ML20248A931
| ML20248A931 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 05/25/1989 |
| From: | Albrecht K, Jensen R NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML112971351 | List: |
| References | |
| NUDOCS 8908090060 | |
| Download: ML20248A931 (125) | |
Text
{{#Wiki_filter: ' Operational Quality Assuranco Plan Rev 13 ,o NORTHERN STATES POWER COMPANY 414 liicollet Mall . Minneapolis, MN 55401 OPERATIONAL OUALITY ASSURANCE PLAN
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REV 13
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O Reviewed By: Date: f>//M /f[9
, Kg/ineth J Albrecht '(f V
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ITirector Power Supply Quality Assurance Approved By: < x= Date: fb / R61and J Jensen i Senior ce President Power Supply 6908090060 PDR 899630 ' A ADOCK 05000263 PDC ,..
' Operational Quality Ascuranca Plan ,
Rav 13 . Table of Contents Pace O 1.0 - Policy Statement ' 1 2.0 l Introduction 6. 3.0 Organization 7 4.0 Operational Quality Assurance Program 22 I 5.0 Modification Control 27 l l 6.0 Procurement Document Control 28 7.0 Instructions, Procedures and Drawings 30 8.0 Document Control 32 9.0 Control'of Purchased Material, Equipment and Services 34 L 10.0 Identification and Control of Materials, Parts and 36
. Components 11.0 Control of Special Processes 37 12.0 Inspection 39 13.0 Test Control 43 14 0. Control of Measuring and Test Equipment 45 10.0 Handling, Storage and Ship,tping 46 16.0 Inspection, Test and Operating Status 48 I
17.0 Nonconforming Materials, Parts or Components 51 18.0 Corrective Action 53 19.0 Quality Assurance Records 54 20.0 Audits 56 l L hppendix A - Monticello Structures, Systems, and Components Subject to Appendix B of 10CFR50 58 1 Appendix B - Prairie Island Structures, Systems, and Components Subject to Appendix B of 10CFR50 63 Appendix C - Nuclear Plant Fire Protection Program 72
) Appendix D - Revision 13 Change Summary 87 l1 ~ '
Op2 ration 31 QuSlity AccurCnca Plcn R::V 13 1.0 Policy Statement l2 m j 1.1 Northern States Power Company (NSP) has established and c is implementing an Operational Quality Assurance Program. This quality assurance program is applicable to NSP nuclear plancs that are regulated utider 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. 10CFR50, Appendix B. l3
- 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 Assurance Plan, Appendix C. l4
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- 7. NSP Plant Security Plans.
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- 8. NSF Radiation Environmental Monitoring Program.
- 9. ANSI N45.2.6-1978, Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants, as modified by Regulatory Guide 1.58, i Revision 1.
- 10. ANSI N45.2.12-1977, Requirements for Auditing of Quality Assurance Programs for Nuolear Power Plants.
- 11. ANSI N45.2.23-1978, Qualification of Quality Assurance Progran Audit Personnel for Nuclear Power Plants, as modified by Regulatory Guide 1.146, August, 1980. l4 1.3 The Operational Quality Assurance Program shall incorporate: (1) the requirements of ANSI N18.7-1976, l4 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.
('S 1. ANSI N18.1-1971, Selection and Training of Nuclear
'"j Power Plant Personnel (Regulatory Guide 1.8, Rev. 1).
- 2. ANSI N45.2-1971, Quality Assurance Program Requirements for Nuclear Power Plants Page 1 of 93
Operatien21 Quality Ancurcnca Plan Rev 13
- 3. ANSI N45.2.1-1973, Cleaning of Fluid Systems and Associated Components During Construction Phase of Nuclear Power Plants h (Regulatory Guide 1.37, 3-16-73).
- 4. ANSI N45.2.2-1572, Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants (During the Construction Phase)
(Eegulatory Guide 1.38, Rev. 2).
- 5. ANSI N45.2.3-1973, Housek'eeping 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 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 During the Construction Phase of Nuclear Power Plants (Regulatory Guide 1.94, Rev. 1).
- 8. ANSI N45.2.8-1975, Supplementary Quality Assurance h '
Requirements for Installation, Inspection and Testing of Mechanical Equipment and Systems for the Construction Phase of Nuclear Power Plants
- 9. ANSI N45.2.9-1974, Requirements for Collection, Storage and Maintenance of Qual.f.y 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, Februa,ry, 1974).
- 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 i Coatings Applied to Nuclear Facilities (Regulatory Guide 1.54, June, 1973).
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Op3raticnul Quality ACCurrnca Plcn . R;v 13 l r [ t 1.4 Management directives and departmental instructions and l procedures shall provide for compliance with
~j 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 th.is 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 I operation. Personnel shall therefore, be familiar with I tha requirements and responsibilities of the program that are applicable to their individual activities and ; interfaces. i 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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optrationni Qunlity Accurznca Pltn R;v 13 Table 1-1 Exception to ANSI N18.7-1976 l
- 1. Documentation required by ANSI N18.7-1976 may be deferred for l5 emergency work. Emergency work is defined 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.
1 2. Exceptions to Regulatory Guid.es and ANSI Standards are acceptable for those principal contractors, retained by NSP, such as NSSS contractors and A/E Firns, which exceptions have been approved by the NRC.
- 3. Section 5.1; delete this section. The provisions associated .l6 with identification of the operational Quality Assurance Program scope are explicity identified in Section 4 of the '
operational Quality Assurance Plan. l 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 l7 fourth sentence.
- 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". l&T
- 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; charge the first sentence to read "... installation or use of such items that serve a safety function". Last paragraph; change " quality" l7 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 utilized 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". l7
- 11. Section 6; delete this section. The referenced cocuments are i
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. Page 4 of 93
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'L ~12.i ' Sections 5.3.9'and 5.3.9.1; delete these sections. Emergency. .
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Operating Procedures.shall be consistent with Supplement'1-to'NUREG .0737 - Requirements for Emergency Response-
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Op3rctien21 Quality Accurenco Plcn RSv 13 2.0 4 droduction l2 i J Northern States Power Company (NSP) is involved in the construction and operation of nuclear, fossil-fueled and l9 hydro power plants. Construction of nuclear plants is conducted under a quality 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 plant operational activities and associated support activitier 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. I l i Page 6 of 93
!, 1 Oparaticnni Qu2lity AICurcnca Plan R2v 13 i
l 3.0 Organization l2 J q. G 3.1 General Requirements
- 1. NSP.shall be responsible for the establishment and execution-of the Operational Quality Assurance ou 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 dnd 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 checking, auditing, inspecting, or otherwise. verifying that an activity has been correctly performed is independent of the individual or group directly responsible for -t}
(,, performing the specific activity. 3.2 Ouality 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 nucleer 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 3.3.2.2.1.1.1,1.
- 3. The Material & 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. l 10
/'S 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. Page 7 of 93
Operatienni Quality A5curenc3 Pltn
'Rav 13. }
l 3.3' Corporate Organization With Operational Ouality Assurance l 11 Proarag Responsibilities 3.3.1 Chairman & Chief Executive Officer l This position is responsible for all NSP activities i including those associated with operating nuclear plants. This responsibility is implemented by assigning responsibility to the corporate officers of the company (See Figure 1, Corporate Organization With 13 Operational QA Responsibilities, Page 21). 3.3.2 Senior Vice President Power Supply This position is designated by the Chairman & Chief l 12 Executive Officer as responsible for the establishment
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e and implementation of an Operational Quality Assurance ! Program. Responsibilities include: l l
- 1. Engineering, construction and operation of all generating facilities.
- 2. Establishment of an Operational Quality Assurance f Plan that governs activities associated with Federal Regulation (10CFR50, Appendix B).
- 3. Establishment of Corporate Nuclear Administrative Control Directives that identify quality assurance requiremecats and positions responsible for implementing those requirements.
- 4. Providing status reports to management.
Positions repErting to the Senior Vice President Power Supply include: Director Power Supply Quality Assurance, Vice President Nuclear Generation, Vice President Combustion & Hydro 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 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 nonconforming l h activities until the adverse conditions have been corrected.
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iOperntionnlLQunlity[Assuranca Plan; Rnv?13: NO . N c . 3 .x 'AssistingiotherLcompany organizations in g~ s ' implementing quality assurance program requiremp~ts. 4., Providing Power. Supply Quality Assurance' status b? r ,-
. Dreport s to appropriate levels of management.
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. . Positions. reporting to this Director include:
Manager Nuclear Projects & Supplier Quality 15 Assurance, and Manager Nuclear Operations Quality 16-Assurance.- 3.3.2.1.1 I Manager Nuclear Projects.& Supplier Quality l 15 Assurance 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: 1.- Inspections-of nuclear fuel suppliers.
- 2. _Q uality assurance audits and qualification of suppliers.
- 3. Review and approval-of A/E, vendor and l 17 O. contractor quality assurance programs.
.4. Quality assurance reviews of nuclear procurement madeLby. general office organizations.
- 5. Quality assurance reviews of_ project specifications and procurement documents.
- 6. Preparation / review ofLinternal quality assurance programs and-procedures.
- 7. Audits / surveillance of engineering, 1 procurement, construction and testing '
activities. 3.3.2.1.2 Manager Nuclear Operations Quality Assurance 16 18 , This position is responsible for quality assurance activities associated with general office
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organizations and internal auditing. Responsibilities include:
- 1. Internal audits of all levels of the operational Quality Assurance Program.
() 2. Review of Corporate Nuclear Administrative Control Directives'and Corporate Nuclear l 19 Administrative Work Instructions. Page 9 of 93 ____ .n_____________ __ _
' Operatien31 Quality Ar:nuranca Plcn RCv 13
- 3. Maintenance of Corporate Nuclear l 19 Administrative Control Directives and Corporate Nuclear Administrative Work Instructions current with corporate commitments and policies.
- 4. Program implementation monitoring and periodic trending.
3.3.2.2 Vice President Nuclear Generation This position is responsible fcz "._. operation and physical control of the company's nuclear generating facilities Responsibilities include:
- 1. Operation of nuclear facilities.
- 2. Maintenance of nuclear facilities.
- 3. Modification of nuclear facilities.
- 4. Nuclear fhcility fuel utilization.
- 5. Operational review of new nuclear facility design.
- 6. Independent review and audit of nuclear plant operations and operating license administration. i
- 7. Corporate security.
Positions reporting to this Vice President include: General Manager Nuclear Plants, General Manager Nuclear Engineering & Construction, General Manager . Headquarters Nuclear Group, Manager Corporate Security, 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.
- 3. Review of Safety Audit Committee reports and recommendations.
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- 5. ' Training support toLall of Power Supply.
S ' Positions reporting to this General Manager I include: Monticello and Prairie Island Plant r Managers, Manager Nuclear Radiological Services $ and1 Manager Production Training. p L y .3.3.2.2.1.1 Monticello and Prairie Island' Plant Managers ( , .These' positions are responsible for ensuring L that activities ar.i operations l comply with L ~ applicable regulatory requirements.- .3 Responsibilities include: !
. J L 11. Responsibilities assigned by the operating '
i: license and the Corporate Nuclear L Administrative Control Directives.
- 2. Plant managerial control system.
- 3. . Plant' operation and maintenance.
'4. Plant. staffing, including qualifications, hiring, training,. discipline, and 4
- administration of the labor' contracts.
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5.. Development and~ implementation of the following programs:
- a. ' Preventive maintenance.
o b. Surveillance. a
- c. Material control.
- d. Operating, maintenance, and testing :l procedural systems. 20 J
- e. Fire protection.
- f. Plant quality assurance and cor..rol.
- g. Operating experience assessment. ,
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- h. Plant physical security and guard force I supervision.
- 6. Coordination of activities performed by non-
-plant staff personnel with plant operation.
- 7. Nondestructive examinations not associated with inservice inspection.
- 8. Technical manual control programs. l 22 )
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Operational Quality Accuranca Flcn
.Rav 13
! The positions reporting to the Plant Manager f having specific quality assurance l 23 responsibilities include: General Superintendent Quality, security & Administration (Monticello only) and 4 Superintendent Quality Engineering (Prairie Island). 3.3.2.2.1.1.1 General Superintendent Quality, Security & Administration (Monticello only). This position is responsible for management direction of the Quality Engineering, Security 24 and Administrative groups of the plant staff. Responsibilities include the development and implementation of the plant Administrative Control program. The position reporting to this General Superintendent having specific quality assurance responsibilities is the Superintendent Quality Engineering (Monticello only). 3.3.2.2.1.1.1.1 Superintendent Quality Engineering (each l 10 plant) . This position is responsible for the & administration of the operational Quality W l Assurance Program requirements at the plant ! level. Responsibilities include:
- 1. Implementation of the plant quality control inspection program (except ISI).
- 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 material packages, emergency equipment, and fire protection).
- 3. Audit of selected plant level activities when determined that the audit will improve plant program implementation.
- 4. Review of plant Administrative Control Directives and Instructions.
- 5. Plant program implementation monitoring and periodic trending.
- 6. Review of procurement documents.
- 7. Receipt inspection performed for plant procured items.
Page 12 of 93
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v . y 8.- Stop work' authority for nonconforming g- . activities at the~ plant.until adverse-conditions have been corrected. 3 , t
, 9. . Providing plant. quality' assurance status.
and-trend' reports;to appropriate levels -l 25 , of management. 1 v 3.3.2.2.1.2 Manager Nuclear. Radiological Services This position is, responsible'for prov'iding. . 1 support to the nuclear plants in the. areas of radiation protection, chemistry, emergency planning and_ radiation environmental monitoring. Responsibilities include:
- 1. Providing a supportive corporate radiation
' protection program.
- 2. .Providing a supportive corporate nuclear chemistry program (BWR and PWR).. l 26
- 3. Providing emergency preparedness management.
- 4. Administering NRC operating licenses and technical specifications for environmental-j-- activities.
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- 5. . Reviewing, coordinating and evaluating company emergency. preparedness and 2
environmental activities required by the NRC.
- 6. Conducting a radiation environmental 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
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training.needs of Power Supply employees, developing programs to meet these needs, and l , providing the necessary instruction.
. Responsibilities include:
- 1. Providing NRC Reactor Operator and Senior Reactor Operator license training programs.
- 2. Providing requested training for Power Supply personnel working at nuclear plants.
- 3. Providing required training to personnel temporarily working at nuclear plants.
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op; rational Quality Accuranca Pltn rov 13
- 4. Managing and operating simulator facilities.
- 5. Providing requested support for Power Supply internal nuclear plant training.
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- 6. Maintaining required training records.
- 7. Providing fire protection training.
- 8. Maintaining INPO accredited training programs.l 27 3.3.2.2.2 General Manager Nuclear Engineering & Construction This position's responsibilities include:
- 1. Design, procurement, manufacture, fabrication, construction, installation, quality control, preoperational testing, and startup of new nuclear generating facilities.
- 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. 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 sites, and for providing craft labor when requested by the plants. Responsibilities include:
- 1. Providing full management direction for all Nuclear Engineering & Construction projects assigned at the plant sites.
- 2. Providing craft labor to support projects done under plant direction and control.
- 3. Assuring procedure adherence by all Nuclear Engineering & Construction and contractor employees at the plant sites for Nuclear Engineering & Con.truction projects.
- 4. Performing Request for Engineering studies as requested by plant management.
3.3.2.2.2.2 Project Superintendent h This position is responsible for quality control for projects assigned to Nuclear Engineering & Construction. Responsibilities include: Page 14 of 93
g -yb , t - - ' g '! b; J (Opdrdtional? Quality lA2suranca Plan-
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4 rov 213 : e - 3-1.. Developing andLi implementing a quality L97-51 control program for all Nuclear Engineering
& Construction projects.
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, 2. ' Performing required-inspections and tests.
3.- Monitoring' contractor quality, control activities. m , .. 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 Nuclear Support Services, Manager Nuclear Analysis, and Manager Nuclear Technical ~ Services. 3.3.2.2.3.1 Manager. Nuclear Support Services
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n .This position is responsible for providing support to the nuclear plants in the areas'of
' licensing. administration and-safety audit and i ' assessment. Responsibilitiestinclude:
. 1. Independent review functions for operating nuclear plants.to verify compliance with operating license requirements as required L by NRC regulations.
2.. License administration for nuclear plants and liaison to the'NRC Office of Nuclear Reactor Regulation. .f f 3. Engineering and technical support to nuclear l 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 adequacy of technical specifications.
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p, , ! Operation 1. Quality Accuranca Plcn R3v 13 l
- 4. Technical expertise, information and l direction to the Fuel Resources Department J in the formulation, coordination and 1 implementation of reload designs, vendor I evaluations and vendor contract negotiations. I l
- 5. Technical expertise and direction to the Power Supply Quality Assurance Department to !
ensure that nuclear fuel meets company and I regulatory agency requirements. {
- 6. Feedback and technical expertise to the Production Training Department to ensure that j necessary expertise, information and computer physics models are available for training.
3.3.2.2.3.3 Manager Nuclear Technical Services l 28 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 include:
- 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 Company's Security l 29 Program, which includes providing trained security personnel at nuclear and non-nuclear plants.
- 2. Nuclear Access Authorization, general employee screening, investigations and other related &
security services. W
- 3. Development and implementation of Nuclear Generation's Fitness for Duty Program.
