criterion used is consistent with previous-HELB evalua tio ns (Ref e rences 7.8-1 & I.8-9) pe rformed for Monticello. Item 1(c) of the Gi ambus so le t te r' ,,

requi res that fo r pipe wh ip , the un res trained pipe is able to whip in any possible direction about the plastic h inge~ fo med at the neares t pipe whip restraint. This evaluation defined the bases for cho osing the neares t plastic hinge and identified that the pipe would whip in '

the plane defined by the pipe geometry and plastic hinge location in a direction opposite of the jet fo rce . . This O is ' consistent with Criterion B.3.a.5 of. BTP MEB 3-1 in S RP 3.'6. 2 ( Re f e rence I . 8-4 ) .

Compliance with Criterion 2:

! Break locations and critical crack locations were chosen for the seismic pipi ng in accordance with this crite-r io n . No guidance was provided for non-seismic piping or seismic piping, for wh ich s tress analysis was not used to de termine break loca tio ns . The ref o re , the criteria contained in S RP 3.6.2 were used and break locations were chosen at each high stress point (e.g.,

j fitting, valve, weld attachment).

Compliance with Criterion 3:

l

The analysis was pe rfo rmed in accordance with this l

criterion with one exception. Footnote 9 indicates that

whipping can occur in any direction normal to the pipe l axis. The analysis assumed pipe wh ip in the dire ction oppos ite the jet within the plane defined by the break lo catio n, plastic hinge, and pipe r outi ng (See response to Criterion 1) .

l 1

iO l

i' I.2-9 REV 6 12/87 l*

t

k MONTICELLO Compliance with Criterion 4:

This evalua tion was more conserva tive than the criterion used in the previous evaluat io n. The an alys is assumed the wh ipping fo rce was equal to a nd op pos ite the jet force without r ega rd to distance to the ta rge t. The analysis also aapumed ~ that steady state conditions we re s eco nds , and that the wh ip pi ng force achieved in 10 existed fo r a sufficient period of time to require the target to ros po nd to the fo r c e .

Compliance with Criterion 5:

The previous evalua tion (Ref erences I .8-1, I.8-7, & I.8-

8) identified those areas where ad dit ion al measures would be requi red to mitiga te the of fects of pipe whip a nd/o r jet impi ngeme n t. Where protect ive restraints or shields were req ui red , they were ide nt if ied , designed, and i nst alled as part of the previous evaluat ion e f fo rt . This evaluation was cons is tent with the referenced criterion.

Compliance with Criterion 6, 7& 8:

l Where additional loads were applied on SSD equipment and Seismic Catego ry I s tructures as a result of HELB interactions, a reanalysis was pe rformed consistent with '

the methods used previou sly to evalua te Mont icello equipment and s tructures, to dete rmine s tress levels and react ions . As such, this evaluation was consistent with the re ferenced crite ria.

Compliance with Criterion 9:

t No new openings in existing structures were required as a result of the HELB evaluat io ns . Therefore, this criterion was not ap plicable .

Compliance with Criterion 10:

The a nalys ie was done in a manner cons is tent with the criterion. A path to sa fe shutdown and mit iga t ion of the ef fects of the break were ide nt if i ed for each break lo ca t io n. Fa ilures of structures, as a result of a pos tula ted H ELB , wh ich would result in a path to safe shutdown not bei ng ava ilab le , have bee n modifi ed to provide a pa th to saf e shutdown.

Compliance with Criterion 11:

The analys is exceeded the Criterion in that a single act ive f a ilure analysis was pe rfo rmed for each break I.2-10 REV 6 12/87

MONTICELLO f3 V which af facted SSD equipment, and a path to safe shutdown was ide nt i fied. In ad dit io n , loss of of f sit e power was assumed fo r all breaks. For any break lo ca tion where a path to sa fe ahutdown c ould n ot be ide nt if ied , a concern was noted and steps take n to resolve the concern.

Compliance with Criterion 12:

None of the ide ntif ied HELB's pos tula ted fo r the analysis af fected the con trol room equipment or its habi tabili ty . Therefore, analysis as identified by this criterion was not required.

Compliance with Criterion 13:

(a) The pos tula ted limiting e nvi ronme ntal condi tions ,

determined for specific compartments, were compared with the pos tula ted HELB's for those ccmpa rtme nt s.

Profiles were gene rated for any compartment which did not have temperature and pressure profiles fo r postulated HELB'S within the compartment.

(b) In. addition, the ons ite eme rge ncy powe r distribu-tion system and diesel ge nera to rs were ide ntified as SSD eq uipme nt . As such, they were required to d be assured operability for each HELB.

Compliance with Criterion 14:

This evalua tion was cons istent with the criterion.

Appendix A shows drawings of the major plant buildi ngs with the compartment boundaries and numbers superimposed on these dr awing s. The locat ions of sa fe shutdown components and their associated piping and cabling are also provided on these drawings. Appendix B shows the routing of' the high energy lines, their high energy boundaries, and the compa rtments they are routed through. Figures I . 5-1 through I . 5-8 show the loca tions of bre aks for those sys tems where break loca t io ns we re determined from the stress analysis. These break loca- '

tions are also provided on the drawing in Appendix B.

Compliance with Criterion 15: ,

This evalua tion is cons is tent with criterion 15. The ef fects of flooding for each HELB were evaluated.

Compliance with Criterion 16:

[ The high e nergy li nes that are loca ted outside of k are of the Primary Ste am, RCIC contai nment parts I.2-11 REV 6 12/87

MONTZCELLO (steam), HPCI (steam), RWCU, Conde ns a te , Feedwater, a nd Sampli ng and Ins t rume n t Se nsi ng systems. Se ct io ns of these systems are des ign ated as ASME Class I or Class II. These se ct ions are nondestructively inspected on a ten year senedule accordi ng to the Monticello Inservice Inspection (ISI) Program Examination Plan which is based on the 1977 Edition (up to and including the 1978 Summer Adde ndum) o f ASME Section XI. The pipe welds on repre-sentative li nes are ins pe ct ed using vo lume tric , surface or vis ual met hods as indicated in the ISI ma nua l . The ten year inspection plan is divided into three 40-month s ubse ct ions so 1/3 of the Clas s I and II pipi ng is ins pe cted during a plant shutdown that occurs during that 40 month period.

When a modification involving welding is done on an ASME Section III, Class I, II, or III piping system, a hydro-static test is required to be performed per ASME Section XI standards to check for le akage . If the piping section des ign is not governed by the ASME code, a hydrostatic or as a minimum an Inservice Leak Test must be done to check for le akage from the modified pipe section.

There are also sections of the high energy lines that do not h ave a n A SME Clas s I or II des igna t io n. Is aks on these li nes would be noted at le as t once eve ry 8-hours when the operating staf f makes their plant rounds. High radi ation areas of the plant (main steam chase and react.or water c lea nup c ompa rtme n t ) are e ntered once a week by the operations staff. Any leaks in high energy lines in these areas would be n ot ed a t that time.

Compliance with Criterion 17:

This evalua tion did not address leak de tection, as le ak detection is not used for HELB mitigation.

Compliance with Criterion 18:

NSP has developed many dif fe rent procedures which would be utilized fo r the mit igat ion of a HELB event a nd the safe shutdown of the unit. Included are Abnormal Proce-dures, Emergency Operating Procedures, and HELB Specific Procedures.

The Abnormal Procedures are procedures for responding to s pecif ic types of events oc curri ng in the unit. They provide spe ci fic inf orma t ion fo r de ali ng with the of fects a nd conseq uences of these pos tulat ed events.

The Eme rge ncy Operating Procedures (EOPs) are sympto-matic eme rgency procudures that h ave been developed I.2-12 REV 6 12/87 t

-~ . , , -

.3 O

MONTICELLO .!

1 q

d- based on the BWR Owner's Group Eme rge ncy Procedures guidelines. In addition, to the Emergency Operating and r' Abnormal Procedures , an operating procedure was written

, in response to the high energy line break event in which the Divis ion I di'esel generator- (which supplies Motor Control Center 13 3) fa ils to start due to cable damage, the . Division II power supply is lost, and there is a concurrent less of off-site power. This procedure outlines an auxiliary source of powe r to start the Divis io n I diesel gene ra tor to supply powe r to safely shutdown the plant.

Compliance with Criterion 19:

This evalua tion is consistent with the criterion.- In orde r to de te rmine break locations, the seismic class and quality classification was ide ntif ied for the high energy piping.

Compliance with Criterion'20:

This evaluation is consistert wi th *.his cri terion. The criteria for estab1!shing jet impingament and pipe whip loads was ide ntifieo and utilized. Also, a single act ive f a ilure was assumed to occur in the equipment, O for which credit was being taken to reach safe shutdown.

Res ponse to Criterion 21:

This evalua tion is cons istent with the criterion.

Structural and mechanical components forming the Primary Containme nt bour firy were cons ide red pipe whip and/o r jet impingement  ::a rgets eve n if they were not SSD camponents.

O f

I.2-13 REV 6 12/87 I

MONTICELLO Oi This Page Intentionally Left Blank O

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1. 2- m REV 6 12/87 e

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~ ~ - -- - - - - - - _ . , . . . _ . . _ _ _ _ , _ , . , _ , _ . _ _ _ _

MONTICELLO O

Nm/ Safe shutdown can be desc ribed as reducing RPV pressure a nd tempera ture from the normal operating conditions of approxima tely 1015 psia and 545*F to a RPV c6cla nt tempe ra ture of 212* F with the reac tor in a subcritical condi tion and no fuel damage or breach of Prima ry contai nme nt. The minimum equipment nec essa ry for safe shutdown includes the following :

1) The Reactor Protection System and Control Rod Drive System (Nega tive Reactivity Funct ion)
2) 3 S/RV's (RPV Ove rpress ure Prote ct ion a nd Decay Heat Removal Functions )
3) 1 RHR Pump (LPCI Mode) or 1 Core Spray Pump and 1 RHR Pump in SPC Mode (Reacto r Vessel Leve l Maintenance and Decay Heat Removal Functions)
4) 1 RHR Se rvice Wa te r Pump (Decay Heat Removal Function)
5) 1 Diesel Ge ne ra to r, D/G . Auxiliaries, 1 D/G Eme rgency Se rvice Wa te r Pump, 1 ECCS Equipment Eme rge ncy Se rvice Water Pump, and 1 Division of p Essential Powe r Dis tr ibution (For all functions ij except Nega tive Reactivity)
6) Shutdown Instrumentation ( from Reference I.8-6) a) Reactor Vessel Level b) Reactor Pressure c) Suppression Pool Temperature d) Suppression Pool Level Evalua tions pe rf o rmed to de termine accep tab le paths to safe shu tdown utilized the above described minimum necessary equipment.

Additional Safe Shutdown Paths can be desc ribed depending on the location of the postulated HELB and the s ingle active failure t ake n . The HPCI and RCIC systems can be used to maintain RPV water level and depressurize the RPV, so that the low pressure high flow pumps (RHR or CS) can be used. Table I.4-1 identifies the sys t ems ava ilable to support each shutdown performance goal.

