ML18018A652

Revised Mgt Capability Rept.

ML18018A652
Person / Time
Site:	Harris
Issue date:	08/04/1983
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML18018A651	List: ML18018A650 ML18018A652
References
NUDOCS 8308100368
Download: ML18018A652 (346)

Text

Carolina Power & Light Company Shearon Harris Nuclear Power Plant Mana ement Ca abilit Re ort Table of Contents Page 1 ~ 0, INTRODUCTION ., .. .. ~...

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2+0 MANAGEMENT OF ROTI¹ NUCLEAR OPERATIONS 2.1 On Site (Harris Plant Staff)............................. 2-1 2.1.1 Harris Plant Staff Organization................... 2-2 2 1 1 1

~ ~ ~ Plant Operations Unit................. 2-2 2.1.1.2 Startupoooee01001001.400000000000000.. 2-7 2.1.1 3 Administration Unit...................

~ 2-8 2~ 1 ~ 1.4 Technical Support U'nit................ 2-9 2.1.1.5 Planning and Scheduling Unit.......... 2-10 2.1 ~ 1.6 Assistant to the General Manager...... 2-10 2.1.2 Plant Staff Personnel Resources................... 2-11 2.1.3 Training.......................................... 2-17 2.1.3.1 Plant Management and Supervisory Personnel Training.............. 2-17 2.1.3 2 Shift Technical Advisor Training.. 2-18 2.1.3.3 On-shift Technical Requirements. ~ ~ 2-18 2.1.3.4 Technical Personnel Training.... 2-28 2.1.3.5 Auxiliary (Non-Licensed)

Operator Training........... 2-30 2.1.3.6 General Employee Training......... 2-30 2 '.3.7 Fire Brigade Training. ~ ~ ~ ~ ~ ... ~ ~ ~

~ 2-30 2.2 Offsite Nuclear Operations Support....................... 2-33 2.2.1 Offsite Organization.............................. 2-33 2.2.1 1 Nuclear Operations Department. ~ ~ ~ ~ ~ ~ t~~~~ 2-35 2.2.1 ~ 2 Operations and Maintenance.... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 2-36 2 '.1.3 Maintenance Support Section... 2-36 2.2.1.4 Technical Services Department. 2-37 2.2.1.5 Corporate Quality Assurance... 2-45 2 '.1.6 Engineering and Construction.. 2-52 2.2.1.7 Corporate Nuclear Safety...... 2-55 2 ' 1 ~ 8 Corporate Health Physics...... 2-58 2.2.1.9 Fuel and Materials Management. 2-60

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r Table of Contents (cont'd)

Page 2.2 2 Offsite Management Resources...................... 2-62 2.2.2.1 Power Supply Group................... ~ ~ ~ 2-62 2.2.2.1.1 Nuclear Operations Department........ ~ ~ ~ 2-62 2.2.2.1.2 Technical Services Department........ ~ ~ ~ 2-63 2.2.2.2 Fuel and Materials Management Group.. ~ ~ ~ 2-63 2' ' ' Engineering and Construction Management.......................... 2-64 2.2.2.4 Ouality Assurance and Quality Control 2-65 2.2.2.5 Corporate Nuclear Safety and Research Department.......................... 2-66 2.3 Senior Management Oversight Functions.................... 2-66 2.4 Offsite Technical Staff Resources........................ 2-68 2.5 Coordination of Interdepartmental Technical Staff Support...o...woo.woo.o.....o...oooo.o.o...woo..o 2-75 2.6 Offsite Staff Training................................... 2-75 2~7 Contract Assistance...................................... 2-76 3.0 EMERGENCY RESPONSE CAPABILITIES 3 ' Emergency Classes............................... 3-1 3' Onsite Resources and Activities............ ...... ... " 3-1 3~2~ 1 Control Room Resources................ 3-2 3.2.2 Technical Support Center (TSC) ........ 3-2 3.2.3 Operational Support Center (OSC)...... 3-2 3.2.4 Plant Emergency Procedures............ 3-2 3.2.5 Emergency Plan Training and Exercise.. 3-3 3.3 Offsite Resources and Activities......................... 3-3 3 '.1 CP&L Fmergency Response Facilities. 3-3 3 3.2 State and Local Facilities......... 3-4 3.3.3 EOF Organization................... 3-4 3.3.4 Corporate Spokesman................ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3-4 3.3.5 Offsite Contractural Assistance and Ag reements.... 3-5 3.4 Conclusion...o.............ooooooo.o.o.oo...o.

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Table of Contents (cont'd)

Page 4.0 HARRIS UNIT 1 STARTUP 4.1 Startup Organization..................................... 4-1 4.2 Harris Pre-Startup Preparations.......................... 4-3 5 ' ADDITIONAL INFORMATION 5.1 Recruiting Program....................................... 5-1 FIGURES 1 12 EXHIBITS 1 4 (7390SNP)

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CAROLINA POWER & LIGHT COMPANY SHEARON HARRIS NUCLEAR POWER PLANT MANAGEMENT CAPABILITY REPORT i+0 INTRODUCTION Over the past years, Carolina Power & Light Company's (hereinafter called the Company) nuclear capabilities have grown significantly. The Company's organizational concepts have changed to accommodate this growth and to meet specific organizational needs as they arose.

The dynamic nature of the Company's organization is in great part the result of experience, growth, and continued dedication to the improvement of the Company's nuclear programs. The Company has always been, and will continue to be, totally committed to safety and quality in the construction and operation of our nuclear facilities. The Company's nuclear program is given top priority on both a daily and long-term basis by the entire management team.

The Company has more than 12 years of experience in nuclear operations. It has safely managed the operation of H. B. Robinson Unit 2 for more than 12 years, Brunswick Unit 2 for over 8 years, and Brunswick U'nit 1 for more than 6 years.

At the present time, the Company has over 8,000 employees. Over 2,000 of these people are dedicated to the nuclear program. This includes personnel at the plants as well as the support personnel provided from corporate organizations. Of these, over 600 are in offsite organizations that provide direct support to the nuclear plants. l The enhancements that have been made to respond to the needs of the Company's increasing nuclear operations activities fall into three areas: Corporate Organizational Response and Commitment, Plant Organization, and Support Programs.

1.1 Cor orate Or anizational Res onse and Commitment One key to the successful management of the Company's nuclear program is direct management involvement. Through organizational changes and greater management involvement, the Company today provides closer management support of its Nuclear Operations activities than was formerly the case.

It is the responsibility of- corporate management to ensure that the Company's nuclear plants are designed, constructed, and operated in a safe and reliable manner. Management's awareness of current concerns that may affect nuclear safety is a critical ".esponsibility. The Company has organized a strong management program consisting of numerous feedback mechanisms such as visits, meetings, and reports. This program ensures that management personnel are fully informed of the status of the nuclear programs and items that need 7390SNP-M

attention. This program has greatly enhanced direct management involvement by allowing management personnel to interface directly with the employees who have first-hand dealings with the programs, projects, and overall operation of the plants.

The Company's nuclear activities are the direct responsibility of the Executive Vice President of Power Supply and Engineering and Construction Groups. Highlights of the Executive Vice President's supporting corporate organization are:

A Corporate Quality Assurance Department reporting directly to the Executive Vice President. All Quality Assurance/Quality Control (QA/QC) functions are consolidated under one manager. There is an on-site QA/QC Unit at each nuclear facility which reports offsite to a Corporate QA/QC Manager.

A Corporate Nuclear Safety 6 Research Department reporting directly to the Executive Vice President. This Department maintains a corporate overview of nuclear safety programs and issues and Health Physics/ALARA Programs. There is an onsite nuclear safety group at each nuclear facility which reports independently, offsite to a Corporate Nuclear Safety Manager.

All operating and maintenance, engineering, and construction activities for the Brunswick Plant are consolidated under the direction of the Vice President of Brunswick Nuclear Project who reports directly to the Executive Vice President.

A Power Supply Group, headed by a Senior Vice President responsible for managing the Nuclear Operations and Technical Services Departments. The responsibilities for nuclear operations of Harris and Robinson fall under the direction of the Vice President of Nuclear Operations. The Technical Services Department, under the direction of a manager, has expanded significantly in its support in a variety of technical-related disciplines.

An Engineering and Construction Group, headed by a Senior Vice President with Nuclear Plant Engineering and Nuclear Plant Construction Departments, each under a Vice President, who both report to the Senior Vice President.

A Fuel and Materials Management Group, headed by a Senior Vice President with Fuel Support Services under the direction of a Vice President.

This corporate organization provides a sound foundation for supporting nuclear activities at each of the plant sites.

1.2 Plant Or anizations Significant enhancements have been made to the Company's operating nuclear plants'rganizations. In addition, planning and implementation of Harris startup has received considerable attention.

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1.2.1 H. B. Robinson Plant Since the early 1970s, there have been significant changes to, and expansion of, the Robinson Plant organization. Some of the significant enhancements are:

Increased staffing and technical support in Engineering, Operations, Maintenance, and Health Physics and Chemistry.

Increased supervisory and managerial positions to maintain appropriate spans of control and management attention-to the various functional areas.

Establishment of a Planning and Scheduling Unit.

Expanded Regulatory Compliance and Administrative Support.

Establishment of OA/QC, Nuclear Safety, and Training groups on site but organizationally separated from the line organization.

Addition of an Industrial Engineering function.

1.2.2 Shearon Harris Plant The Harris Site Project organization is similar to Robinson's organization except Harris has a Manager of Plant Operations. The Harris Site management approach represents the Company's concept of more direct site control of Quality Assurance, Construction, Startup, and prospect support activities. The Company's experience at the Robinson and Brunswick Uhits provided the background needed to develop the personnel and expertise required to assume the direct control of all key functions. Particular strengths in the Harris Project management approach are:

A Quality Assurance/Quality Control organization which maintains a strong, independent reporting chain; in-house, nondestructive examination (NDE) capability; and N-Stamp Certification of Authorization in the Company's name.

P A Design Engineering group located onsite, which interfaces with the Architect/Engineer (A/E), performs some actual design and serves as a foundation for Operating Plant Engineering Support Group.

A Site Construction organization which provides construction management for the contractors hired to do the actual construction work. This is different from many prospects where the major A/E is the construction manager.

The Construction Management Team at Harris has achieved a good record for the project. In addition to the record of construction accomplished, the project has achieved an outstanding industrial safety record.

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In preparing and planning for Harris startup and operation, the Company will:

Ensure maximum utilization of Robinson and Brunswick experience.

Ensure an adequate, well-trained, motivated staff.

Plan for expected staff attrition.

Ensure that quality programs are implemented and procedures are prepared to provide for full regulatory compliance with strong emphasis on compliance with procedures and high discipline of operations.

Ensure a well-coordinated startup program.

The Company is confident that these plans, training, and programs will result in the safe and reliable startup and operation of Harris.

1.2.3 Brunswick Plant Like Robinson, the Brunswick Plant organization has also grown significantly. The basic improvements at Brunswick are the same as those previously discussed for Robinson. In addition, as discussed earlier, responsibilities have been consolidated under the Vice President of Brunswick Nuclear Project. This restructuring of the Brunswick organization improves the channel of communications between the Corporate Office and the Brunswick Pro)ect. It also provides a managerial environment that has reduced the number of external interfaces. These enhancements have lowered decision time and increased overall flexibility in the making of decisions necessary to continue the improvement of operations at Brunswick.

1.3 Su ort Pro rams The Company is justifiably proud of the accomplishments of its organization.

The following are some highlights of support programs, additional improvements, and examples of the Company's continuing success in the management of its nuclear activities.

1.3. 1 ~Tra in in Improvements in training have provided for additional simulator experience for requalification of operators. Simulator training for PPR operators from Robinson and Harris is conducted at the Harris Energy & Environmental Center (HE&EC) located near the Harris Plant. An upgraded PtNR simulator to replace the one currently in operation has been ordered and will include a radiation monitoring system and an emergency response facility information system.

Simulator training for Brunswick operators is currently being conducted at Georgia Power's Hatch facility. A BWR simulator is currently being installed at the Brunswick Plant and will be ready for training in October 1983. The HE&EC is a strong example of the Company's increased capability. The HE&EC serves as the headquarters for the Nuclear Training Section, Environmental 7390SNP-1j

Technology Section, and the Radiological and Chemical Support Section. It also provides modern metallurgical and chemical laboratories that support the nuclear plants.

1.3.2 Emer enc Pre aredness The Nuclear Regulatory Commission (NRC) recently approved the Company's emergency preparedness programs. As a part of its improvement effort, the Company established an Emergency Preparedness thit in the Technical Services Department. The Company has a Corporate Emergency Plan with supporting procedures that ensure that corporate response to a nuclear emergency is well-organized and coordinated. Lessons learned during emergency exercises at Robinson and Brunswick have allowed the Company to further refine its emergency plans and procedures.

1.3.3 Health Ph sics The Company is committed to strong Health Physics and ALARA Programs. This commitment is emphasized in a Corporate Health Physics Policy issued initially in 1977 by the Chairman of the Company. In support of this policy, ALARA Programs exist at the operating plants as well as appropriate support organizations such as Nuclear Plant Engineering and Nuclear Plant Construction projects. The Company currently employs five certified Health Physicists and four employees recently took the certification exam and are awaiting the results. A recent survey of health physicist personnel at other utilities showed the Company ranks among the leaders in certification. The Company's Health Physics and ALARA Programs have received positive reviews both from the Institute of Nuclear Power Operations(INPO) and the NRC.

1.3.4 Cresa McCormick & Pa es (Cresa ) Audit Cresap conducted an eight-month long comprehensive, corporate-wide and independent audit under the direction of the North Carolina Public Utilities Commission during 1982. As a result of this audit, Cresap commended the Company for having:

A well-organized top management team.

Participative management with a commitment to excellence and innovative change.

Sound management approach to the Harris Project incorporating lessons learned from previous generation plants as well as industry experience and competent people from a variety of sources.

Extensive and innovative formal management systems that compare very favorably with those of other utilities they had reviewed recently.

Cresap's overall assessment of the Company concluded that "CP&L is one of the best-managed utilities that we have audited in the past several years".

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1.3.5 Plant Im rovement Pro ram The Company's effort at Brunswick is a key example of the involvement and commitment of Management to the Company's nuclear program. These efforts have involved improvement in Health Physics, Maintenance, Operations, and Ihsign.

These improvement efforts are showing positive results, and the Company is confident that they will result in significantly improved performance at Brunswick in future years. Similar improvement programs and enhancements are in place at both -Robinson and Harris so that the total scope of our nuclear activities will benefit from those efforts.

1.3.6 Essex Review In April of 1980, the Company contracted with Essex Corporation to perform comprehensive human factors engineering evaluations of our nuclear plant control rooms. The primary objective was to improve our control rooms by optimizi'ng the 'man-machine interface. The evaluations were completed in September 1981 for Robinson, Brunswick, and Harris and resulted in detailed data files and summary reports that listed recommendations for improvements.

As a result of these evaluations, the Company completely redesigned the Harris Plant Control Board and rearranged the Control Room to optimize viewing angles. The recommendations for Robinson and Brunswick have been evaluated and implementation is underway.

1.3.7 Industr Involvement A further indication of the Company's commitment is the degree of its industry involvement in major industry efforts. The Company actively participates in and supports industry groups such as the Institute of Nuclear Power Operations (INPO), the ELectric Power Research Institute (EPRI), and the t4estinghouse and Boiling Water Reactor Owners'roups.

The Company's current organization ensures that the Company's other operational activites will not detract appropriate management attention from the safe operation of its nuclear plants. The details of these organizational structures and our capabilities to manage its nuclear plant operations during routine and emergency situations are discussed further in Sections 2 through 4 of this report.

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2.0 MANAGEMENT OF ROUTINE NUCLEAR OPERATIONS 2~ 1 ON SITE (HARRIS PLANT STAFF)

The Shearon Harris Nuclear Power Plant (SHNPP) is a two reactor nuclear power plant. The overall plant staff, when both units are operational, is projected to consist of approximately 700 Carolina Power & Light Company personnel.

However, since Unit 2 is not scheduled for operation until the 1990s, a functional organization has been developed to maximize the efficient operation of Unit 1, while retaining the flexibility to expand to encompass Unit 2 at a later date (see Figure 2). It is recognized that many lessons will be learned in the operation of Unit 1 that can be used to optimize the final two unit organization. For this reason, the following description of plant organization will describe the Unit 1 organization only. At the time that personnel are provided for Vhit 2, an increase to six shift Operating Supervisors and an additional full complement of operators will be added; additionally, the mechanical and electrical maintenance subunits will be duplicated to provide the necessary maintenance for Unit 2.

The organization for the SHNPP has been developed based on experience gained with the startup and operation of Robinson Unit 2 and Brunswick Units 1 and

2. The organization is designed to ensure clear lines of responsibility, authority, and communication among the various units of the organization, and to provide for clear and effective managerial and supervisory control.

Functionally, this organization is similar to the organization for our Brunswick and Robinson plants. As shown on Figure 2, the plant organization is headed by a Plant General Manager, who has the overall responsibility for plant operations. The organization is divided into five units:

Administration, Plant Operations, Planning & Scheduling, Technical Support and Startup & Test.

The Administration thit under the Manager Administration provides the staff support to the entire plant organization. in the areas of cost control, document control, material/stores control, secretarial/stenographical and clerical/mail support, security, emergency preparedness and industrial engineering.

The Plant Operations Uhit is divided into three major subunits. The Environmental and Radiation Control Subunit provides the chemical/environmental services and radiation protection services for all plant areas. The Maintenance subunit provides day-to-day corrective and preventive maintenance for the operating unit and the common support facilities such as radwaste systems, fuel storage facilities, etc. To increase efficiency and allow the personnel to become most familiar with one area of plant maintenance, this subunit has been divided into an Electrical Maintenance Group, a Mechanical Maintenance Group, a Computer Maintenance Group, and the Support Group which will perform maintenance scheduling, develop the work packages and ensure availability of necessary parts for work projects. The third subunit is Operations, which on a shift basis, operates the unit under the direction of the Operating Supervisor, seven Shift Foremen, providing a supervisor each shift and an on-shift crew of two Senior Control Operators, two Control Operators, and five Auxiliary Operators. The sixth 2-1 7390SNP

shift is provided as a training shift; one additional Shift Foreman and three Senior Reactor Operators are available to supplement any shift with additional workers to cover personnel out sick, training requirements, or other situations as needed.

The Startup and Test Lhit is responsible for the initial checkout and preoperational testing of each unit as it is completed by construction forces and prior to turnover to operational forces for initial fuel loading and startup testing.

The Planning and Scheduling Unit provides staff functions to the entire plant for planning and scheduling of maintenance and modifications during regular operation periods and outages.

The Technical Support Unit provides the staff functions to the entire plant for engineering support and regulatory compliance.

In the following sections, each unit, subunit, and the ma)or organizational structure depicted on Figure 2 is described including primary responsibilities, listing of supporting organizations, and to whom the member of the management group reports.

2ol.l Harris Plant Staff Organization The SHNPP organizational structure is based on an analysis of the kinds of key activities needed to achieve the objectives of the organization and the interrelationship of these activities. The organization is based on experience, on accommodating the Three Mile Island Lessons Learned, and on similar organizations at the Brunswick and Robinson plants The Plant General Mana er Harris Plant is responsible for all phases of plant management, including administration, operation, maintenance, and technical support. He manages and controls the organization through personal contact with the five unit heads and through written reports, meetings, conferences, and in-plant inspections. He is responsible for adherence to all requirements of the Operating License, Technical Specifications, Corporate Quality Assurance Program, and Corporate Health Physics and Nuclear Safety Policies. He is supported in these responsibilities by the Manager Plant Operations, Manager - Technical Support, Director Planning & Scheduling, Manager Administration, the Manager Startup and Test, and the Assistant to the General Manager. The General Manager reports directly to the Vice President of Nuclear Operations.

2.1.1.1 Plant 0 erations Unit All key activities that directly contribute to power production are consolidated in the Plant Operations Unit. Operations related activities are decentralized under the Manager Operations, Manager Maintenance, and Manager Environmental & Radiation Control. The Operations, Maintenance, and Environment and Radiation Control (E&RC) subunits are responsible for activities in their respective areas.

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The Mana er Plant erations manages the operation, chemistry radiation control, environmental support, and maintenance support of all operating units. He is also responsible for refueling operations. This is accomplished through a staff which includes the Manager Operations, Manager Maintenance, and Manager Environmental 6 Radiation Control. The Manager Plant Operations reports to the Plant General Manager and assumes all responsibility and authority of the Plant General Manager in his.absence.

0 erations The Mana er erations ensures the safe and efficient operation of the unit and required support facilities. He is responsible for primary and secondary system performance and the timely completion of the scheduled periodic tests, and for adherence to the requirements of the Operating License and Technical Specifications. He is also responsible for coordinating and overseeing the duties of the Operating Supervisor assigned to the plant, the Radwaste Supervisor, and the Principal Engineer Operations. He is responsible for orderly and safe operations, turnovers, and compliance with operating instructions. He is supported in these responsibilities by a staff of the Operating Supervisor, Radwaste Supervisor, Principal Engineer Operations and Engineers/Specialists. The Manager Operations reports to the Manager Plant Operations.

The 0 eratin Su ervisor Unit 1 supervises plant operations during normal day shift. He is responsible for adherence to the requirements of the Operating License and Technical Specifications. He accomplishes this through the various foremen and personnel assigned to him. The Operating Supervisor reports to the Manager - Operations.

The Shift Foremen ensure the safe, dependable, and efficient operation of their assigned shift and are the designated individuals in charge of the plant on that shift unless specifically relieved by the Operations Supervisor or his superior. They are responsible for adherence to the Operating procedures, the Operating License and Technical Specifications. It is the responsibility and authority of the Shift Foreman to maintain the broadest perspective of operational conditions affecting the safety of the plant and to keep this as the highest priority at all times when on Control Room duty. The Shift Foreman, until properly relieved, remains in the Control Room at all times during accident situations to direct the activities of Control Room Operators. He may be relieved only by qualified persons holding SRO licenses. During routine operations when the Shift Foreman is temporarily absent from the Control Room, a Senior Control Operator will be designated to assume the Control Room command function. He is supported by and supervises Senior Control Operators, Control Operators, and Auxiliary Operators. The Shift Foremen report to the Operating Supervisor.

The Radwaste Su ervisor supervises the shift operations of the Waste Processing System. This includes the working procedures and operation of the equipment necessary to generate all the process water utilized within plant systems and to ensure safe and efficient handling and storage of plant generated contaminated wastes until final disposition. He is assisted by the Radwaste Shift Foremen, Radwaste Operators, and Radwaste Auxiliary Operators. The Radwaste Operations Supervisor reports to the Manager-Operations.

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The Shift Foremen Radwaste ensure the safe, dependable, and efficient operation of the Waste Processing System. It is the responsibility and authority of the Shift Foreman Radwaste to direct the activities of the Radwaste Operators to ensure efficient handling, processing, storage, and shipment of plant generated contaminated wastes. They are supported by and supervise Radwaste Control Operators and Radwaste Auxiliary Operators. The Shift Foremen Radwaste functionally report to the Radwaste Supervisor but are under the direction of the Shift Foreman to ensure that Radwaste operations support is comparable with overall plant operations.

The Princi al En ineer 0 erations provides technical and engineering support to the plant operating personnel. He is responsible for the implementation and efficient operation of the shift technical advisor (STA) program at the plant as well as providing direct technical support in the areas of:

(1) Plant Operations; (2) Fire Protection as necessary to support safe, efficient, reliable operations; and (3) reactor core management to meet system load demands and compliance with regulatory requirements. He is assisted by Shift Technical Advisors, a Fire Protection Specialist, and a staff of engineers and technicians. The Principal Engineer - Operations reports to the Manager Operations.