Page 16 of 93
lop 3rctienS11Qunlity A0durcnca Plcn . CR;v : 13. l- o b 3.3.2.3 Vica' President' Combustion & Hydro Operations 9 . M ~This position-is responsible for the maintenance and-Nn)? physical' control of the company's' combustion and ye ~ hydroelectric power plants, and support to the e vi nuclear plants. Responsibilities include:
- 1. Support of nuclear plant maintenance program.
- 2. Cartain electrical maintenance and' material testing activities.
Positions reporting to this Vice President include: l3 General Manager Maintenance & Testing, and' General L Manager Plant Engineering & Construction. 3.3.2.3.1 LGeneral Manager Maintenance & Testing This position's responsibilities include: 1.- Performance of plant electrical equipment maintenance.
- 2. Testing laboratory services.-
- 3. Air filter and battery capacity surveillance testing.
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,', Positions reporting to this General Manager include: Manager Electric Maintenance, Manager 30 Production Plant Maintenance, and Manager Testing. 31~
' Laboratory.
3.3.2.3.1.1 Manager Electric Maintenance This position is responsible for providing administration, supervision, and technical expertise to facilitate maintenance and testing 31
'of Power Production transmission and distribution electrical equipment.
3.3.2.3.1.2 Manager Production Plant Maintenance This-position's responsibilities inclnde:
- 1. Assisting in securing maintenance contracts.
- 2. Providing materials and special process controls.
- 3. Providing technical support and council in assigned areas.
O Positions reporting to this Manager include: Superintendent Materials & Special Processes. Page 17 of 93 _ _ _ _ _ _ = _ - -_ _
bperatien21-Quality Ascurtnca Plcn RSv.13 3.3.2.3.1.2.1 Superintendent Materials & Special Processes 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 and nondestructive examination.
- 5. Providing technical instruction in welding -
and other special processes as required. , 3.3.2.3.1.3 Manager Testing Laboratory This position's responsibilities include:
- 1. Developing and preparing NDE and Testing Laboratory procedures. j 31 i
- 2. Certifying nondestructive examination and other Testing Laboratory personnel and ,
maintaining qualification records. j l
- 3. Providing chemical, nondestructive 1 examination and physical testing and )
inspection services. 3.3.2.3.2 General Manager Plant Engineering & Construction This position's responsibilities include:
- 1. Technical support, drafting, and construction electrical testing services for nuclear plants. l Positions reporting to this General Manager include: Manager Plant Engineering. 32 l
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.oporcticnnl'Qunlity'Accuranco Plcn R;v.13 3.3.2.3.2.1 Manager Plant Engineering l 33 i ,f This position's responsibilities include:
- 1. Drafting services for the plants.'
3.3.2.4 Vice. President Transmission & Inter-Utility Services This position is. responsible for electrical system operation and.the construction, operation, and management of the transmission system. Positions reporting to this Vice President include: 34 General Manager System Operations. 35' 3.3.2.4.1 General Manager System Operations l 35 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.
- 2. Performance of plant electrical equipment .
maintenance. V{} Positions reporting to this General Manager include: Manager Electric Protection Services, Manager Operational Planning and Manager System Control Center.. 3.3.2.4.1.1 Manager Electric Protection Services This position is responsible for the maintenance, proper application and functioning of all relay protection, control, energy management, system computer, communication, and telemetering equipment at all substations, power plants and control centers. 3.3.2.4.1.2 Manager Operational Planning This position is responsible for planned trans- 36 l mission line outages, transmission line con-struction and preventative and restorative maintenance on transmission lines and generating units. I L . 3.3.3.4.1.3 Manager System Control Center ) f /~'} ts j This position's responsibilities include: !1
- 1. Coordinating and supervising operation of NSP's generating and transmission lines.
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op3raticnni Quality Accurenca Plcn Rav-13
- 2. Direct training of subordinates to achieve a 36 high level of performance.
3.3.2.5 Director Fuel Resources O This position is responsible for procurement and disposition of fuel for the company's generating l 37 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 review of nuclear fuel cycle i 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 service contracts.
3.3.2.6 Director Power Supply Support Thi.s position's responsibilities include:
- 1. Procurement for Power Supply. {
[
- 2. Safety support for Power Supply.
- 3. Controlling corporate documents and records including retrievability, storage and 38 disposition.
- 4. Printing and distributing systems.
- 5. Drawing control program.
Positions reporting to this Director include: Manager Power Supply Procurement. 3.3.2.6.1 Manager Power Supply Procurement This position is responsible for Power Supply ! procurement functions for material and services. l 39 3.3.2.7 Director Power Supply Financial Operations This position's responsibilities include: l 40
- 1. Financial operations of Power Supply.
Page 20 of 93
.f i t m ,, e Vd'
' 'se OPERATIONAL QUALITY ASSURANCE PLAN lL REVISION 13 t
_k q '*g
'I , '. 'kW ,_
3- El L ._ 4 4 , 1 ) VICE PRESIDENT DIRECTOR j NUCLEAR CENERATION FUEL RESOURCES 1 1 2.2 1 1 2.5 GENERAL MANAGER GENERAL MANAGER GEI
- NUCLEAR PLANTS E A00uaRTERS MK11AR GROUP - - MAINT(#
112.2.1 3.12.2.3 PLANT MAN..GER MANAGER MJCLEAR
- MONTICELLO ~ SUPPORT SERvlCES - - ELECTE 13.2.2.1.1 13.2.2.3.1 &
GENERAL SlFTRINTEPOENT l MANAGER MANA( QUALITY. SECURITY & NUCLEAR ANALYSIS ADMINISTRATION _ - - PLO 1 1 2.2.3.2 e 3.12.2.1.1.3 ,., SUPERINTENDENT E EM MERIS TECHNICAL SERV!CES -
& SP3
- OV AITY E G NEERING 3.3.2.2.1.1.1.3 13"2*2'3'3 1 KANT MANAGER MANAGER SPECIAL MAZ j - PRAIRIE ISLAND -
""" b
' ' *
- D i 3. 3. 2. 2.1.1
)
MANAGER CORPORATE SUPERINTENDENT SECURITY QUALITY ENGINEERING _ 13.2.2.1.1.1.1 112.2.5 MANAGER NUCLEAR
- RADIOLOGICAL SERVICES 112.2.1.2 DIRECTOR POWER
- PRODUC T RAINING
- 3. .6 13.2.2.1.3 CENERAL MANAGER NUCLEAR MANAGER POWER
- ENGINEERING & CONSTRUC. - SUPPLY PROCUREMENT 112.2.2 112.6.1 MANAGER MONTICELLO
- PLANT PROJECTS 1 1 2.2.2.1 MANAGER PRAIRIE ISLANO
- PLANT PROJECTS 1 1 2.2.2.1 PROJECT SUPERINTENDENT 3.3.2.2.2.2
- e. _ __
l
NAIRMAN & CHIEF , i 3CUTIVE OfflCER 111 !IDR VICE PRESIDENT [POWERSUPPLY _. _1 12 i j i t' ICE PRE 31 DENT COMBUSTION ClRECTOR VICE P'4E31 TENT TRANSM!33 ION
& HYORO OPERATIONS - POWER 1RPPLY CA - & ]NTER-UTILITY SERVICES 3.3.2.3 112.1 j 1 1 2.4 8tAL MANAGEQ GENERAL MANAGER PLANT MANAGER NUCLEAR GENERAL MANAGER ANCE & TESTING ENGIN. & CONSTRUCTION - -
PROJECTS & Siff' LIER OA - SYSTEH OPEFATIONS -
- l. 3. 2, 3.1 3.3.2.3,2 112.1.1 112.4.1 MANAGER HANAGER MANAGER MANAGEF. ELECTRIC IC MAINTENANCE PLANT ENG] NEE 8t1NG - -
MJCLEAR OPERATIONS OA FROTECTION SERV 1CES - i (12.3.1.1 112.3.2.1 112.1.2 1 1 2.d.1.I l {R PRODUCTION $ MANAGER OPERATIONAL MAINTENANCE - p( g jpg _ .62.3.1.2 112. d.1. 2 NOENT M2TER]ALS MANAGER SYSTEM 1AL PCOCESSEQ - CONTROL CENTER - '. 2. 3. 3. 3.1 112. d.1. 3 r,ER TESl!NG l80ROTORY 12.3.1.3 DIRECTOR POWER SUPPLY FINANCIAL OPERATIONS 112.7 Si APERTURE ricune i CARD C0FFORATE ORGANIZATION WITH OPEftA110tW. DA RESPONSIBILITIES Also AvaihMe On Apcrture Card POTEs NUMEPS IN TITLE BOXES It0!CATE SECTIONS DESCRIB3MO RESPONSID1LITIES 8908090060-Q\ Page 21 of 93 m ., j _ I AO M94PT 719 e-_
g __. 7-
- V
;; Op;rctiontiEQh511ty.Accurbnca' Plani rov-13 # 4.0 onarat ional ouality Assurance Proggg3
- 4.1'- General Requirements
- e
, -1.;.The' operational Quality Assurance Program shall'be:
- a. Documented by written Directives, Instructions,.
or Procedures. 'l~41
. 'b. Carried out throughout plant operating' life in accerdance with those Directives,-Instructions,-
or Procedures. l 41
- 2. :Thel Program:shall; include identification of:
4- 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. . ThelProgram 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
}O( under 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 1for the given ,
activity have been satisfied. _
- 5. The Program shall take into account the need for special. controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of. quality by inspection and
. test. -
l 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. O q Page 22 of 93
Operational Quality Accurenc3 Plcn ; R;v 13 I 4.2 General Description
]
1
- 1. The Operational-Quality Assurance Program has been i established to govern the operational activities and I 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 '3 program which governs all safety related, fire protection related and 10CFR71 related activities as they pertain to operating nuclear plants.
- 2. The Program has been initiated by the Chairman & 12 {
Chief Executive Officer of the Company issuing a I single directive to the Senior Vice President Power I J 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.
.s !
- 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 documents (i.e.,
Directives, Instructions, and Procedures). S. For ease of administration, implementing documents shall be issued at the following program levels:
- a. Cornorate: Approving Authority, Power Supply Vice Presidents and Director. l 42
- b. Plants (Prairie Island and Monticello):
Approving Authority, Plant Manager.
- c. Departments Providina Nuclear Plant Suecort:
. Approving Authority, Department Manager.
- 6. It should be noted that the Plant Level Directives 43 are controlled by the Corporate level Directives 44 (i.e., the Corporate level establishes the minimum requirements associated with the Plant level).
Page 23 of 93
Op; ration 01 QuOlity AccurOnc3 Pltn L R;v .13 4.3= Operational Ouality Assurance Plan lI
'J-f 1. The Operational Quality Assurance Plan shall be a document which~ describes in general terms how compliance with the quality requirements presented in 10CTR50, Appendix B and 10CFR71, Subpart H is l4 accomplished with respect to company n,uclear plants regulated by an NRC Operating License.
- 2. The Operational Quality Assurance Plan shall be issued under the authority of the Senior Vice President l Power Supply and shall be reviewed periodically.
l
- 3. The Operational Quality Assurance Plan shall be controlled to 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
~ ,/~'
L(j) 1. Administrative Control Directives (Directives) 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 Assurance 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 44 applicable requirements of 10CFR50, Appendix B and 4 10CFR71, Subpart H. The Directives issued at the plant levels are not expected to satisfy all 10CFR50, Appendix B and 10CFR71, Subpart H l4 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 i) be second tier implementing documents and shall receive a quality review prior to issuance. The quality review shall assure compliance with pertinent Directive requirements and assigned responsibilities.
Page 24 of 93 . l
Oper;ticnni Qu21ity A02ur:nca Plcn Rev 13
- 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 Procedures
- 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 member 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 Procram Administration
- 1. Administration of the Corporate level of the Operational h
. Quality Assurance Program shall be performed by the Director Power Supply Quality Assurance.
- 2. Administration of the Plant level of the Program shall be performed by the Superintendent Quality Engineering.
- 3. Disputes between Quality Assurancs personnel and other organizations relative to Program requirements shall be referred to the Approving Authority (as identified in Section 4.2 of this Plan) responsible for l 45 )
establishing the pertinent requirement. l s
- 4. Program administration shall include the following activities: j
- a. Quality review of Directives.
- b. Quality review of Instructions.
- c. Procurement review.
- d. Performance of required audits.
I
- e. Reporting to management concerning:
- 1. Program status. )
- 2. Program discrepancies including quality trends.
1 Page 25 of 93
~
; o' '
[ [ Operation 51 Quality Ac:Curanca Plcn
,N .g,
- Rav'13 1
4 '. 8 ' Program Boundary I ji js , il.: The structure, systems, components,cand other items
?v~ -
- requiring quality assurance are listed in:
' Appendices ALand B. The Program shall also. include ,
shipment of. radioactive? materials as required by )
'10CFR71 Land systems and activities associated-with j fire protection as'. identified in Appandix c. !
n ,
'l
-2. An index shall.be' established and' maintained.by'the
- Director Power Supply Quality: Assurance which
. identifies'the. Directives andiInstructions that are utilized to..implementithe' requirements of ANSI .
N18.7-1976 that are committed to in Section 1.O.of l 46. :i this' plan and the requirements identified in the remaining sections of this p1an'.- 1 4.3 Ouality Assurance Trainina Training programs shall be established for,those spersonnel performing quality-affacting activities such i that they are knowledgeable-in the quality assurance documents and their' requirements and proficient in; implementing these requirements. .These traindng
; programs shall assure that:-
' 1 '.- ; Personnel responsible for performing quality-affecting
/. . .T.T* activities ~are. instructed as;to the purpose, scope, N -and implementation of the quality-rclated Directives,
-Instructions, and. Procedures.
- 2. Personnel performing quality-affecting activities av trained and qualified, as appropriate, in principles and. techniques of the activity beis.g performed.
- 3. The scope, the objective, and the method'of implementing:the training. programs are documented,
- u. Proficiency of personnel performing quality-affecting'
.e activities is maintained by retraining, re-examination, and/or racertification as appropriate.
l 5. Methods are provided fe' documenting training sessions describing content, ittendance, date of attendance, l and'the results:of the training session, as I appropriate.
- 6. . Fire protection training is accomplished in accordance
-with-Appendix C.
6 Page 26 of 93
, , __ _____ c _ _ _ _ _ _ _ _ _ , }: Rev 13 i r-l' 5.0. Modification control 5.1 General Requirements 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 ProcqSE A' uniform process for controlling modifications to nuclear plants shall be provided in the operational Quality Assurance Program. Measures shell 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.
Page 27 of 93
op;rationni Quality Aaruranco Plcn RV ~ 13 6.0 Procurement Document Control ["') a 6.1 General Requirements Measures shall be established to assure that applicable reg'ulatory 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 B or 10CFR71, }4 Subpart H. 6.2 Technical and Ouality Requirements
- 1. The operational Quality Assurance Program shall contain provisions for control 3ing procurement of material, aquipment, components, and services that are safety related, fire protection related, or l 47 10CFR71 related and utilized at or for an operating nuclear plant.
2, Procurement documents shall contain specffic technical and quality requirements. Rens ial, spare, rS and replacement parts shall be required to meet the x)~ original specification (or properly reviewed and . approved revision) or construction code, quality { assurance documentation requirements, and vendor quality 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 t 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 the package. The procurement documents sball also require copies of all documents referred to in the licenne,
-f7.-) certificates, or other NRC approval au applicable. G Page 28 of 93
f' Operational Quality Assuranco Plan Rov 13 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 Eire Protection Procurement Control The additional procurement controls identified in Appendix C shall be applied to purchasing fire protection systems and equipment. O Page 29 of 93
g-Op3ratien31 Quality AOcurCnca Plcn RV '13 7.0 Instructions. Procedures and Drawinos
/ '
) 7.1 General Requirements v
- 1. Directives, Instructions, Procedurea, and drawings of l 41 a type appropriate to the circumstances shall be prorided for the control and performance of activities which affect quality.
- 2. Diroctives, Instructions, Procedures, and drawings l 41 sball include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.
7.2 Dirgtetives 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 requirement 3.
1
- 3. Procedure content requirements.
7 3) 7.3 Procedures wJ
- 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, nystems, or components.
- 2. The following procedures shall be provided:
- a. Operating procedures.
- b. Emergency procedures.
- c. Surveillance test procedures.
- d. Routine or preventive maintenance procedures.
- e. Calibration procedures. 21
- f. Plant chemistry and count room procedures.
- g. Radiation protection procedures.
- h. Emergency plan procedures.
O 3 (_,/ 1. Special process procedures. Page 30 of 93 ,
L' oparntionni Qunlity Accuranca Plcn *
.RSv 13
- j. Preoperational and operational test procedures. 26
- k. Audit procedures.
- 1. Fire fighting procedures.
- m. Document control procedures. 21
- n. Radioactive material shipment procedures.
1
- 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 related, fire protection related, or 10CFR71 related l 47 structures, systems, or components. i i
) l l
i a I l l 0 Page 31 of 93 I - __- _-__- __ I
50k # Lop;rctionn1' Quality.Accuranca Plan: R;v113 4 8.0 , Document Control-L 8.11 General Requirements
')Yl l%-
- 1. Measures'shall be-established tx) control the issuance p
c of. documents, such as Directives,. Instructions, ) o Procedures, and. drawings,. including changes-thereto, l.41
. which prescribe activities affecting~ quality.