I.4-3 Sa f e Shutdown Sys tem Component to ca t io ns For each of the sys tems ide ntified in Section I.4-1 of this report, the locations of the major components, main p ipe rou t ing s , a nd power and control cable routings are N listed in this section. The info rmat ion is provided by I.4-6 REV 6 12/87

MONTICELLO f Auxiliary Air Compressors are located on the L935' elevation of the Reactor Building, Division I in Compartment I/2B and' Division II-in Compartment II/2C.

heat Each divisions' exchanger discharge pressure  :

control valves (CV-1728-and CV-1729) are also located:in the respeative corner rooms.

Cabling. for the system is routed entirely 'within the Turbine Building.- The - power cabling for the pumps is routed from the respective switchgear. areas (IX/12A and XII/14A) tv- the ' Intake ' Structure (IX/23A). Control-cabling in routed to the Control' Room via the' respective divisional. routing in the Turbine Building. For Division I this would be from the Cable Spreading Area y (VI/8) tc - Compartments IX/19C, IX/13C, IX/13B and to IX/16. For . Division II this would be from the Cable Spreading Area 'to compartments XII/19B, XII/19A,-XII/17 and then into the-Switchgear Area.

I.4-3.9 Shutdown Instrumentation The Shutdown Instrumentation, like the other safety-4 related equipment is divisionalized. The locations for the ccomonents for each division are given below Parameter Equipment f O

Division ' Measured Designation Compartment I Reactor Water Level. LT-2-3-ll2A East side of Reactor .

Bldg. elev.

935' -(I/2B)

I Suppression Pool LT-7338A Torus Area Level (IV/1F)

I Reactor Pressure PT-6-53A East side of Reactor Bldg. Elev.

962'-6" (I/3B) ,

! I Suppression Pool TE-4073A to Torus Area '

i Temperature 4080A (IV/lF) i II Reactor Water Level LT-2-3-ll2B West side of Reactor .

Bldg. Elev.

935'-0

(II/2C) t I.4-12 e

REV 5 12/86 L ____.. _ _. __ - - _ - .

MONTICELLO II Suppression Pool LT-7338B Torus Area Level (IV/lF)

II Reactor Pressure PT6-53B Wes t side of Reactor Bldg. Elev.

962'-6" (II/3C)

II Suppression Pool TE-4073B To ru s Are a Temperature to 4080B (IV/lF)

The cabling from these instruments is routed through the same routing schemes back to the Cable Spreading Area as are the other sa fe ty-re la ted divis ion al separa ted cables. The only excep tion are the Division II Suppression Pool Tempera ture cables and Alternate Shutdown System ( ASDS) Instruments. From the Torus Area they are routed in to the Conde nse r Bay (X/12C) and through Compartment XII/19 A on the 931'-0" eleva tion of the Turbine Buildi ng a nd in to the Eme rgency Filtration Building from there. ASDS ins trume nts that have been rerouted are Reactor Wate r Level LT2-3-ll2B and Suppression Pool Level LT-73388. The rou ting is unde rground ar ound the south s ide of the Reacto r Building to the Reactor Buildings third floor, EFT.

I. 4-3.10 Eme rgency Se rvice Wate r (ESW) System The ESW Sys tem is a 2 divisional system, wh ich supplies coo li ng water to the Diesel Gene ra to rs, the ECCS Pump Room Coole rs , and the RHR & CS Pumps. The system consists of 4 pump s , P-ll l A, B, C & D, located in the Intake Structure (IIX/23A) . Pumps P-lllA and P lllC arts pa rt of Division I, and Pumps P-111B and P-illD are part of Division II. Pump P-lllA is used to supply cooling water to the Divis ion I Diesel Ge ne ra to r. Pump P-lllB supplies cooling water to the Division II Diesel Ge ne ra to r. Pumps P-lllC & D supply cooling water to the respe ctive Division I and II ECCS Pump Room Coolers and the RHR and CS Pumps.

The discharge piping f rom e ach D/G ESU Pump teos in the Intake structure with separate 4" lines routed unde rground to the respective D/G. The other branch of the tee on each line has a closed ma nual valve in the Intake Structure (ESW 61-l&2). The cooling water lines to the D/Gs are crosstied in the Intake Structure. This allows cooling of either D/G from Pump P-lllA or P-lllB.

O I.4-13 REV 6 12/87

[ -

, .- . - .. , . - - - .. . . - .- ... . ~

v

~

g MbNTICELLO

~

i The Division I ESW Line to ECCS Room Coolers , and [

~ Division I: RHR and CS Pumps (ESW1-3"-HF) is routed from the -Intake Structure, through the Intake Structure

. Corridor (IX/23B), in to the Tur bine Buildi ng North

Corridor 1on the 911'-0". elevation (IX/16 ) . .From there, it is ' routed in the North Corridor over into the Reactor r Feed Pump Area (IX/138). Within this - compartme nt - the lines tee _again (Re fettence I .8 -21 a nd I.8-23) with" the piping . line ESWl-3"-HBD routed into the Eme rge ncy F iltratio n Buildi ng . The line to the ECCS . Pump Room _ s Coolers and RHR and CS . Pumps (SW30A-3"-HF is routed from the Main Feed Pump Area in to the Condenser Bay . Area (X/12C) on : the 911'-0" Eleva tion . The pipe is routed i alo ng the east wall over to the . south wall of the room. The piping line exits the Condenser Bay through the South wall and in to the Reactor Building. 'Ihis line ente rs the Reactor Building in the TIP Drive Room 4 (III/2A) a nd is routed from there in to the Torus Area Once in the To ru s Area the line tees aga in (IV/lF).

with pa rt of the flow r outed to the southeast corner room lI/lB). and the rema inde r - to the HPCI Room (II/lE) l for tht Division I HPCI Room Coole r.

The Div!;sion II ESW Line to the ECCS Room Coolers .(ESW5-a parallel route to- its . Division I

. 3"-HF) Milows counterp.t rt to the North Turbine Corridor at 911'-0" elevation (IX/16). From there it is routed to the Compartme nt XII/19 A on the 931'-0" elevation -of Turbine Buil di ng, a nd into Compartment XII/198. In i compartment XII/19B the line tees ( Refe rence I . 8 -2 2 and i I.8-23) with line SW30B-3"-HF ' routed into Compa rtment  ;

! XII/19C and then into the Condenser Bay (X/12C). From the Condense r Bay, the line ente rs the Reactor Building l i

l- in the Main Steam Chase (II/2F) and from there is routed

'< in to the To ru s Are a (IV/lF). Af ter entering the Torus ,

Area, the line tees again with part of the flow going to  ;

the southwest corner room (II/lA) a nd the remainder to

the Division II HPCI Room Coole r in the HPCI Area l (II/lE).

The cabling for the two (2) ESW pumps ( P-lllA and B) .

which cool the D/G's, follows a route ve ry similar to  !

the RHRSW Pumps. This applies to both the powe r and i

control c ables. Power for each ESW Pump comes from the i respective essential MCC of each division with the MCC's i located on dif ferent floors in the southeast corner of  ;

i the Turbine Building. The Div. I cables are routed from i

> MCC-133A unde rground to the intake structure, a nd the -

Div. II cable s from MCC-143A are routed along the east corridor of the Turbine Building through compartmu nts

! XI I/19 B , XII/19C, XII/17, and then down the intake ,

i structure corridor into the intake structure.  ;

i t

I.4-14 REV 6 12/87 s ,

,,..__,.---,_..,_,,,,,,-,-_m,._,.,,.,,,__.,_y,um_.___. m y , , , , , , , , - , - - , . .

MONTICELLO The cable fo r the other two ESW Pumps (P-lllC a nd P-lilD) begins in the EFT Building. The power cable to P-

{' )

111C is routed from MCC-134 in the EFT Building into the Main Fe edpump Are a, IX/13C, alo ng the east corridor at the 911' Eleva tion , IX/13B ,a nd then along the north Turbine Building Corr ido r IX/16 into the intake structure corridor. One control cable is located entirely in the EFT Bu il di ng. The other e n te rs the Turbine Building in the Main Feedpump Area, IX/13C, then is rout ed into IX/19C and up to the Turbine Operating Floor, X/30, before entering the Control Room. For Pump P-111D the power cable is routed from MCC-144 in the EFT Building into the Turbine Buildirq southeast corner on Elevation 931'-0", XI I/19B . It is then routed along the eas t corridor at that eleva tion, frca XII/19B to XII/19C in to the Divis ion II cable way, XI/17, a nd e nding in Divisio n II Essential Switchgear Area, XII/14A. From there the cable is routed outs ide unde rground to the Intake S truc tur e . One control cable fo r this pump is located entirely within the EET Building and the other enters the Turbine Building at the 931'-0" Elevation in compa rtment IX/19B. From there it is routed up to the Turbine Ope ra ting Floor, X/30 before entering the Con trol Room.

I. 4-3.11 Diesel Generators and Auxiliaries There is one Diesel Gene ra to r (D/G) for each Essential Division .at Monticello. The D/G's are loca ted in adjacent compartments with D/G-ll located in XIV/llB and D/G-12 loca systems ted in XIII/15A. The local control panels and air start are also located in the res pective compartment with the D/G. The oil day tanks are loca ted next to Compa rtment XIV/15B in separate day ta nk rooms, XVI/15D - Division I and XV/15C - Division II. Cabli ng for power to the local control and ele c trical cabineto is routed in the same manner as the control cables fo r RHRSW Pump s . The D/G power cab li ng is r'out ed to the respe ct ive es se nt ial swi tchge ar (Division I -

IX/12 A and Division II - XII-14A), which are in adjacent ccmpa rtments in the Turbi ne Building .

I.4-3.12 Auxiliary Power Distribution Systems The Auxiliary Power Distribution System consists of the 4KV swi tchge a r, the 480V load center, Mo to r Con trol Centers and the trans fo rme rs be twe en them. The system powe r is capable of dis tr ibuting the electrical generated by the D/G's under a loss of offsite power condition. For Division I, both the 4KV switchgear and 480V distribution equipment are located in Compartment IX/12A on the 911-0" elevation of the Turbine Building I.4-15 REV 6 12/87

MONTICELLO (W) v in the northwest corner. Th e Division II load centers are located directly above the Division I load center s on the next floor (e leva tion 931'-0")

in Compartment XII/14A. We Essential MCC are located in the southeas t corner of the Turbine Building with Division I MCC-133A located on the 908'-0" elevation in Compartment IX-13C and Division II MCC's -

142A and 143A located on- the 931'-0" elevation in Compar tmen t XII/19B.

Cabling from Load Ce nters to the MCC's is routed in separate compartments. Cables from division I toad Centers are routed from Compartment IX/12A through Compartments IX/16 and IX/13B terminating in Compartmen t IX/13C . Other Division I cables not terminating at the MCC are routed up into Compar tment IX/19C and in to th e cable spreading area.