The Shift Technical Advisor provides accident assessment, technical advice concerning plant safety, and 10 CFR 21 evaluations to shift operations personnel. He accomplishes this by performing engineering evaluations of plant operations, maintaining and broadening his knowledge of normal and off-normal operations, and diagnosing off-normal events. The Shift Technical Advisors report to the Principal Engineer Operations.

Shift eratin Crew The Harris Plant has six Shift Operating Crews assigned to %it 1 and later will have six additional shifts assigned to Unit 2. Each shift for Unit 1 is supervised by a Shift Foreman (SRO license), and as a minimum, is composed of two Senior Control Operators (SRO license), two Control Operators (SRO license), and four Auxiliary Operators. Each Shift Operating Crew is charged with the responsibility of operating its assigned unit in a safe and economical manner within the plant's Technical Specifications, Operating Procedures, Corporate Nuclear Safety Policy, Corporate Quality Assurance Program, Corporate Health Physics Policy, Corporate ALARA Program, and NRC and other regulatory requirements.

Four of the Shift Operating Crews work on three rotating shifts to operate Unit 1, one crew is used as a relief shift for vacationing and sick operators, and the remaining crew is in training. Each shift periodically rotates to the relief or training shift. With the rotating shifts, relief shift, and training shift, there are ample opportunities for all personnel to accomplish training and retraining without any requirements for excessive or unusual working hours. An additional seventh Shift Foreman and three additional Senior Reactor Operators are available to supplement any shift as required.

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Each Shift Operating Crew at SHNPP shall meet the following requirements:

a. When a unit has fuel in the reactor core, there shall be a Shift Foreman with an SRO license on site at all times.

b. For each nuclear power reactor with fuel in the core, there shall be a licensed operator in the control room at all times.

C~ For the reactor that is operating, there shall also be a licensed SRO in the control room at all times.

d. For each control room from which a reactor is operating, there shall be an additional licensed operator to provide relief for the control room operator and to perform duties outside the control room that need to be performed by a licensed operator.

e. For each reactor containing fuel, there shall be an auxiliary operator above the requirements of (a) through (d) above, plus an additional auxiliary operator for each control room from which a reactor is operator.

For all core alterations, there shall be a licensed SRO or SRO limited to Fuel Handling to directly supervise the core alteration. This SRO shall not be assigned any other concurrent operational duties.

ge The Shift Foreman shall be assigned only the minimal administration duties required to operate his shift.

In order to ensure that each onsite crew collectively has the requisite technical qualifications in reactor physics and control, nuclear fuel, thermal hydraulics, transient analysis, instrumentation and control, mechanical and structural engineering, radiation control and health physics, electric power, chemistry, and plant operation and maintenance, an extensive training program has been established and is described in detail in Section 2.1.3.

Maintenance The Maintenance Subunit performs all corrective and preventive maintenance on plant systems and equipment. The Mana er Maintenance is responsible for corrective and preventive maintenance for all units and common support facilities. These include ensuring that equipment, instrumentation, controls, mechanical, and electrical systems of all facilities are maintained at optimum dependability and operating efficiency. He is responsible for the coordination of these functions and for approval of working procedures and standards. He is assisted by the Mechanical Maintenance Supervisor, Electrical Maintenance Supervisor, Project Engineer - Maintenance, Project Engineer Computer, and a staff of foremen, mechanics, engineers, technicians, and specialists. The Manager Maintenance reports to the Manager Plant Operations.

The Maintenance Su ervisor - Electrical ensures that equipment, instrumentation, controls, and electrical systems of Unit 1 are maintained at optimum dependability, safety, and operating efficiency to comply with plant 2-5 7390SNP

Technical Specifications, QA, Security, Radiation Control, plant procedures, and regulatory requirements. He accomplishes this by planning, directing, and controlling a trained staff, inspecting maintenance work, providing effective maintenance procedures and standards, and developing improvements in the Preventive and Corrective Maintenance Program. He is assisted in these functions by a staff of foremen and technicians. The Maintenance Supervisor Electrical reports to the Manager Maintenance.

The Maintenance Su ervisor Mechanical ensures that mechanical systems for Unit 1 are maintained at optimum dependability, safety, and operating efficiency to comply with plant Technical Specifications, plant procedures, QA, Security, Radiation Control, and regulatory requirements. He is responsible for all required painting and pipe covering activities necessary to maintain neat, properly insulated plant systems. He accomplishes this by planning, directing, and controlling a trained staff, inspecting maintenance work, providing effective maintenance procedures and standards, and developing improvements in the Preventive and Corrective Maintenance Programs. He is assisted by a staff of foremen, mechanics, and painter/pipe coverers. The Maintenance Supervisor Mechanical reports to the Manager - Maintenance.

The Pro ect Engineer Maintenance provides technical support to plant electrical and mechanical maintenance and assisting the Manager Maintenance in assuring that plant instrumentation, control, electrical systems and mechanical systems are maintained at optimum dependability, safety, and operating efficiency, while remaining in compliance with all Technical Specifications and regulatory requirements. He is responsible for administration of the Maintenance Management System to accomplish the planning and scheduling of maintenance, ensuring parts availability, and establishing clearances necessary for preplanned work; he is assisted by a staff of engineers, specialists, technicians, and planner/analysts. The Project Engineer Maintenance reports to the Manager Maintenance.

The Pro ect En ineer Com uter provides process computer system maintenance support and technical expertise to ensure that all plant process computer systems are fully operational for the safe, reliable, and efficient operation of the plant. The Project Engineer Computer reports to the Manager-Maintenance.

Environmental and Radiation Control The Mana er Environmental & Radiation Control (E&RC) conducts the plant radiation safety and control health physics) programs, chemical control of plant reactors and systems, and the environmental programs to ensure that environmental and radiation control is maintained in a manner to protect the plant, employees, visitors, general public, and the surrounding community.

His primary responsibility is organizing, planning, and controlling E&RC resources to provide the required support while ensuring compliance with plant Technical Specifications, the ALARA concept, and all applicable state and federal regulations and permit requirements.

Some of his major responsibilities include: (1) ensuring that programs and related procedures are developed and administered to meet plant needs and regulatory requirements; (2) maintaining an awareness of current and pending regulations in the areas of radiation control, chemistry, and environmental 2-6 7390SNP

matters concerning plant operations; and (3) providing adequate documentation pertaining to individual radiation exposures, radioactive effluents, chemical control of plant systems and environmental surveillance and ensuring that these records are maintained in an up-to-date, retrievable manner. He is assisted in these functions by an Environmental & Chemistry Supervisor, a Radiation Control Supervisor, a Project Specialist Environmental and Chemistry, Project Specialist Radiation, and a staff of radiation control and chemistry specialists. The Manager Environmental & Radiation Control reports to the Manager Plant Operations.

The Environmental & Chemistr Su ervisor plans, organizes, and directs chemistry control and environmental surveillance programs, maintains laboratory procedures, test results and records, and adheres to the requirements of the Operating License and Technical Specifications. He accomplishes these responsibilities through Foremen and Technicians. The Environmental and Chemistry Supervisor reports to the Manager - Environmental

& Radiation Control.

The Radiation Control Su ervisor is responsible for the plant Radiation Control Health Physics) Program and for ensuring that all plant activities are conducted in a manner to protect the plant, employees, visitors, general public, and the surrounding community. His primary responsibility is organizing, planning, and controlling Radiation Control Subunit resources to provide the required 'support while ensuring compliance with plant Technical Specifications and all applicable state and federal regulations and permit requirements. He accomplishes this through Foremen and Radiation Control Technicians. The Radiation Control Supervisor reports to the Manager Environmental & Radiation Control.

The Pro ect S ecialist Environmental & Chemistr provides technical advise, recommendations, program enhancement and ensures that the Environmental and Chemistry Programs support efficient, reliable plant operations. He is the Environmental Chemistry technical expert for the Manager E&RC. He is supported by a staff of specialists and technicians and reports to the Manager Environmental and Radiation Control.

The Pro ect S ecialist Radiation Control provides technical advise, recommendations, program enhancement, ALARA, and ensures that the Radiation Control Programs support efficient and reliable plant operations. He is the Radiation Control technical expert for the Manager E&RC he is supported by a staff of Specialists, technicians and clerks and reports to the Manager-Environmantal and Radiation Control.

2.1.1.2 ~Starta The Mana er Startu and Test is responsible for successfully implementing and accomplishing, on schedule, the Shearon Harris Nuclear Power Plant preoperational and startup test program in accordance with the Startup Manual. He is assisted by a staff of supervisors, engineers, and technicians. The Manager Startup and Test reports to the Plant General Manager.

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The Startu Su ervisors are responsible for checking out and starting up on schedule the systems assigned in their areas in accordance with the Startup Manual and regulatory requirements. Each supervisor is assigned engineers and technicians and reports to the Manager Startup and Test.

2.1.1.3 Administration Unit The Administration Unit provides support functions such as cost control, security, office services, document control, emergency preparedness, industrial engineering, and materials requisition, storage, and issue to ensure compliance with regulations and readily available service for the proper functioning of the entire plant staff.

The Mana er Administration provides administrative support to the Plant General Manager; manages plant administrative functions, (including Budgeting and Cost Control, Document Control, Office Services, Stores, and Security);

directs Emergency Preparedness planning and activities; and provides industrial engineering support to the plant staff. He is assisted in these duties by an Administrative Supervisor, Project Engineer Industrial Engineering, Senior Specialist Security, and a Senior Specialist Emergency Preparedness. The Manager Administration reports to the Plant General Manager.

The Administrative Su ervisor manages plant cost control functions; coordination and follow-up of the budget; office services functions; document control functions of receipt, distribution, storage, and retrieval of plant records; and the requisition, receiving, storage, and issue of spare parts, equipment, and material. He is supported in these responsibilities by a Specialist Cost Control, an Office Services Supervisor, a Senior Specialist Document control, and a Materials Supervisor. The Administrative Supervisor reports to the Manager Administration.

The S ecialist Cost Control administers a plant cost control program to assist plant management in budgeting and accounting; controls plant operating and maintenance expenses in accordance with plant, company, and regulatory requirements; plans, directs, and coordinates the cost control staff in accurately recording costs and monitoring plant expenditures; assists with the preparation of Plant Construction Budget Request and the 0&M Budget; provides budget information and costs reports; and processes invoices, petty cash statements, and contractor time sheets. He is assisted by a staff of clerks and technical aides. The Specialist Cost Control reports to t'e Administrative Supervisor.

The Senior S ecialist Document Control provides a document control system which ensures that the plant QA records, procedures, drawings, technical manuals, and other documents are maintained in accordance with plant, company, and regulatory requirements. He ensures that documents can be promptly retrieved and that current documents are provided to appropriate personnel by planning and directing a competent document control staff. The staff is responsible for receiving, microfilming, reproducing, retrieving, distributing, storing, and maintaining accountability of: Plant QA records in the Storage Vault; plant drawings, technical manuals, instructions, etc., in the Plant Library; and master drawings in the Storage Vault. He is responsible for mail pickup, sorting, and distribution. He shall develop and 2-8 7390SNP

implement procedures to ensure security, accountability and retrievability of plant documents and make recommendations for future needs. He is assisted by a staff of Specialists, Technical Aides, and Clerks. The Senior Specialist Document Control reports to the Administrative Supervisor.

The Materials Su ervisor ensures that spare parts, equipment, materials, and expendable supplies are available to support the operation and maintenance of the plant by maintaining the required levels of inventory, warehouse and storeroom procedures and good housekeeping. He also ensures that proper tools are available for all required tasks and ensures that tools are properly secured. He is assisted by a staff of Inventory Control Analysts, Stores Foremen and Clerks. The Materials Supervisor reports to the Administrative Supervisor.

The Office Services Su ervisor provides secretarial/stenographic support for the entire plant staff, including stocking of office supplies for the plant staff and maintaining the personnel folders for the plant staff. He also provides clerical assistance to all plant management'or filing and reproducing paperwork. The Supervisor is assisted by the Plant Office Services Coordinator, Secretary, Stenographers, and Clerks. The Office Services Supervisor reports to the Administrative Supervisor.

The Senior S ecialist Securit develops, implements, and maintains a security program which ensures that the security of the plant is maintained in accordance with regulatory requirements. He maintains a close working relationship with local Law Enforcement Agencies to ensure compliance with existing regulations. He provides input to the Training Unit so that employees requiring access to the plant are properly trained and badged. He ensures that equipment and guards are available and in a state'f readiness.

The Senior Specialist Security is assisted by Technical Aides and a contract security guard force. The Senior Specialist Security reports to the Manager Administration.

The Pro ect En ineer Industrial En ineering develops, implements, and maintains a five-year facility development and utilization plan. He also performs various work measure and productivity studies, assists in economic studies to )ustify various pieces of equipment, determines optimum paper flow systems, and enhances the efficiency and reliability of Plant operations with various Industrial Engineering Programs. He is supported by an Industrial Engineer and reports to the Manager Administration.

The Senior S ecialist Emer enc Pre aredness develops and implements the continuing refinement of the plant Emergency Preparedness Program which ensures that a "state of readiness" is maintained at the plant to cope with any classification of emergency. He incorporates the provisions of the plant Emergency Plan in the program and revises the program and related procedures as changes are made in the plant Emergency Plan. The Senior Specialist Emergency Preparedness reports to the Manager Administration.

2.1.1.4 Technical Su ort Unit The Technical Support Unit provides engineering and regulatory compliance support for the entire plant staff both in terms of investigations of day-to-day equipment and system operation, including routine reportabilities, 2-9 7390SNP

and modification tasks to keep the plant in compliance with new regulations or to improve efficiency of operation.

The Mana er Technical Su ort develops and tests maintenance modifications and provides technical support for plant outages, plant operation, and maintenance as well as managing the plant inservice inspection (ISI) and performance programs. He is also responsible for regulatory compliance activities of the plant. He is supported by the Engineering Supervisors, Principal Engineer Support and the Director Regulatory Compliance. The Manager Technical Support reports to the Plant General Manager.

The En ineeri Su ervisors and Princi al En ineer Su ort are responsible for all plant modifications and provide technical direction and coordination for plant engineering studies. They develop and implement the inservice inspection program and plant performance programs as well as procedures, instructions, and guidelines for plant engineering functions. They are supported in these tasks by a staff of Engineers, Specialists, Engineering Technicians, and draftsmen. The Principal Engineer Support and the Engineering Supervisors report to the Manager Technical Support.

The Director R ulator Com liance coordinates regulatory activities at t'e plant to ensure that commitments, responses, records, and reports are prepared, submitted, and maintained in accordance witn regulatory requirements. He also maintains a tracking system for the resolution of all plant safety and environmental concerns. He serves as the NRC contact on site and provides the License and Technical Specification expertise necessary to support plant activities. He is assisted by a staff of Technicians and Specialists The Director Regulatory Compliance reports to the Manager Technical Support.

2.1.1 ' Plannin and Schedulin Unit The Planning and Scheduling Unit provides planning, scheduling, and coordination for long-term and short-term outages, backfit, and modification work for the entire plant. The unit plans the annual refueling outages and maintenance outages; on a year-round basis the unit plans backfit and modification work to provide the shortest downtime and/or the least interference with plant operations.

The Director Plannin and Schedulin directs the comprehensive and detailed planning, scheduling, and coordination of both long-term and short-term maintenance and refueling outages and backfit and modification work performed at the plant. He coordinates the plant activities with Nuclear Plant Engineering and Nuclear Plant Construction Ihpartment personnel to ensure that work is sequenced into the plant operations and/or the "outage windows." He is assisted in these activities by a staff of specialists and engineering technicians. The Director Planning and Scheduling reports to the Plant General Manager.

2.1.1.6 Assistant to the General Mana er The Assistant to the General Manager is responsible for assisting the General manager in the coordination, control, and centralized services for the operation of the plant. He manages special studies for the Plant General 2-10 7390SNP

manager, prepares speeches, papers and presentations for him, attends designated meetings as his representative, and reviews correspondence for accuracy and completeness. He reports directly to the Plant General Manager.

2.1.2 Plant Staff Personnel Resources The organization shown on Figure 2 for the SHNPP is designed to provide adequate numbers of personnel with the requisite qualifications and experience levels to ensure safe and proper operation and maintenance of the facility at all times, provide opportunity for development of personnel so as to ensure that adequate resources are available to cover future needs, and ensure that major functional areas are organized and staffed in a manner that will provide for a high level of management and supervisory attention.

The following Table 2.1 provides a cross-reference of the minimum training and experience qualifications for the various plant positions:

2-11 7390SNP

TABLE 2.1 2 3 4 5 6 7 8

<N "y

'b o ee e AW We o~ ~% <c a >e Py v c' e G c I ++++Q c'y

<~o+ ~e.~e e

POSITION REMARKS General Manager *Must have exper-ience and training equivalent to that required for SRO license Manager Plant X 4 *Must have exper-

. Operations ience and training equivalent to that required for SRO license.

Manager - Operations X Operations Supervisor X Shift Foreman *May be filled by academic or technical training.

Radwaste Supervisor +May be filled by chemical processing plant experience.

Shift Foreman Radwaste +May be filled by chemical processing plant experience.

Principal Engineer *Must have exper-Operations ience and training equivalent to that required for SRO license.

2-12

TABLE 2.1 2 3 4 5 6 7 0

~v pb

~ e~

Pg Y < 9 /

~O v~o 4 qr g,

e Cf POSITION 4 R1MARKS Mechanics/Electricians Operators Minimum education and exp rienc requ remen s are the c mulat e total of columns 3, 4, and 5. The years of su ervis ry ex erien e lis ed in column 6 are included in columns 4 and 5.

2-13

TABLE 2.1 3 4 5 6 7 POSITION REMARKS Manager Maintenance Maintenance Supervisor Electrical Maintenance Supervisor-Mechanical Project Engineer-Maintenance Project Engineer Computer Manager 'E&RC Environmental & Chemical Supervisor Radiation Control Supervisor Manager Startup & Test Startup Supervisor ll Manager - Technical X Support Engineering Supervisor 2-14

TABLE 2.1 2 3 4 5 6 7 8 0

P

~v 'b oc ee e gN Ne p o

o~

'v4 >a

~e >e

>c P > + G +

~e oe POSITION %MARKS Director Regulatory Compliance Manager Administration *Business Degree is an acceptable alter-nate.

Assistant to General Reports directly to Manager General Manager must have 5 years exper-ience in directly related position.

Administation Supervisor *A Business Degree i an acceptable alter-nate.

Offi,ce Services Superviso X Director Planning 6 Scheduling Engineering Specialists Foremen STAs *Experience and training equivalent to that required for SRO license.

Technicians *Must include 1 year of related technical training.

2<<15

Individuals who do not possess the formal educational requirements specified in Table 2.1 shall not be automatically eliminated where other factors provide sufficient demonstration of their abilities. These other factors will be evaluated on a case-by-case basis. Positive factors which may be considered in making this determination are as follows:

a. High School diploma or GED

b. Sixty (60) semester hours of related technical education taught at the college level (900 classroom or instructor-conducted hours)

Co Qualified by NRC as senior reactor operator at the assigned plant Four (4) years of additional experience in the area of responsibility

e. Four (4) years of supervisory or management experience Demonstrated ability to communicate clearly (verbally and in writing)

g. Successful completion of the Engineer-In-Training examination

h. Professional Engineer License

i. Associate degree in engineering or related science Four-year degree in a related field or other formal studies completed The following definitions sha1.1 also apply for Table 2.1:

Nuclear Power Plant Ex erience Experience acquired in the preoperational and startup testing activities or operation of nuclear power plants.

a~ Experience in design and construction may be considered applicable nuclear power plant experience and will be evaluated on a case-by-case basis.

b. Experience acquired at military, non-stationary or propulsion nuclear plant may qualify as equivalent on a one-for-one time basis up to a maximum of three years.

C~ Experience acquired in non-power plants such as test, training, research, or production reactors may qualify as nuclear plant

.experience on a one-for-one basis, up to a maximum of one year' credit.

'd ~ Training may qualify as nuclear power plant experience acquired in appropriate reactor simulator training programs, on if the basis of one month's training being equivalent to three month's experience, to a maximum of one year's credit.

2-16 7390SNP

e. Training programs, associated with operating nuclear power plants, the culmination of which involves actual reactor operation, may qualify as equivalent to nuclear power plant experience on a one-for-one time basis for up to a maximum of two years'redit.

On-the-job training may qualify as equivalent to nuclear power plant experience on a one-for"one time basis for up to two years'redit.

A Master's Degree may be considered equivalent to one year of professional experience, and a Doctor's Degree may be considered equivalent to two years of experience where course work related to the particular specialty is involved.

2.1.3 Tra~n1ar Plant Staff Trainin Pro ram The objective of the SHNPP training program is to develop and maintain an operating organization capable of and responsible for the safe and efficient operaion and maintenance of the plant. This training program is designed to comply with the intent of the requirements of Regulatory Guide 1.8, Revision 1, "Personnel Selection and Tr'aining." The program provides training based on individual employee experience and intended position in order to fulfillNRC licensing and personnel qualification requirements for the initial plant staff, replacement personnel, and maintenance and upgrading of plant personnel. All personnel attend certain orientation programs and specialized courses; e.g., emergency preparedness, security, health physics, and safety, in addition to receiving specialized training as required in their job skills.

2.1.3.1 Plant .Mana ement and Su ervisor Personnel Trainin I

The formal training program for the plant management and supervisory personnel provides these personnel with the qualifications necessary to assure that the plant will be operated in a safe and efficient manner. Qualifications required by ANSI/ANS 3.1, September 1979 draft, are met at the time of initial core loading or appointment to the position, whichever is later. Managers required to have acquired the equivalent training normally required for a Senior Reactor Operator's license shall participate in training programs described in ANSI/ANS 3.1, September 1979 draft, as necessary to fulfill this requirement (see Table 2-1).

Specialized training courses such as those described below are available to plant supervisory personnel.

~Chenistr A training course taught by the Nuclear Services htvision of Westinghouse Nuclear Energy Services (WNES) or equivalent. This pressurized water reactor (PWR) chemistry course provides PWR systems training and details of routine chemistry surveillance; it updates and extends knowledge in specific areas of chemistry. Topics covered in the program include the PWR, mathematics review, reactor chemistry, radiochemistry (theory), radiochemistry (laboratory procedures), and operating plant training laboratory procedures.

2-17 7390SNP

Instrumentation and Control A training course taught by the Nuclear Services Division of WNES or equivalent. This instrumentation and control engineering course provides and in-depth understanding of the instrumentation and control systems used in the Westinghouse PWR. Topics covered in the program include introduction to nuclear power plants, flux mapping system, nuclear instrumentation system, rod control system, solid state protection system, radiation monitoring system, rod position indication system, process instrumentation, and system interface.

Nuclear En ineerin A training course taught by the Nuclear Services Division of WNES or equivalent. This nuclear engineers'ourse= provides detailed information in those areas for which the Senior Engineer (Operations) is normally reponsible, as well as less detailed discussions of those areas in which he interacts with a remainder of the plant staff. Topics covered in the program include review of reactor physics and reactor systems, fuel consideration, core design, initial reactor start-up program, physics testing, measurement techniques and data reduction, power distribution analysis, plant computer, and load flow.

Mechanical Maintenance A training course taught by the Nuclear Services Division of WNES or equivalent. This maintenance management program provides familiarization with those aspects of maintenance which are significantly different from that of a fossil-fired plant. Topics covered in the program include introduction to nuclear power plants, radiation protection, nuclear power plant equipment maintenance, and maintenance management.

2.1.3.2 Shift Technical Advisor Trainin The Shift Technical Advisors will be provided with training in the following areas, as a minimum: Duties and responsibilities of the Shift Engineer; plant design and layout; accident analysis; thermohydraulics and fluid flow; integrated systems responses; and capabilities and limitation of plant instruments and control.