12 . These. measures shall assure that documents, including changes, are: p ' a '. Reviewed for' adequacy and approved for release by. y authorized personnel, add l 48
- b. Are distributed to and used'at the location where the prescribed activity is performed.
i- 3.. Changes to documents shall be reviewed and approved by'the same organization that performed the original Jreview and approval or.another designated responsible p, organization. c 8. 2- Directive Control l '. Directives issued to implement the Operational Quality Assurance Program shall be controlled to
' -- assure that current copies and appropriate indexes
'. o 'are:made available to personne1' performing the E prescribed activities. Directives shall be reviewed hv quality assurance personnel tx) assure their compatibility with.the Operational Quality Assurance Program objectives and-shall be approved by the designated ranagement.
- 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 controlled to assure that current copies and appropriate indexes are made available to personnel parforming the prescribed activities.
Instructions shall be reviewed by quality assurance p personnial- to assu re 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.
O Page 32 of 93 nu
" 'Opcrzti:n21 Quality-Accurenco Plcn R;v 13 -
I 8.4 Procedure Control l
\
- 1. . Required procedures shall be controlled to assure that-current copies are made available to personnel o {
J performing the prescribed activities. Required pro- ) cedures 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 j indexes of standing procedures shall be formulated and made available to personnel responsible for 3 performing the prescribed activities. j
- 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.
i' 8.5 Drawina Control
- 1. Drawings which rapresent 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. l
- 2. Measures.shall be established for revising plant drawings and for distributing revised drawings.
Proposed revisions to drawings shall be reviewed by a knowledgeable 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 Radioactive Shioment Packace Documents All documents related to a specific shipping package for radioactive material shall be controlled by appropriate l instructions; all significant changes to such documents shall be similarly controlled. 1 8.8 Uedated Safety Analysis Reports 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. L Page 33 of 93
Oper;tirn21 Quality Accurenca Plcn l i RV 13 9.0 Control of Purchased Material. Eculement and Services ;
, )' 9.1 General Requirements
/
- 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 requirements shall be available at the plant site prior to installation or use of such material and equipment. This documentary evidence shall be retained at the plant site and <
shall be sufficient to indicate that the purchased material and equipment meet the specific requirements of the codes, standards, or specifications.
- 3. The effectiveness of the control of quality by vendors shall be assessed at intervals consistent with the importance, counlexity and quantity of the product or service.
/ 'T 9.2 Ouality Review Q ,1
- 1. Documents initiating procurement of safety related, fire protection related, and 10CFR71 related material, equipment and services shall be subject to a quality review to ensure applicable regulatory requirements, design bases, quality assurance, and other requirements are adequately satisfied.
- 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 Verification
- 1. The adequacy of vendor's quality assurance program 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
'r3 be planned and performed in accordance with written
'~'
)
. procedures based upon procurement document requirements.
Page 34 of 93
l optritionni Quality Accuranca Plan
. R;v.13
- 3. Material and equipment ue.y be procured and used based on appropriate certificates of conformance, provided the validity of such certificates are periodically -
evaluated by audits, independent inspection or tests and that such certificates comply with applicable code provisions. 9.4 R_ecelot Inseection
- 1. Material and equipment shall be inspected'upon receipt at the plant site prior.to use or storage to determine that procurement requirements are satisfied. This inspection shall include verification that required documentation is complete.
- 2. Nonconforming material and equipment 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.
O 1 l O i l Page 35 of 93 l L_____
Oparatienni Quality Accuranca Plcn R;v 13 10.0 Identification and Control of Materials Parts and Comoonents m
) 10.1 General Requirements w/
- 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 condition prior to use. 10.3 Material Control Material held in storage for use on safety related, fire protection related, and 10CFR71 related systems, i 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 traceable to the item. The method utilized in identification shall not significantly affect the fit, function, or que.lity of the item being identified. 10.4 Receiot 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 radioactive materials shall be inspected upon receipt to assure that associated procurement document provisions have been satisfied. Measures shall be established for l identifying nonconforming material, parts and components. lO t ) 10.5 Nuclear,ltel Control mJ Measures shall be established to protect special nuclear material against theft or diversion in accordance with applicable NRC regulations. Page 36 of 93 E__.
Op2rction21 Qu211ty Accurtnca Plan l Rav 13 , 11.0 Control of Soecial 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 Weldina 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 appl-icable codes and standards.
- 2. Measures shall be established for controlling welder and brazer qualification and requalification tbat assure qualified personnel are utilized to perform safety related welding and braz.ing.
11.4 Heat Treatina Procedures
- 1. Heat treating shall be performed in accordance with procedures formulated and approved in accordal'ce with applicable codes and standards.
- 2. Measures shall be established for controlling heat treating procedures that assure such procedures are qualified, reviewed, and approved, as required, prior to use.
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 I standards and shall be reviewed to assure their technical adequacy and approved by mar -ement. Page 37 of 93
cpirational-Quality'Accuranca-PlEn
.R;v 13-
- 2. -: Measures'shall'be established for controlling NDE
<- procedures that assure such procedures are reviewed 7.( g . .
and approved, as required, prior.to use. 11'.6 NDE Personnel Qualification g-
- 1. Measures shall be established that assure safety.
related non-destructive examinations (NDE) are performed by personnel 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.
n
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N. 1 I Ii , s Page 38 og 93
- '~'operationni Quality Accurcnca Plcn RCv '
12.0 Insoection 12.1- general Requirements
- 1. Measures shall be established for inspection of activities affecting quality to verify conformance 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 chall 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 monitoring 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 si.all be indicated in appropriate documents.
12.7 Plant Ooeration
- 1. Measures shall be established that assure periodic inspection of safety related and fire protection related systems, components, and structures. Such l 47 inspection of plant systems and equipment shall be performed to assure that such systems and equipment
. are in the required status and configuration.
Routine general inspections of the accessible plant facilities to verify 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 loading or refueling to assure specified fuel and reactor internal configuration.
12.3 Inservice InsoectiGD L Measures shall be established that assure inservice l inspection examinations are performed in accordance l with applicable provisions of the ASME Boiler and Pressure Vessel Code, Section XI as required by , 10CFR50.55 (see Section 13.2 relative to Inservice ! Inspection functional testing). Page 39 of 93 I ' t
Optration21 Quality Ancurcnc3 Plan i R;v 13 12.4 Insoection of Maintenance and Modifications
-f~D Meavires shall be established which assure that 'k l activities associated with plant maintenance and j modifications are inspected when determined 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. Plant equinment is returned to service as required,
('N 10. Activities are appropriately documented, and l 48
-() 11. Redundant equipment is available.
12.5 Modifications and Non-Routine Maintenance l 49
- 1. Measures shall be established which assure that non-routine maintenance and modification receive l 49 prior review by a qualified individual to identify applicable inspections. Such reviews shall include considering: (1) required mechanical inspections, 50 electrical inspections, instrumentation and control 26 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 be 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 l ( surveillance testing, instrument calibration,
\
laboratory services, etc.) are inspected by qualified personnel when determined appropriate by quality or other qualified personnel, j Page 40 of 93
' ~ 0parational Quality Accuranca Plan Rav 13 12.7 Empeiet Inseection Measures shall be established which assure that received items are inspected by qualified personnel (see Section 9.4). l 51 12.8 Vendor Inseection Measures shall be established which assure that inspections and process monitoring specified in appropriate procurement documents for materials, components, and equipment are performed by. qualified personnel. -
12.9 Fire Protection Inspections Measures shall be established which assure that fire l protection inspections required by applicable Technical Specifications are performed by qualified personnel. j 12.10 Radioactive Material Packaces Measures shall be established which assure that packages utilized to ship licensed radioactive material off-site are inspected in accordance with the l 52 applicable provisions of 10CFR71. 12.11 Emeraency Eculement Measures shall be established which assure that emergency equipment required to implement emergency plans is inspected when determined appropriate by qualified personnel. 4 12.12 Handlina Eculement Measures shall be established which assure that plant handling equipment (such as cranes, lift trucks, fuel handling tools) is inspected by qualified personnel and when determined appropriate by quality or other qualified personnel. 12.13 Inseection Procedures
- 1. Required inspections shall be performed in accordance with appropriate instructions, procedures, and checklists. Such instructions, procedures, and checklists shall contain a description of objectives; acceptance criteria and prerequisites for performing 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. l l Page 41 of 93 l t
, ,g u-7
- f. ! ij k' e .
?OpdrationaliQuality Accuranca Plcn s R0vn134 n '
s ,
- 2. 'Where' activities are to be, inspected, the activity
, HP-C %' 'procedurefshalltidentify hold'pointsiin the ,1)"l; activity sequence to permit inspection. The
- activity procedure shall require appropriate i g' approval:for the. work to continue beyond the -
- designated hold point'and identification of.those-LP performing:the-inspection. The inspection procedure or-checklist shall require recording lthe date,
. identification.of those performing the inspection,
, and as-found-condition.
12.14 Personnel Qualification m-1.' Personnel performing required: inspections shall be
, qualified in accordance with applicable codes, i
standards and training programs. Required _ inspections!shall not.be_ performed by' individuals who performed the: inspected activity 1nr directly ~ supervised the inspected activity. 2.. Personnel. performing ~ inspections required by' . LSections 12.2, 12.4, 12.6, 12.10' 12.11, and'12.12.
, l 45; shall:be qualified based.upon experience and 4 training applicable to area of inspected activity p and upon training in inspection methods.
- 3.= Personnel performing' inspections required by Sections. 12.3, 12.5, 12.7, and 12.8 shall be l 45 qualified in accordance with ANSI N45.2.6-1978 as-modified by' Regulatory Guide l'.58, Revision 1.
- 4.: Personnel' performing inspectionsfrequired by Section 12.9 shall.be qualified in accordance with 45 Section 14.0 of Appendix-C.
l I i () ! Page 42 of 93 L ___ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . -
'oparctional Quality Aruranca Plcn R;v 13 13.0 Test Control 13.1 General Requirements
- 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 parformed 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 1 operational tests during nuclear power plant operation, of structures, systems and components ! shall be included as appropriate.
- 2. Test procedures shall include provisions for assuring that all prerequisites for the given test i have been met, that adequate test instrumentation i is available and used, and that the test is !
performed under suitable environmental conditions. Test results shall be documented and evaluated to assure that test requirements have been satisfied. ! i 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 related l l
. structures, systems, and components will perform satisfactorily in service. Surveillance tests ;
shall be identified and performed in accordance l with written test procedures which incorporate { the requirements and acceptance limits contained 1 in applicable documents. The surveillance test i program shall include, as a minimum, those surveillance tests specified in applicable Technical Specifications and functional Inservice l Inspection testing of pumps and valves. Surveillance requirements for fire detection and ; protection systems in other areas of the plants l shall be developed using appropriate NFPA for ) guidance. '
- 2. Surveillance test results shall be documented and evaluated to assure that test requirements have been satisfied or deficient items satisfactorily ,
resolved. Functional Inservice Inspection tests l shall be performed by personnel qualified in accordance with applicable requirements. O Page 43 of 93
[V , ("' .+
' oporationni Qunlity Accurancs Plan j
RLv 13. 13.3 Preocerational and operational Tests I
. r~~s hs,[. Measures shall be established to assure that I appropriate preoperational.and operational tests are f 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 satisfactorily resolved.
13.4 Proof Tests L Measures shall be established to assure that appropriate proof tests are specified in procurer.ent documents for safety related and fire protection related replacement material and equipment and that. such' tests are performed and documented prior to installation. 13.5 Soecial Tests' Measures shall be established that assure safety related tests are reviewed and approved as required by r~g.
- 10CFR50.59 and applicable Technical Specifications.
J ,7 Such tests shall be performed in accordance with appropriate' procedures. Test results shall be documented and evaluated to assure test requirements have been satisfied. f
;(3 Page 44 of 93
f 1 L
(
OpOrction21 Quality A2 urcnc3 Plan p Rav~13 ' -{ 1 14.0 Control of Measurina and Test Eculement- i
~
14.l Ggperal' Requirements i Measures:shall be established to assure that tools, gauges, instruments and other measuring and testing ] i 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 Measurina'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 be traceable to the National Bureau of Standards or where national standards are not available, the basis of calibration shall be documented. Page 45 of 93
v operational Quality Assurance Plan Rov 13 15.0 Handlina. Storace and Shionina 15.1 General Requirements l
- 1. Measures shall be established to control the handling, storage, shipping, cleaning and
, preservation cf 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 Storace Facilities Storage facilities shall be provided at each operating nuclear plant for storage of safety related and fire protecti'on related operating and maintenance supplies, spare parts, replacement parts, replacement equipment, materials and tools. These storage facilities shall 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 snfety related items. 15.3 Nuclear Fuel Storace Areas shall be provided for storage of nuclear fuel which assure physical protection, suberitical arrangement, adequate. cooling, adequate radiation shielding and containment of radioactive material as appropriate for the condition of the stored fuel. 15.4 Radioactive Material Storace
- 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 stored material, and containment of radioactive material as appropriate 4
- 2. Handling, storage, and shipment of radioactive material shcIl be controlled based upon the following criteria:
- a. Established safety restrictions concerning the handling, storage, and shipping of packages for radioactive material shall be followed.
O Page 46 of 93
' ' operational Quality Assrcance Plan Rev 13
- b. h Shipments s'all not be made unless all tests, certifications, acceptances, and final inspections have been completed.
- c. Work instructions shall be provided for 5 handling, storage, and shipping operations.
15.5 Storace Control Stored material, parts and equipment shall be controlled in a manner that assures safe plant ,t 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 Handlina Safety related material, supplies, equipment and parts shall be handled in accordance with procurement I.
- 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 Shineina and Packacino
- 1. Shipping and packaging requirements shall be prepared for material, equipment, and components that are to be shipped off-site and returned for l5 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 Page 47 of 93 A--. _____.--___.r___-__m. - - _ _
Operational Quality Ascurance Plan Rav 13 16.0 Inspection. Test and Ooeratina Status l f 16.1 General Requirements
- 1. Measure.s 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: 1 1 O 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 determino 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, 9
- i. key control and area posting, as appropriate, for the involved activity.
Page 48 of 93
'1
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l 'I 'Op3rctional Quality A;;urrnca, Plan ; l Rav 13- l 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. j l
- 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.
i
- c. Test results are reviewed and' approved as ]
appropriate. l
- 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 Taccina 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 EeE 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 Bvoass, 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.
f'
- c. Removal of bypasses from safety related and fire protection related items are reviewed prior to removal.
Page 49 of 93
Optrction21 Quality Accurenca Plon-ROV 13 d.. Application'of bypasses to safety related'and fire protection related items is authorized by [._') responsible personnel. as/ ,
- 2. The application of safety related and fire l 47 l protection related bypasses shall be considered a temporary measure and shall be reviewed ;
periodically. t
]
- 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.
1(~} 3. Status of non-conforming parts or packages shall be () 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 l 53 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. l h.O . Page 50 of 93
op rrtionni Qunlity Acruranca'Plcn R;v 13 17.0 Nonconforming Materials. Parts or Components 17.1 General Requirements
- 1. Measures shall be established to control materials, I parts, or components which do not conform to i 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 affected organizations.
1
- 2. Nonconformance items shall be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures. )
17.2 -Receint Inspection
- 1. Measures shall be established which assure that safety related and fire protection related=
material, supplies, equipment, and components are inspected to determine that they conform to i specified requirements of pertinent procurement documents upon receipt at the plant site. The , absence of required documentation or discrepant j 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. i 17.3 Maintenance Inscection Equipment, components or parts found nonconforming in a manner that could significantly affect its ability to i fulfill its safety related and fire protection related l 47 function shall be identified as a nonconforming item J 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 1 installation. Provisions shall be made for reviewing the nonconformance and correcting discrepancies by repair or rework if appropriate. O I Page 51 of 93
operational Quality Accuranca Plan Rov 13
- 2. The acceptability of such rework or repair of
- materials, parts, components, systems, and
- structures shall be verified by reinspection and ratesting 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 i documented.
- 3. Normally, nonconforming safety related and fire l4 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 Nonconformance Documentation
- 1. Nonconformance reports shall be initiated for significant deviations from specified requirements.
Such reports shall identify the nonconforming item, describe the nonconformance, the dispositi,on 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 show quality trends and the results of this review shall be reported to the appropriate level of management for review and assessment.
17.6 Reoortina . Measures shall be established which assure that defects as defined in 10CFR21 and failures to comply with the Atomic Energy Act of 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 requirements of 10CFR21. O Page 52 of 93
op rctional Quality Accu?tnco Plan
-R;v 13 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 recurrence. l 54
- 2. The identification of the condition adverse to quality, the cause of the condition, an'd the corrective action taken shall be documented and reported to appropriate levels of management..
18.2 Ooeratina 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 Discrepancies 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 recurrence of the l 54 discrepancies identified and verification of implementation. l Page 33 of 93 l
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% 66 Y y[opnt$1onni" Quality Abcuranca 'Pldn kW R0v> 13. W% m, 19.0; Ouality Assurance' Records i 19.lf General Requirements
- 1. : Sufficient records ~shall,be' maintained to furnish evidence of activities-affacting quality. These records shall include'at'least the following:
~
- a. Operating logs and the.rcsults of reviews, it inspectionah tests,. audits, monitoring of work performance and' material analysis.