  • We Division II cables are routed from the load centers XII/14A through ecmpar tment s XII/17 and XII/19A and terminating in compartment XII/19B. Cables going to the Cable Spreading Area are routed through Compartment XII/19B .

I.4-3.13 DC Power Systems With the exception of the switchgear 125V-DC distribu-tion panel located in each load center corapartmen t, all O the other DC powe r components are located in the O Administration Building or the Emergency Filtration Building. Th e cabling is routed from Cable Spreading Area (VI/8) to the respective components along the divisional, routings described previously. he only exception is the HPCI power cables, which are routed in a similar route to the Division II SPTMOS cables.

However, once routed into the Reactor Building, the HPCI cables are then routed to the HPCI Area (II/lE) .

I. 4-3.14 HVAC Sys tems The only HVAC Equipment required for safe shutdown are the ECCS Room Coolers, V-AC-5 (Division I) and V-AC- 4 (Division II), located in respective Re actor Building corner rooms on the 896'-3" elevation and the EDG supply fan, V-S F- 9 for EDG No. 12 and V-S F-10 for EDG No. 11.

We cooling water for the ECCS Poom Coolers is supplied from the respective Emergency Service Water System pump with V-AC-5 supplied by P-lllC, and V-AC-4 by P-lllD.

Power cabling for the units is routed into the Reactor Building from the cable spreading area in the s ame manner as the other. divisional cables.

I.4-16 REV 6 12/87

MONTICELLO O

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This Page Intentionally Left Blank I 4-17 REV 6 12/87

'MONTICELLO D

U I.6 Ej#.LUATION RESULTS Based:upon the information obtained from the three pre-vious sections, it is possible to identify all the compartments within the Monticello power block requiring r further evaluation. This information is contained in  :

Table I.6-1, which identafies by compartment and systeg l the locations of high energy-piping subject to possible '

pipe rupture. By reading any row, a-determination can be made of all compartments in which . pipe . ruptures of - a specific system are postulated. Also, by examining any column, a. determination can be made if any postulated high energy line breaks are located within aidentifies specific compartment. The last line of the table those compartments which contain

  • safe shutdown equipment.

The following section evaluates the effects of a postulated HELB on safe shutdown in each compartment

' identified in Table I.6-1 as having a postulated break 1- location or has high energy piping transversing the compartment. The evaluations are conducted on a system basis for-each compartment. A coincident loss of of f-site power and single active failure are assumed.

I.6-1 Sincie Active Failure Evaluation For each compartment containing high energy lines, where safe shutdown equipment was adversely affected, a single active failure review was conducted to determine the single active failures that could inhibit safe shutdown of the unit. Determination of the component (s) subject to single active f ailure was - based upon the definition  :'

provided in Appendix A of BTP ASB 3-1 (Reference The I.8-3 single and Section 3.2 of Reference I.8-10).

f ailure~ review was conducted using the criteria given in Section I.2 of this report.

1 The determination of postulated targets and disabled systems revealed that for all HELB's within the plant, the S/RV's and the RPS equipment would not be adversely affected. Also, there are no single active failures within these systems coincident with loss of off-site power (Sections 6.1.5 and 7.5.1 of the USAR-Reference I.8-5) which could prevent these systems from performing i their intended safe shutdown function. Therefore, the reactivity control and Reactor Vessel depressurization i

function could always be achieved.

The Reactor Vessel level control function is achieved by j

l using either the HPCI or RCIC system before RI'V O I 6-1 REV 6 12/87 >

MONTICELLO depressuriza tion and by using one division of the Core Spray or the LPCI Mode of RHR after RP V depr essur iza tion. If neither HPCI nor RCIC is af fected by a HELB, then either sys tem can achieve the level control function, as sumi ng a single active failure in either system. If both HPCI and RCIC are lost through a HELB o r H ELB and s ing le act ive failure, the RPV can be depressurized with the S/RV's a nd either the LPCI Mode of RHR or CS ca n be used to res to re and ma int ain RPV level. The LPCI and CS ars redundant to each other, and each essential division has a CS and LPCI System. There are no single active failurec in the CS, HPCI, RCIC and LPCI Mode of RHR wh ich preve nts the reactor vessel level f unction frce being achieved.

The decay heat removal function is achieved by the HPCI System before RPV depressurization and by the SDC Mode of RHR or the CS, S/RV's and SPC Mode of RHR after depr es s ur izat ion . If the HPCI system is not ava ilable as the result of a HELB or a single active failure, the RPV can be depressurized by using the S/RV's, and decay heat removal established by the SDC Mode of RHR or the CS, S/RV's, and SPC Mode of RHR. Each essential division contains a CS a nd RHR system. Therefore, these systems provide ' redundancy for the decay heat removal SSD function on a systems basis. In addi tion, each RHR Loop is equipped with 2 pumps (only 1 of which is req ui red for SPC or SDC Mods) a nd locked open suction valves to the suppre.nsion pool.

The valves requi red for the SPC Mode and the SDC Mode are dif f cant, so no si ng le .f a ilure of any one valve could pr eve nt the other mode from ope ra tio n. This includes the suction valves to the Suppression Chamber and the SDC connection on the recircula tion line, wh ich are electrically interlocked. If the valves, aligned to the Suppression Pool fail to close, there is no other single active failure and CS and the SPC Mode con be used for decay heat removal. If the single active fa ilure is in the CS system, the SDC Mode of RHR can be used. Therefore, on either ECCS division evnn with a single act ive fa ilure , the decay heat removal f unction is pos sible .

In addi tion to these sys tems , the RHR Service Water

! System is requir ed to remove decay heat from the RHR System and trans fe r the heat to the river. Each RHR Se rvice Wate r System loop is equipped with dual pumps, so a single active failure of either pump does not af fect the abili ty of the ide nt i fied lo op to f unct ion. The only nonredunda nt act ive component in l

l either loop is the con trol valve wh ich r egula te s RHR I.6-2 REV 6 12/87 l

~ - - -. . - - . ...

MONTICELLO J

O h: Se rvice Water pressure. If this valve is tae singlu active failure, suf ficient time axists to man.ially ope n the valve or 0 ce the valve. In this ins tar.ce fo r the HELB's pos tu b h both es se nti al power sys tems or either HPCI ar.) grt not affected. In eithar case, hea t r e je ction to ht rive r would not be r equiced fo r seve ral hours , whi( auld be suf ficient time to witch to the other es s- al division or open the affected valve.

The function of the Eme rge ncy Service Water 'ESW)

System .is to provide cooling to the Diescl Genera tors, the Core Spray and RHR Pump Motors, a nd the ECCS rump Room Cool- ing Units. The only compone nt s on this system subject to a single active failure are the four a pumps (P-lllA, B. C , a nd D ) . If either pump P-lilA or B fails, cooli ng water to both D/Gs can still be supplied from the other pump through the discharge piping cros M ;.e in the Intake Structure. The crosstie also pe rmitt coo ling of the ava ilable D/G by the ESW pump _ from the ot he r division if the single act ive failure is other D/G. Should the single active f ailure be eithe r of the other ESW Pumps ( P-lllC or D), the remaining pump can provide coo ling water to its Os i equipment. Moreove r, cooli ng water can also be supplied to the CS and RHR pump motors and roon coolers by align ing one of the Se rvice Water Punps to the D/G. This realignment can be accomplished for any pos tula ted HELB, becau se of the ESW valve arrangeme nt and the time required for res ta rti ng the pump moto r cooling Gnd room cooling for the C5 and RHR pumps

( approximately . 2-1/2 hours ) .

The rema inde r of che auxiliary sys tems (i.e., D/G Auxilia ry Systems, DC Powe r Systems, and HVAC) can perfo rm their f un ct ion assumi ng a single active fai. ce. the D/G Auxiliaries l include the air start system, cooli ng cys tem, the oil trans fe r sys tem , and the ele ctrical control system. The only act ive component s on the air start sys tem ae the redunda nt air start valves. The cooli ng system has no act ive components excep t fo r the ESW pump, wh ich was discussed previously. The oil day tank for the diesel genera tor contains enough fuel for the diesel to run fo r 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. (Se ction I.8-4.1 - Ro fetence I.8-5). For a a

f ailure of the essential fuel oil trans fe r pumo, gasoline dr i.ven ; ort ab le pump exists to tra ns fer fuel oil from the sto rage tank to either day tank. The electrical con trol power for the diesel gene ra to rs is s take n from the DC power system. D" only act ive l

components on tnis system are the bre ake rs at the s

Essential MCC's. A s i ng ' ,, f ailure of one of theco I.6-3 REV 6 12/87 i: c- -

MONTICELLO b reake rs would not preve nt the functionality of the DC power system. Similarly, for the entire 25 0V a nd 125V DC powe r sys tem, the only active components are the bre ake rs. No si ng le act ive f a ilure can pr eve nt the system frcm f un ct ioni ng , and no components other than the 12 5 DC bre ake rs on MCC's 133 and 143 are subject to H EL B ' s .

The only ins trume ntation required fo r safe shutdown, wh ich is af fected by HELB's, are the Suppression Pool Temperature Monitoring Sys tem (SPTMOS) cable s fo r Divis io n II. Therefore, should the si ng le act ive f ailure be the Division I microprocessor of the SPTMOS, Sup press ion Pool Tempera ture could s t ill be de te rmined by me asuring the ele ct rical res is tance across the RTD le ads of the Divis ion I SPTMOS. This can be done at the Division I SPTMOS Panel in the Eme rgency Filtration Building . For a ny other H ELB, neither division of the SPTMOS is af fected .

The remaining sys tem required is the essential power sys tem co nsisti ng of the two diesel gene ra to rs , 4KV swi tchge ar , 480V load ce nte rs and the es se ntial 460V MCC's. The only potential act ive f a ilur es are the diesel ge nera to rs and their res pe ct ive D/G breaker to each 4 KV bu s . The other switengear and bre ake rs on the power distribution system do not ch ange pos ition and are not subject to an act ive f a ilure .

For the single active failure of one of the diesols, the other diesel alone can supply enough power to its d ivision fo r unit shutdown. If the other division's equipme nt had been damaged by a pos tulated H ELB to th e extent that a path to safe shutdown could not be provided, an ope rating procedure exists, wh ich allows the diesel ge nera to r to be realigned to the other division. This will supply power to that divis ion and it will receive all necessary auxiliary support using only the equipme nt of the opposite division. In this way, eithe r diesel ge nera to r could supply either division with power, mitigate the ef fects of postulated HEL3's, and safely shutdown the unit.

For the pos tula t ed HELB's, it can be conc luded here that there are no coincident single act ive f ailures which would prevent safe shutdown of the unit.

I.6-2 HELB Evalua t io ns by System Evalua tions are provided in this repo rt for each high energy system with high energy piping that could not be excluded from further evaluation based upon the 2%

I.6-4 REV 6 12/87

MONTICELLO

.k criterion or the 1 inch nominal pipe size criterion.