Additionally, Shift Technical Advisors will receive a minimum of two weeks of training on the SHNPP simulator.

2.1.3.3 On-Shift Technical Re uirements Each Shift Operating Crew will have appropriate technical training and qualifications in the areas of reactor physics and control, nuclear fuel, thermal hydraulics, transient analyis, instrumentation and control, mechanical and structural engineering, radiation control and health physics, electric power, chemistry, and plant operation and maintenance. These qualifications will be ensured by virtue of the formal Shift Operating Crew Training Programs, as detailed below.

The formal training program to license the Shift Operator Crews is made up of a series of segments which train personnel with various backgrounds.

2-18 7390SNP

Initial Nuclear Trainin Plant operations and supervisory personnel who must qualify for license examinations are caegorized by experience into the following groups:

a. Individuals with no previous nuclear experience; bo Individuals with nuclear experience at facilities not subject to licensing; C~ Individuals holding or who have held licenses at comparable facilities.

Persons in category a, above, will participate in all portions of the Licensed Operator Training Program.

Persons in category b, above, will receive training as required based upon their experience on a case-by-case basis.

Persons in category c, above, will receive onsite training to prepare them for the NRC license examination.

The training program for the Harris operating crews consists of:

a~ Basic Auxiliary Operator (AO) Program

b. Nuclear Auxiliary Operatory (NAO) Program C~ Control Operator Candidate Training Program d0 SHNPP Cold License Theory Training

e. Cold License Systems Training and Systems On-the-Job Training

f. Cold License Procedure, Theory Review, and Simulator Preparatory Training ge Cold License Simulator Training

h. Cold License Review Series and Audit i ~ Cold License Pre-License Review j ~ Other Cold License Training Required

k. Emergency Plan Training The Basic AO and Nuclear AO programs, a and b above, are presently being taught, supporting the Company's Brunswick, Harris, and Robinson plants.

Items c through j above are specifically tailored to the Harris Plant. The Shift Operating Crews will complete this program prior to initial fuel loading of Nit No. l.

Each Section of the training program and its duration is detailed below:

a~ Basic Auxiliary Operator Training Program This course consists of nine weeks of classroom training interspersed with nine weeks of structured on-the-job/plant-specific systems training. The course is designed to provide theoretical training and in-plant training to provide reinforcement of the basic science and technology or power plant operations. This course is presently available and constitutes a major portion of the training program for Operators at all of the Company's plants, both nuclear and fossil. The topics covered in the course are listed below:

2-19 7390SNP

l. Basic Power Plant Operations 2~ Essentials of Mathematics (review through algebra) 3~ Mathematics II Applications 4~ Plant Science

5. Plant Cycle

6. Plant Auxiliary Equipment

7. Plant Systems 8~ Basic PWR Plant Operation

9. Basic BWR Plant Operation 10 Basic Electricity

'ie Plant Instrumentation 12 Basic Water Chemistry

'3

~ Fuels and Combustion~

14 Boilers*

'5

~ Water Treatment

16. Turbines 17 ~ Environmental Protection Systems*

18. Instrument and Control Systems 19 ~ Power Generation

20. Electrical Systems and Equipment

21. Plant Protection

22. Gas Turbines and Diesels

• Subjects not taken by Nuclear or Radwaste designated operators.

Examinations are given regularly throughout this phase of training to monitor the trainees'rogress. Each trainee, must achieve no less than an 80 percent average grade in this course prior to entering the next phase of the training program.

b. Nuclear Auxiliary Operator Training Program This program is designed to provide those persons with little or no nuclear background with the necessary theoretical knowledge to, become proficient auxiliary operators. The program consists of approximately four weeks of formal classroom training interspersed with on-the-job training at the trainees'ssigned plants. The'opics are listed below:

1~ Math Review 21 Nuclear Theory 3 ~ Heat Transfer

4. Radiation Protection 5~ Instrumentation and Control

6. Reactor Protection C~ Control Operator Candidate Training Program This program is designed to follow the Nuclear Auxiliary Operator Course for all new operator personnel with limited or no nuclear experience in nuclear operations. The program consists of approximately ten weeks of classroom training. The topics are listed below:

l. Math Review 2~ Fluid Flow 2-20 7390SNP

3~ Nuclear Theory 4~ Reactor Theory

5. Chemistry

6. Metallurgy SHNPP Cold License Theory Training This is a formal, approximately ll-week, training program. Reviews and examinations will be given regularly to evaluate the effectiveness of the training. To successfully complete this training requires a minimum average grade of at least 80 percent. The subject areas covered by this training are listed below:

1~ Math Review 2~ Nuclear and Reactor Theory 3~ Heat Transfer, Fluid Flow, and Thermodynamics 4~ Health Physics, Radiation Protection, and Chemistry

5. Pulstar Reactor Training at N. C. State Cold License Systems Training and Systems On-The-Job-Training During this 18-week portion of the Cold License Training Program the students will gain knowledge of actual plant systems configuration and operation. This course consists of nine weeks of systems classroom training alternating with nine weeks of systems research and systems tracing (where possible).

Effectiveness of this training will be monitored through written examinations and systems checkouts. A record systems checkouts will be kept on a Harris Plant Systems Qualification Card which will be completed over the duration of the course. To maintain standardization, Systems Qualification Guidelines outlining specific knowledge required for each system have been provided to all students and training personnel. To successfully complete system training requires a minimum average grade of 80 percent for written examinations. All system checkouts must have a grade of "satisfactory." Plant systems to be covered are listed below:

'lew

~Set em ls Reactor Coolant System 2~ Reactor Vessel and Internals 3~ Steam Generator

4. Pressurizer

5. Reactor Coolant Pumps

6. Chemical and Volume Control System

7. Safety Infection System

8. Residual Heat Removal System

9. Containment Spray System 10 Containment Coolant System Auxiliary Feedwater System

12. Containment Isolation System 13 Component Cooling System

'4.

Normal and Emergency Service Water System 15 ~ Hydrogen Recombiners 16 ~ Post Accident Hydrogen Purge System 2-21 7390SNP

17. Post Accident Hydrogen Monitoring System

18. Cold Leg Accumulators 19 ~ Control Room Ventilation System

20. Fuel Handling Building Ventilation System

21. Auxiliary Building Ventilation System

22. Boron Thermal Regeneration System

23. Fuel Pool Cooling System

24. Instrument and Service Air Systems

25. Fuel Handling and Storage

26. Demineralized Water System

27. Primary Makeup System 28 Boron Recycle System

29. Fire Protection System

30. Communication System

31. Sampling System 32 Trace Heating

'3.

Main Steam System

34. Auxiliary Steam System

35. Condensate and Feedwater Systems

36. Condensate Polishers and Demin

37. Main Turbine and Generator

38. T-G Lube Oil

39. Main Turbine Sealing Steam and Exhaust

40. Generator Gas System

41. Hydrogen Seal Oil System

42. Electro Hydraulic System 43 Turbine Supervisory Control System Main Condenser Evacuation System

'4

'5.

Steam Dump System

46. Moisture Separator Reheaters and Feedwater Heaters

47. Cooling Tower

48. Ultimate Heat Sink

49. Essential Services Chilled Water System

50. Nonessential Services Chilled Water System 51 Waste Process Building Cooling Water System

'2 Circulating Water System

'3 Nuclear Instrumentation System

'4.

Reactor Protection System 55 Steam Generator Water Level Control System

'6.

Pressurizer Pressure Control System 57 Pressurizer Level Control System

'8.

Incore Instruments 59 Steam Dump Control System

'0.

Sequencer

61. Metal Impact Monitoring System 62 ~ Seismic Monitoring System

63. Rod Control System

64. Offsite Power System

65. 6.9 Kv Auxiliary System

66. 480 Volt Auxiliary System

67. 208/120 Volt AC System

68. 120 Volt Uninterruptable AC System

69. Standby AC Power Supply (Diesel)

70. DC Power System 2-22 7390SNP

71. Control Room Area Ventilation System

72. Fuel Handling Ventilation System

73. Auxiliary and Radwaste Area Ventilation System

74. Turbine Building Area Ventilation System

75. Engineered Safety Feature Ventilation System

76. Containment Ventilation System

77. Control Rod Drive Mechanism Ventilation System

78. Containment Atmosphere Purge Exhaust System

79. Diesel Generator Fuel Oil System

80. Diesel Generator Cooling Water System

81. Diesel Generator Air Starting System

82. Diesel Generator Lubrication System

83. Diesel Generator Combustion Air Intake and Exhaust

84. Diesel Engine

85. Liquid Waste Systems

86. Solid Waste Systems

87. Waste Gas System

88. Radiation Monitoring 89 Monitoring System System'ubcooled Knowledge of plant systems will be augmented by participating in procedures development, system acceptance testing, and hot functional testing.

Cold License Procedure, Theory Review, and Simulator Preparatory Training - This approximately five-week course is adminstered prior to going to the simulator. Review and examinations will be conducted regularly to evaluate training effectiveness. To successfully complete this course, the trainee must achieve a minimum grade of 80 percent. Topics covered in this course include: 1) Procedures; 2) Theory Review; 3) Mitigation of Core Damage; 4) Transient and Safety Analysis; 5) Safety and Control Systems Review; and 6) Review of Industry Events.

Cold License Simulator Training The Cold License Simulator Training Program will be approximately nine weeks in length (eight weeks minimum). The training vill include, but not be limited to: 1) control board familiarization; 2) control functions; 3) procedure usage (including Plant Emergency Procedure Implementation); 4) transient and accident analysis; and 5) control manipulations during normal, abnormal, and emergency conditions (including multiple failures). Emphasis will be placed on integrated system response under normal and emergency conditions icluding control room instrument response, diagnostics, and mitigation of core damage. During the training, shift relief will be included in order to provide experience in the areas of total plant operation and control under normal and emergency conditions in a realistic control room environment.

The training staff will monitor progress and performance during the training and instruct as required through periodic critiques. Written and operating examinations patterned after NRC licensing examinations will be administered after completion 2-23 7390SNP

of simulator training to certify cold license candidates at the Reactor Operator and/or Senior Reactor Operator level. A grade of 80 percent is the minimum required for successful completion of this course.

Cold License Review Series and Audit This portion of the Cold License Training Program will be conducted at the SHNPP site during the period between the completion of hot functional testing and the administering of NRC licensing examinations. The review series consists of approximately 2 weeks of instruction including 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> per day of classroom work with the remainder of the day being used for special instruction, plant tours, and individual study. The topics covered in this lecture series include:

1~ Reactor physics and kinetics 2~ Reactor control and protection systems 3~ Health Physics and plant chemistry

4. Technical Specifications 50 Transient, instrument failure, and accident analysis (PTS)

6. Normal and emergency operating procedures 7~ Heat transfer, fluid flow and thermodynamics The audit phase of this portion of the Cold License Training Program will consist of written and oral examinations. The purpose of this audit will be to identify any areas requiring additional training effort. Individual or group weak areas identified by this audit will be corrected by intensive training efforts for those involved and training program modifications to minimize recurrence in future classes.

Cold License Pre-Licenses Review This approximately four week phase of training is designed to improve the weak areas brought out from audits and to bring the License Candidates to a peak knowledge level for the NRC examinations. Plant procedures and subjects are listed below:

l. Procedures 2~ Theory Review 3~ Mitigating core damage 4~ Transient and safety analysis

5. Safety and control systems review

6. Review of industry events 7~ Review of plant and procedure changes since initial training

8. Simulator review Other Cold License Training Required Cold License Candidates will receive training in the following areas:

1. Pire Brigade Training 2~ Emergency Plan Training 2-24 7390SNP

3. Security Training

4. Management Training for Licenses Supervisors (for personnel requiring SRO Licenses)

a. Leadership

b. Interpersonnel Communication

c. Command Responsibility

d. Motivation of Personnel

e. Problem and Decisional Analysis

f. Administration Requirements

g. Aberrant Behavior Training

5. Training During Low Power Testing

a. Each licensed Reactor Operator and Senior Reactor Operator will participate in or observe the initiation, maintenance, and recovery from natural circulation.

6. Specific Plant Experience

7. Fuel Handling Operations Training Auxiliary Operators will participate in the Basic Auxiliary Operator Training Program and the Nuclear Auxiliary Operator Training Program as described above. This training, along with a qualification card system, will provide sufficient training and evaluation for these individuals to be become qualified Auxiliary Operators.

Re lacement and Retrainin Following initial licensing of operating personnel for the Harris Plant, an on-going training program will be utilized to maintain the proficiency of the plant operating organization after initial plant start-up. This training program will include, as described below, requalification training for licensed personnel, and training for replacement personnel. This program will be based in concept on similar programs already in effect at the Brunswick and Robinson plants.

a. Licensed Operator Requalification Training Following the initial licensing of cold license candidates, a requalification training program will be initiated to maintain and demonstrate the continued competence of all licensed personnel. This requalification training program will be conducted on an annual basis and will include pre-planned lectures, on-the-job training, and regular and continuing operator evaluation. The SHNPP simulator will be used to fulfill appropriate portions of this retraining program.

b. Lectures A minimum of six pre-planned lectures will be presented during each requalification cycle. These lectures will be scheduled throughout the year taking into account heavy vacation periods and infrequent operations such as refueling periods and forced outages. Lectures may be deferred due to unanticipated shutdowns. However, these lectures shall be conducted as soon as practicable thereafter. Content of the lectures shall take into consideration the categories as listed in 10 CFR Part 55, Apendix A: heat transfer, fluid flow, thermodynamics, mitigation of accidents involving a degraded core, operating experiences 2-25 7390SNP

from similar plants, and the results of the annual examination.

Training aids such as films, video tapes, and slides may be used and some self-study may be required in conjunction with the lectures. An instructor will present or attend as an auditor at least 50 percent of the lecture series.

All licensed individuals will be required to attend every pre-planned lecture except those specifically exempted. Exemptions will be allowed only for individuals scoring greater than 80 percent in the corresponding area on the previous examination.

c. On-the-Job Training f

The on-the-job training portion of the requalification program will consist of the following:

Licensed operators shall participate in a minimum of ten reactivity changes during each annual cycle Reactor operators shall accomplish this by manipulation of the station controls; Senior Reactor Operators will accomplish this by manipulating or directing or evaluating the manipulation of the station controls. These manipulations may consist of any of the following, providing that asterisked items are performed annually and all other items are performed on a two-year cycle:

• 1) Start-up to the point of adding heat

2) Orderly shutdown

• 3) Manual steam generator control during start-up and shutdown

4) Boration and/or dilution during power operation

• 5) Any significant (>10 percent) power changes in manual rod control

6) Turbine start-up and shutdown

• 7) Loss of coolant

a. Including significant steam generator leaks

b. Large and small including leak rate determination

c. Resulting in saturated RCS

8) Loss of instrument air

9) Loss of electrical power and/or degraded power sources

• 10) Loss of forced coolant flow/natural circulation

11) Loss of circulating water/condenser vacuum

12) Loss of service water

13) Loss of shutdown cooling

14) Loss of component cooling system or CCW to an individual component

15) Loss of normal feedwater or normal feedwater system failure

• 16) Loss of all feedwater (normal and emergency)

17) Loss of protective system channel

18) Control rod misalignment or drop

19) Inabili,ty to drive control rods

20) Conditions requiring emergency boration

21) High activity in reactor coolant

22) Turbine or generator trip 2-26 7390SNP

23) Malfunction of automatic control system(s) which affect'eactivity

24) Malfunction of CVCS system

25) Reactor Trip

26) Main steam line break (inside or outside containment)

27) Nuclear Instrumentation failure(s)

These control manipulations may be performed on the SHNPP simulator.

2~ Knowledge of Plant Systems - Individuals licensed as Reactor Operators and Senior Reactor Operators shall demonstrate an understanding of the operation of controls and equipment and shall be familiar with the operating procedures in each area for which they are licensed.

Demonstration methods may include any of the following:

a) Manipulation of the systems and their associated equipment.

b) A walk-through of the procedural steps required to start, stop, or change conditions of the system.

c) Use of the SHNPP simulator 3~ Knowledge of Facility Design, Procedures, and Facility License Changes: Licensed Reactor Operators and Senior Reactor Operators shall be made aware of safety-related facility design changes that affect station operation, operating procedure changes, and facility license changes.

Demonstration methods include any of the following:

a) Brief lectures conducted by the Operating Supervisor or other appropriate personnel.

b) Staff Meetings c) Written communications to each licensed individual from facility management Explanation of ma)or changes as part of the pre-planned lecture series 4~ Knowledge of Emergency Operating Procedures: Licensed Reactor Operators and Senior Reactor Operators shall review the contents of emergency operating procedures periodically such that knowledge of these procedures is maintained.

Demonstration methods may include any of the following:

a) Actual performance under emergency conditions b) Drills using the SHNPP simulator c) A walk-through of the procedural steps necessary to cope with the situation d) Brief lectures conducted by the Operating Supervisor or other appropriate personnel e) Self-study combined with items a) through d) above.

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Evaluation The evaluation program for licensed personnel will include the following:

a. Annual Vritten Ex amination. An annual examination based on performance feedback and needed upgrade and comparable in scope and degree of difficulty to an NRC examination shall be given to each licensed Reactor Operator and Senior Reactor Operator. The examination will contain categories of examination questions as appear on the NRC examination and appropriate to the performance feedback bases.

A grade of less than 70 percent in any category shall require accelerated requalification in that category. A grade of less than 80 percent overall requires accelerated requalification in all categories graded less than 80 percent.

b. Annual Observation and Vritten Evaluation: Observation and evaluation of the performance of licensed Reactor Operators and Senior Reactor Operators by supervisors or training staff members will include evaluation of performance during actual or simlated emergency conditions. Observation and evaluation of the performance of licensed personnel during simulated emergency conditions may be conducted by simulator training staff personnel. Discussions of actions taken or to be taken during emergency situations may be used as evaluation tools in lieu of or in addition to the above methods. Any licensed Reactor Operator or Senior Reactor Operator given an unsatisfactory overall evaluation shall require accelerated requalification.

Accelerated Re uglification Persons requiring accelerated requalification as a result of annual evaluation shall not perform licensed duties until successfully completing the program.

Accelerated requalification shall be given in the categories required or areas identified in the annual observation and written evaluation. The Director Training will tailor the scope and duration of the accelerated program to the individuals'emonstrated deficiencies. Successful completion of the program shall be measured by a reexamination, of individual categories, repeating an entire written annual examination, or reevaluation by observation or oral examination. Successful completion of an accelerated requalification program shall be by the grade criteria as defined in the evaluation paragraph above.

2.1.3.4 Technical Personnel Training Technical personnel who require specialized training to properly perform in their areas of responsibility will attend formal training courses in their particular specialties as well as receive on-the-gob training at the plant site prior to startup. This training is described below:

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Radiation Control and Test (RC) and Environmental & Chemistr (E&C) Technician

~Trainin RC and RliC Technicians will be required Co complete Che applicable I training programs described below.

a~ Basic Course Basic Series RC and E&C Technician Chemistry 2 weeks III Basic Health Physics - 2 weeks Basic Counting Room - 1 week Basic Environmental 1 week

b. Intermediate Course Series RC and E&C Technician II Intermediate Chemistry 1 week Intermediate Health Physics 1 week Intermediate Counting Room 1 week Intermediate Environmental 1 week At and above the RC and E&C Technician I level, specialized training will be provided as necessary by the Company or by vendors. Radiation Control and Environmental & Chemistry personnel will also receive on-the-job training by participating in systems checkout and start-up, preparing the laboratories for service, participating in initial radiation surveys, and participating in the writing, review, and study of radiological and chemical procedure manuals.

Instrumentation and Control (Z&C) Technician Training I&C Technicians not having the appropriate qualifications will be required to complete the applicable training programs described below prior to appointment to their respective positions.

a. Basic I&C Course Series I&C Technician III Basic Electronic Instrumentation 1 week Basic Pneumatic Instrumentation 2 weeks Basic Electromechanical Devices 1 week

b. Intermediate I&C Course Series I&C Technician II Intermediate Electronic Instrumentation

• Intermediate Pneumatic Instrumentation

• Intermediate Electronmechanical Devices *

• This series includes approximately four instructional weeks.

At and above the I&C Technician I level, specialized training will be provided as necessary by Carolina Power & Light Company or by vendors. Additionally, Instrumentation and Control Technicians will receive on-the-)ob training prior to startup by participating in checkout and testing of control. circuits, annunciator responses, computer inputs, calibration of controls and instruments, and troubleshooting various equipment problems.

Mechanic and Electrician Trainin Mechanics and Electricians will be required to complete 4~eek basic and/or 4-week intermediate courses in their respective crafts. Additionally, Mechanics and Electricians will receive on-the-job training with the equipment on the plant site. Mechanics and 2-29 7390SNP

Electricians may receive advanced or specialized training for their individual functions as necessary through attendance at Carolina Power 6 Light Company or vendor courses.

Radwaste erator Trainin Radwaste Operations personnel will be required to complete a training program in radwaste operations. The program will consist of Radiation Control 6 Chemistry courses augmented by classroom and structured on-the-job training in the areas of radwaste systems and procedures and related technical specifications. Initially, the Radwaste Operators will participate in the Basic Auxiliary Operators course. Additionally, radwaste operations personnel will receive on-the-job training in their area of responsibility through participating in system checkout and startup. A qualification card system has been developed and will be utilized by all Radwaste Operators.

2.1.3.5 Auxiliary (Non-Licensed) Operator Training Auxiliary Operators will participate in the Basic Auxiliary Operator Training Program and the Nuclear Auxiliary Operator Training Program. This training, along with a qualification card system, will provide sufficient training and evaluation for these individuals to become qualified Auxiliary Operators.

2.1.3.6 General Employee Training All permanently employed plant personnel (those assigned on a day-to-day basis) will participate in a General Employee Training Program consi,sting of, but not limited to, Radiological Health and Safety, Quality Assurance, Industrial Safety, Plant Security, Emergency Plan, Fire Protection, and other appropriate plant plans and procedures. General employee training will be provided to Company personnel at the time of employment at the plant or as soon thereafter as practicable. This training is designed to qualify personnel to be badged for unescorted entry into various parts of the operating plant and to be able to function safely and recognize problems that need to be reported within these areas. There will be annual requalification training and testing to ensure that all plant personnel remain current in the areas of plant plans and procedures.

The Nuclear Operations Department has an orientation program for all new employees that is designed to acquaint the new employee with the policies, procedures, practices of the Company and the Department. Included in this orientation are review of the "Corporate Quality Assurance Program Policy Statement," "Corporate Nuclear Safety Policy," and "Corporate Health Physics Policy." These policy statements are contained in an orientation program manual that is provided to each new employee.

2.1.3.7 Fire Brigade Training Fire Bri ade Members

a. Instruction Instructions in the topics listed below will be administered to each individual prior to assignment as a fire brigade member. The instructions will include:

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Identification of the location and types of fire hazards that could produce fires within the plant, including identification of the areas where breathing air will be required.

2~ Identification of the location of installed and portable fire fighting equipment in each area, and familiarization with the layout of the plant, including access and regress routes to each area.

3~ Proper use of available equipment, and the correct methods of fighting the following types of fire: electrical, cable and cable trays, hydrogen, flammable liquids, waste/debris, and record file.