- b. The records.shall also' include closely related-data such as qualifi:.:ations of personnel, procedures, and. equipment.
4
- c. Inspection and test records shall, as a uninimum, 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.
O-19s2 Oneratina Records
-Measures shall-be established which assure that records as they apply to plant operation.are generated and retained as required by the Technical Specifications or other: regulatory requirements.
19.3 ' Plant Modification Records Measures shall'he established which assure that adequate records are generated and retained to reconstruct plant modi.fications that are safety related orifire protection.rez.ated. 19.4 Elant Maintenance Records Measures shall be established which assure that records e' pertaining to maintenance of plant safety related and fire protection related structures, equipment and components are generated and retained.
-19.5 Personnel Qualification Recordis Measures shall be established which assure personnel qualification records are generated and retained.
, Page 54 of 93
.. ~ _ __ _ _ - - _ _ _ _
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Op3rctional QuSlity Accurcnca Plcn R;v ' 13 i 19.6 Procurement Records J Measures shall be established which assure that safety 4 related, fire protection related, or 10CFR71 related I procurement documents and associated documents are generated c.nd retained. 19.7 Surveillance Test Records Measures shall be established which assure that records associated with Surveillance Testing, including l4 Inservice Inspections, are generated and retained. 19.8 Audit Reoorts Measures shall be established which a.=sure 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 Drawinas Measures chall 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 Manacement
- 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 _ vast be stored in insulated filing devices (rated 350-1 hour by UL as to fire resistarse u iv) located in areas having combustible loading c2 less than 5 l 55 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 record.s 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 i program implementation are collected, stored, and maintained in accordance with good records management practices. Such systems shall assure chat these records are made available to auditars and inspectors in a timely manner.
hj j i 1 1 Page 55 of 93 I
Oparctionni Qunlity A ;urtnco Plcn R;v 13. 20.0 Audits
<~T j
) 20.1 General Requirements l
A comprehensive system of planned and periodic audits shall be carried out to verify compliance with all l aspects 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. Follow-up action, including re-audit of discrepant areas, shall be taken where indicated. 20.2 Recuired Audits l 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 Assurance Plan. This overall audit shall be performed by an organization other than that responsible for fs administration or implementation of the program. V 20.3 Audit Schedules
- 1. Required audits shall be performed each year except that' this tima period may be extended to not more than two years provided such extensions are justified based upon past experience. Special audits ray be scheduled on the initiative of quality a ssurance personnel based upon suspected or known discrepancies or as directed by management.
- 2. Appropriate audit schedulos shall be prepared each year.
2(,4 Audit Procedgra 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. ("h
\j Page 56 of 93
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L Op ratirnal Qunlity Ancurenca Plnn R;v 13 20.5 Audit Reports
- 1. Reports of the results of each audit shall be prepared. These reports shall include a f description of the area audited, identification of individuals responsible for implementation of the l audited provisions and for performance of the audit, identification of discrepant-areas, and recommended corrective action as appropriate.
- 2. Audit reports shall be distributed to the appropriate management level and to those individuals responsible for implementation of audiced provisions.
1
- 3. Audit reports and associated nonconformance reports j shall be periodically analyzed for quality trends I and the results reported to the appropriate level of management for review and assessment. l 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.
O Page 57 of 93 l l
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Op2 ration'1' n Quality Assuranca Plan
;Rev 13- Appendix A APPENDIX A l
~ ^# Monticello Structures. Systems, and Components Subiect to Accendix B of 10CFR50
.l. STRUCTURES-Reactor. Building L Plant Control ~and Cable Spreading Structure Off Gas Stack Intake. Structure (Service Water pump area)
' Diesel Generator Building Diesel Fuel Oil 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)
Emergency Filtration Train (EFT) Building l 56
- 2. MECHANICAL SYSTEMS AND CONDONENTS COMPONENTS Feactor 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 Discharge Piping
- Feedwater Piping (to and including outermost containment isolation valve)
Control Rod Drive Housing Supports Reactor Vessel Interne,lg Fuel Assemblies Core Support Structure Jet Pumps Control Rods Liquid Poison Pipe Core Spray Sparger Control Rod Drive System Control Rod Drives Control Rod Drives Accumulators Scram Discharge Volume Scram Piping
!O Page 58 of 93
f i (- OpSrationsi Qunlity A3 urenca Plcn Rev 13 Appendix A COMPONENTS Standbv Licuid 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 including outermest 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 sice) Core Sorav System Core Spray Piping, Pumps, and Valves . 1 Hich Pressure Coolant Iniection 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 Beactor Core Isolation Coolina Svstem RCIC Steam Piping and Valves Inside Containment (to and including outermost isolation valve) RCIC Steam Supply and Exhaust Piping and Valves (outside containment) RCIC Pump-Turbine RCIC Injection Piping and Valves RCIC Suction Piping O' Page 59 of 93
-Opiraticnal Qunlity Accurenca Plsn Rav 13 App 3ndix A
' Service Water System
[ .
;. ,]
y() Emergency Service Water Pumps Emergency Service Water Piping and Valves RHR Service Water Pumps RHR Heat Exchanger (tube side) RHR Service Water Piping and Valves . l 26 EFT Emergency Service Water Pumps, Piping, and Valves ] i Reactor Water Cleanuo System RWCU Piping and Valves
'(to and including outermost isolation valve)
Soent Fuel Storace Systems Spent Fuel Pool Diesel Generator Succort 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 Heatina and Ventilating System
,s Emergency Filtration Train (EFT) for Control Room and
( EFT Building Combustible Gas Control System V)- 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 16 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 CRD Pump Feed Breaker 152-606
,6) x Page 60 of 93
Op;rctional Qunlity Accurenca Plan Rev 13 Appendix'A 4160 Volt Bus 15 Breaker 152-502, 501, 509, 511, Feed Breaker 308 I. RHR Service Water Pump A Motor RHR Service Water Pump C Motor RHR Pump A Motor j RHR Pump C Motor Core Spray Pump A Motor 480 Volt Load Center 103 No. 11 CRD Pump Feed Breaker 152-506 fjfL Volt Switchgear Load Center 104 480 V HCC 142 (1) (Essential) 480 V MCC 143A (1) 480 V MCC 143B (1) 480 V MCC 144 (1) 480 Volt Switchaear Load Center 103 480 V MCC 133A (1) ' 480 V MCC 133B (1) 480 V MCC 134 (1)
. Diesel Generator No. 11 (1)
Diesel Genere. tor No. 12 (1) 250 V DC (Division 11 Distribution Panel D31 (1) 250 V DC (Division 2) Distribution Panel D100 (1) 125 V DC Distribution Panel D-11 (1) 125 V DC Distribution Panel D-21 (1) 120/240 Veit 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 l 57 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.
I 3. Certain components are excluded from the quality assurance l 23 I l program requirements if they meet the criteria described in Section 5, q l Page 61 of 93 l
l 1 Op;rction:1 Quality Accurancs Plcn j Rev;13 Appendix A .y INSTRUMENTATION SYSTEMS AND COMPONENTS I((~5{ : 4'.
'~'
Reactor Protection System
-Primary Containment Isolation System
{ High Pressure Coolant Injection System Initiation and .
' Isolation Reactor Core Isolation Cooling System Initiation and Isolation Core Spray System Initiation Low Pressure Coolant Injection System Initiation
-Automatic Blowdown System Neutron Monitoring System (IRM and APRM) I 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 'bove, .
. quality, assurance instrumentation program requirements are l 57 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 quality assurance l 23 program. requirements if-they meet the criteria described in Section 5.
- 5. ELECTRICAL AND INSTRUMENTATION SYSTEM CO}RQE,ENT EXCLUSION CRITERIA
- 1. Any component of an electrical sysc.em in Section 3 or 58 instrumentation system in Section 4 is excluded from the 58 quality assurance program requirements if it' meets the 59 following criteria:
- a. A failure of the component by electrical shorting, open circuiting, grounding or mechanical 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
.gs nonsafety-related and then qualified for use in equipment
't requiring quality assurance. Examples of such items are l 57
%' resistors, capacitors, switches, indicators, coils, wire, connectors, solid state devices and miscellaneous hardware.
Page 62 of 93 , i i
Op;rctionni Quality AScuranca Plan
'Rev 13' Appendix B APPENDIX B
. f~
\~ .Erairie Island Structures. Systems, and comoonents Subiect to Anoendix 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 and shell side) l 26 Pressurizer, including instrumentation, piping, and components Reactor coolant hot and cold leg piping, fictings Surge pipe, fittings Loop bypass line Temperature detector bypass manifold Reactor coolant thermowell Reactor coolant thermowell. boss Safety valves Relief valves Reactor. coolant' system boundary valves
~
s Control rod drive mechanism head adapter plugs Reactor coolant pump (J'- Pump casing Main. flanges 1 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
/~N Page 63 of 93
op3rotional Qu211ty Arsurcnc3 Plan e R;v 13 Appandix B
- 2. REACTIVITY CONTROL SYSTEMS Drive mechanisms incNding: ;
Control rod clus drive shaft assembl'y, including latch. assembly Re-4 tor trip breakers
- Co. trol rods and rod cluster assemblies Control rod guide tube Control 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 volume control tank Positive displacement charging pump and motor Seal water filter Letdown orifices and letdown valves Excess letdown heat exchanger Seal water heat exchanger Boric acid tanks Boric acid transfer pump
~ ,
Boric acid filter Reactor coolant pump seal and bypass orifice . Piping, inboard of isolation valves Electric medules with safety function Cable with safety function Heat tracing
- 4. INCORE INSTRUMENTATION Thimble guide tubes Seal table
- 5. BOPON 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 l Page 64 of 93 l
f-K m a - _ ~ - - - - . . _ _ . _ _ _ .
,- 4:
C ,Oparation21; Quality A25uranca Plan
~Rev 13 Appendix _B p . gg -7. CONTAINMENT SPRAY SYSTEM
'i N' f~ 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. 3ESIDUAL HEAT REMOVAL SYSTEM Pumps and motors ,
Heat exchanger Piping and valves with satety function 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 FAN COOLER SYSTEM Ductwork
- gg' Fans
't ) 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 containment boundary l Systems handling gaseous radioactive materialo- !
Electric modules with' safety function , Cable with safety function The Waste Gas Disposal System shall be maintained in accordance with the guidance established in Regulatory Guide 1.143, Revision 1, October, 1979. l4
- 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 isolation valve from other safety related systems A.
s Page 65 of 93 j L___---_--_--
op3 rational Quality Assuranca Plan
- R;v 23 -App 2ndix B
- 13. ETEAM GENERATOR BLOWDOWN
' Piping from steam generator to containment isolation valves ,
- 14. REACTOR PROTECTION SYSTEM Electrical modules Cable
- 15. PROCESS RADIATION MONITORS Radiation Monitors, including electric modules and cable with a safety function, associated with the Shield .
Building, Auxiliary Building, Spent '.uel Pool and Control Room Ventilation Systems
- 16. CONTAINMENT HYDROGEN CONTROL SYSTFH Piping and valves with safety function Electric modules and cable
- 17. REACTOR VESSEL SERVICE EOUIPMENT Containment polar crane vessel head handling equipment Crane structural supports Crane electrical, cable, controls and instrumentation with safety function
- 18. REFUELING.EOUlPMENT Spent fuel cask Auxiliary Building crane Auxiliary Building crane structural supports Crane electrical, cable controls and instrumentation with safety function Fuel transfer tube Spent Fuel Bridge Crane Manipu.lator Crane
- 19. FUEL STORAGE New fuel racks Spent fuel racks Spent fuel pool structure and enclosure
- 20. CONTROL ROOM PANELS Electric modules, with safety function Cable with safety function O
Page 66 of 93
L_ l"' Op3rstional Qtte.lity Accurcnca Pltn Rev 13 Appendix B l -S 21. LOCAL PnNELS AND RACKS t
) Electric modules with eafety function Cable with safety function
- 22. EMN STEAM SYSTEtf Main steam piping and valves from steam generators up to and including piping restraints downstream of the main steam isolation valves 1 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 discharge Electric modules with safety function Cable with safety function
- 23. FEEDWATER_S,YSIDi 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 -
g) 24. AUXILIARY FEEDWATER SYSTEM ,
.j 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 j cooling water system Electric moduler, with safety function - Cable with safety function
- 25. COOLING WATER SYSTRME 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 Valves, other, with safety function Electric modules with safety function L l Cable with safety function l Cooling water systems (essential) Fiping (except to turbine building and non-essential equipment) (~s Diesel engine pumps I l ()' Etrainers, with safety function Valves, isolation . Page 67 of 93 l r C_
L 'Op rctional Quality Assurancei Plan '
.Rsv 13. App:ndix B Valves, other, with safety function l Screen wash systems, with safety function l Traveling screens, with' safety function i t
Electric modules with safety function Cable with safety function Diesel engine pump auxiliaries as feows: Diesel oil storage tanks , Day tanks Fuel oil transfer pumps and motors Fuel oil piping and valves with a safety function Starting air compressors j Air receivers ! Starting air piping and valves with a safety function i Cooling water piping and valves with a safety function Electric modules with safety function Cable with safety function Diesel engine, lubricat2.ng oil and jacket cooling systems Diesel fuel oil
- 26. INSTRUMENT AIR SYSTEM <
l Piping and valves associated with containment penetrr. tion j i
- 27. DIESEL GENERATOR Diesel oil storage tanks 3 1
Day tanks i Pumps and motors, fuel oil transfer ' Diesel filter valves, with safety function Piping except vent and fill piping downstream of lest 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 GRHjl3ATOR AIR STARTING SYSTEM Compressor Air receivers l '
Piping and valves from receiver to Ciesel generator Piping between compressor and receiver l J Page 68 of 93 e1 l l _ _ _ _ _ _ _ _ _ _ _ _ l
c opbratienti Qunlity Accuranca Plcn H Rev 13 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 point of interface with Class lE breakers 60 which. isolate main Class lE onsite buses from the offsite power system; and including components and circuitry interfaces th'at affect the proper performance of such interfacing breaker).
4,160 - 480 V.switchgear from engineered safety systems (ESF), including ESF buses 4,160 - 480 V transformers (ESF load centers) 480 - 120/208 V transformers (control room and ESF area emergency 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 dc panels and switchgear (vital de power distribution) 120- V ac instrument bus panels (vital instrumentation ac power dist:ibution)
Containment' penetration assemblies Main ~ control board j e-C Radiation monitor panel
! Hot shutdown panel
\') Control room air conditioning control panel
. Post LOCA-Hydrogen control panel Emergency lighting r Emergency communications Diesel generator.and accessories Diesel generator control' panels ~
Relay boards and racks-Wire and cable raceway system I Underground electrical duct' bank system l Cable system (power, centrol and instrumentation) l: Instrument racks ! Electrical supports I Heat tracing / freeze protection i
- 30. INSTRUMENTATION AND CONTROL SYSTEM COMPONENTH Reactor trip system Engineered safety features (ESF) actuation system
[. Systems required for safe shutdown l Safety relatec instruments, tubing and fittings LO L I
- Page 69 of 93 E __ _ _ _- _ _
Op;rcticnni' Quality Accurenca Plcn .
- ROY 13 Appendix B 1
- 31. HEATING. VENTILATION AND AIR CONDITJOEING SYSTEMS (HVAC) '
i Control and Rolay 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 Fans, ductwork and dampers Filters 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 i Fans, ductwork and dampers Auxiliary Building Normal Ventilation System Ductwork and dampers associated with steam exclusion Turbine Building Ventilation System ,
.Ductwork and dampers associated with steam exclusion i f
Electric modules with safety function , Cable with safety function j HVAC sensors and monitors having safety function
- 32. FIVIL 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 Tarbine Building (housing emergency diesel generator, cooling vater pipes, batteries, safeguards switchgear, auxiliary feedwater pumps) I i Page 70 of 93
]
l . _ _ _ _ _ _ _ _ _ )
r .... - , y ,
, p.
4 , J (op3 rational Quality-Agguranca Plan 1 Revjl3 Appendix B j
- 33. OTHER i
' 3/(/ -
;s
'~' A. Fire protection system; piping associated with'the. l:61 f v safeguards ventilation exhaust filters and l j
- , j ,
containment penetration B. : Turbine building crane l 61 Crane structural support Crane electrical, cable, controls and instrumentation with safuty function a:--
. c.
J' l o 7-
'(x Page 71 of 93
- s
'I t
Op;rction21 Quality A?rursnca plcn Rav 13 App:ndix C s APPENDIX C i (_I Nuclear Flant Fire Protection Procram 1.0 Policy Statement Northern States Power Company (NSP) has established a system of Administrative Control Directives (ACDs) that implement the OperationalfQuality Assurance Plan. This system shall be used to implement the requirements of the operating nuclear power plant fire protection program. The basic requireraents 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 establishing and implementing the fire protection program, or any part thereof, but shall retain responsibility for the program.
- 2. The authority and duties of persons and 7,
organizations involved in the fire protection
! j program shall be clearly established and delineated
' in writing.
- 3. To assure adherence to the fire protection program, ;
management metsures 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 pertcin to the fire protection program.
- 1. Director Pcwer Supply Quality Assurance
- a. Scheduling and assuring completion of independent off-site fire protection insn:ctions and audits.
- b. Reviewing non-plant (other than Nuclear l 62 73 Engineering and Construction) purchase
( requisitions related to fire protection.