For eacn s ys t em , an evaluat ion ' is provided for each compa rtme nt in wh ich HELB's are pos tula ted . . The evalua tions descr ibe the. ef fects of pipe wh i p , je t l impingeme nt, ccupa rtme ntal pres sur iza tion , flooding, and e nviro nment al effects. Info rma t ion on paths to safe shutdown is given, if the pos tula ted HELB's -damage SSD equipment.

-I.6-2.1 Main Steam Sys tem The high energy lines fo r the Main Steam System are lo cat ed in four c ompa rtme nt s, the Main S te am Chase (II/ 2F) in the Reactor Building, the Conde ns e r Area (X/12C), the Steam Jet Air Ejector (SJAE) Room (X/12E),

and the Turbine Operating Floor (X/30) in the Turbine Building.

The high energy lines on the Main Steam System i nelude the four main steam lines from the drywell penetrations to the equalization li nes , the turbine bypass lines to the condense r, primary steam to SJAF. line, and Primary Steam to Steam Seal System. 3 I.6-2.1.1 Main Steam Chase (II/2F)

The Main Steam Chase contains the four main steam lines and the as so ciated drain piping. Pipe whip from the main steam lines is not cons ide red a problem, since these lines are restrained at several locations in this compa rtme nt and pipe whip reactions are not toward any '

SSD equipme nt. Since the o nly pos tulated break ics c a--

tions are at the pe netra tions to the drywell and the pene tration acts as an anchor point, only circumfere n-tial breaks are requi red to be pos tula ted . . The resulta .' jets from the cir cumfere ntial breaks . do not impact ,

SSD equipment except for a small portion of the jet wh ich would hit the c ompa r tment ce ili ng.

Imbedded in the ceiling are Division II cables of SSD equipment , but these cableo would be unaf fected because.

of the concrete ce iling reinf orceme nt below the con-duits. Because of the s te am in the compartment, both HPCI and RCIC would be lost due to steam line iso la-tion. However, all other SSD systems of both divisions would be ava ilable . Depres suriza tion is accomplished by S/RV's, and LPCI and CS can be used for decay heat removal and RPV level maintenance. Therefore, safe shutdown can be accomplished.

The postulated main steam line break would cause a peak l compa rtment pressure of 21.7 psia (Reference 1.8-11),

rupturing the blowout panels to the Turbine Operating I.6-5 REV 6 12/87

M0.iTZCELLO Floor (X/30), a nd f aili ng the door to the vest side of the Reactor Buildi ng at elevation 935'-0" . The ef fects of the pres suriza tion would be from the same circum-forential break, a nd the identical SSD equipment would be af fe ct ed . No addi tional SSD equipment would be adversely af fected as a result of the pr ess ur iza t io n .

The ref o re , safe shutdown could be accomplished in the same manner as described above.

Floodi ng in the area from a main steamline break would cau se the b ot tom of the c ompa rtment to flood to a height of 1 foot, as suming the entire mass of steam leaving the break condensed in the area. This would be extremely conservative due to the loss of stee a through the blowout pan els a nd door. With this pos tulated flood he igh t, no additional SSD equipment other than HPCI and RCIC would be af fected. Wate r would not le ave the area since the bot tom of the door opening is 4 feet above the compartment floor elevation.

The peak tempe ratur e in the room would be 29 8'F and the relat ive humidi ty would immediately go to 10 0% The e nvironme ntal of fects of the Main Steam line break in this compartment were used for equipment qualif icat ion '

purposes. The re f o re , no additional SSD equipment will be cd ve rsely af fected, a nd the previou sly ide nt ified path to saf e shutdown can be utilized.

Bre aks on the Prima ry S te am Drain Line (PS15-3"-EB) we re not* eva lua ted , since both containment isola tion valves are closed during normal operation.

I It is, therefore, concluded that a path to safe shu tdown exists for any pos tulated Primary Steam break in the Main Steam Chase (II/2F).

I . 6-2.1. 2 SJAE Room ( X/12E)

The primary steam piping line (PS9-3"-ED) is routed to the SJAE Room from the Conde nse r Area. No concern exists with respect to pipe whip and je t impi ngeme nt, there is no SSD equipmen t in the area. An since evalua tion was pe rf ormed to assess the consequences of floodi ng in the c ompa rtme n t. It was assumed that all i

steam released conde ns ed in the compa rtme nt and that all the water remained in the SJAE Ro om. The result was a flood height of 8 inches. Since there is no SSD equipment in the room, there was no n egat ive of fect.

Also, the drains in the room can drain the entire volume to the sump area. In ad dit io n, the door to the SJAE Roan is air tigh t and any le akage as a result of l

the flooding woulu be neg ligible.

I.6-6 REV 6 12/87 l

l l

i MONTICELLO r

b Compa rtme ntal pressuriza tion and envi ronme ntal ef fects from this steamline bre ak wouta not ad ve rsely of fect any SSD equipment. The SJAE has adequate vent areas to the Condense r Bay (X/12C) a nl to the 931'-0" elevat ion of the Turbine Building. The door to the SJAE Room has also been modified to withsta nd the pr ess ur iza t ion .

The peak pressure in the SJAE Room Tnd peak temperature from the ste amline break were bou nded by the peak pressure and temperature ge nerated by a Main Steamline break in the ad joini ng rendense r Bay. The only SSD equipment that may be ad ve rsely af fected is the essential switchgear in Compa rtmen t ( IX/13B', . The pressure and tempe rature transient in the Swi tchge ar area is bounded by the same Main Steamline bre ak in the Conde nser Bay, for which the swi tchge ar is qualified.

Cons eq uen tly, the SJAE Room break does not ad ve rsely a f fect any SSD equipment and paths to safe shutdown e xis t.

I . 6-2.1. 3 Conde nse r Area ( X/12C)

The bulk of Main Steam piping is within the conde nser area. Pipe whip from each of the main steam lines ca n

" " " ' ""*' -' '"* S ** '9 * " Y S*""' * "*' " (SS")

A pipe wh ip of (v'/ lines (SW30A-3"-HF and SW30B-3"-HF).

line PS4-18"-ED could addi tionally damage RHR Se rvice Water line ( SWS -18 "-G F ) . The most critical break would be in the steam bypass line (PS7-10"-ED). A wh ip from this line could damage the Eme rgency Se rvice Water line SW30B-3"-HF power cables to the HPCI System, and cables of one divis ion of the sup press ion Pool Temperature Moni tori ng System (SPTMOS). The pipe whip ef fects of the other primary steam piping within this compartment could not cause any damage to safe shutdown equipment.

For jet impi ngeme n t, the wo rs t case event would be a longitudinal break in the bypass steam line (PS7-10"-

ED), which could impinge the ESW line SW30B-3"-HF and whose pipe reaction would damage the HPCI a nd SPTMOS cables on the other s ide of the line. All other pos tula ted Main Steam breaks would damage individual piping of safe shutdown systems, but loss of any one line would not result in loss of the safe shutdown capability. The reason for this is only one safety division would be af fected , and safe shutdown can be achieved from the other sa fe ty divis io n, assumi ng a single active f ailure (see Section I.6-1) .

For the bre ak that damages ESW Line SW30B-3"-HF, the HPCI Powe r Cables, a nd one division of SPTMOS, a pa th O- to safe shutdown can still be a chi eved .

4 Following the I.6-7 REV 6 12/87

MONTICELLO b re ak , RPV water level can be maintained by RCIC and the pressure maintained by the SRV's. The remaining ESW li ne ( SW30A-3"-HF) can be used for the Division "I" RHR and Core Spray pumps and room coolers once the RPV is depressurized. Should the single active f ailure be on the Divis ion "I" diesel genera tor, ESW Pump P-111C, or some other component which does not allow ope ration of the Divis ion "II" RHR and CS System, the Divis ion "II" RHR and CS Pumps can be used for up to 2 1/2 hours without coo ling water. Duri ng this pe r iod a Se rvic e Water Pump can be aligned to either D/G to provide coo ling water to these pumps. If the si ng le act ive f ailure is in the RCIC system, then both D/G's are ava ilable , and P-111C is also ava ilab le. Hence, safe shutdown would be accomplished using the S/RV's to depress ur ize the RPV and CS/RHR fo r RPV water level control and decay heat removal.

The flooding that would result from the Main Steamline break would not affect any SSD equipment, as there are no SSD components in the bottom of the Condense r Bay.

Mo reove r, if all wa ter conde ns ing from the released steam were to remain in the Condense r Bay, the flood water would not af fect any other SSD equipment. This is because of two bay areas that are at a lower [

e leva tion with free vo lumes subs tantially la rge r than C the volume of water gene rat ed by conde nsing the steam f rom a Main Steamline break.

The peak compartr.antal pressure from the Main Steamline Bre ak is 15.4 psia and the temperature would be 20 6*F (Reference I.8-ll). The peak temperature in the Switch gea r Are a (IX/13B) would be 93*F and the pressure would tube be 15.4 psia. Tb doors and the conde nse r knockout blocks do not fa il under this co ndit io n The doors to the Conde nse r Bay (Reference I.8-ll).

h ave been modified to prevent their fa ilure under a Main Steamline break. The analysis that was performed showed that with the vent area in th e Co nde nse r B ay ,

the knock out blocks would not fail. The calcula ted tempe ratures and pressures in the Switchgear Area are bounded by the temperatures and pressures for which the swi tchgear is qualified. Co ns eq ue n tly , no othe r SSL equipme nt is adve rs ely af fe ct ed other than the SSD equipment directly damaged by the pipe whip a nd jet impingeme nt .

With the exception of the break on the Turbine Bypass line ( PS7 -10 "- ED ) , which damages an ESW line, the HPCI l powe r cabling , and one division of SPTMOS, all other postulated HELB's on steamlines in the Condenser Bay do not adve rsely af fect more than one division cf SSD I. 6-8 REV 6 12/87

MONTICELLO equipment. The path to safe shutdown for the Turbine Bypass line bre ak has been' previously descr ibed , and all other breaks are less severe. Therefore, a pa th to -

safe shutdown exists for a ny pos tulat ed . H ELBs in this compartment from any steamline other than mains team.

I . 6-2.1. 4 Turbine Operating Floo r ' ( X/3 0 )

The only pos tula ted HELB's in this compartment from the Primary Steam System ~are bre aks at the inlet to the High Pressure Turbine. A break at this location would not expose any SSD equipment to either pipe whip or jet' impingeme nt . Also, the environme ntal ef fects would be the same as a bre ak in the Conde nse r Bay (Reference I.8-11), and any water condensing would drain either to the Conde nse r Bay or lower elevations of the Turbine Buildi ng to areas where the water could be drained.

The SSD equipment in -these areas would not be adve esely af fected by the water produced.