4. Indoctrination to the plant fire fighting plan, with coverage of each individual's responsibilities.

5. Proper use of breathing, communication, lighting, and portable ventilation equipment.

6. A detailed review of procedures, with particular emphasis on what equipment must be used in particular areas.

7. A review of the latest modifications to the facility, procedures, fire fighting equipment, and fire fighting plan.

8. The proper method of fighting fires inside buildings and tunnels.

Refresher instructions will be provided to all fire brigade members on a regularly scheduled basis of not less than four sessions a year with sessions to be repeated at an interval of not more than two years. Instructions will be provided by qualified individuals knowledgeable and experienced in fighting the fires that could occur in the plant with the equipment available at the plant. Special instructions will be provided for fire brigade leaders in directing and coordinating fire fighting activities.

b. Practice Sessions Practice sessions will be held for fire brigade members to teach them the proper method of fighting various types of fires and to provide them with practice in extinguishing actual fires. These sessions will be conducted at facilities sufficiently remote from the nuclear plant so as not to endanger safety-related equipment, with the sessions provided at regular intervals not exceeding 1 year. These practice sessions will be conducted requiring fire brigade members to don protective equipment, including emergency breathing apparatus.

c. Drills Drills will be performed in the plant so that the fire brigade will remain proficient in fire fighting techniques. These drills will include:

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The simulated use of equipment for the various situations and types of fires which could reasonably occur in each safety-related area.

20 Conformance, where possible, to the established plant fire fighting plans.

30 Operation of fire fighting equipment, where practical, including self-contained breathing apparatus, communication equipment, and portable and installed ventilation equipment.

Drills will be performed at regular intervals, not to exceed three months, for each fire brigade to allow members of the brigade to train as a team. At least one drill per year for each fire brigade will be unannounced to determine the fire readiness of the plant fire brigade and plant fire protection systems and equipment. Drills will be planned to establish training objectives and will be critiqued to determine how well the training objectives were met. This critique will, as a minimum, assess: fire alarm effectiveness; response time; selection, placement and use of equipment; the fire brigade chief's direction of the fire fighting effort; and each fire brigade member's response to the emergency.

A drill will be held annually at which offsite fire department participation will be requested.

Other Plant Em lo ees

a. Instruction for All Non-Fire Brigade Members Once a year all employees will be instructed on the fire protection plan, evacuation routes, and procedures for reporting a fire. Security personnel will be instructed in entry procedures for offsite fire departments, crowd control for people exiting the stations, and procedures for reporting potential fire hazards observed when touring the facility. Instruction will also be given to all shift personnel who will assist the fire brigade in the event of a fire. Temporary employees will be given instructions to familiarize them with the plant' evacuation signals, evacuation routes, and procedures for reporting fires.

b. Drills A plant evacuation drill will be performed annually.

Fire Protection Staff Fire protection staff members will be introduced to a program of specialized training. Instructions for the staff will include:

Analysis of building layout and system design with respect to fire protection requirement, including consideration of potential hazards associated with postulated design basis fires.

bo Design and maintenance of fire detection, suppression, and extinguishing systems.

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Co Fire protection techniques and procedures.

d. Training in manual firefighting techniques and procedures for plant personnel and the fire brigade.

Offsite Fire De artments In accordance with commitments for the use of offsite fire departments, the training offered these offsite fire fighting personnel will include courses in basic radiation principles and practices and. emergency response. Additional training will be offered to familiarize them with typical radiation hazards that may be encountered when fighting fires at a nuclear power plant.

Construction Personnel Training for construction personnel will include instructions in reporting fires, responding to alarms, and locating evacuation routes.

Initial Trainin The initial fire protection training program will be completed prior to receipt of fuel at the site. The Emergency Plan implementing procedures for fire protection will be completed at least three months prior to receipt of fuel. Sufficient fire protection drills will be performed immediately prior to fuel receipt to provide assurance that the plant staff is adequately trained to cope with fire-related emergencies.

2+2 OFFSITE NUCLEAR OPERATIONS SUPPORT 2.2.1 Offsite Or anization The Company's nuclear plants are supported by an extensive offsite organization that provides expertise in a variety of areas. For the most part, the offsite organizations are structured to focus nuclear activities within separate departmental and organizational structures. This philosophy ensures that the Company's other, nonnuclear activities will not detract appropriate management attention from safe nuclear operation. As shown on Figure 4, the offsite Corporate support for nuclear operations is managed by Mr. E. E. Utley, Executive Vice President Power Supply and Engineering &

Construction Groups who reports to Mr. Sherwood H. Smith, Jr.,

President/Chairman. Reporting to Mr. Utley are four officers and a department manager whose organizations further subdivide offsite technical and managerial support into five areas: Mr. L. W. Eury, Senior Vice President Power Supply Group; Mr. M. A. McDuffie, Senior Vice President Engineering & Construction Group; Mr. J. M. Davis, Jr., Senior Vice President Fuel & Materials Management Group; Dr. T. S. Elleman, Vice President Corporate Nuclear Safety and Research Department; and Mr. H. R. Banks, Manager Corporate Quality Assurance. Mr. P. W. Howe, Vice President Brunswick Nuclear Project, also reports to Mr. Utley; he is located at the Brunswick site.

The Corporate Nuclear Safety and Research Department conducts the independent safety review of nuclear operations, and the corporate health physics program and assessment. The Vice President Corporate Nuclear Safety & Research also 2-33 7390SNP

provides the Senior Management oversight function by formal feedback mechanisms with Senior Management up to and including the President/Chairman.

The Corporate Quality Assurance (COA) Department is responsible for quality assurance (QA) and quality control (QC) within the Company, including engineering, construction, and operations. In addition, the CQA Department is responsible for QA audit functions. This department was formed in early 1981 to provide more efficient and effective QA/QC within the Company by consolidating the QA/QC functions that had been under the Technical Services, Nuclear Safety and Research, and the Nuclear Operation Department.

Conducting independent nuclear safety, health physics assessment, and quality assurance oversight functions in organizations separate from the Power Supply Group and Engineering 6 Construction Group provides effective mechanisms for strong independent assessments at the Senior Management level of how well these programs are working. These offsite technical resources and Senior Management oversight mechanisms are described in detail in Sections 2.3 and 2.4.

The major offsite support organization for nuclear operations is the Power Supply Group, which includes the Nuclear Operations Department support staff, the operating and maintenance support staff of the Fossil Operations Department, the Technical Services Department, and the Maintenance Support Section.

The Power Supply Group provides offsite technical and managerial support to the nuclear plants in the areas of environmental and radiation control, emergency planning, training and retraining, industrial security, maintenance support and planning, plant chemistry and radiochemistry, regulatory compliance, outside contractural supervision, operational management, environmental monitoring, meteorological and seismic monitoring, and nuclear licensing activities.

The concentration of the above listed offsite power supply activities in organizational groups separate from the nuclear safety, quality assurance, and engineering and construction support activities, provides a strong, independent management focus on safe and reliable nuclear operations support. Offsite management in the Power Supply Group is organized to minimize conflict with the application of other resources applied to other nuclear and nonnuclear activities within the. Company. It includes a manageable scope of support activities which is not diluted by the need for management attention to engineering and construction activities and corporate nuclear safety and quality assurance assessment, which is managed by others who report to an equal senior management level.

The Fuel and Materials Management Group provides offsite technical and managerial support in the areas of nuclear fuel procurement, refueling operations support, and plant procurement support.

The Engineering & Construction Group provides a source of offsite technical resources to assist and support the operating plants in areas of civil design, instrumentation and controls, computers, mechanical, electrical, nuclear engineering, metallurgical analysis, and construction.

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Figure 5 outlines the offsite support organization of the Power Supply Group, Fuel and Materials Management Group, and the Corporate Quality Assurance Department. The offsite organization for the Corporate Nuclear Safety and Research Department is shown on Figure 8. The Engineering and Construction Group is shown on Figure 6.

The Company has established clear lines of responsibility and effective interface mechanisms between the Engineering 6 Construction Group, the Power Supply Group, and the fuel and Materials Management group which ensures that independent management attention to each of these support activities complements and strengthens the total attention to offsite support of nuclear operations.

Resources of each of the units or sections shown are available to the plants or other units as needed. Routine lines of communication exist between the personnel of each offsite unit and their appropriate counterparts at each nuclear plant. Support of each nuclear plant is the function of the o'ffsite organization, and assistance requested by the plants is routinely provided without a requirement for formal management action. In the event that high level coordination is required for some prospects, the Section managers maintain frequent communication, and the Vice President Nuclear Operations and/or the Vice President - Brunswick Nuclear Project is fully informed.

2.2 ~ 1 ~ 1 Nuclear 0 erations De artment The Nuclear Operations Department operates within the guidelines of the Corporate Policy Statements on Radiation Protection and Nuclear Safety, and the Corporate Quality Assurance Manual. The Vice President Nuclear Operations is responsible for all aspects of operation and maintenance of the Harris and 'Robinson nuclear plants.

Some of the Vice President's functional responsibilities include:

aO The establishment and approval of the qualification requirements for all plant staff positions and the personal review of the qualification of specific personnel for managerial and supervisory positions.

b. Review and concurrence of plant programs established to implement requirements in radiation protection, industrial security, quality assurance, fire protection, training, operations, and maintenance.

C~ Review of the regulatory compliance record of the nuclear power, plant including NRC Inspection and Enforcement inspection reports (IER), Licensee Event Report (LER) performance history, and review of correspondence with the NRC Office of Nuclear Reactor Regulation and the Office of Inspection and Enforcement concerning operating nuclear power plants.

'd ~ Review of reports from the various independent company audit and review groups, such as Corporate Nuclear Safety or Corporate Quality Assurance. In order to stay in touch with plant performance and personnel, the Vice President Nuclear 2-35 7390SNP

Operations communicates with the Plant General Manager on a daily basis and visits each plant on a frequent basis, normally at least once a month. In addition, members of Senior Management visit each plant periodically.

Administrative and Technical Su ort Section This section is headed up by the Assistant to the Vice President Nuclear Operations and provides support to the Vice President Nuclear Operations in all functional areas. This section also provides support to the plant managers of the Harris and Robinson nuclear plants for prospects involving more than one plant. This section is composed of the Technical Support %it and the Administrative Support Unit.

2.2.1.2 0 eration and Maintenance Section Generation Operation and Maintenance Section, Fossil Operations Department provides staff engineering to all generating plants and the Vice Presidents of Nuclear Operations, Brunswick Nuclear Project, Fossil Operations and Technical Support in the areas of: 1) operations and maintenance 2) plant efficiency and reliability and 3) fossil fuel analysis and performance. This support also takes the form of assisting the plants in the development of CPM charts or other scheduling methods to ensure a coordinated outage program. By providing available statistics of manpower and time utilization from previous maintenance operations, the section assists in the proper scheduling of maintenance.

The continuing analysis of current maintenance problems permits the improvement of maintenance methods and cost reductions. Current methods are compared with methods used by other utilities or with manufacturers recommendations, and suggestions for improvement of maintenance methods are recommended to the plant. The accumulation and analysis of data on plant availability and equipment reliability is used to assist plant operating staffs in developing and implementing programs to improve plant reliability.

2~2~ 1 ~ 3 Maintenance Su ort Section The Maintenance Support Section provides support to the maintenance programs at the Company's operating power generating plants. These functions include coordinating the scheduling of generating equipment outages with the System Operations Department and providing maintenance manpower and technical support to generating plants during outages and other periods of high maintenance activity. The Maintenance Support Section was created in 1982 and reports to the Senior Vice President Power Supply.

The outage coordinating function consists of an operations engineer who obtains plant equipment outage requirements on long-term, shore-term, and emergency bases. These requirements are used to coordinate with System Operations to establish a long-term outage plan, a short-term outage schedule, and to provide for emergency outages as economically as possible while meeting the system power requirements.

The maintenance manpower function consists of approximately 200 craftsmen, 17 foremen, and six technical specialist. The crews and specialists are home 2-36 7390SNP

based at three strategic locations, the Northern, Eastern, and Southern areas of the Company system, and report to an area superintendent. Each crew is equipped with a van-type trailer which is moved to the plant site at which the crew is assigned and contains foreman's office, crews'ockers, work benches, and special tools.

Each area superintendent maintains close communication with,the maintenance supervisors of the generating plants in his respective area in order to stay cognizant of their requirements for maintenance support. The s'chedule for crews and specialists to work at various plants is approved by the Section Manager and is based on input from the area superintendents and operating plants with special emphasis on criticality of work and priority of the unit for system generation. Personnel from all three areas work at any plant as needed to provide as much of the outside requirements for manpower, supervision, and technical support as possible. Maintenance Support craftsmen receive the same craft training as nuclear plant craftsmen.

During 1983, the Maintenance Support Section will assume the responsibility, for coordination of contractors'anpower utilized by generating plants.

2.2.1.4 Technical Services De artment The Technical Services Department provides Operations support in the areas of environmental and radiation control, plant chemistry and radiochemistry, licensing, health physics, environmental technology, emergency preparedness, and training.

The Department was moved from the Engineering & Construction Group to the Power Supply Group in early 1981. In order to enable the Vice President Nuclear Operations to devote more time to plant operations, many of the operating support staff functions were transferred from the Nuclear Operations Department to the Technical Services Department. The Technical Services Department contains: The Licensing & Permits Section, the Environmental Technology Section, the Nuclear Training Section, the Radiological and Chemical Support Section, and the Emergency Preparedness Unit. The responsibilities of each of these groups are discussed below, and the department is structured as shown on Figure 5.

Licensin & Permits Section The Licensing & Permits Section acts as the Company's interface with the NRC Offices of Nuclear Reactor Regulation and Inspection and Enforcement, and provides special technical expertise in the areas of seismology and meteorology. The section is organized into units with the following function responsibilities:

a. Nuclear Licensin Unit The Nuclear Licensing Unit is responsible for coordination of all Office of Nuclear Reactor Regulation (ONRR) activities affecting the Company's five nuclear units This includes the coordination and preparation of responses to all ONRR activities, and the preparation of license amendment requests and licensing documents such as the Harris FSAR. It is responsible for the maintenance of operating licenses and revisions to the Technical 2-37 7390SNP

Specifications, updating of FSARs; it advises-management on critical licensing issues and ensures that all incoming NRC correspondence is routed properly and responses are prepared to accurately address licensing issues.

b. S ecial Nuclear Pro rams Unit The Special Nuclear Programs unit is responsible for coordination of generic licensing issues. This includes coordination and preparation of responses to generic Office of Nuclear Reactor Regulation (ONRR) activities affecting the Company's five nuclear units. In addition, Special Nuclear Programs coordinates the Company's involvement in industry organizations including INPO, AIF, EEI and EPRI. This unit also participates in the various owner's groups. Special Nuclear Programs also supports other special projects of a technical or regulatory nature as required.

c. Permits Uhit The Permits Unit establishes and operates the Harris seismic monitoring program and the Harris, Brunswick, and Robinson meteorological data collection programs and coordinates efforts to obtain the National Pollutant Discharge Elimination System (NPDES) permit and any federal, state, and local permits not required by the NRC.

Environmental Technolo Section The Environmental Technology Section conducts the Company's environmental monitoring assessments and conducts plant analytical chemistry and metallurgical laboratory services for the Company at the Harris Energy &

Environmental Center (HE&EC) in New Hill, North Carolina. The Analytical Chemistry, Air Quality, Biology, and Metallurgy laboratories are capable of providing an array of services and technical support to generating plants, engineering activities, quality assurance, and construction programs within the Company. In addition, one subunit of the Biology Unit is located at BSEP.

a~ Anal tical Chemistr Laborator The Analytical Chemistry Laboratory (ACL) provides chemical technical support for the Company's programs through a broad spectrum of activities. Specific capabilities and other features are listed below:

1 Water Analysis 2~ Tissue Analysis 30 Transformer Oil Analysis

4. Lubricating and Hydraulic Fluid Analysis 50 Metal Analysis

6. Air Quality

7. Special Analysis Capabilities

b. Metallur Laborator The Metallurgy Laboratory at the Harris Energy & Environmental Center is equipped to provide metallurgical analysis, mechanical analysis, and spectrochemical analysis for three ma)or 2-38 7390SNP

purposes: Failure analysis, materials evaluations, and quality assurance and control. The Metallurgy Laboratory has been certified by the Construction Quality Assurance Section as a supplier of laboratory services in accordance with the nuclear and nonnuclear sections of the American Society of Mechanical Engineers'ASME) Boiler and Pressure Vessel Code. Testing in accordance with the requirements of the American Welding Society (AWS), the American Society for Testing Materials (ASTM), and the American National Standards Institute (ANSI), etc., is also available.

C~ Biolo Laboratories The majority of the work performed by these laboratories is in support of environmental programs required by regulatory agencies related to licenses and permits for construction and operation of various facilities of the Company and to respond to plant personnel requests regarding biofouling and emergencies relating to biological organisms.

The staffs of these laboratories have expertise in various disciplines of ecological sciences, including aquatic biology (phytoplankton, zooplankton, benthos, and freshwater and marine

'fisheries biology), toxicology, bioassay studies, wildlife biology, and botany. Technical support includes investigations and recommendations regarding fish kills, biofouling and flow studies in circulating water systems, fish diversion devices around intake structures, fishery management programs, weed growth problems in cooling lakes and ponds, wildlife and aquatic insect pest problems, erosion control, threatened or endangered species, water quality assessment, and bioassay studies to allow onsite investigation of plant effluents.

Nuclear Traini Section The Company has established a comprehensive training program to provide required training for. both licensed and unlicensed nuclear plant operations personnel such as Auxiliary Operators, Control Operators, Senior Control Operators, Shift Foreman, Shift Supervisors, and Shift Technical Advisors.

'Qe program also provides for the training of other craft and technical personnel such as Radiation Control, Environmental and Chemistry Technicians, Instrumentation and Control Technicians (I&C), and Mechanics. The training program makes use of a combination of offsite training conducted at the HE&EC combined'ith on-the-job training at appropriate generating plants. Off-system training such as vendor training programs is utilized where appropriate. Within the Technical Services Department, the training programs is coordinated by the Nuclear Training Section under the direction of the Manager Nuclear Training.

The Manager Nuclear Training provides staff support to the Nuclear Operations Department in the areas of Operations, Technical and Craft Training and the operation of the simulator and other training facilities at the HE&EC, and at the operating nuclear plants. The primary purpose of the Nuclear Training Section is to assure that we have highly qualified personnel 2-39 7390SNP

available to, maintain and operate the nuclear generating plants in a safe and efficient manner. These responsibilities and services are provided from an organization consisting of eight units which support nuclear power plants:

The Nuclear and Simulator Training Unit, the Fossil Operator Training Unit, the Craft Technical Training lhit, the Administrative thit, and the Curriculum Development Unit at the HE&EC; and the Robinson Training Unit, the Brunswick Training thit, and Harris Nuclear Power Plant Training Unit located at their respective nuclear plants.

The Nuclear Training Section units at the HE&EC are responsible for continuing support of present and future department needs, assisting plants in determining training needs and performing task analysis, monitoring regulatory and audit requirements, evaluating and providing simulator training, and providing support to plants in areas such as instructor training, evaluation, and consultants. The 'training director at SHNPP is responsible for training conducted for plant employees served by the Nuclear Training Section. Plant and Corporate line management will render their support to the execution of that responsibility. The training director works closely with all other plant management personnel to identify, develop, and "present specialty training for plant employees. The Training Director develops, coordinates, and arranges for the presentation to Company and contractors'mployees various training and indoctrination programs, i.e., security and health physics. The plant training director is also responsible for maintenance of training documentation and records for all plant personnel.

Formal training work sessions are held at least quarterly to ensure timely and deliberate interface and control of assignments between the plant training supervisors and HE&EC training section personnel to further coordination, communications, and consistency in all training-related matters, and the timely implementation of the respective responsibilities an assigned tasks.

In addition to the quarterly training work sessions, a Training Evaluation &

Planning Committee has been established. This committee is charged with evaluating existing training programs and planning future programs. It provides an opportunity for personnel from the plants and from the HE&EC to jointly participate. The committees'valuations are broad in scope and include operation and administration of training programs, suitability of programs to meet regulatory and/or legal requirements, program effectiveness, or any other aspect of training that affects plant operation or overall training efficiency. The planning activities of the committee focus on the general program mix that will be appropriate to meet training and retraining needs in the next year. Additionally, the committee is charged to promote and assure consistency with regard to qualification programs for personnel in Operations, Radiation Control, Environmental. and Chemistry, I&C, and other Craft and Technical disciplines; to improve communications and share ideas in regard to training; and to monitor the effort to staff while maintaining a qualified staff at each of the other operating nuclear plants. The Manager Nuclear Training is the Chairman of this committee.

The Nuclear Training Section staff works to maintain academic quality in all training programs by providing qualified instructors from its own staff or, when necessary, contracting for instruction services with universities and technical colleges or other vendors. For example, in the fall of 1980 local 2-40 7390SNP

universities were contracted to provide credit courses at an operating nuclear plant in the areas of mathematics and nuclear engineering; also, a local university was contracted to teach mathematics at the HE&EC to students in our Nuclear Auxiliary Operator Training Program.

The Nuclear Training Section includes the following units:

a. Nuclear Simulator Trainin Unit The Nuclear and Simulator Training Wit is responsible for providing both classroom and simulator training at the HE&EC for nuclear plant operations personnel. This training is coordinated with that portion of the overall training conducted by the training supervisors at each nuclear plant. The unit's responsibilities include providing training for auxiliary operators, simulator certification for RO and SRO license candidates, STA simulator training, and simulator retraining for plant operations personnel.

The Pressurized Water Reactor (PWR) simulator at the HE&EC, an integral part of the Company's operator training program, was declared operational in February 1978. It has been continually improved and updated since that time. During 1979, model changes were made to improve simulation of the Three Mile Island incident and the computer capacity was increased. A major modification to the Radiation Monitoring System of the simulator is scheduled for completion by the end of 1983. A new SHNPP simulator has been ordered from Westinghouse which incorporates the human factors engineering studies and has enhanced modeling. This simulator is scheduled to be operational by the end of 1985. The Training Center at the HE&EC was constructed with adequate space for three power plant simulators and their associated computers. The Company accepted delivery on a Boiling Water Reactor (BWR) simulator in July, 1983, in order to further improve operator training. This simulator is undergoing installation and checkout and should be operational by October, 1983. This new simulator simulates Brunswick Unit 2.

Operator training is a combination of classroom, in-plant, and simulator training which requires approximately 30 months for licensing an individual with no prior experience and approximately 18 months for a person with considerable experience such as ex-Navy Nuclear operators. The in-plant training is provided at the plant for which the individual will be licensed.

A portion of the classroom and all PWR simulator training is provided at the HE&EC; BWR simulator training will continue to be provided at vendor simulators until the completion of the BWR simulator in October of 1983. The Nuclear & Simulator Training Vhit coordinates the operator training program, ensuring that on-the-job, in-plant, classroom, and simulator training are welded into a cohesive program.

b. Craft Technical Traini Unit The Craft Technical Training Unit is responsible for the training of both craft and technical personnel. Major responsibilities in support of nuclear generating plants include the initial, specialized, and advanced training for personnel in Radiation Control, Environmental and Chemistry, I&C, mechanical, electrical, and radwaste operator disciplines. This unit has already provided training for personnel working on the Harris Safety Analysis Report.

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There are laboratory and shop facilities as well as classrooms in use by this unit to provide both practical theory and hands-on instruction. The Instrumentation and Control Training Laboratory has equipment such as oscilloscopes, dead-weight testers, power plant instrumentation, and hand tools. The Health Physics and Chemistry Laboratory contains equipment such as dose-rate meters, air samplers, balances, sulfur analyzer, and neutron detectors. A Craft Training Building housing equipment such as welding machines, lathes, radial drills, motor control trainers, electric power supply, hand tools; and a hydraulic crane has been completed and is in use.

There are separate laboratories for mechanical, welding, and electrical training as well as three classrooms and support areas. These facilities are designed to provide well qualified craftsmen with the required skill for all "the power plants.