)
Page 72 of 93
r-i Operational Qu;11ty Accurcnca Plcn - R;v 13 Appindix C
- 2. Plant Managers
- a. Routine inspection of the plant for fire hazards. g]
- b. Establishing plant fire brigades.
- c. Procurement of equipment for the fire brigades.
- d. Ensuring that fire brigade members receive required training and physical evaluations,
- e. Coordinating fire drills and determining their effectiveness,
- f. Establishing coo'peration with the local fire department, including joint drills 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.
- i. Establishing a system to con. trol nonconforming items.
F j. Reviewing required work processes for fire ll 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 i required).
Reviewing purchase requisitions initiated by the m. 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 dating a fire.
- p. Preparing news release information for NSP's Communications Department.
O Page 73 of 93
' operational Quality Assurance Plan Rev 13 Appendix C
- 3. Manager Production Training
[ a. Establishing a training program for the fire brigades. 2.3 Fire Protection Engineer
- 1. A fire protection engineer (or engineering consultant) shall be used to provide the followir g types of services:
- a. Review of design for a significant modification to a fire protection system.
- b. Review of proposed plant modifications which would intr 1 duce major hazards not analyzed in the Fire rnzards Analysis,
- c. 3riennial independent fire protection inspections (see section 14.2). l 45
- 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; Er l 63
- b. A membet- in the 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, l4 Section III.H, and the requirements listed below:
- a. Fire brigade composition may be less than the l
minimum requirements for a period of time not to exceed 2 hours, 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 minimrm shift crew () complement needed to safely shut down the unit (s). Page 74 of 93
operational Quality Assuranco Plan Rav 13 Appandix C
- a. Fire brigade composition may be less than the .
minimum requirements for a period of time not to I exceed 2 hours, 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.
- 2. Each fire brigade shall have an appointed leader.
This leader shall not be the Shift Supervisor (the Unit No. 1 Shift 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 follow-up physical examinations shall includo l 64 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 tasting, 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 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 hatards.
l 1) Common fire hazards. ' 21
- 2) Combustibles, general.
l l 3) Flammable liquids.
- 4) Flammable gases.
- c. Fire detection systems.
Page 75 of 93
Operational Quality Assurance Plan Rav 13 Appendix C
- d. Types of extinguishing systems.
- e. Special fire hazards associated with nuclear-21 power plants.
- f. Emergency Plan with emphasis on fire emergency.
- 2. .Bcsic instruction in fire protection shall be given to contractor personnel before granting them unescorted access to safety related areas of the plant.
4.2 Monticello
- 1. Fire brigade trainin'g shall be conducted in accordance with the requirements of Appendix R, l4 Section I.
4.3 Prairie . Island
- 1. Level II fire protection training s' hall 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 In addition, the subject matter in Level I.
- following items shall be covered:
O' 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 emargency breathing equipment.
l h. Considerations of radiation and contamination in fire breas. O . Page 76 of 93
Operationni Quality Annuranca plcn - R;v 13 App:ndix C
- 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 followirig additional material:
- 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 hazarde.
- 3. Training Documentation Classroom training sessions, practica sessions, and drills for the fire brigade shall be documented.
The following should be included in the documentation for persons participating:
- a. Name.
- b. Date. I 21
- c. Summary of what was done.
- d. Evaluation by observer.
5.0 Drills and Practice 5.1 Monticello
- 1. Fire protection drills and practices shall be conducted in accordance with the requiremeists of Appendix R, Section I and the requirements listed l4 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 contrc3 room operators in their use shall be tested during drills,
- c. Each year, one dz 11, conducted with any fire brigade, shall im olve the local on-duty fire department.
93 O Page 77 of
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' Operatien211 Qunlity A2Curcnco Plan pu ', < ,
R;v 13- App:ndix C $ , .. 5.2 IPrairie Island
;7\
!( s_ ) -
- 1. Drills
- a. Drills shall be scheduled so that each fire brigade a
will participate in at least four drills'per year. The following types of drills shall be scheduled:
- 1) Involve local on-duty fire departz nt. 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 independent of NSP's staff.
- 3) Back shift, conducted by the fire brigade leader on duty at the time. This shall be scheduled at least once per year for each brigade.
- 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 r'N 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. Thise 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 equipment under strenuous conditions.
- 3) Extinguishment of fire.
/~' 4) Best method by which te approach each type
( )( of fire, to the extent possible. 78 of 93 i Page l-1. l
op;rctirn21'Qunlity-Accur2nco Plan i R;v 13' App;ndix C y
- b. Brigade members' missing a practice session shall 4'
.be rescheduled to attend a la.ter session with
~
another brigade. If this is not possible, they f shall be required to review the training material covered during the practice session.
.6.0 Control of Combustibles and Ignition Sources 6.1 Monticello E t
- 1. Control'of combustibles and ignition sources shall l I l. be in accordance with the requirements of Appendix R, l4 {
f Section K, and the requirements listed below: j
- a. All areas containing safety related equipment or f cables shall be surveyed once each working day t for fire hazards by a member of the plant staff.
- b. Storage of combustible materials shall be permitted only in posted areas or in approved cabinets and containnrs.
! c. Transient combustibles in any safety related l l area, or area containing safe shutdown equipment, shall be limited to the equivalent of
- 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, i grinding and open flame work is involved.
r-I
- 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. Snoking shall be prohibited in all safety r: elated areas, except those specifically d esignated 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: g Page 79 of 93 i
L_.n_
op2rctional Quality A3:uranca Plan .l Rav 13 App 2ndix C .!
- 1) Govern the handling and limitation of the use of ordinary combustible materials, g[~~$ combustible and flammable gases and liquids, pg ,/' high efficiency particulate air and charcoal e
filters, dry ion exchange resins, or other combustible supplies in safety related areas. Govern the removal from the crea of all 4 ' 2) waste, debris, scrap, oil spills, cr 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 ac combustible and flammable materials during maintenance, modification, or refueling operations.
- 4) Govern the use of specific combustibles in safety related areas.
a) All wood used in safety related areas during maintenance, modification, or refueling operations (such as laydown blocks or scaffolding) shall be treated with a flame retardant. b) All untreated wood in safety related
/')
\s,/ 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. Hewever, all combustible materials shall be removed from the area immediately following the unpacking, d) Lc ge amounts of combustible material shall not be left unattended during lunch breaks, shift changes, or other similar periods. l e) Loose combustible packing material such as wood or paper and excelsior shall be placed in metal containers with tight-fitting self-closing metal covers.
O 93 Page 80 of
Op3ratienni Quality Accurcnca Pltn ' RLv 13 Apprndix C
- 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 cont 11ning 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.
- 2. Cutting, Welding, Grinding and Open Flame
- a. Cutting, welding, grinding ar.d 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 management.
- 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. l 65 7.0 Fire Fighting Procedures 7.1 Monticello Fire fighting procedures shall be established in accordance with the requirements of Appendix R, l4 Section K. Page 81 of 93 O
'l Op3 rations 1 Quality Accur:nca Pltn RCv 13 App:ndix C ]
7.2 . Prairie Island Fire fighting procedures or instructions shall be (' ') 1. developed to cover the following areas:
- a. Discovery of fire including:
t
- 1) Notification.
21
- 2) Attempts to extinguish fire.
- b. Action of Control Room operator including:
- 1) Announcement.
Sounding of fire alarm. 21 2)
- 3) Who to notify.
- c. Selection and delineation of responsibilities of fire brigade members.
- d. Coordination of off-site fire departmant activities.
- e. Actions of security guards during a fire emergency.
- f. Delineation of responsibilities of other plant
/)i
(, 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 suppression 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.
- 6) Identification of radiological and toxic hazards in the area.
G l )
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Page 82 of 93 (
.op; rational Quality Accurencs plcn
. R;v 13 App;ndix C
- 7) Ventilation system lineups to minimize spread of smoke and to remove smoke from the area.
(
- 8) Identification of actions which must be i coordinated with operations personnel.
- 2. Instructions and preplanned strategies shall be tested during drills.
8.0 Modification Control (Monticello and Prairie Island) l 26 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 making 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?
8.2 If a fire protection review is required, the individual lk assigned to perform the review shall use the 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. i Page 83 of 93 l
x, - - - _ _ - _____ - __ - _ _ _ _ _ _ _ _ _ __ _ l'
.g Ni #
\ ;, L. (Revf13 Lapp ndix C-
" E9;04l Procurement Control-(Monticello and Prairie Island) l 26
[' ' ~ 9.1: .. Underwriters Laboratories (UL) and' Factory Mutual,(FM) t*
' directories shall be reviewed to determine if the item is listed'es: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 notLbuying an item that is UL or FM listad is if it is a replacement of original equip-k ment or: NSP standard type, then it.shall be identified uas such and procured from the original supplier or manufacturer. As a minimum, the item or equipment 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'avaluation shall be'made to determine the M' compatibility of the item to the existing system or component, car l 63
- 2. If'the item has been manufactured for a long period of time, ADA l 48 '
a., The item is standardized, and l 48 (h -b. The item has a satisfactory performance history, A/ and l 48
- c. Appropriate receipt inspection is identified in the procurement documents, then an evaluation is unnecessary. . The fact that the supplier and item meets these requirements.shall be documented in the procurement files. .
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.
- 2. The model number of the component *or equipment shall be identified.
- 3. The specific part number shall be identified.
- 4. Documentation frum the supplier shall be requested that indicates the part-delivered meets the specifi-cation of the part used in the original component or equipment. If the part has been changed, the manufacturer shall be asked to indicate any changes in the operation of the component or equipment. In ]
lieu of this documentation, the acceptance of the O part shall be based on inspection or testing. Page 84 of 93 a
Oparationni Qunlity AScuranca Plan - Rav 13 App:ndix C 9.4 -All purchase requisitions pertaining to fire protection systems and equipment shall be reviewed. Plant requi-
- sitions and Nuclear Engineering and Construction g requisitions shall be reviewed by an individual desig-nated by the Plant Manager. Non-plant requisitions l 62 shall be reviewed by an individual designated by the Director Power Supply Quality Assurance.
10.0 Instructions, Procedures, and Drawings (Monticello and l 26 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 issued 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 and Prairie Island) l 26 11.1 The Technical Specifications specify the surveillance ; and inspection requirements for the fire protection system. Surveillance shall be scheduled, per'urmed, 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. { i 12.0 Conditions Adverse to Fire Protection (Monticello and Prairie l26 Island) 12.1 Administrative Control Directives shall establish criteria for housekeeping. 12.2 . Work control process procedures shall be used to correct equipment failure, 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 ' l report them to their supervisor. 13.0 Records (Monticello and Prairie Island) l 26 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. ll Page 85 of 93
i=
; operational Quality A2suranca Plan
- R v.13 " App 3ndix C
;j, 14 .' O . Audits (Monticello!and Prairie Island)~ l 26' hk 14'.1 In addition to normal quality assurance audits (at-
.least. biennial), an independent fire protection and loss prevention inspection and audit shall-be o performed annually-at each plant utilizing either.
qualified off-site NSP personnel or an outside fire protection engineer or engineering consultant-(annual l 66 independent inspection). E14.2 An1 inspection and audit by an outside qualified fire protection engineer or engineering consultant (see 66. Section 2.3) shall be performed at each plant at.least. 45 every three years (triennial independent inspection). 14.3 ' Inspection and audit'results shall be reported to
. levels of management having fire protection program responsibilities in those areas audited or inspected.
LO Page 86 of 93
P, f0p2 rational' Quality A22uranca Plan l
.Rcv : 13 ' App:ndix D APPENDIX-D
[ _
-Revision 13 Chance Summarv l1 f
'This appendix summarizes the changes made in Revision 13 to the l1 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.
' Chance Number identifies the change number next to tite sideline on the affected pages.
EAEA identifies the page numbers containing the change. Reason (R) identifies the reasen for the change. Basis (B) 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 Number Pace (s) Basis 1 .1, 87 El Revision number updated to 23. El Not required; editorial item. s 2 1, 6, 7 El Underlined section title. ( ,) 31 Not required; editorial item. 3 1, 4 El Removed spaces (s). Di Not required; editorial item. 4 1,24,28,55,65, El Comma (s) added. 74,76,77,79,81 Hi Not required; editorial item. 5 4 El Added the "N"'in ANSI N18.7-1976. El Not required; editorial item. 6 4 El The lower case used for the word
" delete".
El Not required; editorial item. 7 4 El The quotation marks were moved inside of the period. El Not required; editorial item. 8 5 El The word " Requirements" was capitalized. R1 Not required; editorial item. 9 6 El The words " fossil-fueled" were hyphenated. El Not reqcired; editorial item. j O Page 87 of 93
Op raticnni Quality A;curanca Pltn RV 13 App:ndix.D Change Reason ' Number Pace (s) Basis
-10 7, 12 El Section reference renumbered to reflect the addition of the General Superintendent Quality, Security & Administration at the Monticello plant.
El This is a reference c.rrection only. It does not reduce previous commitments. 11 8 Hi The word "With" was capitalized. 31 Not required; editorial item. 12 8, 23 El Title change from President & Chief Executive Officer. This is a title change only. It El does not reduce previous commitments.
'13 8 El The reference to the organizational chart title and location was corrected.
El Not required; editorial item. 14 8 Bi The hyphen was removed from the word " nonconforming". 31 Not required; editorial item. Title change from superintendent i 15 9 El Nuclear Projects & Supplier Quality Assurance. This is a title change only. It El does not reduce previous commitments. 16 9 El Title change from Superintendent Nuclear Operations Quality Assurance This is a title change o,nly. It 31 does not reduce previous 1 commitments. i 17 9 El The word " selected" was removed. El This is an improvement of the QA Program description. It does not reduce previous commitments. 18 9 El Section number corrected. 31 Not required; editorial item. 19 9, 10 El The words " Corporate Nuclear" were i added. 31 Not required; editorial item. , I 20 11 El Double spaced program listing. l Not required; editorial item. l Hi Page 88 of 93 I
9 9 'ophrstionsi QualityiAssurnnes Plan Rav 13. App 3ndix D l. i ' Change Rea52n
. ,_.\ :
./
~ Number Pace (s) Basis 21 11,23,30,31 El Period placed at end of each item.
-75,76,77,82 . AL- Not required; editorial item. ~
j 22 11 El The responsibility for technical i manual control programs changed to. 1 Plant Managers. 31 This is a responsibility l ' description only. It doen.not-L l reduce previous commitments.
'23 12, 61,.62 E1 . change from "QA" to " quality i ass'urance".
Hi Not required; editorial. item. 24 12 B1 Position of General Superintendent Quality, Security & Administration added at Monticello. El This is a responsibility description only. It does not { reduce previous. commitments, i
)
.25 13 B1 The responsibility to provide ;
4 trend reports was added to the Plant Superintendent of Quality
; Engineering.
O) ks/ Hi This is an improvement of the QA Program description. It does not reduce previous commitments. 26 13,31,40,60 El Change from "&" to the word 63,83,85,86 "and". Di Not required; editorial item. H .27 14 El The responsibility to maintain INPO accredited training programs added to Manager Production Training. El This is an improvement of the QA Program description. It does not reduce previous commitments. 28 16 El Section number cor*:ected. 11 Not required; editorial item. 29 16 El The word " Corporation's" changed to " Company's". 31 Not recuired; editorial item. 1 O Page 89 of 93
~oparctional Quality Accurcnca Plan Rev.13 Appendix D Change Reason Number Pace (s) Basis ,
f 3C 17 El The Manager Production Performances & Services has been removed from the QA Program. The responsibilities have been reassigned to the Plant Managers and the Director Power Supply Support. 31 This is a responsibility description only. It does not reduce previous commitments. 31 17, 18 B1 Positions of Manager Electric Maintenance and Manager Testing Laboratory added as positions with
' QA responsibilities.
El This is an improvement of the QA Program description. It does not reduce previous commitments. 32 18 El Listing of positions reporting to General Manager Plant Engineering
& Construction added.
R1 This is a correction of an omission from Revision 12. It does not reduce previous ' commitments. 33 19 B1 Title char.ye from Manager Technical Services. Di This is a title change only. It does not reduce previous
, commitments.
34 19 El Added the word " General". . El The word " General" was inadvertently left off of Revision 12. It does not reduce previcus commitments. 35 19 R1 An "s" was added to the word
" operations".
Hi Not required; editorial item. 36 19, 20 B1 Delineated responsibilities of positions reporting to General Manager System operations. Si This is an improvement of the QA Program description. It does not reduce previous commitments. 37 20 El The lower case used for the word
" company's".
31 Not required; editorial item. 1 Page 90 of 93 I
JOp;rstional. Quality AIguranca Plan L R v.13f Appandix D l:
-Change' .
Reason . j
,e '; Number Pace (s) Basis i Further delineated the l ; ,i -- - 38' 20. El- '
responsibility'ies of the Director l' - Power Supply Support. This is an improvement of the QA 1 Bi Program description. It does not reduce-previous commitments.
'39 20 El The word " purchasing" changed to
" procurement".
El Not required; editorial item. 40 20 El Drawing control responsibility ! removed from Director. Power Supply Financial Operations and assigned to Director. Power Supply Support. 4 El This is a responsibility description only. It does not reduce previous commitments. 41 22, 30, 32 Bi The word " Procedures" was capitalized A;. Not required; editorial item. 42 23 El Removed the letter "s" from the word " Director". rs 31 The Director Power Supply. Quality Assurance is the only Director () authorized to approve Corporate level implementing procedures. This does not reduce previous commitments. 43 23 El The word " Plant" was capitalized. R1 Not required; editorial item. 44 23, 24 El The lower case used for the word
" level".