~

I.6-2.2 Feedwater System The high energy Feedwa ter System piping (FW2A-14"-EB and F B 28-14 "- EB ) begins in Compa rtmen t IX/13B on e leva tion 911'-0" of the Turbine Building at the .

O di scha rge nozzle main Feedwater of feed pumps P-2A and P-2B.

lines, FW2A-14" and FW2B-14", pass The two.

through Compartment IX/13C, up into Compartment IX/19C a nd then into the Conde ns e r Area X/12C. Before entering the Reactor Building Steam Chase (II/2F), each Feedwa ter line is connected to its res pe ct ive high pressure feedwater heate rs (E-14A & B and E-15A & B) on the Turbine Operating Floor (X/30) .

The two Feedwater lines and the Feedwater r egulating station piping were seismically analyzed. Break loca-tions were sele cted based upon-the seismic analysis of the piping and the break location criteria established for seismic Category I piping. All four inte rmediate break loca tions for the Feedwa ter system were identified in Compartment IX/13B, the Reacto r Feed Pump Area at Eleva tion 911' . There were no break locations in the Turbine Buildi ng Pipe Chase (IX/19C). An addi tional break location was chos en in the Condenser Are a ( X/12C ) , as a result of the seismic analysis.

I.6-2.2.1 Reactor Feedwater Pump Area - Compartments IX/13B and IX/13C For pos tula ted HELB's on Feedwater piping in these com-the only pipe whip or jet impi ngemen t O pa rtment s, targets are MCC 133, Division I cables, the compartment I.6-9 REV 6 12/87

. .- - a..

l MONTICELLO c o ili ng , and ESW lines ESWl-3"-HBD a nd SW30A-3"-HF.

Loss of th e MCC a nd the cables does not pr event safe I shutdown, since the Division II equipment with HPCI and RCIC could be used for this. If the ce ili ng to this ccmpa rtment is damaged, essential MCC 143 on the 931'-

0" elevation could be adve rsely af fected, and a path to safe shutdown would not exist. To prevent damage to the ce iling from a Feedwater line bre ak, seve r al pipe whip restraints and a jet impingeme nt shield have been added to the area. Therefore, a path to sa fe shutdown will exist.

If the ESW lines, ESWl-3"-HBD or SW-30A-3"-HF'are also a ta rge t, then cooling to the Divis ion "I" RHR and CS pumps would be lost along with the MCC 133. Safe shut-down would be acc omplished by using HPCI or RCIC for RPV level control and the S/RV's for RPV pressure control. Decay heat removal would be accomplished by u sing the Division "II" CS and RHR Pumps and an RHR SW Pump. The worst single active failure would be fa ilure of the Division "II" D/G, its bus tie breaker, or ESW Pump P-lllB. For any of these failures, the capability exists to use the Division "I" D/G to powe r the Divis ion , "II" equipme n t. Hence, sa fe shutdown can be a ccomplished fo r any of these sing le act ive failures.

With a ny othe r sing le , act ive f a ilure the D/G to l l Division "II" is available and multiple pathways within Division "II" exist for achieving safe shutdown.

Flooding in this compa rtme nt is not a concern. The only equipment poss ibly af fected would be MC C 133, which would already be los t due to the other HELB effects. For the purpose of this e valuat io n , it was assumed that the entire hotwell vo lume of 80,000 gallons was pumped in to this area. No othe r SSD equipme nt- would be af fected , and mos t of the water would be contained below th e MC C in the flood cavity.

The re.naining water would be trapped in the compartment I due to a 6-inch curb in the corr idor to compa rtmen t X/16 in the northeas t corner of the Turbine Building at 25,000 this eleva t io n. In ad dit io n, approximately gallons would drain to the Mechanical vacuum Pump Area.

l Enviro nme ntal qualif icat ion of SSD in other af fected compa rtme nts was ba sed upo n the pos tula tion of HELB of breaks on the Fcedwater piping. Table I.3-1 Reference I.8-ll gives a pe ak pressure and tempe ra tu re of 15.1 psia a nd 212*F in this c ompa rtme nt and also i ndica tes a pe ak tempe ratu re in the MCC 143 area of 104*F. The MCC's are qualified for that tempe ra tur e .

Therefore, no addi tio nal SSD equipment is adve rs ely af fected.

I.6-10 REV 6 12/87

MONTICELLO O The only SSD. component potentially af fected by ccrapart-mental press ur ization is MCC 14 3, a nd only if the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire barrier between the compa rtme nts fail.

Evaluat ions were conduct ed to assess whe ther the fire barrier rema ined intact. The res ult s (Page 3-4 of Reference I.8-ll) obtained showed that .the fire barrier was not breached with a dif ferential pressure equa?- to across fire barrier. The r ef or e ,

0.4 psi the compa rtme nt al pressurization does not adve rsely af feet additional SSD equipment.

I. 6-2. 2. 2 Turbi ne Building Pipe Chase (IX/19C)

There are no pos tulated Feedwater Line bre aks within this compa rtme nt . Thus, pipe wh ip , jet impingeme nt ,

flooding a nd c ompa rtme nt al pressurization is not a concern. The maximum tempe rature and pressure resulting from a Feedwater Line bre ak in the Feedwater Pump area would be 14.7 psia and 104*F.

I 6-2. 2. 3 Condense r Area ( X/12C)

The break loca tion fo r Feedwater Line FW2A-14" is

/] located at an elbow be tween column lines 6 and 7 at the

\J centerline eleva tion of 934'-10". The Division ' II

- Eme rgency Se rvice Water line (SW30B-3"-3HF) is the only pipe whip or jet impingement targe t, and the ef fects of a loss of this line h ave been discussed previously in Sect ion I . 6-2.1.

The iesultant flood of the 80,000 gallons of water would f ill the r.ow Pressure Heate r Drain Bay a nd the Conde nser Pit. The ramaining water would drain to the Mechanical Vacuum Pump Are a , in which the sumps are located. There would be no standing water above the 911'-0" elevation to adversely affect any additional SSD equip.ne nt .

The previous analysis used to determine the pressure, temperature, a nd humidi ty transients for variou s pos tula ted pipe ruptures (Reference I.8-11) included the rupture of the Fe edwa te r line in the Conde nse r Area. Howeve r, the conditions generated by a Main Steamline break bound those of the Feedwater Line Break no

  • (see Table 3-1 of Reference I.8-ll). Hence, additional SSD equipment is ad ve rsely af fected by the Feedwater Line break.

Since the only SSD equipment adversely af fected by the Fe edwate r Line bre ak is the ESW line SW30B-3"-HF, it O can be concluded that saf e shutdown can be achieved for I.6-ll REV 6 12/87

MONTICELLO the pos tula ted Feedwater li ne break in the Conde nse r Area.

I.6-2.2.4 Turbine Operating Floor (X/30)

Fe edwater lines FW2A and FW2B have break lo ca t io ns at the terminal ends of the inlets and ou tle ts and the Feedwater heate rs in this area. The pos tulated H ELB 's have no pipe wh ip or je t impingeme nt targe ts. The resultant flooding would drain back to the Co n'de nse r Area (X/.1';) with the same areas flooded as a break in the Co nde nse r Area. The ca scadi ng water would not adve rsely af fect any SSD equipment. The pressure, tempera ture , a nd humidi ty tra nsie nts pos tulat ed as a result of pipe raptures included the Fe edwa te r Line Bre ak in this compa rtment. The resultant peak pressure was calcula ted to be 14.9 psia and the peak tempe ra ture 18 9 ' F . This tra nsie nt was bounding for the Turbine Operating Floor. Therefore, these conditions were used for qualifying SSD components, and no SSD equipment is adversely af fected by a break of the Feetwater Line on the Turbine Operating Floor.

I.6-2.2.5 Main Steam Chase (II/2F) ,

Break loca tions in this area cons is t of the terminal ends at each primary containment pene trat ion a nd one intermediate break location on each Fe edwa te r line.

For a bre ak at the te rmin al point on line FW2A-14-ED, RCIC flow would be los t because the RCIC injection line (FWS-4"-ED) conne cts ups tre am of the pos tulated break. Also, the intermediate break loca tion for line FW2A-14"-ED is at the weld to valve FW-94-1, wh ich is also downstream of the RCIC injection point, and RCIC would again be lost. Fo r Feedwater lin e FW2B-14"-ED, a break at the terminal end would cau se the loss of the HPCI System, because the inject-ion point is upstream of the break. However, HPCI flow is not af fected for the postulated break at the inte rmediate location, because the break location is upstream of the HPCI injection point a nd the check valve on the Feedwater line (FW2B-14-ED). No safe shutdown eq uipme nt can be adve rs ely af fected by pipe wh ip or direct jet impi ngeme nt from any pipe break of the Feedwa ter System in the Reactor for conse rvatism, both Building Steam Chase. Howe ve r, HPCI a nd RCIC are assumed los t due to floodi ng (described below) and poss ible jet impi ngement due to compa rtme nt ge cme try .

Flooding in the Reactor Buildi ng Steam Chase, as a result of a Feedwater line break, would cause the loss of both HPCI a nd RCIC, since the injection valves fo r I.6-12 REV 6 12/87

MONTICELLO i P-

[

these sys tems (MO-2068 and MO-2107) would be subme rged I above the motor opera to rs unt il the door to the compa rtme nt f a iled . Water exiting the steam chase would flow along the floor of the Reactor Building 93 5' e leva tion . Some of the water would also flow down to the compa rtment containing the Control Rod Drive Pumps and into the HPCI pump room. Additional safe shutdown equipme nt would not be af fe cted by the water exiting the steam chase, because no SSD equipment is located in the af fected areas. There fore , the flooding from the postulated Feedwater line break would adversely af fect HPCI, RCIC, and the CRD Pumps. The path to sa fe shut-down would be the same as that desc ribed in Section I.6-2.1.1 with the except ion that an RHR or CS Pump would be required to maintain RPV level.

As in the case of the Conde nse r Bay (X/12C), the e nvironme ntal condi tions produced as a result of the liain Steam Line break would be boundiag to a Feedwater Line break. Since a path to safe shutdown was dertonstrated for the Itain Steam Line break (see Section I.6-2.1), the Feedwater Line break wauld be mitiga ted and safa shut down conducted in the same manner as for

- the Main Steam Line break. ,

In summa ry , a Feedwater Line break in the Main Steam Chase (II/2F) could ad ve rsely af fect HPCI, RCIC, and the CRD Pumpa. The Eme rge ncy. Se rvice Water Line, SW30B-3-HF, would not be adve rsely af fected from a Feedwa te r Line break. Paths to safe shutdown would include usi ng the S/RV's to depress ur ize the RPV and u sing the RHR and CS for RPV level maintenance and decay heat removal. . No pos tulated single act ive f ailure prevents the other division from achieving safe shutdown.