Course series for all crafts are divided into three levels basic, intermediate, and specialized/advanced. In general, the basic and intermediate series for each craft are four to six weeks in duration. The specialized/advanced training is normally given after the completion of basic and intermediate training, and is accomplished "as-needed," as defined by the plant training supervisors and training sections. Vendor schools are used whenever appropriate. The general subject areas for each craft are:

Radiation Control-Chemistry, Health Physics, Environmental Counting Room

2. ISC-Pneumatics, Electronics, Electrical 3~ Radwaste Operators-Auxiliary Boilers, Off-Gas, Chemical/Liquid Treatment, Solid Waste, Radwaste Management

4. Mechanics-Welding, General Shop, Plant Mechanical Systems

5. Electricians-Practical Theory, Wiring Practices and Electrical Distribution, Plant Electrical Systems, Electromechanical Controls, Motors, and Generators

c. Curriculum Develo ment Unit The Curriculum Development Unit is responsible for providing assistance in the development of educational programs and means for evaluating the effectiveness of programs offered. This unit also provides administrative support in long-range planning, projection of training requirements, scheduling of students to attend training programs, and in the development of training aids. Of key and equal importance is the unit's participation in task analysis, the performance of employee follow-up evaluations after completion of training, and assistance to advisory committees on craft training course content.

The unit has conducted task analyses for each of the sections'raft and technical training programs in order to validate these courses of study. To further the depth and to provide third-party input into the contracts with university faculty in such 'areas as thermodynamics sections'rograms, and hydraulics, private consultants with extensive operator training and 2-42 7390SNP

evaluation expertise, and liaison with the community colleges and technical institutes are actively pursued. Maintenance of a number of active contracts with outside contractors ensures continuing access to qualified personnel when and if they are needed to supplement the unit's own resources.

Radiolo ical and Chemical Su ort Section The Manager of the Radiological and Chemical Support Section, reporting to the Manager Technical Services, provides staff support in the areas of health physics, chemistry, and environmental activities and for the effective operation of the environmental dosimetry and chemistry laboratory. The Radiological and Chemical Support Section (R&CSS) has responsibilities identified in the Corporate Emergency Plan to provide health physics and environmental support to the nuclear plants in the event of an accident.

These responsibilities and services are provided from an organization consisting of three units, each headed by a principal specialist.

a. Health Ph sics Unit The Health Physics thit provides support and services to the plants in the area of Health Physics and Radiation Piotection. This support includes advice and guidance in implementing and maintaining the ALARA program evaluation of the plant health physics programs, assistance in developing and implementing health physics procedures, assistance in selecting and procuring monitoring equipment, technical support to assist in evaluating and resolving health physics related problems, and providing professional and technical support for ma)or outage activities.

This unit is responsible for administering the personnel monitoring program.

This responsibility includes reading of all TLDs and maintaining Company records on radiation exposure. The Health Physics Units is also responsible for maintaining the Radiation Control and Protection Manual which defines how radiation protection procedures and programs are implemented at our nuclear facilities.

The Health Physics Kit also provides support in the development of the emergency response/recovery organization to support the Cor orate Emer enc Plan. This includes onsite support'uring emergency situations in the areas of environmental assessment and dose calculations and progections.

ln carrying out these responsibilities, personnel from the Health Physics thit work closely with cognizant personnel on the plant staffs as well as the Corporate Health Physics Section of the Corporate Nuclear Safety & Research Department. They maintain an awareness of current and proposed regulations and work closely with plant and Department management to ensure that requirements are understood and effectively implemented. Periodic meetings are held with the plant Environmental and Radiation Control Supervisors for the purpose of reviewing programs and problems related to health physics and approaches for resolution of those problems.

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bi Environmental Unit The Environmental Unit provides support to the plants through the Radiological Environmental Lab, located at the Harris Energy & Environmental Center. The Radiological Environmental Lab is well equipped with laboratory equipment such as a Low Beta Counting System, Liquid Scintillation Detection System, Ge Li System, Beta Gamma Counting System, and a ND 4400 Counting system. 'Ibis equipment is used by this unit to provide radiological environmental surveillance services to the operating plants and preoperational surveillance for the Harris Plant. This unit performs radiochemical analyses of environmental monitoring samples, prepares and distributes routine and special radiological environmental monitoring reports, and plans and implements radiological environmental monitoring programs for new plants.

The Environmental Unit maintains a close working relationship with each plant by making routine visits to each facility. These visits are made to monitor plant programs and procedures and to assist plants with problems. One example of this type of program is the manner in which the unit assists the plant in maintaining the radwaste system by reviewing plant programs and procedures and making recommendations for improving systems operations. In addition to the permanent staff, this unit also utilizes support from various consultants.

This unit is also responsible for assisting the plants with the implementation and maintenance of programs relating to the Radiological Effluent Technical Specifications.

In order to maintain a high standard of excellence, the Environmental Lab participates in Quality Assurance Programs in conjunction with State and Federal agencies. This program consists of analysis of spiked samples issued by the agencies, or the analysis of split samples and the comparison of laboratory results. A laboratory quality assurance program is also conducted by the Company, which consists of the analyses of unknown samples and interlaboratory comparison of results.

The Environmental Unit maintains a cognizance of existing and proposed regulations and coordinates closely with plants and Departmental management to ensure that regulations are understood and effectively implemented.

The Chemistry lhit provides staff chemistry support to the operating nuclear and fossil plants and operates the HE&EC radiochemistry lab in support of the plants.

The Chemistry Kit staff assists the plants in developing chemistry programs, procedures, and specifications. Continuous review of the plant chemistry programs is maintained through review of weekly and/or. monthly chemistry data and reports from each plant. These reports are also transmitted to consultants who are experts on the various plants, and the are coordinated by the Chemistry Unit staff. The staff periodically consultants'eviews visits the plants to discuss the reviews of chemistry programs and procedures, 2-44 7390SNP

to inspect the plant for chemistry conditions, or to coordinate recommendations from the staff and/or consultants. The Chemistry Unit staff also maintains industry contact through various organizations such as EPRI, the EEI Chemistry Committee, and ASTM.

The radiochemistry lab provides services to the plants by performing chemistry analyses which may be beyond the capability of the plants or are special one-time studies for which the plants are not equipped or staffed. Examples are radwaste evaporator bottoms studies, quality assurance checks on bulk chemicals specifications, and chemical cleaning solution analyses.

The radiochemistry lab conducts a round robin program where standard unknown solutions are sent to the plant labs for analysis. The results are analyzed by the Chemistry Lab Supervisor and returned to the plants with recommendations for improvements in procedures or techniques. This program includes samples for standard chemical analysis and samples for radioisotope analysis. This serves as an important quality assurance check for the plant labs.

Emer enc Pre aredness Unit The Emergency Preparedness Unit of the Technical Services Department is responsible for: Directing and coordinating Corporate Emergency Planning to ensure regulatory compliance; assessing the readiness of all CP&L emergency plans and programs; serving as interface with regulatory agencies on emergency preparedness matters; providing emergency preparedness support for CP&L nuclear plants; maintaining training qualifications of plant personnel in emergency response; testing emergency preparedness by preparing and conducting exercises; ensuring the availability and operational readiness of emergency facilities, equipment, and supplies; developing dam failure emergency plans for the hydro plants; and for providing coordination with Federal, State, and local agencies.

2.2.1 ' Cor orate Oualit Assurance De artment The Corporate Quality Assurance Department was formed in 1981 to consolidate the quality assurance, quality control, and audit functions which were previously performed separately for engineering and construction activities, operations activities, and corporate quality assurance audit activities. Each nuclear plant site now has on-site OA/OC staff to oversee OA/OC activities for engineering, construction and operations. The Section Manager responsible for QA/QC activities at the Brunswick and Robinson plants is located in the Corporate Offices with on-site Directors reporting to him. The Section Manager responsible for QA/QC activities at the Harris Plant Site was recently relocated from the Corporate Offices to the Harris Plant. Those functions that are not plant specific, such as vendor surveillance, auditing training, administrative support and engineering support for the Corporate Office, were consolidated under the Manager, Quality Assurance Services Section. These Section Managers report to the Manager, Corporate Quality Assurance Department, who reports directly to the Executive Vice President Power Supply and Engineering & Construction. In this manner the Corporate OA Department Manager oversees the OA/OC activities of both the Power Supply and the Engineering & Construction Groups while maintaining independence from any responsibilities within those groups.

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The Corporate Quality Assurance Department is organized into three functional entities: QA/QC Harris Plant, QA/QC Brunswick and Robinson Plants, and Quality Assurance Services, as described below.

A/ C Harris Plant Section The QA/QC Harris plant section under the direction of the Manager, has the responsibility for assuring proper application of quality codes, standards, practices and procedures throughout engineering, construction, startup and operations of SHNPP. The Manager QA/QC Harris Plant and staff are responsible for:

aO Reviewing design specifications, and procurement documents to ensure inclusion of QA/QC requirements. Revisions to these documents which alter QA/QC requirements will also be reviewed.

b. Reviewing the Corporate Quality Assurance Program and propose revisions, as appropriate.

C~ Assuring timely resolution of concerns and identified nonconformances.

d. Providing construction site QA/QC inspection activities with the exception of those inspection activities delegated to the Site Manager as defined under his responsibility. Inspections performed by the Construction Inspection group will be monitored by on-site QA/QC personnel.

e. Reviewing field purchase documents for inclusion of QA/QC requirements.

Reviewing applicable construction procedures for compliance with specific QA/QC requirements.

go Training, qualification, and certification of Nondestructive examination (NDE) personnel.

h Providing NDE procedures.

Providing, as necessary, Level III expertise which includes interpretation of test data.

Reviewing applicable Contractor NDE procedures involving code class work.

k. Maintaining Radiation Isotope License issued by the State of North Carolina.

Providing stop work authority to:

(a) the Principal QA Engineer (b) the erector QA/QC, Harris Plant (c) the Principal QA/QC Specialist NDE 2-46 7390SNP

m. Stopping maintenance or modification work which does not meet requirements'.

Reviewing modification and plant maintenance documents and plant procedures and instructions, as defined to operations, to assure that quality requirements are adequately prescribed.

0~ Ensuring holdpoints have been inserted in work control documents and conducting inspections and witness points for maintenance and modification of the plant.

po Coordinating/conducting surveillance of ongoing plant activities, reporting results to the appropriate Plant Supervisor and following up to assure that timely corrective action is taken, when appropriate.

q. Providing QA/QC services at the plant ~

r. Reporting quality-related problems for correction.

s~ Verifying acceptability of items and conditions by means of inspections, examinations, or tests.

Providing guidance or check lists for accumulation of documentary evidence or quality and other QA records for retention.

The Director A/QC Harris Plant, who reports to the Manager OA/QC Harris Plant, conducts Harris Plant QA/QC activities in accordance with the Corporate Quality Assurance Program, the Company ASME OA Manual, and QA/QC procedures. He supervises the QA/QC staff assigned to the site and has stop work authority.

The Princi al OA En ineer, who reports to the Manager OA/OC Harris Plant, is responsible for implementation of the onsite Harris Plant OA engineering program to ensure design, construction and modifications meet applicable regulations and codes.

The Princi al OA/OC S ecialist NDE, who reports to the Manager OA/OC Harris Plant, is responsible for implementation of NDE QA procedures at the Harris Plant and other Company projects.

A/ C Brunswick 6 Robinson Plant Section The QA/QC Brunswick and Robinson Plants Section, under direction of the Manager, assures proper application of quality standards, practices, and procedures associated with plant operations, maintenance, and modification.

The Manager QA/QC Brunswick and Robinson Plants and staff are responsible for:

aO Providing OA/QC services at the plant.

b. Reporting quality-related problems for correction.

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Co Stopping maintenance or modification work which does not meet requirements and documenting this action.

d. Reviewing modification and plant maintenance documents, the Plant Operating Manual, and other plant procedures and instructions, as appropriate, to assure that quality requirements are adequately prescribed.

e. Ensuring holdpoints have been inserted in work control documents and conducting inspections and witness points for maintenance and modification of the plant.

Verifying acceptability of items and conditions by means of inspections, examinations, or tests.

Providing guidance or check lists for accumulation of documentary evidence of quality and other OA records for retention.

h. Coordinating/conducting surveillance of on-going plant activities, reporting results to the appropriate Plant Supervisor and following up to assurance that timely corrective action is taken, when appropriate.

Providing procedures or instructions necessary for the accomplishment of QA/QC activities.

Reviewing purchase requisitions and ensuring that OA/OC requirements are specified, except when reviewed by Ouality Assurance Services.

k. Reviewing Nuclear operations generated Architect-Engineer contracts and NSSS inquiries and contracts to ensure inclusion of necessary QA/QC requirements.

For Nuclear Operations generated contracts, maintaining liaison with Architect Engineer (A/E) and NSSS Supplier in accordance with this Corporate Quality Assurance Program to keep up to date on QA/QC activities and status, and to assure timely resolution of quality-related problems.

Reviewing site-generated design specifications and procurement documents to ensure inclusion of QA/QC requirements. Revisions to these documents which alter QA/QC requirements will also be reviewed.

n. Reviewing the Corporate QA Program and proposing revisions, as appropriate.

The Director OA/OC thit at Brunswick and Robinson Plants, reporting to the Manager QA/QC Brunswick and Robinson Plants, is the Senior QA/OC representative at the plant. He is responsible for conducting site OA/OC activities in accordance with the Corporate Quality Assurance Program and QA/QC procedures. He supervises the OA/OC staff assigned to the the plant and has stop work authority.

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The Qualit Assurance Services Section, under direction of the Manager, assures proper application of quality standards, practices, and procedures during engineering, construction, operation'nd modification of nuclear power plants. The Manager Quality Assurance Services and staff are responsible for:

a ~ Reviewing A/E and NSSS inquiries and contracts to ensure inclusion of necessary QA/QC requirements.

b. Maintaining liaison with the A/E and NSSS Supplier in accordance with this Corporate Quality Assurance Program to keep up to date on prospect QA/QC activities and status, and to a assure timely resolution of quality-related problems.

Co Reviewing design specifications and procurements documents to ensure inclusion of QA/QC requirements. Revisions to these documents which alter QA/QC requirements will also be reviewed.

d0 Reviewing the Corporate Quality Assurance Program and propose revisions, as appropriate.

e. Assuring timely resolution of conerns and identified nonconformances.

Establishing the qualification of Vendor and Contractor Quality Assurance programs by conducting facility surveys, when required. The actual function of conducting these surveys may be delegated to others, such as the A/E. When this is the case, OA Services Section will monitor and may participate in surveys.

go Establishing, maintaining, and issuing an Approved Suppliers List. This will include the evaluation and determination for periodic audits.

h~ Conducting inspections and product acceptance activities (shop inspections) at Vendor facilities. The actual function of conducting these inspections may be delegated to others, such as the architect engineer. When this is the case, QA Services Sect'ion will monitor and may participate in inspections.

Maintaining liaison with the field QA/QC representative of the A/E and NSSS Supplier and the Company Site Manager (or Construction Manager, if contracted services are used) to assure prompt interchange of quality-related problems. This function includes setting up definite communication lines Conducting an independent Corporate Quality Assurance Audit Program. Auditors shall have no responsibility for quality achievement nor for quality assurance other than auditing. They shall be training in preparing, conducting, reporting, and following-up of audits to assure timely corrective action of conditions, practices and items that could degrade quality.

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k. Maintaining QA/QC procedures for corporate and/or field use, including document control and coordination of preparation of revisions.

Maintaining the Corporate Quality Assurance Manual, including document control Distributing and maintaining controlled documents, as required.

n. Assisting other Corporate QA Department Sections in developing, implementing, and maintaining training programs to quality and upgrade QA/QC personnel.

o. Maintaining current codes and standards.

po Controlling issuance of code and standards and their update or revision.

Providing company interpretation of codes and standards, when required.

Reviewing applicable general office procedures for compliance with QA/QC requirements.

The Princi al A S ecialist Performance Evaluation, who reports to the Manager Quality Assurance Services, implements the independent Corporate Quality Assurance Audit Program for nuclear activities at operating plants and construction projects and performs periodic audits of vendors. This audit program is conducted in accordance with the Corporate Quality Assurance Program, the Company ASME Quality Assurance Manual, the Operating Plant Technical Specifications, QA/QC procedures, and authorized commitments by the Company. Responsibilities include:

aO Implementing the Corporate Quality Assurance Program of periodic audits.

b. Ensuring timely reporting of Quality Assurance Audits to audited management, and to other upper level and designated management.

Co Providing special investigation teams to investigate, evaluate and report QA/QC problems or violations as requested.

Evaluating and analyzing internal, NRC, and other audit reports for trends and corrective action to prevent recurrence.

The Princi al OA S ecialist Trainin & Administration, who reports to the Manager Quality Assurance Services, is responsible for:

Developing and maintaining the Corporate Quality Assurance Program Manual including commitments to Regulatory Guides and ANSI standards.

b., Maintaining current codes and standards, controlling issuance, and providing company interpretation when requested.

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Co* Assisting other Corporate QA Department Sections in developing, implementing and maintaining training programs to qualify and upgrade QA/QC personnel.

'd ~ Determining, developing and upgrading standardized procedures to be utilized at construction and operating sites. Preparing and maintaining procedures for corporate and field use.

e. Distributing and maintaining controlled documents, as required.

Preparing and maintaining departmental procedures which support corporate and department policies and guidelines.

g. Performing administrative duties, as assigned.

The Princi al ualit Assurance En ineer, who reports to the Manager Quality Assurance Services, is responsible for:

a~ Implementation of the engineering quality assurance program for nuclear projects to ensure power plant design, construction and modifications meet applicable regulations and codes.

b. Reviewing engineering and procurement documents such as specifications, contracts, and purchase requisitions for quality assurance requirements.

C~ Reviewing applicable general office procedures for compliance with QA/QC requirements.

d. Reviewing A/E and NSSS inquiries and contracts to ensure inclusion of necessary QA/QC requirements.

e. Performing qualification audits of A/E and NSSS Suppliers.

Performing periodic evaluations and determining need for periodic audits of A/E and NSSS Suppliers.

The Princi al Vendor Surveillance S ecialist, who reports to the Manager >>

Quality Assurance Services, is responsible for:

a~ Establishing the qualification of Vendor and Contractor Quality Assurance programs by conducting facility surveys, when required. The actual function of conducting these surveys may be delegated to others, such as the A/E. When this is the case, Vendor Surveillance personnel will monitor and may participate in surveys.

b. Establishing, maintaining, and issuing an Approved Suppliers List. This will include the evaluation and determination for periodic audits.

C~ Conducting inspections and item acceptance activities (shop inspections) at Vendor facilities for procurement and ensuring timely resolution of identified concerns and findings.

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d. Evaluating Vendor's corrective action to prevent recurrence of nonconformance identified during shop inspections.

e.. For procurement by the A/E, the actual functions of conducting shop inspection's and evaluating corrective action are performed by the A/E. The Vendor Surveillance Kit will monitor and may participate in these inspections.

When requested, conducting and/or participating in audits of quality-related activities of Vendors for compliance with the approved QA program as specified by contract.

g. Preparing and maintaining Vendor Surveillance QA procedures.

It is intended that performance of certain QC inspection functions will be assigned to the Nuclear Plant Construction Department. When such assignments are made, they will be mutually agreed upon and documented by the Vice President Nuclear Plant Construction and the Manager Corporate Quality Assurance. For those inspection activities performed by the Nuclear Plant Construction Department, the Director QA/QC Harris Plant shall perform a monitoring function to ensure that requirements of this program are satisfied.

2.2.1.6 En ineerin and Construction

'he Company's nuclear project management experience began with Robinson.

Robinson Unit 2 was a "turn key" project, whereby the initial responsibility for design and construction rested with the prime contractor, Westinghouse Electric Corporation, and the Company's participation in construction and site quality assurance was limited. The Company did, however, take active roles in the area of licensing, startup testing, and operation, and retained ultimate responsibility for the entire plant.

This approach proved to be sound for that period as shown by the fact that Robinson has been one of the most reliable nuclear power plants constructed in the whited States. Acceptable proposals for turn key plants diminished, however, and the need for more direct control by the Company in future nuclear construction became apparent. Consequently, at Brunswick, the Company assumed greater engineering and construction management responsibilities. For example, separate contracts were awarded for engineering and construction with the Company assuming more responsibility for coordination of the total pro)ect than it had at Robinson.

As Brunswick progressed, we realized that in the future even more direct management responsibility for engineering and construction of our nuclear plants would be essential. This resulted in a change in the Company's construction management policies and led to our creating an organization which could assure direct control of construction management activities for Harris. Our experience during Robinson and Brunswick construction provided the background we needed to develop the personnel and the corporate management expertise required for the Harris Project. Specifically, for the Harris 2-52 7390SNP

project we engaged an A/E (Ebasco) for design. The Company provides the construction management for the contractors hired to do the actual construction work.- In this case, Ihniel Construction Company has the prime contract for construction of the power block and associated facilities.

As construction manager, the Company is responsible for job coordination and communication, planning, cost control, inspection, quality assurance, accounting, warehousing, procurement, field engineering, milestone scheduling, and establishing and monitoring the master schedule. The master construction schedule is expanded by the Company into a critical path network by use of PMS-IV, a computerized project management tool. This exercise of construction management by th'e Company aids in making the actual determination as to the rate and sequence of construction, as well as the determination as to which portions of work will be accomplished by Daniel and which portions are better handled by separate contracts let by the Company. In addition, the Company retains and exercises authority to approve or disapprove prime contractor recommendations on construction methods and force levels, provides the communications link between the designer (Ebasco) and builder (Daniel and others), and controls site delivery dates. By retaining these responsibilities, the real burden of construction management remains with the Company rather than being delegated to a constructor, as was our past

, practice.

The above experience support the contention that we have the management resources necessary to safely design and construct Harris Unit 2, while providing the required technical backup resources for Harris U'nit 1 operations.

En ineerin & Construction Grou Technical Staff Resources The Engineering and Construction Group supports nuclear operations in the areas listed on Figure 4.

En ineerin Resources Engineering support of the Harris Project is provided by the Nuclear Plant Engineering Department which is organized into two sections: The Harris Plant Engineering Section and the Engineering Support, Nuclear Plants Section.

The Harris Plant Engineering Section is responsible for providing the design and engineering for the Shearon Harris Nuclear Power Plant, including engineering support of site activities. This includes: Evaluation and preparation of responses to the NRC I&E Bulletins and Orders; engineering of plant modifications; plant system problem resolution; and expertise in the nuclear, civil, instrumentation and controls, electrical, and mechanical disciplines. The section generates engineering documents, manages the contract for A/E service, provides direction for project design, performs appropriate design activities, provides resolution of field problems, and manages the procurement of engineered equipment.

This Section is organized into three units: Mechanical, Electrical, and Civil. Each of these units is headed by a Principal Engineer. The units are organized into subunits to handle specialized areas of each discipline such as power and I&C, radwaste, nuclear steam supply, and hangers and restraints.

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Additional support and 'performance organizations provide interface and schedule control.

The Harris Plant Engineering Section is located at the Harris Site and works on a day-to-day basis with the site construction organization, site startup organization, and Harris Plant operations organizations to assure overall success of the project. The on site Company staff is supplemented by A/E and contract engineers working under the direction of, and in accordance with our QA and Design Control program.

The Engineering Support, Nuclear Plant Section provides engineering support for the Company's operating nuclear plants. The Section is organized into three units: a Mechanical (Systems) Unit, a Mechanical (Components) thit, and an Electrical Unit. These units are under the direction of Principal Engineers. In addition, the section has the capability and is experienced in adding contract engineers and designers to supplement its forces, working under Department supervision and to Company procedures, as well as utilizing outside A/E organizations.

These units are located in the General Office, but spend substantial time in the field in support of operating plant modification.