Si Not required; editorial item. 45 25,42,74,86 El Tne word "Section(s)" was capitalized. El Not required; editorial item. 46 26 El Section number reference changed from "1" to "1.0". El Not required; editorial item. 47 28,31,39,43, El The word "related" was added 50,51,52,55 after the words " safety" and/or
" protection".
, 31 Change made to provide consistency throughout the document only. It
,f does not reduce previous 2 f., commitments.
Page 91 of 93
Op;ratienEl Qu21ity A2curanca Plcn , R;v 13 App;ndix D Change Reason Number Pace (s) Basis 48 .3 2 , 40, 84 El 'The word "and" was underlined. Hi Not required; editorial item. 49 . 40 El The word "non routine" was hyphenated. ' Hi Not required; editorial item. 50 40 El Changed the comma after the word l "considering" to a colon. 31 Not required; editorial item. 51 41 El The lower case used for the word "see". 31 Not required; editorial item. 52 41, 47 El The word "off-site" was hyphenated. El Not required; editorial item. 53 50 El Comma removed after word
" components". '
i Hi Not required; editorial item. 54 53 El The word " repetition" was changed to " recurrence".
. 31 Change made to provide consistency throughout the document only. It does not reduce previous commitments.
55 55 El The word "an" was removed. El Not required; editorial item. 56 58 El The Emergency Filtration Train Building was added to the Monticello list of structures subject to Appendix B of 10CFR50. El This is an improvement of the QA program description. It does not reduce previous commitments. 57 61, 62 El The lower case used for the words
" quality assurance".
31 Not required; editorial item. 58 62 B1 The word "in" was placed before the word "Section". l Hi Not required; editorial item. 59 62 El The lower case used for the words
" quality assurance program".
Hi Not required; editorial item. o
~
Page 92 of 93
L.Op3rstional= Quality.Accurance' Plan:
~ R3v 13. App 3ndix D.
'.N (
' Q)
Change, Number Pace (s)
. ~ Reason Basis comma removed after the word 60 69.: RI.
" systems" and parentheses placed around'the explanation.
JLt. Not requred; editorial item. 61- 71 El. Paragraph identifying scheme changed'from numbers to letters. 11, Not required;-oditorial. item. 62 72, 85 R;; The word "non-plant" was hyphenated.
- 11. 'Not required; editorial item.
-63 t 74, s4 El. The word "or" was underlined.
- 11. - . lot required; editorial item.
~
64 75 Bj. The word " follow-up" was hyphenated.
- 11. Not required; editorial item.
65' 81 ' El. The wcrds " shift supervisor" were capitalized. H1. Not required; editorial item. 66- 86 El. The parentheses were removed from. around the phrase "or engineering
- 31. Nt r d; editorial item.
Page 93 of 93
MONTICELLO 9 DESIGN AND FABRICATION REQUIREMENTS Tbe Monticello reac or vessel was designed, fabricated, inspected, ard tested in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section III, Nuclear vessels 1965 Edition and Addenda to and including Summer 1966 Addenda, and the foll; wing additions:
- 1. ASME SA533 plate and Incor.el material per Summer 1967 Addenda.
- 2. Main closure flange material per Code Case 1332-3.
- 3. Studs and nuts material for main closure flanges per Code case 1335-2.
- 4. Main closure flange and stud shank transition radius per Code Case 136
- 5. Bearing stresses for stabilizer brackets and coefficients of thermal expansion per Winter 1967 Addenda.
- 6. Magnetic particle and liquid penetrant examination per Winter 1966 Addenda.
The date of the contract between the Buyer, General Electric Company, Atomic Power Equipment Department, San Jose, California and the Seller, 9 Chicago Bridge and Iron Company, San Francisco, California, was July 18, 1966. There are no deviations to the formal code throughout the design, fabrication, inspection and testing of the reactor vessel. Design fabrication, inspection, and test requirements in addition to those required by the B&PV code were required by the Buyer's vessel purchase specification 21A1112 (Exhibit 1) . These include but are not limited to the following pertinent inspections and/or tests:
- 1. Established specific maximum nil ductility transition temperatures for the main closure flanges and the shell and head materials connecting to these flanges (+10*F NDT temperature) and elsewhere (+40*F NDT temperature).
- 2. A fabrication test program on vessel shell material which included testing of large size tensile specimens (80" of the vessel wall thickness in diameter) both plain and welded samples. (See Exhibit 3).
- 3. Provisions are made for determining the effects of nuclear radiation upon the reactor vessel structural materials by supplying specimens of the vessel material to be exposed to the core irradiation at the vessel wall inside of the vessel.
Pertinent certifications are contained in Exhibit 2, Manufacturer's Data Recort and Vessel Certification, Chicago Bridge 6 Iron Company. 9 The summary of results of the detailed stress analysis is contained in Exhibit 4 H.1-1 RE'l a 12/S5
MONTICELLO Plans for the vessel fabrication and assembly were described in Amendment 2 to the FDSAR, " Design Fabrication and Ereccion of the Reactor Vessel." Actual fabrication and assembly was in accord with Section IV of Amendment 2 except for minor modifications as listed in Exhibit 5 of this report. The GE quality control of the reactor vessel was essentially as described by General Electric Quality Control Plan, Section IV of Amendment 2 to the FDSAR, except the Domestic Turnkey Projects organization of General Electric Co. also nade an independent QC audit. A detailed seismic analysis of the Reactor Pressure Vessel was prepared by John Blume & Associates and was included in Appendix A along with other seismic analyses. In 1977, repairs were made to the reactor pressure vessel feedwater nozzles and safe ends to minimize damage to the feedwster nozzles due to thermal cycling. The repairs consisted of removing cladding from the nozzle blend radius and bore and the installation of a feedwater sparger interference fit thermal sleeve with a piston ring seal. These design changes invalidat.ed the
" Summary of Stress Analysis for the feedwater Nozzles" shown on page 4-14 of Exhibit 4. Details of this repair and design are contained in Exhibit 7. . Also in 1977, a design change modified the CRD return line because of its susceptibility to intergranular strces corrosion crecking. The 3" CRD return line and the reactor vessel nozzle safe-end forging were removed and the nozzle was capped using a 4" diameter schedule 120 pipe cap. Tius design change eliminated the imposed mechanical loading for the nozzle, creating a much less severe condition than the nozzle was originally designed for. As a result of this modification, the " Summary of Stress Analysis for the 3" CRDHSR Nozzle" shown on page 4-18 of Exhibit 4, is invalidated. Details of the modification and new stress analyses are contained in design change 77M069.
In 1981, new feedwater nozzle safe ends featuring a tuning fork design with a welded in thermal sleeve were installed and a section of piping upstream of each nozzle was replesced with piping of a different material. These modifications were performed to provide a significant reduction in thermal cycling of the feedwater nozzle area. The new stress analyses that replaced the " Summary of Stress Analysis for the Feedwater Nozzle" shown on page 4 14 of Exhibit 4, are contained ir E).hibit 8 and Exhibit 9. In 1984 several modifications were incorporated to provide greater resistance to intergranular stress corrosion cracking. The core differential pressure and standby liquid control safe end was replaced using a safe end of similar design, but with different materials. The new stress analyses are contained in General Electric Stress Report No. 23A4115, included in Design Change No. 832049C. H.1-2 REV 7 12/88
MONTICELLO The jet pump instrumentation safe end and penetration seal was replaced with the jet pump instrumentation nozzle penetration seal, using low carbon 315 to replace the original ASTM A508 Class II material. The new stress analyses are contained in General Electric Stress Report No. 23A1939, also included in Design Change No. 83Z049C. The core differential pressure and etandby liquid control, and the jet pump instrumentation modifications invalidated the " Summary of Stress Analysis for Core Differential Pressure and Liquid Control Nozzle, Head Cooling Spray and Instrumentation Nozzles, Vent Nozzle, Instrumentation Nozzles, Jet Pump Instrumentation Nozzlss, Drain Nozzle, High Pressure Seal Leak Detector Nozzle and Low Pressure Seal Leak Detector Nozzle" shown on page 4-28 of Exhibit 4. Also, in 1984, a corrosion resistant cladding overlay was applied to the inside diameter of the RV head vent nozz2e and PV head cooling spray and instrumentation nozzles. The weld overlay of 308L isolated the IGSCC susceptible existing weld butter located in the weld residual stress area from the reactor coolant. As documented in General Electric Stress Report No. 23A4280, part of Design Change No. 84Z068, stress calculations performed originally at this location are still valid. The recirculation inlet and outlet nozzles were both modified during the 1986 General Electric Stress Report No. 23A1627, part of Design Change outage. No. 83Z049A, documents the analysis of the redesign and replacement of the recirculation inlet nozzle safe ena and thermal sleeve, including the attachment weld and the weld overlay to the recirculation inlet nozzle. This design change invalidated the " Summary of Stress Analysis for Recirculation Inlet Nozzle" shown on page 4-22 ef Exhibit 4. Eechtel Stress Report No. SR-10040-SS2 (Rev. 3), also part of Design Change No. 832049A, documents the analysis of the replacement of the recircula-tion outlet nozzle safe end fitting, a forged and machined component made of SA 358 Type 316 stainless steel. The " Summary of Stress Analysis in Recirculation Outlet" shown on page 4-24 of Exhibir 4 has been invalidated by this change. In 1986, new core spray safe ends featuring a tuning fork design with a thermal sleeve were installed along with a section of piping upstream at each nozzle. This modification was performed to minimize the chance of IGSCC from occurring in the Core Spray System. The new stress analyses is documented by Bechtel Document 301-P-5. Also in 1986, the CRD return nozzle, previously capped in 1977, was again modified. The purpo:.e of the modification was to remove that portion of the existing weld butter layer susceptible to IGSCC, and re-clad the weld prep area with corrosion resistant cladding and install a new nozzle cap. General Electric Stress Report No. 23A5553, included as part of Design Change No. 86Z016, documents the analysis. H.1-3 REV 7 12/88
tv j S , i l i I i THIS PAGE INTENTIONALLY LEFT B W T O'
x MONTICELLO-i'
.; > s[J <( ' \- TABLE OF. CONTENTS (Continued)
(g Pace I.4-1.2 High Pressure Coolant Injection I.4-2 (HPCI) I.4-1.3 Safety / Relief Valves (S/EV's) I.4-3 I.4-1.4- .Reacto r 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) 1.4-4 I.4-1.7 RHR Service Water (RHRSW) I.4-4 I.4-1.8 Shutdown Instrumentation I.4-5 I.4-1.9. Auxiliary Support Systems I.4-5 I.4-2 Description of Safe Shutdown Paths I.4-5
.I.4-3 Safe Shutdown System Components I.4-6 Lo cation 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 Hign Pressure Coolant Injection I.4-8
/ (HPCI) System
( , I.4-3.4 I.4-3.5 Safety / Relief Valves (S/RV's) Reactor Core Isolation Cooling' I.4-8 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-ll I.4-3.9 Shutdown Instrumentation I.4-12 I.4-3.10 Eme rge ncy Se rvice Wa ter I.4-13 (E SW)
- System I.4-3.ll Diesel Generators and Auxiliaries I.4-15 I.4-3.12 Auxiliary Power Distribution I.4 ,
Sys tems I.4-3.13 DC Power Systems I.4-16 T,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 y I.5-1.2 Feedwater System I.5-2 I.5-1.3 Condensate system I.5-2 I.5-1.4 HPCI ( S te am) System I.5-2 I 5-1.5 RCIC (Steam) System I.5-3
'[\ Reactor Wate r Clea nup ( RWCU) i I.5-1.6 I.5-3 1( / System I-v REV 6 12/87 L=.____--_________-________ _ . . _ _ _ _ .
MONTICELLO' TABLE OF CONTENTS (Continued) Pace I.5-1.7 Instrument and Sample Lines I.5-3 I.5-1.8 Core Spray System I.5-4 I.5-1.9 Residual He,at 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.5-5 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 1.5-3 Description of Preak 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 1.5-3.6 Reactor Water Cleanup Break I.5-23 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 System 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.6-17 j I.6-2.6 Reactor Water Cleanup System I.6-18 1 I.6-2.7 Core Spray System I.6-21 I.6-2.8 Residual Heat Removal ~ System I.6-23 I.6-2.9 HPCI (Water) System I.6-23A I.6-2.10 RCIC (Water) System I.6-23A I.6-2.11 Standby Liquid Control System I.6-23A I.6-2.12 offgas System I.6-23A I.6-3 Table of System Effects I.6-4 I.6-23B Use of RCIC in Safe Shutdown Sequence I.6-27 I- vi l REV 7 12/f8 w _-___ _ _ _ _ -_
MONTICELLO criterion or the 1 inch nominal pipe size criterion. n For each system, an evaluation is provided for each HELB's are postulated.
) compartment in which The v evaluations ' describe the effects of pipe whip, jet impingement, compartmental pressurization, flooding, 1 and environmental effects. Information on paths to safe shutdown is given, if the postulated HELB's damage <
SSD equipment. I.6-2.1 Main Steam System The high energy lines for the Main Steam System are located in four compartments, the Main Steam Chase (II/2F) in the Reactor Building, the Condenser 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 include the four main steam lines from the drywell penetrations to the eg'talization lines, the turbine bypass lines to the condenser, primary steam to SJAE line, and Primary Steam to Steam Seal System. I.6-2.1.1 Main Steam Chase (II/2F) The Main Steam Chase contains the four main stea.m lines [] G/ , and the associated drain piping. Pipe whip from the main steam lines is not considered a problem, since these lines are restrained at several locations in this compartment and pipe whip reactions are not toward any SSD equipment. Since the only postulated break loca-tions are at the penetrations to the drywell and the penetration acts as an anchor point, only circumferen-tial breaks are required to be postulated. The resultant jets from the circumferential breaks do not impact any SSD equipment except for a small portion of the jet which would hit the compartment ceiling. Imbedded in the ceiling are Division II cables of SSD equipment, but these cables would be unaffected because of the concrete ceiling reinforcement below the con-duits. HPCI and RCIC would both be available to support safe shutdown, since the steamline isolation valves and injection valves are qualified for the anticipated steam environment. In addition, all other SSD systems of both divisions would be available. Depressurization is accomplished by the HPCI or the l S/RV's, and LPCI and CS can be used for decay heat removal and RPV level maintenance. Therefore, safe shutdown can be accomplished. m I.C-5 i REV 7 12/88
MONTICELLO The postulated main steam line break would cause a peak compartment pressure of 21.7 psia (Reference I.8-11), 3, rupturing the blowout panels to the Turbine Operatin ; Floor (X/30), and failing the door to the west side ci the Reactor Building at elevation 935'-0". The effects L of the pressurization would be from the same circum-ferential break, and the identical SSD equipment would be affected. No addi.tional SSD equipment would be adversely affected as a result of the pressurization. Therefore, safe shutdown could be accomplished in the same manner as described above. Flooding in the area from a main steamline break would cause the bottom of the enmpartment to flood to a height of 1 foot, assuming the entire mass of steam leaving the break condensed in the area. This would be extremely conservative due to the loss of steam through the blowout panels and door. With this postulated flood height, no additional SSD equipment would be affected. The HPCI and RCIC injection valves would remain available due to the height of the operators above the floor (~5 ft.). Water would not leave the area since the bottom of the door opening is 4 feet above the compartment floor elevation. The peak temperature in the room would be 298'F and the relative humidity would immediately go to 100%. The environmental effects of the Main Steam line break in this compartment were used for equipment qualification purposes. Therefore, no additional SSD equipment will be adversely affected, and the previously identified path to safe shutdown can be utilized. Breaks on the Primary Steam Drain Line (PS15-3"-EB) were not evaluated, since both containment isolation valves are closed during normal operation. It is, therefore, concluded that a path to safe shutdown exists for any postulated Primary Steam break in the Main Steam Chase (II/2"). 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 Condenser Area. No concern exists with respect to pipe whip and jet impingement, since there is no SSD equipment in the area. An evaluation was performed to assess the consequences of flooding in the compartment. It was assumed that all steam released condensed in the compartment and that
'11 the water remained in the SJAE Room.
. The result was a flood height of 8 inches. Since there is no SSD I.6-6 REV 7 12/88
t-MONTICELLO equipment in the room, there was no negative effect. n Also, the drains in the room can drain 'he entire L volume to the sump area. In addition, the >r to the
,U SJAE Room is air tight and any leakage as a result of the flooding would be negligible.