I.6-2.3 Condensate System The high energy Condensate System li nes are located in the following compartments:

(1) Condenser Area (X/12C)

(2) Turbine Building ~ Pipe Chase Area (IX/19C)

(3) Reactor Feedwater Pump Area (IX/138 and IX/13C)

The high energy Conde nsate System piping inc ludes the main Conde nsate lines (C4A-16"-GB a nd C48-16"-GB) , the Fe edwate r Pump minimum flow lines (C4A-2"-EB and C4B-O 2"-EB), a nd the Conde nsate cros s-tie (C7-16"-GB) .

I.6-13 REV 6 12/87

MONTICELLO Break loca tions were sele ct ed based upo n a seismic analysis of the pipi ng a nd bre ak lo cat ion criteria e stablished fo r seismic Catego ry I piping . The pipe runs extended from the te rmin al points on the third s tage intermediate heaters to the suct ion nozzles of the Reac to r Fe edwa te r Pumps. The inte rmedia te break loca tions on line C4A-16"-GB are in the Feedwater Pump Are a (IX/13B), and the inte rmediate break locations for C4B-16"-GB are in the Condenser Area (X/12C) . Be cau se the condensate cross-tie line (C7-16"-GB) a nd Fe edwate r Pump minimum flow lines (C4A-2"-EB and C4B-2"-EB) a re routed totally within the IX/13B Compartment, all break loca tio ns for these lines are in Compartment IX/138.

I . 6-2. 3.1 Condense r Are a (X/12C)

For the pos tula ted break loca tions in the Conde ns er Bay, the only pipe wh ip or jet impi ngement targets would be either of the two ESW lines (SW30A-3"-HF or SW30B-3"-HF). Loss of one of either of these lines and the pa ths to safe shutdown are discussed in Sect ion I.6-2.1.3 for a postulated Primary Steam Line bre ak .

The flooding caused by a Condensate Line break would be the same as the Fe edwate r Line bre ak, a nd the associated environmental conditions would be . bounded by the Main Steam Line bre ak in the same compartment. The floodi ng , compa rtme ntal pres sur iza tion , and environmental ef fects caused by a postulated Condensate Line break would adve rsely affect no other SSD equipment other than the ESW lines dire ctly af fected.

It can be concluded that paths to safe shu tdown exist for postulated Condensate HELB's in this compartment.

I .6-2. 3. 2 Turbi r i Buildi ng Pipe Chase ( IX/19C)

There are no pos tulated Condensate line HELB's within this ccmpartme nt.

I.6-2.3.3 Reactor Feedwater Pump Area (IX/13B and IX/13C)

The postulated Condensate line HELB's in these compart-ments would not af fect any addi tional SSD equipment other than the SSD equipme nt described for a Feedwater line break in the same ccmpa rtme nt . The of fects of pipe whip, jet impi ngeme n t, flooding, c ompa rtmental pres suriz a tion , and environme ntal ef fects are less adve rse than for the Feedwater line bre ak. Therefore, the same evalua tic as , as provided in Sect ion I. 6-2. 2.1 of this report, apply, a nd the same paths to safe l shutdown can be used.

I I.6-14 REV 6 12/87

MONTICELLO V' I.6-2.4 High Pressure Coolant Injection (Steam) System The steam supply line (PS18-8"-ED) to the HPCI Turbine begins at the drywell penetration located in the Steam Chase (Compa rtmen t II/2F). The s te am supply line ente rs the Torus Area (IV/lF) and then the HPCI Compartmen t Area, (II/lE). This steam supply line is a high e nerg y line from the drywell pe netra tion to the steam supply valve located on the HPCI Turbine.

I.6-2.4.1 Main Steam Chase ( II/ 2F)

Pos sible pipe wh ip targe ts include the Feedwater and Main Steam lines wh ich are ausumed to be unaf fected, becau se they are la rge r and thicke r walled than the HPCI steam line. No other SSD equipment is af fected by a HPCI steam line pipe whip.

Jet impingeme nt targe ts include the ce ili ng through which Division II embedded conduits are r outed. The e f fe ct s of jet impingeme nt on the ceiling is discussed in Se ctio n I . 6-2.1.1. However, the ef fects from a HPCI steam line break are less severe than a break of one of the Main Steam lines. Eme rgency Se rvice Wate r Piping

/ (SW30B-3"-HF) a nd the RCIC Steam line (PS17-3"-ED) can

( be damaged by jet impingeme nt from a lo ngi tudin al line break. This results in loss of HPCI, RCIC and the Division II ESW.

For this pos tula ted H EL B , the path to safe shutdown would consist of using the S/RV's for RPV depressuriza-tion and Division I RHR and CS to maintain reactor wa te r level a nd remove. decay he a t. With the presen t arrangeme nt of the ESW System and capability to have either diesel power either es se nti al bu s , no si ngle active f ailure can prevent the use of either Essential Division to safely shutdown the unit. If the Division II CS and RHR Pumps are required, as stated before, ap proximately 2-1/2 hou rs of pump ope rat ion is accep tab le be fo re pump and Room Cooli ng is req ui red .

Therefore , ample time exists to i sola te the broken ESW Line (SW308-3"-HF) a nd start a SW Pump to provide the cooling wate r.

Compa rtme nt al flooding would not af fect any other SSD equipment other than the piping lines directly af fected by the jet impingeme nt . Assuming all of the steam mass from the HPCI line were to conde nse in the compartmen t, the result ing flood would cove r the floor to a height of approximately 1 fo ot . Since the bottom of the door to the room is 4 feet above the floor, no water would e scape from the Main Steam Chase. Thus, no other SSD i

I.6-15 REV 6 12/87 j

MONTICEiLO equip;ne nt would be af fected other than the equipme nt descr ibed ab ove , fo r wh ich a path to sa fe shutdown ha s been demonstrated.

The compa rtme ntal pressuriza tion and peak temperatures associated with the HPCI steam line bre ak in the Ma!.n Steam Chase a r. ! - f fe ct s in the ad joini ng compa rtme nts are bounded by Main Steam line break in the Main Steam Chase. Therefore, no additional SSD equipment would be ad ve rsely af fe cted.

I . 6-B. 4. 2 Toru s Area ( IV/lF)

No intermediate brean loca tions were pos tula ted on the HPCI Steam line in this compa rtment.

I.6-2.4.3 HPCI Room (II/lE)

Only HPCI Sys tem components are located in this area.

Any postulated HELB on the HPCI Steam Line would af fect only itself. The 4KV power cables for the Division II Core Spray and RHR pumps in the adjoining Equipment and Floor Drain Tank Room (II/lD) a nd the C RD Pump in Compa rtmen t (II/lG) are qualified for the expe cted t

e nvi ronme ntal cond i tie ns . Even if this equ .pment were los t, sa fe shutdown could st ill be acc omplished using the Divis io n I equipment with any pos tula ted sing le act ive f a ilure . Also, a ny flooding in tne HPCI Area could possibly adve rsely af fect only the 4KV cables to the RHR and CS pumps of Divis io n II. Sa fe shutdown could be accomplished using the same path to safe shut-down as described above.

I.6-2.5 RCIC (Steam) Sys tem The ide nt ified high energy piping for the RCIC System is the ste am supply line (PS17-3"-ED) to the RCIC Turbine from the drywell penetration. This line begins in the S te am Chase (II/2F), runs down in to the To ru s Area (IV/lF) a nd into the RCIC Compartme nt (III/lC).

I.6-2.5.1 Main Steam Chase (II/2F)

For a pos tula ted HELB on the RCIC Steam line , the only SSD equipment af f e ct ed by either pipe wh ip or je t impingement would be the ceiling to the compartment, in which Division II control and power cables are located, the HPCI Steam line ( PS 18 -8" -ED ) a nd the ESW "B" line (SW30B-3"-HF). The ef fects on the ce ili ng are bounded by ef fects from a pos tulated Main Steam line break, a nd so have no impact on safe shutdown. The steam escaping from the RCIC line break would isola te both HPCI and I.6-16 REV 6 12/87

MONTICELLO b RCIC. Loss of both HPCI and RCIC and the ESW "B" line

(/ was discussed in the se ct ion on the HPCI Steam lin e breaks in the main Steam Chase (I.6-2.4.1). The path to safe shutdown would be the same.

Floodi ng from the RCIC Steam line break in the main Ste am Chase, assuming all steam condenses a nd remains in the comp a rtme nt, would result in approximately 1 inch of wate - at the bot tom of the c ompa rtme n t. The e nvironme ntal condi tions ge nera ted both in the Main Ste am Chase a nd in ad joini ng compa rtme nts would be bounded by the ef fects of a pos tula ted Prima ry Steam line bre ak. Therefore, no new adve rse conditions have been produced, and paths to safe shutdown described for the pos tulat ed Main S te am line bre ak are applicable here.

I. 6-2. 5. 2 To ru s Are a ( IV/lF )

The only pipe wh ip or jet impingeme nt targe t in this compa rtment is the ESW "B" line (SW30B-3"-HF). HPCI would not be af fect ed because it is not a target and the steam line is not i so lat ed . Consequently, the path to safe shutdown would be the same as described for the Main. Steam Chase with the ad dit ion that HPCI could be p]

t used.

The pos tula ted floodi ng in the Toru s Area from a RCIC Steam line break would be less than 1 inch in heigh t, and no SSD equipment would be af fected. The SSD equip-ment in the Torus Area a nd the ad joini ng corner rooms is qualified for either a HPCI Steam line or RCIC Steam line break with the HPCI Steam line bounding. The SSD equipment is qualified fo r any anticipa ted pressure, temperature, and humidity e nvironment as a result of an RCIC break in this compa rtme nt, and no additional SSD equipment is adve rsely af fected.

I . 6-2. 5. 3 RCIC Compa rtme nt (III/lC)

The only SSD components in this compa rtme nt are the RCIC equipment with the exception of the Division I RH R and Core Spray pumps' 4KV power cables. It is possible due to the RCIC pipe ge ome t ry, to c au se a jet impingeme nt on these cab le s . If these cable s are damaged, only the Division I CS and RHR pumps, and the RCIC system are lost. By using the rema ining SSD equipme nt, with any one single f a ilur e , a path to saf e shutdown exists. For the wors t case, los s of the Division II Diesel ' Genera tor, the Divis io n I Diesel O could be cross-tied to power the Division II equipment V and safe shutdown would be a chi eved. For flooding, I.6-17 REV 6 12/87

MONTICELLO compartmental pressurization, and environmental ef fects (

only these power cables would be af fected. Since a l path to safe shutdown has already been demo ns tr ated ,

assuming a loss of the Divis ion I ECCS pumps, i evaluation of the other affects is not required. l l

I.6-2.6 Reactor Wate r Clean Up ( RWCU) System The Reactor Water Cleanup high energy line (REW3-4"-ED) begins at the drywell pene trat ion and ou tboa rd RWCU isola tion valve. This line supplies reactor water to the RWCU heat exchanger a nd the RWCU pumps. The RWCU return line (REW6-3"-ED) retur ns the water to the reactor coolant system. All high ene rgy lines are in the RWCU compa rtme nt (II/3D), except the return line which connects to RCIC a nd HPCI injection lines in the steam chase (II/2F) a nd is routed through the MG Set Rocm (V/3A) a nd the no rt hwes t s ide of the 935'-0" of the Reactor Building ( II/ 2C) .