The Vice President - Nuclear Plant Engineering also has on his staff a Director Nuclear Engineering Safety Review, who is responsible for assuring that operations and engineering feedback on both internally and externally generated nuclear plant safety issues are incorporated into new plant design and into modifications to operating plants. He assures that a program is in place and implemented to ensure that Departmental personnel are trained in their responsibilities to meet NRC regulations, codes, standards, and other commitments made by the Company, and in particular that ALARA considerations are factored into all design activities.

En ineerin Su ort for Harris Unit 1 Startu and 0 erations Engineering Support of the testing, startup and operation of Harris is provided by the Nuclear Plant Engineering Department through the Harris Plant Section located at the Harris Site. 'ngineering During the testing, startup and operation of Harris, the Harris Plant Engineering Section will maintain its discipline structure. The responsibility of this organization will be to provide engineering modifications and design configuration control for the operating unit. The Harris Plant 'Engineering Section will produce detail design modifications required by the plant. The section will focus on generation and maintenance of design documents (drawings, specifications, design basis documents, etc.). Technical support will be provided to the operations organization as required in areas such as spare parts, Q list equipment, equipment qualification, etc. Harris Plant Engineering personnel will be available to participate in the review of plant operating, maintenance and surveillance procedures, as requested. A major benefit of this process will be that the same technical staff that administered the design of the Harris, Plant during construction of the plant will be responsible for the technical support of plant operations.

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Construction Su ort Resources The Nuclear Plant Construction Department has the responsibility for construction of all nuclear generating facilities and associated systems required for the production of nuclear power by the Company and for providing assistance as requested on operating plant modification. Currently, there are three sections which are staffed and operational: The Harris Site Management Section, the Robinson Construction Management Section, and the Construction Procurement and Contracting Section.

The Project General Manager of the Harris Site Management Section is responsible for construction management of the Harris site and for the control over the constructor and contractors of the plant site. He supervises professionals in the engineering support, inspection, cost control. and project accounting areas. He is responsible for the review of design drawings and specifications to ensure ease of construction. He is also responsible for the administration of contracts, the coordination of Company-owned tools and equipment, participation in construction methods, selection planning, and direct supervision and inspection of the constructor and contractors.

The Robinson Construction Manager is responsible for construction and engineering for the Robinson generation facilities as required'o complete project assignments in a manner which minimizes impact on plant operations, keeps management informed, and meets all requirements. Site staffs have been established at the Robinson Plant to perform construction work assigned to the Nuclear Plant Construction Department. These staffs are termed the Robinson Construction Management Section. The work being performed by this section includes major modifications or additions to the existing plants which are beyond the scope that can be conveniently handled by the plant organizations and which are specifically assigned by the Nuclear Operations Department.

The Robinson Construction Management Section manages, administers, supervises, and inspects both firm-price and reimbursable contractors performing modifications and additions for the respective plants. Activities include review of design drawings to ensure ease of construction, implementation of the construction ALARA program, cost control, schedule, and coordination and direction of contractor activities. This section participates in selection of construction methods, planning and scheduling, providing engineering support, and in direct supervision and inspection over several contractors.

The Manager Construction Procurement and Contracting provides all procurement and contracting activities required to support the completion of construction project assignments. He supervises professionals in 'providing both firm-price and reimbursable contracts, site procurement, expediting, and construction equipment and tool management. Site procurement staffs have been established at the Harris, Robinson, and Brunswick Plants to perform procurement of materials and equipment to support construction assigned to the Nuclear Plant Construction Department.

2.2.1.7 Cor orate Nuclear Safet (CNS) Section The CNS Section monitors the Company's operating nuclear plants to assure the associated nuclear safety programs are carried out in an effective manner.

Independent Review (IR) and the Independent Safety Engineering Group (ISEG) 2-55 7390SNP

functions make up the primary areas of responsibility. The CNS organization consists of over 30 professionals allocated among four units; one offsite in the General Office and three at the nuclear plants, one at each site.

The offsite unit (Nuclear Safety Review) is responsible for the IR program as well as providing general evaluations of safety related systems.

The CNS independent review activity addresses the following:

a~ Procedure and changes meeting 10CFR50.59 review criteria

b. Licensing actions C~ Test or experiments not described in the facility FSAR

d. Plant operational occurrences (LERs)

e. Regulatory violations (IE Reports)

f. Technical Specification changes ge Plant Nuclear Safety Committee (PNSC) meeting minutes

h. Any item deemed appropriate for review relative to safe operations Required reviews are processed in the manner described below:

Safety related items are evaluated by the responsible members of the respective plant staff. If the item contains an unreviewed safety question, Technical Specification change, FSAR change, or is deemed safety-significant by the Plant General Manager, it is forwarded to the CNS Section for independent review. Upon receipt, the item is logged in and sent to the NSR Director. Depending on the extent of the item and the disciplines or areas involved, the package is as'signed to one or more of the Project Engineers in the IR Subunit. Considerations to be included in the review are specified along with the time frame in which the review is to be completed'ith respect to both time and detail, the Project Engineer has a significant amount of latitude in carrying out- the assignment. Sometimes additional details that need to be considered are uncovered in the review, or a given review may produce unforeseen complexities that require more time than first estimated. In these instances the scope and time are adjusted as necessary with the concurrence of the Principal Engineer. Once the assignment is finished, the comments are documented and the package is sent to the Principal Engineer for concurrence. The Principal Engineer evaluates the package, and if satisfied, sends it to the Director Nuclear Safety Review for final approval and filing. If not satisfied, the Principal Engineer returns the package to the reviewer with specific comments that need resolution before approval. For each item reviewed at least three signatures of qualified individuals are required to show that the items has been adequately evaluated. The final signature is normally the Director Nuclear Safety Review. In all cases where Technical Specification changes are submitted to the NRC or where a modification or test constitutes an unreviewed safety question, formal approval must be obtained from CNS prior to implementation'n the case of modifications which do not constitute unreviewed safety questions, an approval memorandum is required, but the modification work can proceed before receipt. Formal approvals are issued under the signature of the Manager CNS.

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If a specific item .cannot be fully closed out, yet the open aspect does not constitutes an immediate safety problem, a "follow-up item" may be issued. In these cases the review item is- closed out and a specific follow-up,document is initiated to trace and eventually close out the long-term open issue. These follow-up items may be generated by any CNS fhit, are serialized, updated, and closed out in a time frame consistent with their priority. The respective Unit Director reviews the status of outstanding items to assure they are being handled in a timely manner.

Most items are resolved via direct contact between the CNS Engineers and the appropriate individuals of the plant staff. However, if such efforts are not successful and it is determined that further action is required to enhance plant safety, a formal concern or recommendation is issued. A recommendation results from a decision that a specific course of action is desired to improve nuclear safety margins. A concern is initiated when the course of action is unspecified but an identified deficiency requires recertification to improve nuclear safety margins. Formal correspondence describing the concern or recommendation is initiated and sent to the appropriate Vice President in the nuclear operations organization for resolution. Target dates for resolution and final corrective action are established consistent with the safety implications. If the problem is of immediate safety concern, it is verbally communicated to the Plant General Manager and respective Vice President for prompt resolution. Additional reports to Senior Management are discussed in Section 2.3.

A significant number of the items reviewed in the IR program are Licensee Event Reports (LERs). These LERs are examined for safety implications, accuracy of room cause identification, and adequacy of corrective action to prevent recurrence. Repetitive events are examined closely for safety significance and proper corrective actions. Trends of LERs relative to equipment, procedure, and personnel deficiencies are also examined to highlight areas which may needs attention.

The second major responsibility of the NSR Unit is the evaluation of plant safety-related systems to assess the overall performance. This activity is carried out by gathering and compiling data generated by tests, modifications, and repairs of the system; conducting interviews with operators; reviewing industry practice; and integrating this information 'into an overall performance summary. Periodic reports are issued providing the nuclear operations personnel with an outline of the evaluation, conclusions, and any appropriate recommendations and/or concerns.

The NSR Unit also monitors unresolved safety issues and is developing capabilities in the transient analysis area. The main thrust of these programs is to act as the primary technical contact on key genesis issues affecting CPRL nuclear plant operations and gaining the ability in-house to thoroughly evaluate and resolve issues insofar as practicable On-site Nuclear Safety lhits fulfill the role of the ISEG as outlined in NUREG-0737. These units are located at each CP&L nuclear plant site and have a relatively high degree of flexibility in carrying out their tasks. Major functions included in the ISEG role are:

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a0 Operating experience feedback program bo Procedures and modification reviews Co Evaluation of transients and safety system challenges d~ Direct observation of plant activities

e. Special investigations The operating experience feedback function consists of screening assigned documents for applicability to the plant. Documents such as plant reports, NRC issuances, Company reports generated external to the plant, and industry assessments of operating experience are included in the system. The ONS Unit determines if an item is applicable and then routes it to the cognizant plant personnel with specified action to be taken. A document control system is used to assure proper action.

As opposed to the 10 CPR 50.59 review conducted by the NSR Lhit on selected procedures and modifications, the task carried out by ONS under this heading is directed toward an "on-line" technical adequacy check. Emphasis is placed on identifying programmatic deficiencies and establishing a good feedback loop to ensure. corrections are applied'o the procedure/modification development process. The same focus is placed on the assessment of plant transients and safety system challenges, i.e., "real time" assessment that thoroughly address the initiating events and apply the lessons learned.

In the routine conduct of CNS Section activities, informal contacts are required throughout the Operations Groups of the Company. The Section's members are unconstrained in these contacts and have the organizational freedom, authority, and independence to contact any person regarding safety matters. Other than members of the plant organizations, frequent contacts are made with the support functions in the Nuclear Operations and the Engineering and Construction Departments.

2.2.1.8 Cor orate Health Ph sics The Corporate Health Physics Section, under the supervision of the Nanager-Corporate Health Physics, consists of personnel with education and/or work experience in fields of radiation hygiene or health physics. The section is also responsible for the formulation and recommendation of corporate level health physics policies and programs, evaluation of health physics programs and recommending any needed improvements and modifications, and providing health physics expertise throughout the Company.

The Corporate Health Physics Section formulates and recommends corporate level health physics policies and programs, evaluates existing health physics programs and recommends any needed improvements to these programs, provides support to the licensing and corporate nuclear safety activities of the Company, development and distribution of the Corporate ALARA Program, and periodic assessment of various ALARA programs developed to comply with the Corporate ALARA Program.

The section staff has the flexibility to review all aspects of the Company's health physics programs including staffing, training, procedures, equipment, management support, etc. The staff conducts a range of activities to accomplish this review including assignments at operating nuclear plants to assist in health physics activities, attendance at meetings with health 2-58 7390SNP

physics personnel associated with nuclear plant operations, and review of NRC inspection and enforcement reports, Corporate QA audit reports, and incoming NRC Correspondence.

The section staff review proposed NRC and EPA regulations, NRC regulatory guides, and industry standards pertaining to health physics activities to assess the potential impact on Company operations. Company management is advised of the results of these evaluations as appropriate to assist management in planning to meet any future requirements. These proposed requirements are circulated to appropriate Company health physics personnel for, their review and information. The Manager Corporate Health Physics participates in several health physics task forces and ad hoc industry groups to assist in efforts to assess nuclear industry health physics activities.

This information is used to keep appropriate Company personnel informed about the activities for possible utilization in Company operations. Discussions with health physics personnel at other utilities and visits to their facilities are used to assess and incorporate, as appropriate, information obtained from other utilities.

The Manager - Corporate Health Physics investigates known or suspected radiation overexposures as defined in applicable state of federal regulations and reporting the results of the investigation to the Executive Vice President Power Supply and Engineering & Construction including steps that have been taken to prevent similar incidents in the future. The staff also reviews other matters of noncompliance pertaining to Company health physics activities and has the organizational freedom to review in-depth the circumstances surrounding the noncompliance and report the results of the in-depth investigation to management levels as determined by the Manager Corporate Health Physics.

The Manager Corporate Health Physics is also responsible for the development, implementation, and maintenance of the Corporate ALARA program.

The Corporate ALARA Program defines the scope and requirements for individual ALARA programs for Company health physics activities as well as activities which affect health physics programs, such as nuclear plant engineering and construction. The line organizations are required to implement appropriate ALARA programs that comply with the Corporate ALARA Program.

The Manager Corporate Health Physics has been assigned the responsibility for conducting a periodic management review of the quality assurance audit program. The review of this program is conducted twice each year and results are documented in reports and discussed with the Vice President Corporate Nuclear Safety and Research and the Executive Vice President Power Supply and Engineering & Construction. The reports are also sent to the Chairman/President of the Company.

The Manager - Corporate Health Physics has complete freedom to discuss health physics matters with anyone within the Company including the Chairman/President. The Corporate Health Physics Section has no line responsibility for engineering, construction, or operation. The section is primarily responsible for reviewing the Company health physics programs to determine if adequate Company resources are being expended fn these programs. If problems are identified, the staff has the organizational freedom to discuss the problems with the appropriate levels of management to 2-59 7390SNP

resolve the problems. If the problems are not resolved at one level of management, the matter is pursued with higher management until resolution satisfactory to the Manager Corporate Health Physics is obtained. These Senior Management feedback mechanisms are discussed in Section 2.3 Strong interfaces exist between the Corporate Health Physics Section staff, the Nuclear Plant E&RC Manager and their staffs, and the Manager Radiological and Chemical Support and his staff. ,Additional interfaces with the Nuclear Plant Engineering Department and Technical Services Departments have been established to assist them in their health physics related activities.

2~ 2~ 1 ~ 9 Fuel and Material.s Mana ement Grou The Nuclear Fuel Section of the Fuel Department provides operations support to the Company's units by giving technical assistance in four key areas:

refueling support, startup support, analytical support, and fuel performance monitoring.

Prior to plant refueling outages, Nuclear Fuel Section personnel work with the plant staff to develop fuel shuffle procedures to assure proper core loading and to minimize critical path downtime. During the physical fuel movement activities, Nuclear Fuel Section personnel recommend and assist in various irradiated fuel inspection programs and shipping procedures. Technical recommendations are given for test result interpretations and, if required, fuel reshuffling plans are developed. In the event a fuel assembly is damaged, the Nuclear Fuel Section is responsible for taking corrective action and coordinating activities regarding a replacement bundle.

Nuclear Fuel Section personnel are on site during start up physics testing to provide assistance in test procedure development and review, to assist in performing the physics tests, and to give technical assistance in data reduction and interpretation. Comparisons made between measured and predicted results with any discrepancies analyzed and explained in detail.

Nuclear Fuel Section personnel provide analytical support for all activities relating to core management and operations. The Section has primary responsibility for fuel cycle planning and procurement. Analytical support is provided for determining batch size and enrichment requirements, loading patterns, and the merits of proposed changes in fuel design. Plant licensing and core physics characteristics are reviewed for each core loading. Vendor supplied process computer data constants are verified, and process computer backup support is provided. The Nuclear Fuel Section provides recommended rod patterns for startups, conducts in-core flux map analyses, analyzes xenon transients, provides shutdown margin calculations, and determines estimated critical positions. Analytical support is provided for special operating conditions, such as operation with a stuck rod and end-of-cycle coastdown.

Nuclear Fuel Section personnel provide procedures for special measurements and transient parameter tests. Topical reports on the Company's BWR steady state neutronics methodology are under review by NRC. Work is progressing toward steady state neutronics methodology for PWRs and transient analysis methodology for both BWRs and PWRs.

The Company maintains an ongoing fuel performance monitoring program in which 2-60 7390SNP

routine reports are generated, detailing statistics on past generation, remaining fuel generation capability, power and exposure distributions, margins to Technical Specification limits, and reactor coolant activity levels. Startup physics test results are documented and compared to predictions. Fuel performance information is also collected and transmitted to the fuel supplier for his review and analysis.

In addition to the above operations support activities, the Nuclear Fuel Section coordinates the Company's special nuclear material accountability program and monitors and supports the Company's spent fuel transportation and management activities.

In carrying out its operations support responsibilities, the Nuclear Fuel Section interfaces with other Company departments with primary contact being made with Nuclear Operations Department and Technical Services Department personnel. In addition to those activities described above, interaction with the Nuclear Operations Department involves periodic planning for optimum fuel operating strategies and core control. Daily contact is maintained between the Nuclear Fuel Section staff and nuclear plants'ngineering and operations staff.

Continuing interactions with the Nuclear Licensing thit of the Technical Services Department provides for detailed reload fuel licensing review and coordination.

Materials Mana ement De artment The Purchasing Section of the Materials Management Department provides procurement support to the operating nuclear plants. Within the Purchasing Section, the Generation Procurement Ihit Nuclear is responsible for the purchase of operating materials and spare parts for the Harris plant. The Expediting thit is staffed to provide support for expediting shipment of material to operating nuclear plants.

The Materials Control Section is assisting in layout of warehouse and cataloging of materials and implementation of inventory control computer systems for the Harris plant.

Interface with the Nuclear Operations Department includes coordination with nuclear plants'taff and Operations Quality Assurance personnel. The interface with the nuclear plants'taff involves procurement of standard and quality assurance related materials and services, clarification of quality assurance requirements, problem resolution, and securing necessary documentation. The interface with Operations Quality Assurance involves resolution of problems relative to supplier qualification, quality assurance requirements, internal and external quality assurance audits and training coordination.

Purchasing interfaces with the Corporate Quality Assurance Department in areas of quality assurance audit and supplier qualification. The audit is performed to ensure compliance with the Corporate Quality Assurance Program and other related programs and manuals.

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2.2.2 Offsite Mana ement Resources The Company's offsite management is well qualified to direct the design, construction, and operation of the Harris Plant. The Company has continuously increased its offsite management involvement in nuclear plant design, construction, and operations from 1966, when the Robinson project commenced, to the present, when five nuclear units are under Company construction and operations management.

The Company offsite management is experienced in coordinating complex nuclear operations which require close coordination between design, construction, and operating forces. An example of these types of activities is the startup and operation of Brunswick Unit 2 while Brunswick Unit 1 was still under construction, with security separation of common control rooms. This experience, as well as our additional experience with Harris, demonstrates the Company's capability to safely manage all nuclear activities simultaneously.

These management resources are described in more detail in the following sections.

The Executive Vice President Power Supply and Engineering & Construction Groups, Mr. E. E. Utley, manages offsite nuclear operations support activities in the areas of power supply, engineering support, nuclear fuels and materials, nuclear safety, and quality assurance. Mr. Utley has 31 years of experience with the Company, including serving as superintendent of three plants, and manager of power supply activities for the past ten years. The structure of these groups is shown on Figure 5 and 6.

2.2.2.1 Power Su 1 Grou The Senior Vice President Power Supply Group, Mr. L. W. Eury, reports to Mr.

Utley, and within his group provides the primary offsite management resources for nuclear operations. Mr. Eury holds a bachelors degree in electrical engineering, is a registered professional engineer, and has 23 years of experience in various engineering and management positions in the Company.

The departments and section management within this group are described below.

2.2.2.1.1 Nuclear erations De artment The Vice President Nuclear Operations Department, Mr. B. J. Furr, holds a bachelors degree in mechanical engineering and has 20 years of engineering experience, 15 of which is nuclear power experience. He has been Plant Manager of both the Robinson and Brunswick Plants. He has held an SRO license on the Robinson plant and was involved on both Robinson and Brunswick during startup testing and operation.

Administrative and Technical Su ort Section The Assistant to the Vice President Nuclear Operations, Mr. J. L. Harness, heads up the Administrative and Technical Support Section of the Nuclear Operations Department. He holds a Bachelor of Science degree in Physical Science and a Master's degree in Radiation Biology. He has 26 years of experience in the nuclear field and has held engineering positions at both the Robinson and Brunswick nuclear sites.

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2.2.2.1.2 Technical Services De artment The Manager Technical Services Department, Mr. W. J. Hurford, holds a bachelor of science degree in Metallurgical Engineering and a Master's degree in Industrial Management., He has 34 years of experience in the nuclear field. Mr. Hurford served as the manager of the Light Water Breeder Reactor Core Activity at the Westinghouse Bettis Laboratory; he was the Vice President of Corporate Production of Wyoming Mineral Corporation; and he was the Manager of Production for the Western Zirconium Division of Westinghouse. He is a member of the American Society for Metals and joined the Company in January, 1983.

Licensin 6 Permits Sections The Manager Licensing & Permits Section holds a bachelors, degree in engineering, is qualified as a reactor operator in the U. ST Naval Nuclear Power Program, is a registered professional engineer, and has 20 years experience in 'nuclear engineering fields.

Environmental Technolo Section The manager, Environmental Technology Section, holds a bachelor and master of science degree in biology, a doctorate degree in zoology, and has 17 years of experience in environmental assessment and monitoring field.

Nuclear Trainin Section The Manager of the Nuclear Training Section has a masters degree in nuclear engineering and 34 years of nuclear experience, including three years as professor at the Air Force Institute of Technology, 13 years as professor of Nuclear Engineering at North Carolina State University, with emphasis on nuclear operations engineering and systems design; and seven years as special consultant to NRC Operator Licensing Branch and DOE Safety Division. He is a registered Professional Engineer. He has been a Project Engineer on the design, construction, and operation of three major nuclear facilities.

Radiolo ical and Chemical Su ort Section This section is headed by a manager with a bachelors degree in physics and 24 years of experience in the area of health physics, with 20 of those years associated with power or research reactors. He is certified by the American Board of Health Physicists.

J Emer enc Pre aredness The Director of this unit has a bachelors degree in industrial engineering, has U.S. Naval Nuclear Power Program experience, is a registered professional engineer, and has 15 years of experience in nuclear engineering.

2 ' ' ' Fuel and Materials Mana ement Grou The Senior Vice President Fuel and Materials Management Group, Mr. J. M.

Davis, Jr., reports to the Executive Vice President Power Supply and 2"63 7390SNP

Engineering & Construction Groups. Mr. Davis holds a bachelors degree in mechanical engineering, is a registered professional engineer, and has 24 years of experience in engineering and management.

The Vice-President Fuel Department, Mr. R. A. Watson, holds a bachelors degree in nuclear engineering, a masters degree in physics, is a registered professional engineer, and has 25 years of experience in nuclear engineering activities. Reporting to the Vice President Fuel Department, is the Manager Nuclear Fuel Section, who holds a bachelors degree in nuclear engineering, is a registered professional engineer, and has 16 years of experience in nuclear engineering activities.

2.2.2.3 En ineerin 6 Construction Mana ement As shown in Figure 6, the Senior Vice President Engineering 6 Construction Group has two Department Managers, five Section Managers and a Director Nuclear Engineering Safety Review, who provide the management for the Company's nuclear plant engineering and construction. The Senior Vice President, Mr. M. A. McDuffie, holds a bachelors degree in civil engineering, is a registered professional engineer, and has over 30 years of experience in power plant engineering and construction, including 16 years experience in nuclear plant construction and engineering.

The Vice President Nuclear Plant Engineering, Mr. A. B. Cutter, holds a bachelors degree in chemical engineering, a masters degree in nuclear science and engineering, and has over 20 years of experience in a combination of Navy nuclear engineering and operations, light water reactor and breeder reactor engineering. He is a registered professional engineer. The Nuclear Plant Engineering Department is divided into two sections: the Harris Plant Engineering Section and the Engineering Support, Nuclear Plants Section. In addition to the two sections, the department Vice President has a Director Nuclear Engineering Safety Review on his staff. The Manager Harris Plant Engineering Section, who is responsible for providing the design and engineering for Harris, including engineering support of site construction, startup, and operating activities, holds a bachelors degree in electrical engineering, a professional degree in nuclear engineering, is a registered professional engineer, and has 19 years of engineering and power plant operations experience, 13 years of which are nuclear engineering experience. The Manager Engineering Support, Nuclear Plants Section, who is responsible for providing engineering support for t'e Company's Brunswick and Robinson nuclear plants, holds a bachelors degree in engineering, a masters degree in nuclear engineering, is a registered professional engineer, and has 23 years of experience in nuclear engineering and related fields. The Director Nuclear Engineering Safety Review holds a bachelors degree in nuclear engineering, industrial engineering, and engineering math, is a registered professional engineer, and has 23 years of experience in nuclear operations, nuclear regulation and nuclear power plant engineering management. He is responsible for engineering safety review of the department's activities and for processing operations and engineering feedback to prevent known problems from occurring in the design of the Harris plant or in modifications to operating plants He also assures that a departmental training program is developed and implemented to provide department personnel knowledge of the responsibilities of their duties, and that the ALARA program is effectively implemented.