Compartmental pressurization and-environmental effects from this steamline break would not adversely effect any SSD equipment. The SJAE has adequate vent areas to the Condenser Bay (X/12C) and to the 931'-0" elevation of the Turbine Building. The door to the SJAE Room has also been modified to withstand the pressurization. The peak pressure in the SJAE Room and peak temperature from the steamline break were bounded by the peak pressure and temperature generated by a Main Steamline break in the adjoining Condenser Bay. The only SSD equipment that may be adversely affected is the essential switchgear in Compartment (IX/13B). The pressure and temperature transient in the Switchgear area is bounded by the same Main Steamline break in the Condenser Bay, for which the switchgear is qualified. Consequently, the SJAE Room break does not adversely affect any SSD equipment and paths to safe shutdown exist. I.6-2.1.3 Condenser Area (X/12C) [dD The bulk of Main Steam piping is within the condenser area. Pipe whip from cach of the main steam lines can affect either of the Emergency Snrvice Water (ESW) lines (SW30A-3*-HF and SW30B-3"-HF). A pipe whip of line PS4-18"-ED could additionally damage RHR Service Water line (SU9-18"-GF). The most critical break would
-be in the steam bypass line (PS7-10"-ED). A whip from this line could damage the Emergency Service Water line SW30B-3"-HF power cables to the HPCI System, and cables of one division of the Suppression Pool Temperature Monitoring System (SPTMOS). The pipe whip effects of the other primary steam piping within this compartment could not cause any damage to safe shutdown equipment.
For jet impingement, the worst 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 and SPTMOS cables on the other side of the line. All other postulated 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 affected, and safe shutdown can be I.6-7 Rl:.V 7 12/88
I . 1
. MONTICELLO achieved from the other safety division, assuming a single active failure (see Section I.6-1).
For the break that damages ESW Line SW30B-1"-HF, the HPCI Power Cables, and one division of SPTftOS, a path to safe shutdown can still be achieved. Following the break, RPV water level can be maintained by RCIC and the pressure maintained by the SRV's. The remaining ESW line (SU30A-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 fallure be on the Division "I" diesel generator,.ESW Pump P-lllC, i or some other component which does not allow operation of the Division *II" RHR and CS System, the Division "II" RBR and CS Pumps can be used for up to 2 1/2 hours , without cooling water. During this period a Service Water Pump can be aligned to either D/G to provide ; cooling water to these pumps. If the single active i failure is in the RCIC system, then both D/G's are l available, and P-111C is also available. Hence, safe shutdown would be accomplished using the S/RV's to depressurize the RPV and CS/RHR for 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 Condenser Bay. Moreover, if all water condensing from the released steam were to remain in the Condenser Bay, the flood water would not affect any other SSD equipment. This is because of two bay areas that are at a lower elevation with free volumes substantially larger than the volume of water generated by condensing the steam from a Main Steamline break. The peak compartmental pressure from the Main Steamline Break is 15.4 psia and the temperature would be 206*F (Reference 1.8-11). The peak temperature in the Switch gear Area (IX/13B) would be 93*F and the pressure would be 15.4 psia. The doors and the condenser tube knockout blocks do not fail under thit condition (Reference I.8-11). The doors to the Condenser Day have been modified to prevent their failure under a Main Steamline break. The analysis that was performed showed that with the vent area in the Condenser Bay, the knock out blocks would not fail. The calculated temperatures and pressures in the Switchgear Area are bounded by the temperatures and pressures for which the l switchgear is qualified. Consequently, no other SSD 1 equipment is adversely affected other than the SSD 1 equipment directly damaged by the pipe whip and jet impingement. I.6-8 REV 7 12/88
)
MONTICELLO With the exception of the break on the Turbine Bypass
/^Y line (PS7-10"-ED), which damages an ESW line, the HPCI V , power cabling, and one div!sion of SPT!!OS , all other postulated HELB's on steaml'.nes in the Condenser Bay do not adversely affect more than one division of SSD equipment. The path to safe shutdown for the Turbine Bypass line break has ,been previously described, and all other breaks are less severe. Therefore, a path to safe shutdown exists for any postulated HELBs in this compartment from any steamline other than mainsteam.
I.6-2.1.4 Turbine Operating Floor (X/30) The only postulated HELB's in this compartment from the Frimary Steam System are breaks 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 impingement. Also, the environmental effects would be the same as a break in the Condenter Bay (Reference I.8-11), and any water condensing would drain either to the Condenser Bay or lower elevations of the Turbine Building to areas where the water could be drained. The SSD t.quipment in these areas would not be adversely affected by the water produced. I.6-2.2 Feedwater System n ( The high energy Feedwater System piping (FW2A-14"-EB and FB2B-14"-EB) begins in Compartment IX/13B on elevation 911' -0" of the Turbine Building at the discharge o :le of feed pumps P-2A and P-2B, The two main FeecNater lines, FW2A-14" and FW2B-14", pass through Compartment IX/13C, up into Compartment IX/19C and then into the Condenser Area X/12C. Before entering the Reactor Building Steam Chase (II/2F), ecch Feedwater line is connected to its respective high pressure feedwater heaters (E-14A & B and E-15A & B) on The Turbine Operating Floor (X/30). The two Feedwater lines and the Feedwater regulating station piping were seismically analyzed. Break loca-tions were selected based upon the seismic analysis of the piping and the break location criteria established for seismic Category I piping. All four intermediate break locations for the Feedwater system were identified in Compartment IX/13B, the Reactor Feed Pump Area at Elevation 911'. There were no break locations in the Turbine Building Pipe Chase (IX/19C). An additional break location was chosen in the Condenser j Area (X/12C), as a result of the seismic analysis. l O V I.6-9 l REV 7 12/88
MONTICELLO I.6-2.2.1 R]Octor Foodwater Pump Arca - Compartm2nto IX/13B and IX/13C For postulated HELB's on Feedwater piping in these com-partmenta, the only pipe whip or jet impingement targets are i1CC 133, Division I cables, the compartment ceiling, and ESU lines ESul-3"-HBD and SU30A-3"-HF. Loss of the MCC and the cables does not prevent safe shutdown, since the Division II equipment with HPCI ani RCIC could be used for this. If the ceiling to this compartment i t, damaged, essential !!CC 143 on the 931'- 0" elevation could be adversely affected, and a path to safe t.hutdown would not exist. To prevent damage to the ceiling from a Feedwater line break, several pipe wnip restraints and a jet impingement shield have been added to tne area. Therefore, a path to safe shutdown will exist. If the ESW lines, ESWl-3"-HBD or SW-30A-3"-HF are also a targut, then cooling to the Division "I" RER and CS pumps would be lost along with the MCC 133. Safe shut-down would be accomplished by using HPCI or RCIC for RPV level control and the S/MV's for RPV pressure control Decay heat removal would be accomplished by using Oe Division "II" CS and RHR Pumps and an RHR SW Pump. The worst single active failure wou'.d be failure of the Division "II" D/G, its bus tie breaker, or ESW Pump P-IllB. For any of these failures, the capability exists to use the Division "I" D/G to power the . Division "II" equipment. Hence, safe shutdown can be accomplished f or any cf these single active failures. With any oth2r single active failure the D/G to Division "II" is available and multiple pathways within Division "II" exist for achieving safe shutdown. Flooding in this compartment is not a concern. The only equipment possibly affected would be MCC 133, which would already be lost due to the other HELB effects. For the purpose of this evaluation, it was ' assumed that the entire hotwell volume of 80,000 gallonb was pumped into this area. No other SSD equipment would be affected, and most of the water would be contained below the MCC in the flood cavity. The remaining water would be trapped in the compartment due to a 6-inch curb in the corridor to Compartment X/16 in the northeast corner of the Turbine Building at this elevation. In addition, approximately 25,000 gallons would drain to the Mechanical Vacuum Pump Area. Environmental qualification of SSD in other affected compartments was b ned upon the postulation of HELB breaks on the Teodwater piping. Table I.3-1 cf I.6-10 REV 7 12/88
MONTICELLO Reference I.8-11 gives a peak pressure and temperature s of 15.1 psia and 212*F in this compartment and also
~'
J- indicates a peak temperature in the MCC 143 area of 104'F. The MCC's are qualified for that temperature. Therefore, no additional SSD equipment is adversely affected. The only SSD component potentially affected by compart-mental pressurization is MCC 143, and only if the 3 hour fire barrier between the compartments fail. Evaluation's were conducted to assess whether the fire barrier remained intact. The results (Page 3-4 of Reference I.8-ll) obtained showed that the fire barrier was not breached with a differential pressure equal to 0.4 psi across the fire barrier. Therefore, compartmental pressurization does not adversely affect additional SSD equipment. I.6-2.2.2 Turbine Building Pipe Chase (IX/19C) There are no postulated Feedwater Line breaks within this ' compartment. Thus, pipe whip, jet impingement, flooding and compartmental pressurization is not a concern. The maximum t.emperature and pressure resulting from a Feedwater Line break in the Feedwater Pump area would be 14.7 psia and 104*F.
-,a 1 I.6-2.2.3 Condenser Area (X/12C)
G' The break location for Feedwa t.e r Line FW2A-14" is
' located at an elbow between column lines 6 and 7 at the centerline elevation of 934'-10". The Division II Emergency Service Water line (SW30B-3"-3HF) is the only pipe whip or jet impingement target, and the effects of a loss of this line have been discussed previously in Section I.6-2.1.
The resultant flood of the 80,000 gallons of water would fill the Low Pressure Heater Drain Bay and the Condenser Pit. The remaining water would drain to the Mechanical Vacuum Pump Area, in which the sumps are located. There would be no standing water above the 911'-0" elevation to adversely affect any additional SSD equipment. The previous analysis used to determine the pressure, temperature, and humidity transients for various postulated pipe ruptures (Reference I.8-ll) included the rupture of the Feedwater line in the Condenser Area. However, the conditions generated by a Main Steamline break bound those of the Feedwater Line Break (see Table 3-1 of Reference I.8-ll). Hence, no i(. L ' V} I I.6-11 l l
- f. REV 7 12/88 E__ . - - - _
MONTICELLO 4 additional SSD equipment is adversely affected by the Feedwater Line break. Since the only SSD equipment adversely affected by the Feedwater Line break is the ESW line SW30B-3"-HF, it can be concluded that safe shutdown can be achieved for the postulated Feedwater line break in the Condenser Area. I.6-2.2.4 Turbine Operating Floor (X/30) Feedwater lines FW2A and FW2B have break locations at the terminal ends of the inlets and outlets and the Feedwater heaters in this area. The postulated HELB's have no pipe whip or jet impingement targets. The resultant flooding would drain back to the Condenser Area (X/12C) with the same areas flo>ded as a break in the Condenser Area. The cascading water would n'ot a(versely affect any SSD equipmen:. The pressure, temperature, and humidity transients postulated as a result of pipe ruptures included :he Feedwater Line Break in this compartment. The resultant peak pressure was calculated to be 14.9 psia and the peak temperature 189'F. This transient was bounding for the Turbine Operating Floor. Therefore, these conditions were used for qualifying SSD components, and no SSD equipment is adversely affected by a break of the Feedwater Line c. the Turbine Operating Floor. I.6-2.2.5 Main Steam Chase (II/2F) Break locations in this area consist of the terminal ends at each primary containment penetration and one intermediate break location on each Feedwater line. For a break at the terminal point on line FW2A-14-ED, RCIC flow would be lost because the RCIC injection line (FWS-4"-ED) connects upstream of the postulated break. Also, the intermediate break location for line FW2A-14"-ED is at the weld to valve FW-94-1, which is also downstream of the RCIC injection point, and RCIC would again be lost. For Feedwater line FW2B-14"-ED, a break at the terminal end would cause the loss of the HPCI System, because the injection point is upstream of the break. However, HPCI flow is not affected for the postulated break at the intermediate location, because the break location is upstream of the HPCI injection point and the check valve on the Feedwater line (FW2B-14-ED). No safs shutdown equipment can be adversely affected by pi;. : whi9 or direct jet impingement from any pipe break of tho Feedwater System in the Reactor Building Steam Chase. However, for conservatism, both ! HPCI and RCIC are assumed lost due to flooding I.6-12 l REV 7 12/88 _---_ _ _ _ - _ _- )
MONTICELLO (described below) and possible jet impingement due to compartment geometry.
,c \
U Flooding in the Reactor Building Steam Chase, as a result of a Feedwater line break, would cause the loss of both HPCI and RCIC, since the injection valves for these systems (MO-2068 and !!O-2107) would be submerged above the motor opera. tors until the door to the compartment failed. Water exiting the steam chase would flow along the floor of the Reactor Building 935' elevation. Some of the water would also flow down to the compartment containing the Control Rod Drive Pumps and into the HPCI pump room. Additional safe shutdown equipment would not be affected by the water exiting the steam chase, because no SSD equipment is located in the affected areas. Therefore, the flooding from the postulated Feedwater line break would adversely affect
~ HPCI, RCIC, and the CRD Pumps. The path to safe shut-down would be the same as that described in Section I.6-2.1.1 with the exception that an RHR or CS Pump would be required to maintain RPV level.
As in the . case of the Condenser Bay (X/12C), the environmental ccrditions produced as a result of the Main Steam Line break would be bounding to a Feedwater Line break. Since a path to safe shutdown was' demonstrated for the Main Steam Line break (see Section
,ew I.6-2.1), the Feedwater Line break would be mitigated i ) and safe shut down conducted in the same manner as for the Main Steam Line break.
In summary, a Feedwater Line break in the Main Steam Chase (II/2F) could adversely affect HPCI, RCIC, and the CRD Pumps. The Emergency Service Water Line, SW30B-3-HF, would not be adversely affected from a Feedwater Line break. Paths to safe shutdown would include using the S/RV's to depressurize the RPV and using the RHR and CS for RPV level maintenance and decay heat removal. No postulated single active failure prevents the other division from achieving safe shutdown. I.6-2.3 Condensate System The high energy Condensate System lines 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/13B and IX/13C) {s ( I.6-13 REV 7 12/88
MONTICELLO The high energy Condensate System piping includes the main Condensate lines . (C4A-16"-GB and C4B-16"-GB ) , the Feedwater Pump minimum flow lines (C4A-2"-EB and C4B-2"-EB), and the Condensate cross-tic (C7-16"-GB). Break locations were analysis of selected based upon a seismic the piping and break location criteria established for seismic' Category I piping. The pipe runs extended from the terminal points on the third stage intermediate heaters to the suction nozzles of the Reactor Feedwater Pumps. The intermediate break locations on line C4A-16"-GB are in the Feedwater Pump Area (IX/13B), and the intermediate break locations for C4B-16"-GB are in the Condenser Area (X/12C). Because the condensate cross-tie line (C7-16"-GB) and Feedwater Pump minimum flow lines (C4A-2"-EB and C4B-2"-EB) are routed totally within the IX/13B Compartment, all break locations for these lines are in Compartment IX/13B. I.6-2.3.1 Condenser Area (X/12C) For the postulated break locations in the Condenser Bay, the only pipe whip or jet impingement 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 paths to safe shutdown are discussed in Section I.6-2.1.3 for a postulated Primary Steam Line break. The the flooding same caused by a Condensate Line break would be as the Feedwater Line break, and the associated environmental conditions would be bounded by the Main Steam Line break in the same compartment. The flooding, ' compartmental pressurization, and environmental Line effects caused by a postulated Condensate break would adversely affect no other SSD equipment other than the ESW lines directly affected. It can be concluded that paths to safe shutdown exist for postulated Condensate HELB's in this compartment. I.6-2.3.2 Turbine Building Fipe Chase (IX/19C) There are no postulated Condensate line HELB's within this compartment. I.6-2.3.3 Reactor Feedw'ater Pump Area (IX/13B and IX/13C) The postulated Condensate line HELB's in these compart-ments would not affect any additional SSD equipment other than the SSD equipment described for a Feedwater line break in the same compartment. The effects of pipe whip, jet impingement, flooding, compartmental pressurization, and environmental effects are less I.6-14 - _ _ _ _ _ _ _ _ . _ _ _ .. aan_1 s a na .
MONTICELLO adverse than for the Feedwater line break. Therefore, the same evaluations, as provided in Section I.6-2.2.1 this report, and the same paths to safe (~ of apply, ()/ shutdown can be used. 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 (Compartment II/2F). The steam supply line enters the Torus Area (IV/lF) and then the HPCI Compartment Area, (II/lE). This steam supply line is a high energy line from the drywell penetration to the steam supply valve located on the HPCI Turbine. I.6-2.4.1 Main Steam Chase (II/2F) Possible pipe whip targets include the Feedwater and Main Steam lines which are assumed to be unaffected, because they are larger and thicker walled than the HPCI steam line. No other SSD equipment is affected by a HPCI steam line pipe whip. impingement targets include the ceiling through Jet The which Division II embedded conduits are routed. effects of jet impingement on the ceiling is discussed in Section I.6-2.1.1. However, the effects from a HPCI o steam line break are lessEmergency severe than a break of one of Service Water Piping Q the Main Steam lines. (SW30B-3"-HF) and the RCIC Steam line (PS17-3"-ED) can be damaged by jet impingement from a longitudinal line break. This results in loss of HPCI, RCIC and the Division II ESW. - For this postulated HELB, 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 With the reactor present water level and remove decay heat. arrangement of the ESW System and capability to have no single either diesel power either essential bus, active failure 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, is approximately 2-1/2 hours of pump operation acceptable before pump and Room Cooling is required. Therefore, ample time exists to isolate the broken ESW Line (SW30B-3"-HF) and start a SW Pump to provide the cooling water. Compartmental flooding would not affect any other SSD equipment other than the piping lines directly affected by the jet impingement. Assuming all of the steam mass (q) I.6-15 REV 7 12/88
MONTICELLO from the HPCI line were to condense in the compartment, the resulting flood would cover the floor to e height of approximately 1 foot. Since the bottom of tue door to the room is 4 feet above the floor, no water would escape from the Main Steam Chase. Thus, no other SSD equipment would be affected other than the equipment . described above, for which a path to safe shutdown has been demonstrated. . The compartmental pressurization and peak temperatures associatei: with the HPCI steam line break in the Main Steam Chase and effects in the adjoining compartments are bounded by Main Steam line break in the Main Steam Chase. Therefore, no additional SSD equipment would be adversely affected. 1.6-2.4.2 Torus Area (IV/lF) No intermediate break locations were postulated on the HPCI Steam line in this compartment. I.6-2.4.3 HPCI Room (II/lE) Only HPCI System components are located in this area. Any postulated HELB on the HPCI Steam Line would affect 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) and the CRD Pump in Compartment (II/lG) are qualified for the expected environmental conditions. It is possible that the postulated HELB in the HPCI Room could cause isolation of the RCIC steamline due to the resulting temperature transient in the Torus Area (IV/lF). However, the RCIC System could be restarted as soon as the high tempera-ture isolation signal resets (at approximately 187'F), and a local verification has been made of the RCIC steamline to confirm that it has not ruptured in the Torus Area. The high temperature in the Torus Area would last for only about 2 minutes and the RCIC could be manually initiated after the local verification. Even if the RHR and CS Division II pumps were lost, safe shutdown could still be accomplished using the Division I equipment with any postulated single active failure. Also, any flooding in the HPCI Area could possibly adversely affect only the 4KV cables to the RHR and CS rumps of Division II. Safe shutdown could be accomplished using the same path to safe shutdown as described above. I.6-16 REV 7 12/88
MONTICELLO I.6-2.5 RCIC ( S tean') System 3 The identified high energy piping for the RCIC System (
'a
) is the steam supply line (PS17-3"-ED) to the RCIC Turbine from the drywell penetration. This line begins in the Steam Chase (II/2F), runs down into the Torus Area (IV/lF) and into the RCIC Compartment (III/lC).