I .6-2.6.1 RWCU Are a (II/3D)

Pipe wh ip targe ts consist of condui ts wh ich supply mot ive power to the Divisio n II reac tor sample line isola tion va l ve', both Division II core spray outboard i nje ction valves, an RHR containment s pray valve and the Prima ry Containme nt Atmospheric Control (PCAC) i so lation valves. Redundant Divis io n I inboard containment isolation valves loca ted ins ide the primary containment are ava ilable to i so la te th e CS aCore nd reacto r s ample lines for safe shutdown. The Spray i so lat ion valves and the PCAC valvec are no rmally closed. Los s of powe r to these valves would cause no change of state and create no adve rse concerns.

Jet impingeme nt in the RUCU compa rtme nt ca n targe t any of the above valves. Ad dit ional jet impingemen t t arge ts include both Division II Core Spray injection valves, an RHR containment s pr ay i so lat ion valve, the Reactor Sample line and isola tio n valve, and the Prima ry Containme nt Atmosphe ric Con trol i so la t ion valves. The RWCU outboard containment isola tion valve is not a jet impi ngeme nt ta rge t bec au se of pipe geometry.

Redunda nt valves ins ide the cont ainme nt or located out l

s ide this compartment, mitigate any concerns on loss of the above components with one exception: the loss of air supply to the PCAC i so la tion valves A0-2386 and A0-2387. These valves have air infla ted seals, which l

deflate on loss of air. The loss of these seals does not create an ad ve rs e condi tion fo r a RWCU break, becau se the le akage across the se als is within I.6-18 REV 6 12/87

MONTICELLO

()

s Technical Specification limits. Therefore, the unit can be safely shutdown even with the PCA C i so la t io n valves seals de fla ted . For the other valves af fected ,

safe shutdown can be accomplished using the Division I equipment and the HPCI System. Su f ficient redunda ncy exists to be able to shutdown the unit assumi ng a single active failure.

Compartmental flooding would not be a problem, since no othe r SSD equipment would be ad ve rsely af fe cted by flooding other than the SSD equipment directly af fected by pipe whip or jet impingement. The flood he ight in the RWCU Area would be no more than 5 inches.

The effects of c ompa rtme nt al press ur iza t io n, tempe ra-ture, and humidi ty were calcula ted as part of the e nviro nme nt al q ualif ication analysis of Monticello (Reference I.8-ll). Any SSD component affected by the RWCU line bre ak is qualified for the pos tulated e nvironme nt .

Outside the RWCU compa rtme nt in compa rtme nt II/3D are two Division II c able tr ays which are jet impingement R t a rge ts . Should these cables be damaged, safe shutdown of the unit can be accomplished by using only Division p I equipment and the,HPCI Sys tem with power supplied by Q- either diesel gene ra tor . The c able tr ays are approximately 5 feet from the floor and would not be af fected by a ny flooding. In fact, no additional SSD equipme nt would be af fected by any floodi ng frcm a pos tulated RWCU HELB in the nort hwes t corner of the Reactor Building. The same analysis (Reference I.8-11), wh ich pos tulat ed 'a n RWCU HELB in the RWCU Room also assumed a break outside the .coom and used the maximum pressure, temperature, and humidity conditions developed by the analysis. Table 2-2 of Reference I.8-11 designates that open areas of the Reactor Building were given the same peak pressures and temperatures as the RWCU Area for e nviro nme nt al q ualif icat ion purposes. No other SSD equipment is adversely af fected from a postulated HELB in this area.

)

I . 6-2. 6. 2 MG Se t Room (V/3A)

Jet impingeme nt targe ts are limited to the two powe r distribution panels for the RHR air compressors and the two wall mounted Containment Atmos phe ric Monitoring panels, wh ich are not r equired for safe shutdown. No p ipe wh ip targe ts are located in this compa rtme nt because the closes t SSD component is farther than the pipe whip mome nt arm (Ref e rence I.6-18). Loss of both p RHR air compressor power fe eds is not a concern, since d

I.6-19 REV 6 12/87

\

MONTZCELLO 1

the air receive rs are not af fected by the je t. &I Additionally, the RHRSN valves could be opened manually W (

if compressed air is un ava ilable . With the rema inde r of the SSD systems ope r ab le , . ap pro xima t ely 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> would be ava ilable to open the valve. With these 2 components lost a nd a ny othe r si ngle fa ilur e , numerou s paths to safe hutdown would still exist.

Because the room has both a 6 inch curb around the room a nd floor drains, flooding in this ccmpa rtme nt would remain in the room and not af feet any other SSD equip-ment. The same would apply to the e nvi ronme ntal co nditions ; i.e. , no other SSD other than the equipment within the MG Set Room would be af fe cted by a break in this room.

I.6-2.6.3 Reactor Building Open Area 935'-0" Wes t Side (II/2C)

The poss ible pipe wh ip or jet impi ngeme nt ta rgets in this area are a series of Division II control cables located in the nort hwes t corner of the Reacto r Build-i ng . Safe Shutdown would bo accomplished in the same manner as that descr ibed i .: Se ct io n I.6-2.6.1 of this r epo rt . The floodi ng would not ad ve rs ely af fect any ad ditional SSD c orpone nt 3 in this compa rtment a nd in the ad jo ining compa rtme nts (Main Steam Chase -

II/2F a nd th e C RD Pump Room - II/lG) bec au se all components a re loca ted above the floor and the depth of the flood would be less than 2 in che s . Like the c ompartme nt s described in Sect ion I.6-2.6.1, envi ronme ntal qualification of equipment in this compartment is based upon maximum pressure, tempe ra tu re and humidi ty co ndit io ns developec by. a RWCU HELB in the RNCU Are a a nd the H ELB of a Main Steam Line break in the ad joini ng Main Oteam Chase. The re for e , no ad ditional SSD equipment is ad ve rsely af fe ct ed by the e nviro nme nt al co ndit ions gene rat ed by pos tula t ect RWCU H ELB 's .

I.6-2.6.4 Main Steam Chase (II/2F)

There are no pipe whip targets in this area. Any other concerns in this area are bounded by other la rger pipes discussed in Sections I.6-2.1, I.6-2.2, and I.6-2.4.

Check valves at the RWCU i n je ct ion points will prevent loss of both HPCI and RCIC flow from a break at either e ns ure the te rmin al end. Also, the check valves availability of both HPCI and RCIC for a break on the nonseismic po rt ion of the RWCU line in the Steam Chase area, as suming the systems are not iso la ted on high tempera ture in the compa rtment resulting from the

! )s tula ted RWCU HELB .

I.6-20 REV 6 12/87

MONTICELLO Safe shutdown would be accort.plished in the manner de-scr ibed in Section I.6-2.1, assuming HPCI & RCIC are not ava ilable .

I.6-2.7 Core Spray System Within the Core Spray (CS) system, only a small portion of each injection line was determined to be high ene rgy piping requi ring further evaluation. The portion of the i nje ct ion lines identified as h igh e ne rgy , runs from the containment penetration out to the firs t normally closed Core Spray injection valve. Line TtC-8"-ED is located entirely within the Reactor Water Clea nup compart cent (II/30) . The remaining Core Spray line TW i l- 8 "-E D is located within a block wall compa rtmen t (I/3B) above the Reactor Bu ilding 962'

?.ev e l . For both lines, the break loca tions were sbs tulated at the (i.e., at ethe te rmin al nds of the high ene rgy drywell penetration portion of the line and at the outboard containment isolation valve) .

I . 6-2.7.1 Cca.pa rtme nt Above 9 62 ' Elevation (I/38) g The only jet impi ngement ta rgets in this c ompa rtment are the block walls sur rounding Core Spray Injection Valve MC-17 5 3 a nd pipi ng lin e TWi l- 8 " - ED . A break on this line may result in block wall missiles which could pote nt ially impact the i ns t rume nt conduits fr o., r ack s C55 and C 5 6. Howeve r , since the amount of energy released due to a HELB is limited by the inboard check valve, and since the phys ical sepa ra tio n of these conduits a nd the blo ck wall is co ns ider able (The conduits are above the block wall and 20' away), damage to the conduits is not cons idered feas ible.

Compartment pressurization .will be insignificant due to the limited available ene rgy in the line and the avail-able vent area for the compa rtme nt . The RHR Head Spray Inje ction Valve MO-2026 a nd the i ns trument lines at pene tra tio n X-29 will be subject to the envi ronme ntal ef fects because of their locat ion within the compart-me nt . Howeve r, these components are qualified for the e nviro nme nt. Flooding is not a concern due to the small volume of water contained be twe en the valves

( ap proximat ely 50 gallons), and bec ause of the available drain area. The drain pa a of water would also not adve rsely af fect any SSD equipment.

There are no pipe whip targets in the proximity of this line.

I.6-21 REV 6 12/87

l l

! MONTICELLO I

I .6-2.7. 2 Reactor Water Cleanup Compartment ( II/ 3D)

! Break locations for the Core Spray Injection lin e TW7 -

8"-ED are pos tula ted at the drywell penetration and at L the no rmally closed Core Spray i nje ct ion valve MO-l 1754. Because of the normally closed inboard check valve, the ene rgy ava ilab le in the pipi ng line is l This pos tula ted line break would not be as limited.

seve re as a pos tulat ed RWCU line break. Any effects l f rom a pos tulated Core Spray line break are, therefore, bounded by a postulated break of the RWCU li6e. There {

are no pipe wh ip or jet impingeme nt t a rge ts fo r the l postulated break locations on the CS line. {

I.6-2.8 Residual Heat Removal System The only port ion of the RHR System piping req ui ring f urther evaluation are the po rt ions of the LPCI i nje c-tion lines, TW20-16"-DB and TW30-16"-DB between the ou tboa rd containment i so lat ion valves a nd the prima ry containment penetrations. These two lines have normally closed check valves ins ide the containment to preve nt reve rse flow. Therefore, the e nergy release for these pos tula t ed line bre aks is negligible because cf the small volume of fluid contained be twe e n the l inboard a nd ou tboa rd contai nme nt i so lat ion valves l

(approximately 17 0 gallons of water).

I . 6-? . 8.1 RH R Val ve Compa r tme n t (I/2G) l The postulated break locations include the terminal end of line IW30-16"-DB at d rywell pene tra t ion X-13B and the terminal end at the LPCI injection valve MO-2014.

Because of the la rge ve nt a nd drain areas a nd the limited energy release, there are no floodi ng or com-l pa rtme nt press ur izat ion concer ns.