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The Vice President Nuclear Plant Construction Department, Mr. Sheldon D.

Smith, holds a bachelors degree in civil engineering, is a registered professional engineer, and has 34 years of construction experience.'ll Nuclear Plant Construction Department construction activities at the Harris Plant are managed by the Project General Manager, who holds a bachelors degree in civil engineering and has 16 years of experience in nuclear power plant construction management. The nuclear construction activities relating to Robinson are managed by the Robinson Construction Manager who holds a bachelors degree in mechanical engineering, is a registered professional engineer, and has 35 years of construction experience, including 13 years of nuclear construction management. Construction procurement and contracting activities relating to the Harris, Brunswick, and Robinson Plants are managed by the Manager - Construction Procurement and Contracting, who holds a bachelors degree in science and has 32 years of construction experience, including 17 years of nuclear construction.

2.2.2.4 ualit Assurance and Qualit Control As shown on Figure 1, the Manager Corporate Quality Assurance, who reports to Mr. Utley, manages all QA/QC functions within the Company except certain construction inspection activities that have been delegated to the Harris Site Management Section. In such cases the Corporate QA Department will conduct surveillance of those activities. The Manager of this Department, Mr. H. R.

Banks, has 35 years of experience, which includes 20 years of Navy experience and 15 years in various engineering management positions with the Company.

His experience includes 24 years in the nuclear power field. The Corporate Quality Assurance Department is organized into three separate functions, as shown below QA/ C Brunswick and Robinson Plant Section This Section Manager has 17 years of engineering experience, 11 of which are with the Company. He has held a variety of engineering and management positions at the General Office including assignments in nuclear licensing, siting, system planning, and corporate communication. He has a bachelors degree in engineering, a masters degree in nuclear engineering, and is a registered professional engineer.

ualit Assurance/ ualit Control Harris Plant Section This Section Manager has 30 years of QA experience, ten of which are with the Company. Twenty nine of these years were in managerial positions. He is a registered professional engineer.

ualit Assurance Services Section This Section Manager has 31 years of experience, one of which was in a managerial position with the Company and 30 years in the Navy, 23 of which were in the Navy Nuclear Power Program with a major portion of that time involving duties that required direct supervision of the construction, operation, and management of nuclear propulsion plants. He has a bachelors degree in Marine Engineering.

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2.2.2.5 Cor orate Nuclear Safet and Research 1e artment Cor orate Nuclear Safet The independent safety review function is accomplished by the Corporate Nuclear Safety and Research Department. The Vice President Corporate Nuclear Safety and Research Department, Dr. T. S. Elleman, holds a doctorate degree in physical chemistry and has 27 years of nuclear engineering experience, most recently as head of the nuclear engineering department at North Carolina State University. He is also a Certified Health Physicist.

As shown in Figure 7, the Corporate Nuclear Safety Section and the Corporate Health Physics Section report directly to the Vice President Corporate Nuclear Safety and Research. These sections conduct the independent nuclear safety reviews and health physics assessments of the Company's nuclear facilities.

The Manager Corporate Nuclear Safety, Dr. J. D. E. Jeffries, holds a bachelors degree in engineering, a masters degree and a doctorate degree in nuclear engineering, is a registered professional engineer, has four years of experience in the U. S. Marine Corps, and thirteen years of experience in nuclear power plant engineering and safety review.

The Section is composed of a Nuclear Safety Review Unit (offsite) and three On-site Nuclear Safety Units, one at each plant site. The qualifications of the SHNPP ONS Unit are outlined below:

The IKrector On-site Nuclear Safety (SHNPP), Mr. H. W. Bowles, holds a bachelors degree in engineering physics and a masters degree in economics, has three years of engineering experience relative to naval nuclear propulsion plants, and nine years of experience with nuclear power plants including nuclear. safety review. The three engineers reporting to him have a combined experience of approximately 45 years. -This Unit has been relocated at the plant site since early 1983 and has the responsibility of implementing the nuclear safety review program (ISEG) for SHNPP from preoperations testing through commercial operation.

There is also an offsite Nuclear Safety Review Uhit located in the General Office which reports directly to the Manager Corporate Nuclear Safety. This unit consists of a Director and eleven engineers.

Cor orate Health Ph sics The IIanager Corporate Health Physics, is responsible to the Executive Vice President - Power Supply and Engineering & Construction through the Vice President Corporate Nuclear Safety and Research. He holds a bachelors and masters degree in nuclear engineering and has 18 years of experience in nuclear power, of which 15 years are utility nuclear experience.

2 3 Senior Mana ement Oversi ht Functions The Vice President Corporate Nuclear Safety and Research Department provides Senior. Management,'ncluding the Chairman/President and the Board of Directors, a continuing assessment of current nuclear safety programs.

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Additionally, should any nuclear safety or quality assurance issue require immediate attention, the Vice President Corporate Nuclear Safety and Research has the authorized organizational freedom to contact anyone with the Company, including the Chairman/President and the Board of Directors, in order to resolve such concern to his satisfaction. Figure 8 shows these administrative channels and direct communications channels. The Corporate policies for nuclear safety and quality assurance are set forth in Exhibit 1 and 1A.

The Vice President Corporate Nuclear Safety and Research Department meets quarterly (or as immediate attention dictates) with the managers of OA/OC at Brunswick, Robinson, and Harris and the Manager of the COA department to discuss safety and QA issues. He also meets with the Plant General Managers at Brunswick and Robinson and will begin to meet with the Plant General Manager at Harris on a quarterly basis. If the Vice President Corporate Nuclear Safety and Research is absent, or in the event he does not,respond satisfactorily, each of the listed Managers and Directors are required to communicate directly with the Chairman/President.

In addition, the Vice President of Corporate Nuclear Safety and Research meets approximately twice a year with the Company Chairman/President (exhibit 3) to review actions related to nuclear safety at the plants and to present a nuclear safety summary. An annual presentation to the Board of Directors is made on Nuclear Safety of the operating plants. If the Vice President-Corporate Nuclear Safety and Research is not satisfied with the Chairman/President's response, he is directed to meet with a member of the Board of Directors to express his concerns.

This strong program which implements an effective Senior Management feedback system, independent of the normal administrative channels, ensures that those people within the Company who administer nuclear safety and quality assurance programs maintain complete organizational freedom to act with written Corporate authority, A similar Corporate Policy exists with respect to health physics and radiation protection (see Exhibit 4, memorandum from J. A. Jones to members of Senior Management), in which the goals and objectives of the radiation protection programs are established and mechanisms for Senior Management oversight are delineated. Additionally, documentation which implements this policy is included in Exhibit 4.

Additional Senior Management oversight is effected by routine reports. Senior Management receives copies of the major reports generated by Corporate Nuclear Safety. These" include a bimonthly summary of outstanding items, a summary of LER trends by categories, and activities reports that summarize operations experience feedback and review activities. As appropriate, detailed assessments of system performance may also be distributed to senior management. A "management advisement" may be included in the report commitment does not appear to be sufficient to resolve a problem by the if a established target date. Items of interest regarding safety areas are also briefly summarized as a means of keeping senior management apprised of ongoing activities of importance.-

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Senior Management also receives direct feedback on the results of quality assurance audits. An audit report includes the audit scope, participants, results, recommendations, comments, and details of nonconformances. There are two types of nonconformances that are included in the report: findings and concerns. A nonconformance requires documented corrective action for resolution A comment is a weakness which may present a potential problem and may not require documented action. The report is addressed to the Company President/Chairman, the Executive Vice President Power Supply and Engineering & Construction, and the Senior Vice President Power Supply, with copies to the activity which was audited. Within 30 days after receipt of the report, the audited activity must formally indicate the corrective action and schedule for resolving each finding. The Manager reviews these responses for adequacy and contacts the responsible managers to resolve any problems.

Zn addition, monthly reports containing the status of all quality -assurance audits are sent to Senior Management. The key to the success of this audit program is its innate ability and freedom to identify deficiences to Senior Management at the same time they are identi,fied to middle management.

Senior Management responsible for nuclear plant operations is keenly interested in all aspects of plant operations and maintains an up-to-date awareness of any concerns that ma'y affect nuclear safety. On a daily basis, the Executive Vice President Power Supply and Engineering & Construction receives written reports of the status of operating units, load reductions, and reasons for those reductions. Through direct discussions and staff meetings, the Vice Presient Nuclear Operations and the Vice President Brunswick Nuclear Project keep the Senior Vice President Power Supply advised of any significant concerns affecting the nuclear plants.

On a periodic basis, the Executive Vice President Power Supply and Engineering & Construction Groups, the Senior Vice President Power Supply, and the Vice President - Nuclear Operations, meet at the operating nuclear plants to review matters related to nuclear plant operations with plant management personnel.

2.4 Offsite Technical Staff Resources The functional organization of the offsite groups which support nuclear operations is described in Section 2.2.1. Within this organization are personnel with the training, experience, and expertise to ensure that technical capability is maintained in the following areas:

a~ Nuclear, mechanical, structural, electrical, thermal-hydraulics, and fluid systems

b. Metallurgy and materials; instrumentation and controls engineering C~ Plant chemistry and radiochemistry

'd ~ Health physics

e. Fueling and refueling operations support 2-68 7390SNP

Maintenance support

g. Technical and Engineering support

h. Operations management

i. quality assurance Fire protection

k. Emergency Preparedness The distribution of areas of expertise and responsibility within the functional organizaiton is depicted on Figure 4.

In addition, the Company's staff resources rely upon other technical resources such as the HENDEC. THE HENDEC provides modern labs that support the Company's staff resources in a variety of ways. A detailed discussion of that support can be found in Technical Services Department offsite support (Section 2.2.1.4).

In order to demonstrate the staff personnel resources in a clear and concise manner, a figure and several tables are provided. Figure 9 shows authorized and actual staffing levels of professionals, including degreed engineers, and the total years of experience in each depicted department.

Table 2-2 shows the total numbers of college graduate personnel within nuclear support organizations. Table 2-3 and 2-4 provide a breakdown of professional engineers and engineers in training who support nuclear operations. Table 2-5 presents a breakdown of degrees pertinent to nuclear operations; the distribution of these persons within the departments.

it depicts The technical staff resources described above, coupled with the Company's nuclear experience involving one unit at Robinson and two units at Brunswick, demonstrate the Company's capability to properly support all of its nuclear activities.

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TOTAL NUCLEAR SUPPORT STAFF GRADUATE PERSONNEL SRBGLRY AS OF DECEMBER 31 1982 GRADUATE TOTAL DEGREES DEGREES/

PERSONNEL BS BA MS MA PHD TOTAL PERSON Brunswick Nuclear Pro). ] 65** 140 33 19 2 1 195 1 ~ 18 Corp. Nuc. Safety & Res. 29 28 1 18 0 3 50 1.72 Corp. Quality Assurance 43 38 7 3 0 0 48 1.12 ENGINEERING & CONSTRUCTION GROUP Eng. & Const. Sup. Svcs. 22* 20 2 0 0 0 22 1.00 Fos. Plant Eng. & Const. 62 63 03 14 0 2 82 1 32 Nuclear Plant Const. 109 104 13 12 0 0 129 1.18 Nuclear Plant Eng. 88 87 5 15 1 1 109 1.24 281 274 23 41 1 3 342 1.19 FUEL & MATERIALS MANAGEMENT GROUP Contract Services 10* 10 0 0 0 0 10 1.00 Fuel 42 34 9 19 5 0 67 1.60 Materials Managment 32 24 10 5 0 0 39 1 22 TOTAL FUEL & MATo MGT e 84 68 19 24 5 0 116 1.38 POWER SUPPLY GROUP Maint. Sup. & Admin. 13* 12 2 2 0 0 16 1 '3 Nuclear Operations 226 200 32 33 3 3 271 1.20 Technical Services 156 143 22 54 1 7 227 1.46 TOTAL POWER SUPPLY 395 355 56 89 4 10 514 1.30 TOTAL OPER. GROUPS 997 903 139 194 12 17 1,265 1.27

• Includes Group Executive

• • Includes Assistant to Exec. Vice Pres.

TABLE 2-2 (7 492M SG1 cv) 2-70

NUCLEAR SUPPORT STAFF PROFESSIONAL ENGINEERS AS OF DECEMBER 31 1982 There is a total of 150 registered professional engineers in the operating groups.

ENGINEERING/PHYSICS DEPARTMENT PE'S DEGREES / PE'S Brunswick Nuclear Project 14 136 10.3 Corporate Nuclear Safety & Research 11 47 23.4 Corporate Quality Assurance 9 30 30.0 ENGINEERING & CONSTRUCTION GROUP Engineering & Construction Sup. Svcs. 1* 6 16. 7 Fossil Plant Engineering & Construction 27 69 39.1 Nuclear Plant Construction 20 86 23.3 Nuclear Plant Engineering 25 94 26.6 TOTAL ENGINEERING & CONST. GROUP 73 255 28.6 FUEL & MATERIALS MANAGEMENT GROUP Admin. & Contract Services 1* 5 20.0 Fuel 13 48 27 F 1 Materials Management 1 6 16 '

TOTAL FUEL & MATERIALS MAN GROUP 15 59 25.4 POWER SUPPLY GROUP Admin. & Maint. Support 4* 8 50.0 Nuclear Operations 18 198 9.1 Technical Services 6 130 4.6 TOTAL POWER SUPPLY GROUP 28 336 8.3 TOTAL OPERATING GROUPS 150 863 17.4

• Includes Group Executive TABLE 2-3 (7492MSGlcv)

NUCLEAR SUPPORT STAFF ENGINEERING IN TRAINING AS OF DECEMBER 31 1982 There is a total of 135 Engineers in Training in the operating groups.

De artment Prospect EITs Brunswick Nuclear 19 Corporate Nuclear Safety & Research 3 Corporate Quality Assurance 2 ENGINEERING & CONSTRUCTION GROUP Engineering & Construction Support Services 0 Fossil Plant Engineering & Construction 9 Nuclear Plant Construction 20 Nuclear Plant Engineering 31 60 TOTAL ENGINEERING & CONSTRUCTION GROUP FUEL & MATERIALS MANAGEMENT GROUP Admin. & Contract Services 1 Fuel 7 Materials Management 0 TOTAL FUEL & MAT+ MANAGEMENT GROUP 8 POWER SUPPLY GROUP Nuclear Operations 33 Technical Services 10 TOTAL POWER SUPPLY GROUP 43 TOTAL OPERATING GROUPS 135 TABLE 2-4 (7492MSGlcv) 2-72

BREIKNMN OF TECHNICAL DEGREES BRUNSWICK CORPORATE CORPORATE NUCLEAR NUCLEAR NUCLEAR TECHNICAL NIj LEAR NlCLEAR QUAL I TY FUELS PLANT PlANT OPERATIONS SERV ICES PROJECT SAFETY ASSISANCE, CONSTRlCTION ENGINEERING TOTAL DEGREE 8 M D 8 M 0 8 M D 8 M 0 8 M 0 8 M D 8 M 0 8 M D 8 M D AEROSPACE ENG ANIMAL SCI ~

8 IOLOGY 40 2 32 4 I 108 6 I BOTANY 2 2 2 2 CHB41 CAL ENG. I 24 CHEMISTRY 12 2 11 2 I 20 I I 48 5 C I VIL ENG. 34 3 15 'I 3 I 64 6 COMPUTER SC I ~

CONST, ENG. 2 I ELEC ENG, 2 I 20 5 26 I 4 I 79 8 ENG. GEN, 15 I I ENG, MATH, ENG. MECH, ENG OPS, 6 I ENG. PHYSICS ENG, TECH ENTOMOLOGY ENV, ENG, 2 2 ENV IRON. HEALTH 3 I 2 I 5 3 ENV TOXICOLOGY F ISH, SCIEICE 3 3 GEOLOGY GEOPHYS ICS HEALTH PHYS ICS 2 3 2 17 8 ~ BACHELORS DEGREE M ~ MASTERS DHREE 0 ~ DOCTORATE AS OR DECEMBER 31~ 1982 TABLE 2-5 2-73

BRUNSW I CK CORPORATE CORPORATE NU LEAR NlCLEAR NUCLEAR TECHNICAL NUCLEAR NUCLEAR QUALITY FUELS PLANT PLANT - OPERATIONS SERVICES PROJECT SAFETY ASSURANCE CONSTRIJ TION ENGINEERING TOTAL DEGREE 8 M D 8 M D 8 M D 8 M 0 8 14 0 8 M D 8 M D 8 M D 8 M D I NOUSTR I At. ENG. 9 I 22 1 I NDUSTRIAL TECH.

L IMtIOLOGY MACHINE DES IGN MARINE ENG MATERIALS ENG. I I 5 2 MATHEMATICS 6 I 22 I MECHANICAL ENG. 11 I 5 2 14 I 24 5 1 99 14 MECHANICAL TECH.

MEDICAL TECH.

METALLURGY ENG, 3 I 3 I I METAL LIRGY 2 I 3 I METEOROLOGY MINING ENG NUCLEAR ENG, 10 7 6 2 13 13 2 2 8 4 28 10 10 4 78 43 2 NlCLEAR TECH PHYS ICS 10 2 I 8 I I 30 3 2 RAD, HEALTH I I I 2 SANITARY ENG, SCI EtCE 6 I 14 2 STATISTICS SYST&IS ENG.

WELDING ENG, WI LDL IF E ZOOLOGY 12 2 I 15 2 I TOTAL 698 137 11 8 ~ BACHELORS DEKEE M ~ MASTERS DEGREE D ~ DOCTORATE AS OF. DECEMBER 31, 1982 TABLE 2-5 (Cont'd) 2-74

2.5 Coordination of Interde artmental Technical Staff Su ort Coordination of technical staff support for nuclear operations is the ultimate responsibility of the Vice President Nuclear Operations Department. Written memoranda define each department's functional responsibilities for support activities; however, ultimate responsibility for safe nuclear operations is assigned to the Vice President Nuclear Operations.

During routine operations, when the Nuclear Operations Department requires technical assistance, the appropriate manager initiates a written Task Assistance Request (TAR) to the department within the Company which is assigned responsibility for those activities. Typically, the Nuclear Plant Engineering Department is requested by the Nuclear Operations Department to provide engineering assistance to resolve a plant design problem by initiation of a TAR. The Nuclear Plant Engineering Department performs the necessary engineering and recommends a solution to the Nuclear Operations Department.

The modifications are designed to comply with the Corporate QA Program, the plant Technical Specifications, and to other appropriate design codes. The Nuclear Plant Construction Department may be requested by the Nuclear Operations Department to perform the installation, or the modification may be installed by the Nuclear Operations Department. All modifications to the operating plant are approved by the appropriate management in the Nuclear Operations Department prior to installation. Similar definitions of interdepartmental responsibilities exist between the Fuel Department, Technical Services Department, and Nuclear Operations Department to ensure that nuclear refuelings and spent fuel movements are coordinated properly, nuclear licensing interface with NRC is coordinated with the plants, and that meteorological, seismological, and environmental monitoring programs are carried out as required by each Plant Operating License.

Should situations arise at nuclear plants requiring immediate technical staff support, the Vice President Nuclear Operations Department requests immediate assistance from appropriate department heads throughout the Company or from outside contractors and vendors, if necessary.

2.6 Offsite Staff Trainin Training for the offsite technical support personnel is primarily supplied in the course of routine operations. Written procedures for handling incoming NRC correspondence ensure that appropriate offsite support staff are aware of current NRC directives and issues. Plant events and modifications would routinely be discussed with appropriate offsite personnel by the plant staff, and this interaction is an excellent mechanism to keep the offsite personnel informed. Many of the offsite personnel who deal with the nuclear power plants have spent time at the plant sites and have, therefore, received plant-specific health physics and security indoctrinations. As new personnel join the offsite support groups, they will also receive this training as their job assignments require it. Expertise in specific areas is maintained by continual on-the-job training as routine job assignments are carried out.

Offsite support staff are active participants in many industry groups which deal with generic interests and problems of power plants such as the Edison Electrical Institute, the Southeastern Electric Exchange, and the Atomic Industrial Forum. In addition, this staff routinely participates in utility owners'roups in order to respond to hardware or regulatory concerns such as 2-75 7390SNP

BWR Pipe Cracking or the PWR Steam Generator issues. New information is passed onto the others within the Company who could benefit by briefings by those who first come into contact with the new information.

The Nuclear Training Section provides a formal, professional orientation program for all newly assigned Operations Department professional staff.

Depending on the new employee's experience, this program may consist of up to five phases. In addition, supervisory/management courses are presented by the Employee Relations Department in the area of Human Resources Development; Organizational Development Specialty Training Courses are routinely utilized for employee upgrading and specialization.

2.7 Contract Assistance The Fuel and Materials Management Group maintains a large number of active contracts for labor and services in order to ensure availability of qualified personnel when they are needed by the nuclear plants to augment the Company's resources. These contracts provide the capability to obtain additional maintenance personnel, such as mechanics, electricians, I&C Technicians, and technical specialists, such as RCPT Technicians, on an expedited basis. The Company also has contracts with engineering and consulting firms for the purpose of having immediate access to specialized technical expertise when becomes necessary to supplement Company technical capabilities.

it For example, contracts are in effect with the NSSS vendors, Westinghouse and General Electric, as well as architect engineers, such as Ebasco and debited Engineers. In general, contracts are written so that the General Managers of the nuclear plants can obtain the services and resources provided by these contracts as deemed appropriate. This mechanism ensures that contractual assistance for nearly every conceivable task is available and ready when needed.

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3.0 Emer enc Res onse Ca abilities The Company has developed a Corporate Emergency Plan which, in con)unction with the Shearon Harris Nuclear Power Plant (SHNPP) Emergency Plan, describes the onsite and Corporate level emergency response for the SHNPP.

3.1 Emer enc Classes The Corporate Emergency Plan makes provision for emergency response to four different emergency classes which in order of increasing severity are:

Uhusual Event

. Alert Site Emergency

. General Emergency The categorization or classification of events according to one of the four emergency classes is done on the basis of guides (specific sets of plant conditions, instrument readings, etc.) called Emergency Action Levels (EALs) ~ The EALs are published in Plant Emergency Procedures'he initial responsibility for recognizing off-normal plant conditions and categorizing them within the parameters of the four emergency classes is assigned to the Shift Foreman/Operating Supervisor who has the authority and responsibility to immediately and unilaterally activate the Plant Emergency Plan and initiate any emergency actions. A clear line of succession to the Shift Foreman/Operating Supervisor is established in the Plant Emergency Plan.

Upon declaration of an "Unusual Event," the Shift Foreman/Operating Supervisor shall cause appropriate onsite personnel to be notified and shall begin notification of offsite resources as deemed necessary to assist in the response effort. Upon declaration of the "Alert" emergency condition, an accident is presumed to have occurred or to be in progress which involves an actual or potential substantial degradation of the level of safety of the plant and the response actions as detailed in the Plant Emergency Plan are initiated.

A "Site Emergency" is declared when events are in progress or have occurred which involve actual or likely major failures of plant functions needed for the protection of the public. A "General Emergency" is declared when events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity.