I.6-2.5.1 Main Steam Chase (II/2F)' For a postulated HELB on the RCIC Steam line, the only SSD equipment affected by e'ither pipe whip or jet impingement would be the ceiling to the compartment, in which Division II control and power cables are located, the HPCI Steam line (PS18-8"-ED) and the ESW "B" line (SW30B-3"-HF). The effects on the ceiling are bounded by effects from a postulated Main Steam line break, and so have no impact on safe shutdown. Due to the poten-tial jet orientation, the loss of the HPCI steamline, and hence, the HPCI system was assumed. Loss of both HPCI and RCIC and the ESW "B" line was discussed in the section on the HPCI Steam line breaks in the main Steam Chase (I.6-2.4.1). The path to safe shutdown would be ' the same. Flooding from the RCIC Steam line break in the main Steam Chase, assuming all steam condenses and remains
,, in the compartment, would result in approximately 1 inch of water at the bottom of the compartment. The (V) environmental conditions generated both in the Main Steam Chase and in adjoining compartments would be bounded by the effects of a postulated Primary Steam line break. Therefore, no new adverse conditions have been produced, and paths to safe shutdown described for
'the postulated Main Steam line break are applicable here.
I.6-2.5.2 Torus Area (IV/lF) The only pipe whip or jet impingement target in this compartment is the ESW "B" line (SW30B-3"-HF). HPCI would be affected because the resulting temperature transient would cause the HPCI steamline to isolate, when the HPCI temperature sensors tripped. However, the elevated temperatures would exist for approximately . 2 minutes, and the HPCI System could be initiated once the high temperature isolation resets (~187'F), and a local verification has been made that the HPCI steam-line has not ruptured in the Torus Area. Consequently, the path to safe shutdown would be the same as : described for the Main Steam Chase with HPCI available l l k,_ I.6-17 R'EV 7 12/88 L ________ ____
MONTICELLO approximately 10 minutes after the initiation of the HELB. The postulated flooding in the Torus Area from a RCIC Steam line break would be less than 1 inch in height, and no SSD equipment would be affected. The SSD equip-ment.in the Torus Area and the adjoining 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 for any anticipated pressure, temperature, and humidity environment as a result of an RCIC break in this compartment, and no additional SSD equipment is adversely affected. I.6-2.5.3 RCIC Compartment (III/1C) The only SSD components in this compartment are the RCIC equipment with the exception of the Division I RHR and Core Spray pumps' 4KV power cables. It is possible I due to the RCIC pipe geometry, to cause a jet impinge-ment en these cables. If these cables are damaged, only the Division I CS and RHR pumps, and the RCIC system are lost. The HPCI system would remain avail-able, since the resulting temperature transient in the torus area would not cause a trip of the HPCI steamline temperature sensors and subsequent isolation of the HPCI steamline. However, even if the HPCI System isolated because of a temperature transient, it could be restarted approximately 10 minutes after the initia-tion of the event after the HPCI high temperature isolation resets (~187'F), and a local verification has been made of the HPCI steamline to confirm that it has not ruptured. By using the remaining SSD equipment, with any one single failure, a path to safe shutdown exists. For the worst case, loss of the Division II Diesel Generator, the HPCI system and the SR/V's could be used until the Division I Diesel could be crcss-tied to power the Division II equipment and safe shutdown would be achieved. For flooding, compartmental pres-surization, and environmental effects only these power cables would be affected. Since a path to safe shut-down has already been demonstrated, assuming a loss of the Division I ECCS pumps, evaluation of the other affects is not required. I.6-2.6 Reactor Water Clean Up (RUCU) System The Reactor Water Cleanup high energy line (REW3-4"-ED) begins at the drywell penetration and outboard RWCU isolation valve. This line supplies reactor water to the RWCU heat exchanger and the RWCU pumps. The RUCU return line (REW6-3"-ED) returns the water to the I.6-18 REV 7 12/88
MONTICELLO reactor coolant system. All high energy lines are in the RUCU compartment (II/3D), except the return line } /~') which connects to RCIC and HPCI injection lines in the C,/ steam chase (II/2F) and is routed through the MG Set Room (V/3A) ano the northwest side of the 935'-0" of 1 the Reactor Building (II/2C). I.6-2.6.1 RUCU Area (II/3D) . Pipe whip targets consist of conduits which supply motive power to the Division II reactor sample line isolation valve, both Division II core spray outboard injection valves, an RHR containment spray valve and the Primary Containment Atmospheric Control (PCAC) isolation valves. Redundant Division I inboard containment isolation valves located inside the primary containment are available to isolate the CS and reactor sample lines for safe shutdown. The Core Spray isolation valves and the PCAC valves are normally closed. Loss of power to these valves would cause no change of state and create no adverse concerns. Je.t impingement in the RWCU compartment can target any of the above valves. Additional jet impingement targets include both Division II Core Spray injection valves, an RHR containment spray isolation valve, the Reactor Sample line and isolation valve, and the Primary Containment Atmospheric Control isolation A valves. The RWCU outboard containment isolation valve is not a jet impingement target because of pipe geometry. Redundant valves inside the containment or located out side this compartment, mitigate any concerns on loss of the above components with one exception: the loss of air supply to the PCAC isolation valves A0-2386 and A0-2387. These valves have air inflated seals, which deflate on loss of air. The loss of these seals does not create an adverse condition for a RUCU break, because the leakage across the seals is within Technical Specification limits. Therefore, the unit can be safely shutdown even with the PCAC isolation valves seals deflated. For the other valves affected, safe shutdown can be accomplished using the Division I equipment and the HPCI System. Sufficient redundancy exists to be able to shutdown the unit assuming a ! single active failure. Compartmental flooding would not be a problem, since no other SSD equipment would be adversely affected by flooding other than the SSD equipment directly affected by pipe whip or jet impingement. The flood height in the RUCU Area would be no more than 5 inches. V I.6-19 REV 7 12/88
I MONTICELLO The affects of compartmental pressurization, tempera-ture, and humidity were calculated as part of the environmental qualification analysis of Monticello (Reference I.8-11). Any SSD component affected by the RUCU line break is qualified for the postulated environment. Outsido tne PMCU compar'tment in compartment II/3D are two Divis ion II cable trays which are jet impingement targets. Should these cables be damaged, safe shutdown of the unit can be accomplished by using only Division I equipment and the HPCI System with power supplied by either diesel generator. The cable trays are approximately 5 feet from the floor and would not be affected by any flooding. In fact, no additional SSD equipment woul; be affected by any flooding from a postulated RWCU HELB in the northwest corner of the Reactor Building. The same analysis (Reference I.8- j 11), which postulated an RWCU HELB in the RWCU Room also assumed a break outside the room 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 environmental qualification purposes. No other SSD equipment is adversely affected from a postulated HELB in this area. I.6-2.6.2 MG Set Room (V/3A) Jet impingement targets are limited to the two power distribution panels for the RHR air compressors and the two wall mounted Containment Atmospheric Monitoring panels, which are not required for safe shutdown. No pipe whip targets are located in this compartment because the closest SSD component is farther than the pipe whip moment arm (Reference I.6-18). Loss of both RHR air compressor power feeds is not a concern, since the air receivers are not affected by the jet. Additionally, the RHRSW valves could be opened manually if compressed air is unavailable. With the remainder of the SSD systems operable, approximately 8 hours would be available to open the valve. With these 2 components lost and any other single failure, numerous paths to safe shutdown would still exist. Because the room has both a 6 inch curb around the room and floor drains, flooding in this compartment would j remain in the room and not affect any other SSD equip- ' ment. The same would apply to the environmental conditions; i.e., no other SSD other than the equipment I.6-20 REV 7 12/88
MONTICELLO within the MG Set Room would be affected by a break in this room. I.6-2.6.3 Reactor Building Open Area 935'-0" West Side (II/2C) [] L/ The possible pipe whip or jet impingement targets in this area are a series of Division II control cables located in the northwest corner of the Reactor Build-ing. Safe ' Shutdown would be accomplished in the same manner as that described in Section I.6-2.6.1 of this report. The flooding would not adversely affect any additional SSD components in this compartment and in the adjoining compartments (Main Steam Chase - II/2F and the CRD Pump Room - II/lG) because all components are located above the floor and the depth of the flood would be less than 2 inches. Like the compartments described in Section I.6-2.6.1, environmental qualification of equipment in this compartment is based upon maximum pressure, temperature and humidity conditions developed by a RWCU HELB in the RWCU Area and the HELB of a Main Steam Line break in the adjoining Main Steam Chase. Therefore, no additional SSD equipment is adversely affected by the environmental conditions generated by postulated RUCU HELB's. I.6-2.6.4 Main Steam Chase (II/2F) (3 There are no pipe whip targets in this area. Any other (j concerns in this area are bounded by,cther larger pipes discussed in Sections I.6-2.1, I.6-2.2, and I.6-2.4. Check valves at the RWCU injection points will prevent loss of both HPCI and RCIC flow from a break at either terminal end. Also, the check valves ensure the availability of both HPCI and RCIC for a break on the nonseismic portion of the RUCU line in the Steam Chase area. l Safe shutdown would be accomplished in the manner de- . scribed in Section I.6-2.1, with potentially either I HPCI or RCIC not available. 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 energy piping requiring further evaluation. The portion of the injection lines identified as high energy, runs from the containment penetration out to the first ; normally closed Core Spray injection valve. Line TW7- l 8"-ED is 1ccated entirely within the Reactor Water I Cleanup compartment (II/3D). The remaining Core Spray i I.6-21 REV 7 12/88
MONTICELLO lino TW11-8"-ED is located within a block wall compartment (I/3B) above the Reactor Building 962' level.- For both lines, the break locations were postulated at the terminal ends of the high energy portion of the line (i.e., at the drywell penetration and at the outboard containment isolation valve). I.6-2.7.1 Compartment Above 962' Elevation (I/3B) i The only jet impingement targets in this compartment l are the block walls surrounding Core Spray Injection j Valve MO-1753 and piping line TWil-8"-ED. A break on l this line may result in block wall missiles which could potentially impact the instrument conduits from racks C55 and C56. However, since the amount of energy released due to a HELB is limited by the inboard check valve, and since the physical separation of these l conduits and the block wall is considerable (The l conduits are above the block wall and 20' away), damage to the conduits is not considered feasible. Compartment pressurization will be insignificant due to the limited available energy in the line and the avail-able vent area for the compartment. The RHR Head Spray Injection Valve MO-2026 and the instrument lines at penetration X-29 will be subject to the environmental effects because of their location within the compart-ment. However, these components are qualified for the environment. Flooding is not a concern due to the small volume of water contained between the valves (approximately 50 gallons), and because of the available drain area. The drain oath . of water would also not adversely affect any SSD equipment. There are no pipe whip targets in the proximity of this line. I.6-2.7.2 Reactor Water Cleanup Compartment (II/3D) Break locations for the Core Spray Injection .line TW7-8"-ED are postulated at the drywell penetration and at the normally closed Core Spray injection valve MO-1754. Because of the normally clesed inboard check valve, the energy available in the piping line is limited. This postulated line break would not be as severe as a postulated RWCU line break. Any effects from a postulated Core Spray line break are, therefore, bounded by a postulated break of the RWCU line. There are no pipe whip or jet impingement targets for the postulated break locations on the CS line. I.6-22 REV 7 12/88
W MONTICELLO I.6-2.8 Residual Heat Removal System The only portion of the RHR System piping requiring (~T further evaluation are the portions of the LPCI injec-() tion lines, TW20-16"-DB and TW30-16"-DB between the outboard containment isolation valves and the primary containment penetrations. These two lines have normally closed check valves inside the containment to prevent reverse flow. 'Therefore, the energy release for these postulated line breaks is negligible because of the small volume of fluid contained between the inboard and outboard contaf ment isolation valves (approximately 170 gallons of water). I.6-2.8.1 RRR Valve Compartment (I/2G) The postulated break locations include the terminal end of line TW30-16"-DB at dry.well penetration X-13B and the terminal end at the LPCI injection- valve MO-2014. Because of the largs vent and drain areas and the limited energy release, there are no flooding or com-partment pressurization concerns. The only pipe whip target consists of a radwaste line not required for safe shutdown. The only SSD component that is a possible impingement target is the RHR shut-down cooling suction line, REW10-18"-ED. However, failure of this line will not prevent decay heat removal, since the SRV's and Core Spray are () es still available to remove the decay heat from the RPV and transfer it to the Suppression Pool. The Suppression Pool Cooling mode of RHR is unaffected and can be used to remove decay heat in the Suppression Pool. I.6-2.8.2 RHR Valve Compartment (II/2H) The postulated break locations include the terminal end of line TW20-16"-DB at drywell penetration X-13A and the terminal end at the LPCI injection valve MO-2015. l There are no additional safe shutdown components in j this compartment other than line TU20-16"-DB and valve ' MO-2015. Therefore, pipe whip, jet impingement and an adverse environment 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 j portion of the line (Approximately 160 gallons) and the J large vent and drain areas in the compartment. 1 (~T I.6-23 U 1 12/88 REV 7 l
MONTICELLO I.6-2.9- HPCI (Water) System The high energy portion of this system consists of the HPCI injection line TU3-12"-ED from the normally closed HPCI injection valve (MO-2068) to the Feedwater line. This portion is located entirely within the Steam Chase i Area (II/2F), and the effect of a pipe break on this line is bounded by the, other high energy lines pre- - viously described for this area (See Section 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 injectior. line, FW5-4"-ED, from the normally closed RCIC irjection valve (MO-2107) to the Feedwate-line. This portion is located entirely within the e Steam Chase Area (II/2F), and the effect of a pipe break on this line is bounded by.the other high energy lines previously (escribed for this area (see Sections I.6-2.1 and I.6-2.2). I.6-2.11 F .andby Liquid Control mine CH2-1 1/2"-DC from containment penetration X-42 to check valve XP-6 is not routed in the vicinity of any safe shutdown components. Two inboard check valves will limit any energy release, and because of the small volume of water contained between the inboard contain-ment . check valves and valve XP-6 (approximately 20 gallons), compartment pressurization and flooding are not a concern. The compartment (I/2B) has an extensive drain rystem and a large free volume. I.6-2.12 Off-Gas System ~ The steam supply line SHP101-4" is routed from the Condenser Bay Area through the SJAE Room. The effects of postulated breaks on this line in the Condenser Area are bounded by the postulated breaks on other lines such as Feedwater and Primary Steam (see Sections I . 6 2 . .'. and I.6-2.2). A transient analysis has been l recently completed for the SJAE Room, and the results have shown that the Main Steam Line break in the Condenser Bay is bounding for pressure, temperature, and other environmental effects. There is no SSD equipment in the SJAE Room to be affected by a pipe whip or jet impingement. Postulated breaks on line SHP101-4" are bounded by other postulated HELB's, for which paths to safe shutdown have been demonstrated (see Sections I.6-2.1 and I.6-2.2). l I.6-23A l REV 7 12/88
[, I.6-3 Table of Synte g g g o F Table I. 6-2 shows the effect of specific high energy r- x line breaks by compartment and system. This table
!'~') includes the required auxiliary systems which are considered potential HELB targets. Any system not affected by a postulated HELB is to be considered available to support safe shutdown. The meaning of the letter codes,used in the~ table are as follows:
F - Primary failure as a direct result of a line break. A - System is unaffected by a line break and is avail-able to support safe shutdown. U - The system is unavailable due to the failure of a required function or component associated with another system.
,e \.j ,e I.6-23B l ! T wJ REV 7 12/88
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