The only pipe wh ip targe t consists of a radwaste line not required for safe shutdown. The only SSD component that is a possible impingeme nt t arge t is the RHR shut-down coo li ng s uct ion line, REW10-18"-ED. Howeve r ,

f ailure of this line will not preve nt decay heat removal, since the S RV 's and Core Spray are s t ill ava ilable to remove the decay heat from the RPV and tra ns fer it to the Suppression Poo 1. The Suppression Pool Cooling mode of RHR is unaf fected and can be used t to remove decay heat in the Suppression Pool.

I.6-2.8.2 RHR Valve Compartment (II/2H)

The postulated break locations include the te rminal end of line TW 2 0- 1 6" -DB at drywell pe netra tio n X-13A and I.6-22 REV 6 12/87

MONTICELLO f~) the terminal end at the LPCI injection valve MO-2015.

C/

There are no ad dit ion al safe shutdown c ompone nts in this compartment other than line TW20-16"-DB and valve MO-2015. The re fore, pipe whip, jet impingement and an adve rso environme nt are not a concern. Flooding and compartment pressurization are not a concern due to the low volume of water contained in the high energy line portion of the line ( Approximately 160 gallons) and the large vent and drain areas in the compartment.

I.6-2.9 HPCI (Wate r ) System The high energy portion of this sys tem cons ists of the HPCI inject ion line TW3-12"-ED from the normally closed HPCI inj ection valve (MO-2068) to the Feedwater line.

This portion is located entirely within the Steam Chase Area ( II/ 2 F) , a nd the effect of a pipe break on this line is bounded by the other h igh energy lines pre-viously described fo r this area (See Sect ion I.6-2.1 and'I.6-2.2).

I.6-2.10 RCIC (Water) System The high energy portion of this system consists of the RCIC inje ct ion line, FWS-4"-ED, from the no rmally c losed RCIC injection valve (MO-2107) to the Feedwater line. This po rt ion is lo cat ed entirely within the Steam Chase Area (II/2F), and the ef fect of a pipe break on this line is bounded by the other high ene rgy lines previously described for this area (see Sections I . 6-2.1 a nd I . 6-2. 2 ) .

I.6-2.11 Standby Liquid Control Line CH2-1 1/2"-DC from containment penetration X-4 2 to check valve XP-6 is n ot routed in the vicinity of any s af e ' shutdown compo nent s . Two inboa rd check valves will limit any energy release, and because of the small volume of water contained be twe en the inboard contain-ment check valves a nd valve XP-6 ( ap proximately 20 gal lons ) , compa rtme nt pr es suriza tion and flooding are not a concern. The compa rtment (I/2B) has an extensive drain sys tem and a large f ree volume.

I.6-2.12 off-Gas System The steam supply line SHP101-4" is routed from the Condense r Bay Area through the SJAE Ro om. The ef fects cf postulated breaks on this line in the Condenser Area are bounded by the pos tulated breaks on other lines O such as Feedwater and Primary Steam (see Sections I.6-I.6-23 REV 6 12/87 l

MONTICELL0 2.1 a nd I.6-2.2). A transient analys is has been i recently completed for the SJAE Ro om , a nd the result s j have shown that the Main Steam Line break in the (

l Conde nse r Bay is bounding for pressure, tempe ra tur e ,

and other e nvironme ntal ef fects. There is no SSD equipment in the SJAE Ro om to be af fected by a pipe whip or jet imping eme nt . Fos tula ted breaks on line S HP101-4" are bounded by other pos tulat ed HELB's, fo r which paths to safe shutdown have been demons tra ted

( se e Se ct io ns I . 6-2.1 a nd I . 6-2. 2 ) .

I.6-3 Table of System Ef fects Table I.6-2 shows the ef fect of speci fic high energy line bre aks by compa rtme nt a nd system. This table includes the req uired auxiliary systems wh ich are co ns idered pote nti al HELB targets. Any system not affected by c pos tula ted HELB is to be cons ide red ava ilable to suppo rt safe shutdown. The meaning of the letter codes used in the table are as follows :

F - Prima ry failure as a direct result of a line b rc ak .

A- System is unaf fected by a line break and is available to support safe shutdown.

U- The system is unavailable due to the failure of a required f un ct ion or c omponent as so cia ted with anothe r sys tem.

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MONTICELLO V I 6-4 Use of RCIC In Safe Shutdown Sequence Based upon the discussions in Sections I .6-1 and I.6-2, it is appropriate to discuss the need for the RCIC system for the attainment and maintenance of a safe shutdown. The RCIC system is not required for any pos tula ted HELB in any of the ide ntif ied compartments. Howeve r, for most postulated H ELB'S RCIC is ava ilable to support safe s hu tdown . This is the result of the following evaluation.

For any pos tula ted HELB, either HPCI is or is not ava ilable. If HPCI is ava ilab le , RCIC is not required, since HPCI could perform the same function. If HPCI is not ava ilable due to a si ng le act ive failure or damage from the pos tula ted HELB, the unit is shutdown using the SRV's (RPV depressurization), Core Spray (RPV water level maintenance), and RHR (decay heat . removal and RPV water level maintenance). No single act ive fa ilure cotild ef fect more than 1 SRV, 1 CS division or 1 RHR divis ion.

For the case of the loss of either ESW line and HPCI as result of a postulated H ELB, both divisions of RHR and CS would be available for up to 2 1/2 hours. This time

(. interval is more than suf ficient to isolate the damaged s

ESW lines and trans fe r eme rge ncy powe r to a Se rvice Water Pump. The Service Wa te r Pump will provide the necessary cooling water to the RHR and CS pump rooms to allow continuous use of these pumps for the duration of the safe shutdown sequence . Hence, safe shutdown can be accomplished for a ny pos tulated H ELB without requiring the RCIC System to support the safe shutdown.

O I.6-27 REV 6 12/87

MONTICELLO I.7 CONCLUSION A systematic review of the effects of pcatulated high energy line breaks was performed for the Northern States Power Monticello Na: lear Generating Plant. It is cord uded that for every postulated break of a high energy line coincident with loss of of fsite oower and a single active failure, a path to safe shutdow'n exists.

O l

l I

I.7-1 REV 5 12/86

p MONTICELLO l l

( 'I . 8 REFERENCES I.8-1 Postulated Pipe Failures Outside Containment, Monticello.

Nuclear Generating Plant, Monticello, Minnesota, Sep'tember 9, 1973 (File 4 NSP730.0008).

I.8-2 Letter from A. Giambusso, Deputy Director for Reactor Projects, to Northern States Power Company, .

Subject:

Hign inergy Breaks Outside of the Containment, December 18, 1972 (File i Ndr730.02?ai I.8-3 Standard Review Plan 3.6.1, Plant Design for Protection Against Postulated Piping Failures in Fluid Systems outside Containment, Rev. 1, July 1981 (File i NSP730.0009).

I.8-4 Standard Review Plan 3.6.2, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping, Rev. 1, July 1981 (File i NSP730.0009).

I.8-5 Updated Safety Analysis Report for the !!onticello Nuclear Generating Plant, Northern States Power Company, Rev. 3, December 1984.

4 I.8-6 NSDO-22087, Fire Protection and Safe Shutdown Systems s Analysis Report, Monticello Nuclear Generating Plant, Northern States Power Company, June 1982 (File i NSP730.0014).

I.8-7 NSP-30-102, Review of High Energy Line Analysis for the Monticello Nuclear Generating Station, Rev. O, June 1986 (File i NSP730.0102).

I.8-8 NSP-32-101, ADDENDUM: Analysis of High Energy Lines Not Considered in the 1973 Report, Monticello Nuclear Generating Plant, Rev. 0 ,'- July 1986 (File i NSP732.0101).

I.8-9 United States Atomic Energy Commission -

Safety Evaluation by the Directorate of Licensing, Docke t No.

50-263, Monticello Nuclear Generating Plant -

"Analysis of the Consequences of High Energy Piping Failures outside Containment", July 29, 1974 (File i NSP730.0009).

I.8-10 ANSI /ANS-58.2-1980, American National Standard, Design Basis for Protection of Light Water Nuclear Power Plants Against Effects of Postulated Pipe Rupture, December 31, 1980 (File i NSP730.0018).

O I.8-1 REV $ 12/86

MONTICELLO I.8-11 Repo rt , 01-000910-1137, Monticello Nucle ar Generating Plant Environmental Ef fects Due to Pipe Rupture, Rev. 2, September 19 86 ( File # NSP740.0007 ) .

I.8-12 Updated Fire Haza rds An alys i s , No rther n States Powe r Company, Monticello Nuclear Generating Pla nt , June 1985 (File # NSP73 0.0014 ) .

I.8-13 Monticello Nuclear Generating Pla nt , Re-exami na tio n of Ap pe ndix R Separation Analysis, S ept embe r 1985 (File #

NSP730.0014).

I.8-14 Speci f ica tion fo r Piping Materials, Classif ica tio n and Sta nda rds for Monticello Nuclear Generating ?lant, 5828-12-4 0, August 1957 (File # NSPl701.0lll).

I.8-15 NSP letter to Te r ry !!axey ;

Subject:

Evaluat ion of RHR Pressure a nd Tempe ra ture Increase fo r Environme ntal Qualification Impact, June 198 5 (File # NSP73 2.00ll) .

I.8-16 Safety Concerns Associated with Pipe Breaks in the BWR Scram System, NRC Generic, Ja nua ry 1986 (File #

N SP7 3 2. 0 013 )

I.8-17 Scram Disch arge Volume Des ign Ch'a nge Re comme nda tio ns ,

SIL No. 331, Octobe r 19 8 0 ( F ile # NS P73 2. 0 012 )

I.8-18 Survey Book, High Energy Lines in Systems cons ide red in 19 7 2 H ELB An alys is, May 198 6 (File # NSP730-0013).

I.8-19 Survey Sheets and Survey Notes for Ide nt if ied Non-Be ch t el High Ene rgy Lines, Jun e 1986 (File #

N SP7 3 2. 0010 )

I.8-20 Calcula tion NSP730.0101-001 "Determination of High Energy Pipe Iccat ions for the Systems, Bechtel Icoked at in 19 73" ( File # NSP730.0101-0 01, Rev . 0).

I.8-21 Drawing, N D-10 0 4 8 2, ESW-EPT ESW "A" Divis io n Cr oss tie Piping ISO, Revision 0 10/14/87 (File i NSP762.0002).

I.8-22 Drawing, N D-10 0 4 8 3, E SW-E rr E SW "B" Divis io n Crosstie Piping ISO, Revision 0 10/14/87 (File # NSP762.0002).

I.8-23 Drawing, NH-11145, E SW-E rr E SW Crosstie Division A ESW SW30A-3"-HF Rou ting (Issued for Construct ion), Revis ion 0, 10/14/8 7 (File #NSP7 62.0012) .

I.8-24 Drawi ng , N H-l ll4 6, ESW-EFT ESW Cros stie Division B ESW SW30B-3"-HF Routing (Issued fo r Construction), Revis ion 1

0, 10/14/87 ( File #NSP7 62.0012) . l l l

I.8-2 REV 6 12/87 i

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