3 ' Onsite Resources and Activities The Plant Emergency Plan provides for an onsite emergency organization to cope with any accident situation. This emergency organization will utilize the existing onsite organizational structure to the greatest extent practical to take advantage of established lines of communication and responsibility The organization provides for management personnel to be assigned responsibility for specific functions. The onsite emergency organization, headed by the Site 3-1 7390SNP

Emergency Coordinator (Plant General Manager or alternate), will be responsible for all site operations necessary to establish safe plant conditions.

The onsite emergency organization includes resources discussed in the following paragraphs.

3.2.1 Control Room Resources When an emergency first occurs it is assumed that only the minimum required shift personnel will be available. The Shift Foreman/Operating Supervisor will be in charge and will serve as the site Emergency Coordinator until relieved. The shift licensed operating staff shall consist of personnel identified earlier in Section 2.

The shift organization will be augmented when requ1red as additional personnel are called into the plant 1n accordance with Plant Emergency Procedures.

3.2.2 Technical Su ort Center (TSC)

The TSC located in the Fuel Handling Building will be activated in the event of an "Alert" or more serious emergency class to provide added support to Control Room personnel. The onsite TSC will be under the supervision of the Site Emergency Coordinator (Plant General Manager or his designated alternate) to control all site operations necessary to establish safe plant conditions.

Four emergency response positions, the Plant Operations Director, the Emergency Repair Director, the Logistics Support Director and the Rad1ological Control Director, report to the Site Emergency Coordinator in the TSC when is activated.

it The TSC has been designed to function in accordance with NRC requirements as set forth in NUREG 0737 Supplement 1 and plant emergency procedures .

Personnel to staff the TSC may come from the plant staff, the offsite organization and/or, through contract arrangements, other organizations as described 1n the Plant Emergency Plan. These personnel or their equivalent replacements will be available continuously for the duration of the accident condition.

3.2.3 erational Su ort Center (OSC)

An onsite OSC will be established in the Plant Service Building inside the Protected Area. It is separate from the control room and the TSC and is the place in which the operations support personnel will assemble for assignments in an emergency situation. Communications with the control room and the TSC will be provided.

3.2.4 Plant Emer enc Procedures Emergency and accident procedures w111 include:

a~ A clearly established chain of command which identifies the individuals and their alternates, their responsibilities and authority, and lines of communication during an emergency 3-2 7390SNP

b. Adequate communication systems and detailed instructions for contacting the plant personnel necessary to meet the Plant Emergency Plan and for contracting the applicable offsite personnel to implement offsite response plans.

C~ Control room access shall be limited to onsite operators and personnel requested by the Site Emergency Coordinator and/or Shift Supervisor.

d. Clearly specified lines of authority and special training required.

e. Provision for emergency classification, notification, activation of emergency personnel and facilities, accident assessment, radiological dose projection and monitoring, public warning and notification, and protective actions.

3.2.5 Emer enc Plan Trainin and Exercises Initial training and annual retraining in the implementation of the emergency procedures is required for all onsite personnel with emergency assignments.

Additionally, drills or exercises are held to ensure that all the personnel with emergency assignments are capable of performing their emergency duties.

3.3 Offsite Resources and Activities 3.3.1 Emer enc Res onse Facilities The Corporate Emergency Plan provides for the establishment of a number of special emergency facilities which include the Emergency Operations Facility, a Plant Media Center, a Corporate Emergency Operations Center and a Corporate Headquarters Media Center.

The Company is constructing an Emergency Operations Facility (EOF) located at the, Shearon Harris Energy and Environmental Center (HE&EC) approximately 2 miles from the Plant, to house offsite Corporate, A/E, NSSS vendor, and other personnel who will provide emergency support and long-tenn recovery support to the plant under the direction of the Emergency Response Manager. The EOF, when activated, is responsible for performing offsite notifications and environmental monitoring, assessing consequences of radiation releases to the environment and recommending offsite protective actions to governmental agencies.

The Plant Media Center will also be located at the HE&EC. The Plant Media Center will be operated under the direction of the Site Public Information Coordinator. Center personnel will assist the various media representatives in obtaining accurate information regarding the emergency. The Corporate Spokesman will work out of the Plant Media Center and participate in news conferences with representatives of the NRC and state and local governments .

Another Company emergency response facility is the Corporate Emergency Operations Center (CEOC). The CEOC is an information collecting and emergency support facility located at the Corporate Headquarters in Raleigh, NC. The 3-3 7390SNP

CEOC provides an interface between facilities at the plant site and Company top management as described in the Corporate Emergency Plan and Implementing Procedures.

The Corporate Headquarters Media Center, also located in Raleigh, provides media related support to the plant.

3.3.2 Emer enc erations Facilit Or anization (EOF)

The EOF organization consists of the Emergency Response Manager, three managers of support functions responsible to the Emergency Response Manager, and supporting personnel as shown in Figure ll. This organization may be mo'dified during an emergency to better respond to the conditions of that emergency.

Emer enc Res onse Manager The Hnergency Response Manager directs and controls the emergency response organization in the EOF to assist and support the plant during the emergency. He will activate the EOF in accordance with the Corporate Emergency Plan and assist the Plant Staff in ending the emergency. Concurrently, the emergency response organization will begin efforts to determine the cause of the incident, develop a recovery plan, and procure any company and outside services and equipment necessary to complete needed repairs.

Technical Anal sis Mana er The Technical Analysis Manager is responsible to the Emergency Response Manager for technical analysis of plant conditions and for the development of plans for bringing the plant to a safe shutdown condition. He will coordinate the receipt and assessment of technical information onsite and offsite related to plant systems and facility operations, and submit timely recommendations to the Emergency Response Manager for implementation.

Administrative and Lo istics Mana.er The Administrative and Logistics Manager provides administrative, logistics, communications, and personnel support for emergency and recovery operations. He will provide assistance to the Emergency Response Manager in the short-term planning, scheduling, and expediting of recovery operations.

Radiolo ical Control Mana er The Radiological Control Manager is responsible for providing support to the Plant in the areas of environmental monitoring and offsite radiation dose projection and for recommending offsite protective actions to the Emergency Response Manager.

3.3.3 Cor orate S okesman The Corporate Spokesman for the Company will have primary responsibility for all public statements to the media at the site concerning the cause and effects of the emergency, and steps the Company is taking to mitigate it.

3.3.4 State and Local Facilities The State Emergency Response Team (SERT) Headquarters, located in Raleigh, NC, will house the necessary facilities for state and federal agencies responsible for directing all offsite activities required for protection of the public 3-4 7390SNP

health and safety. A Company representative will be assigned to the SERT Headquarters to provide liaison between Company management onsite and the State of North Carolina and other outside agencies operating from that location.

Each County within the 10-mile Emergency Planning Zones will provide an Emergency Operations Center from which County officials can coordinate'he emergency activities under local control. Emergency information will be provided to them directly from the Company until activation of the SERT.

3.3.5 Offsite Contractual Assistance and reements A number of active outside contracts are maintained in order to ensure continuing access to qualified personnel when and if they are needed to supplement Company resources in the event of an emergency. These contracts provide the capability of obtaining, on an expedited basis, additional support personnel; see paragraph 2.7.

INPO serves as a clearinghouse for industry wide support during an emergency. The Company is a signatory to the mutual assistance agreement development by INPO for utilities in the nuclear industry.

In addition, the Plant Emergency Plan contains written agreements between the Company and others such as hospitals, ambulance squads, doctors, and local government agencies that may be required to provide support to the plant in the event of an emergency.

3.4 Conclusion The above described emergency organization, facilities, communications, plans, and training activities have been designed by the Company to maintain a state of readiness that provides reasonable assurance that adequate protective measures can and will be taken in the event of an emergency.

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4.0 Harris Lbit 1 Startu 4.1 Startu Or anization The Harris Startup Organization, which will be responsible for the startup of both Harris units, consists of 42 people, including a Manager - Startup and four Startup Supervisors, each of whom supervises the efforts of seven senior engineers and two engineering technicians. The responsibilities and authorities of each are as follows:

Mana er Startu aO Supervise the Activities of the Startup Organization through the Startup Supervisors.

b. Prepare and update the startup schedule.

C~ Assign overall test responsibility to the Startup Supervisors.

d~ Review and approve requests for vendor assistance as recommended by the Startup Organization.

e. Review and approve/recommend approval of test procedures, test procedure modifications,'nd test data in accordance with the Startup Manual instructions.

Review and recommend approval of startup requests for construction and engineering modifications or changes required during the test program.

go Issue periodic progress reports and work schedules for the Startup Organization.

h~ Issue special reports concerning startup activities as deemed necessary.

Review progress of startup activities with contractors, vendors, and Company management.

Maintain liaison with the plant management, keeping them informed of the test program status, and coordinate with them the activities of plant personnel assigned to startup activities in conjunction with their training program.

k. Represent the Startup Organization on interdepartmental and interorganizational committees associated with the test program.

Maintain liaison with contractors and vendors to coordinate their activities relating to the test program.

Responsible for the preparation and maintenance of the Startup Manual.

n. Accept Release for Tests from Company Construction.

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Startu Su ervisor aO Assign a cognizant Startup Engineer for each test identified to be required on assigned systems, and periodically review test assignments to maintain an even distribution of work load.

b. Supervise the activities of and provide guidance to the Startup Engineers reporting to him and assure that their operations are conducted in accordance with SHNPP Startup Manual instructions.

co Supervise the preparation of test procedures as assigned to the individual Engineer.

d. Provide technical guidance and assistance in the preparation of test procedures.

e. Recommend plant scheduling changes as necessary to support the testing effort.

Review and recommend approval of test procedures, test procedure modifications, and test data in accordance with the Startup Manual procedures.

Recommend approval of and schedule vendor representative assistance.

h. Recommend changes in plant design and/or construction to facilitate testing, operation, and maintenance.

Review periodic progress reports and work schedules.

Assist in the preparation of special reports concerning startup activities when required.

Startu En ineer aO Conduct all work assignments in accordance with the Startup Manual and other Startup procedures/instructions.

b. Prepare and recommend for approval assigned test procedures.

C~ Conduct all assigned tests and prepare test reports.

d. Review engineering drawings and documents and prepare requests for construction and engineering changes to facilitate both operation and maintenance.

C

e. Recommend approval of system Release for Tests.

Define system and subsystem Release for Test boundaries.

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go Conduct an inspection of assigned systems prior to system Release for Test acceptance and recommend the acceptance of systems from construction for testing.

h. Coordinate and supervise activities of personnel assigned to support the test program.

Ebasco Services Inc.

Ebasco Services is the A/E for the SHNPP. As the design organization, Ebasco will provide the design bases to be used in the evaluation of appropriate safety related and balance-of-plane preoperational test procedures and results. Ebasco will be consulted, as necessary, during the procedure development and evaluation of test results to ensure that the test program verifies plant design parameters.

Westin house Electric Cor oration Westinghouse provides onsite technical assistance to the Company during the installation, startup, testing, and initial operation of each nuclear steam supply system (NSSS). Westinghouse onsite personnel review and comment on NSSS preoperational test procedures to assure that test acceptance criteria and test objectives are appropriate and consistent with safe operation.

4.2 Harris Pre-Startu Pre arations As a result of the experience gained at Robinson and Brunswick, the Company is aware of the value of early pre startup planning for Harris Unit 1. Several management actions have been taken with the objective of effecting a smooth, orderly transition from construction through startup to operations:

aO The construction schedule reflects a consideration of plant startup requirements. A Release for Test (RFT) schedule has been developed jointly by Startup, Engineering, and Construction which defines the sequence and timing of each RFT turnover to the Startup organization to support an orderly testing process. The various buildings and piping systems have been scheduled in a manner that will allow early startup involvement. This will allow leveling the startup manpower requirements and will permit the startup and testing of the entire system to begin as early as practicable. As equipment and systems are completed, they are transferred to the Company Startup Organization, which implements the testing program. This jurisdictional transfer consists of a series of formal system/component and facility/structures Release for Test until eventually all plant system/components and facility/structures have been formally released to the Startup Organization. This transfer (RFT) only represents a transfer of jurisdiction and not a transfer of work responsibility that is normally performed by Nuclear Plant Construction Department (i.e., identified on plant design documents and punch lists).

All work performed by the Nuclear Plant Construction Department after release to Startup will be performed in accordance with the Startup Manual.

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A similar logic prevails with respect to the electrical portions of the plant. The construction schedule places emphasis on early completion of those portions of the common building which contain

'he control room and the cable spread rooms. The intent is to construct these facilities as soon as possible in order to allow installation of the relay racks and control boards in the cable spread room and the control room. Again, releasing startup and test work as early as possible will provide the maximum amount of time for an orderly transition to the operating phase.

b. Construction management action was also taken in a second area of scheduling in order to contribute to a smooth running RFT Program and startup of the plant. Facilities were provided at the Harris site to permit the Startup Organization to move to the site earlier than the norm, thus providing opportunity for the Startup Engineers to become involved in the day-to-day construction of the various system for which they are responsible. This assignment of plant operating personnel will facilitate their becoming familiar with the plant and its systems while they write pre-operational and testing procedures and observe the construction in progress. The familiarity thus gained will simplify problems during turnover for startup.

Co Definition of the boundaries of the RFT packages was started in 1979 so that, construction could maintain a continuing cognizance of the boundaries of the startup systems and the sequence in which the systems would be required to be completed and turned over to plant operations. These boundaries allow construction to plan work on the systems needed first, and permits Quality Assurance personnel to collect, check, and package documentation in volumes that are grouped in anticipation of future RFTs.

d. The organization at the Harris site has been structured such that there are fewer organizations involved in the RFT process as compared to our other nuclear plants. Therefore, the process of construction completion, construction checkout, QA checkout, and Release for Test to Startup flows through fewer organizations resulting in a better understanding and closer contact with the system being completed.

As individual systems or portions of systems are completed, the on site Corporate Quality Assurance/Quality Control Unit performs final system inspections. Any nonconforming or incomplete work is identified and documented for resolution. Subsequent inspections are performed to ensure proper resolution of nonconformances and incomplete work. Any outstanding work (i.e., incomplete work, nonconformance, etc.) is identified in the RFT package as exceptions. The RFT package is then forwarded to the Startup Organization for acceptance. The site Quality Assurance/Quality Control Unit will continue to provide Quality Assurance coverage for any further work (e.g., resolution of exceptions) performed by the Nuclear Plant Construction Department subsequent to RFT and provide QA/QC coverage for the testing, operation, and maintenance activities performed by the Nuclear Operations Department after RFT.

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As an aid in developing complete turnover packages, a Material Trackin S stem has been initiated for pipe spool, hanger, and electrical cable installation. The purpose of this system is to track the completion status of these bulk items from the delivery phase through installation and Release for Test to the Startup group in a consistent and efficient manner. The objective of this system is to establish a common data base from which material status will be extracted by all functional groups, thus reducing the creation of duplicate files with their potential for inconsistent information.

Construction Control S stem The concept of this module of the Material Tracking System consists of an on-line computerized system with a common data base from which engineering, construction, quality assurance, and operations (startup) can extract pertinent data relative to the status of material during the erection, inspection, testing, and RFT phases of the project. One major component of the Construction Control System is an RFT Tracking System which provides the status of systems for startup and testing. Additional components include:

a. Startup work list This component was implemented in 1982 and is the master punch list of remaining work on systems released to the Startup group for testing. The list is subdivided by system and includes all RFT exceptions, work items identified after RFT as result of testing, and design change issued after RFT.

b. Field Change Request (FCR)/Design Change Notice (DCN) Tracking System This system was implemented in 1983 and tracks the status of field initiated design changes. (FCRs and A/E initiated DCNs). The system also provides a cross reference between PCRs/DCNs and Release for Test boundaries to aid in completion systems for RPT.

C~ Mechanical Status Reporting System This component of the Construction Control System was implemented in 1979. The system tracks the status of piping spools, hangers, restraints, and valves. It is currently being updated to track entire RFT packages and their related documents.

d. Electrical Status Reporting System This component of the Construction Control System was approved and implemented in early 1981. This system tracks the status of cables, cable trays, conduit, and terminations. it is also designed to interface with the A/E computerized Cable Routing System which produces raceway, cable pull, and cable termination cards. In September 1981, this system was converted to an on-line system.

e. Welder Qualification System This system, a manually operated system for monitoring welder qualification, has been in operation since the Construction Permit was received, and was computerized in 1980. It tracks the qualification status and performance of each welder on the SHNPP project, and alerts project management when qualifications on each welding procedure is subject to expire.

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The Harris Startup Manual addresses the plant startup program. Startup includes the period from completion of construction (on a system or equipment basis) through the completion of the Power Test Program. There are two major periods of activity and two major areas of responsibility as follows:

Construction Phase/CP&L Nuclear Plant Construction De artment Area of Jurisdiction: A period that begins with receipt of equipment on site and ends with system/component releases to the Harris Startup Group.

Initial Test Phase/CP6L Startu Grou Jurisdiction: A period that commences with system/component release to the Startup Group and terminates with the completion of the Power Test Program.

The Startup Manual, Section 5, "Construction Completion Activities," is used by the Nuclear Plant Construction Department as a guide in performing necessary checks prior to releasing the equipment to the Startup Group.

The Startup Manual, Section'6, Construction/Startup Interface, describes in detail the interface procedures and controls used from initiation of RFT (Release for Test) from the Nuclear Plant Construction Department until a system is placed on the Operational Systems List after completion of testing. It should be understood that this transfer (Release for Test) only represents a transfer of jurisdiction and not a transfer of work responsibility that is normally performed by the Nuclear Plant Construction Department. The significant point is that although work will be processed through and controlled by the cognizant Startup Engineer or Supervisor after release to Startup Group, the RFT exception work and any required modifications will be performed by construction forces already on the site.

This approach along with several significant factors related to craft labor will allow the construction startup group to be adequately staffed. The significant craft labor factors which will exist during the startup period are:

aO The peak requirement for mechanical and electrical crafts for Unit 1 will have passed. Since work will be winding down and the number of craftsmen reduced, the constructor will be in a position to retain those with higher skill levels.

b. There will be a sizeable pool of workers from which to draw.

Co The craftsmen devoted to the effort will be those who made the initial installation.. This familiarity of installation and site specific procedures, requirements, etc., will expedite the work.

d. The twin-unit site offers additional advantages:

The pool of experienced manpower working on Unit 2 could be shifted to Unit 1 if unforeseen circumstances arise; 2~ With Unit 2 construction manpower requirements building up, as Unit 1 requirements decrease, we can offer continuous employment to qualified craftsmen and minimize the tendency for craftsmen to being looking for outside employment as Unit 1 winds down.

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The startup of SHNPP Unit 1 will be the fourth nuclear unit on which Company professional and technical personnel have participated. As we progress through construction, the expertise level of existing employees will rise and the addition of new employees will add to that experience level.

Another advantage of the twin-unit plant is that the learning curve from Unit 1 to Unit 2 results in less demand on senior site personnel's time, allowing them to devote more of their time to startup support.

Construction Plant Securit Duri Unit 1 0 erations Based on our experience during construction of Brunswick Unit 1 while Brunswick Unit 2 was operating, the SHNPP security plan has been formulated and is continuing to be reviewed and updated as necessary. Details of the plan are designated safeguards information pursuant to 10 CFR 73.21 and require protection from unauthorized disclosure. For these reasons, the following information pertaining to construction support of plant security does not address specific details of the plant security plan. When nuclear fuel is received for SHNPP Unit 1, portions of the plant security program will be implemented to provide necessary security. This plan will be administered by the Nuclear Operations Department and is anticipated to have minimal impact upon the Nuclear Plant Construction Department's activities directed toward completing final construction requirements and plant startup. The operating plant security plan will have a phased implementation process which covers the initial security for startup with increasing requirements until meets the it security needs for operating Unit 1.

While construction forces are completing the work required to startup Unit 1, and at the same time continuing work on Unit 2, a construction security force will continue to be maintained on the plant site. This construction security force will continue to control the entrance and exit of construction personnel, equipment, and materials to and from the plant site in construction areas. Prior to receipt of the Operating License for Unit 1, the operational plant security organization will be trained and established in accordance with the provisions of the Operating License, and will have the following general responsibilities: controlling all access to Unit 1 areas, monitoring alarm and surveillance systems, providing foot and vehicle patrol of protected and vital areas and the Owner Controlled Exclusion Area, and maintaining liaison with local law enforcement agencies. Construction supervision will cooperate with plant security forces as necessary to provide security orientations, briefings, and training to construction craft, craft supervision, and constxuction management personnel.

During the time when both construction security and plant security forces are on site, each force will have very specific duties with necessary coordination achieved between them.

Construction forces will comply with applicable "Security Procedures" which are adopted for the operating plant.

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5.0 Additional Information 5.1 Recruitin Pro ram The overall proposed staff for SHNPP is now estimated to be 700 plant personnel by the end of 1988. The anticipated buildup of this organization will extend over a ten-year period, as shown on Pigure 12.

The necessity of a well developed, detailed, and productive recruiting program to accomplish this type of hiring program has been recognized. In order to take advantage of the existing plants as training bases for the buildup of the Harris staff, beginning in 1979, personnel have been hired into the Harris organization but temporarily assigned to various operating plants for training purposes. At various times, these personnel or other trained personnel from those plants will be transferred to the Harris Plant organization on site.

Details of this proposed program are described below.

The Company has developed an aggressive recruiting program at colleges and universities. During the academic" year 1982-1983, 25 visits were conducted at 20 colleges and universities. In the academic year 1983-1984, visits have been scheduled at four-year engineering institutions. In addition to visits at education institutions, naval bases will continue to be visited in 1984 in order to add to the complement of experienced nuclear personnel.

The Company is also seeking to recruit technician level personnel from technical schools and community colleges. In 1983, 26 community colleges and technical institutes were visited for recruiting purposes. The Company actively participates in programs designed to enhance the Company's image on two-year campuses. For example, in 1983, personnel participated in "caredr days" at the community colleges and technical institutes.

The Cooperative Education Program, established at eight four-year and six two-year institutions, and summer employment provide vocational training to students, and serves as a means of identifying potential employees.

Advertising on 'the local and national levels in newspapers and professional/technical publications provide emphasis on recruiting engineering and technical personnel.

The Company's Wage and Salary Program is competitive with other utilities and industries on the local and national levels, which attracts entry level and experienced personnel for operating plants. An annual review is made of the Salary Program in order to reflect changes in the cost of living and other factors to keep the program competitive. Entry level professional and technical salaries are evaluated twice annually and ad)ustments made over the first two-year period.

With the recent reorganization in the Employee Relations Department, the Area Personnel Relations function at the Harris Plant has been reassigned to the Personnel Relations-Nuclear Plants Section. This unit will provide overall Personnel Relations supporting the recruitment of construction and operations personnel. The unit will provide the onsite coordination of all recruitment activities at the Harris Nuclear Prospect. The Area Personnel Relations office at Harris will be involved in recruitment efforts at college campuses, 5-1 7390SNP

0 0

technical schools, and naval installations. As a result, more attention will be focused on the recruiting efforts to provide a quicker response to filling a vacancy and to provide improved coordination with onsite managers and corporate recruiting.

College recruiting is an:on-going program designed to attract qualified college graduates for entry level engineering, technical, and professional non engineering positions. From September 1982 to April 1983, Company recruiters have visited the following college and universities: Mississippi State, Alabama, Kentucky, Pennsylvania State, North Carolina State, North Carolina, Virginia, Virginia Polytechnical, South Carolina, Clemson, Purdue, Memphis Sate, Duke, Florida, North Carolina Central, Georgia Tech., Tennessee, North Carolina A&T, St. Augustine, 'and UNC at Wilmington.

Technical school and community college recruiting will assure the Company of a continuous supply of qualified technical graduates with at least two years of training. Schools in the following states are visited on a routine basis:

North Carolina, South Carolina, Virigina, Tennessee, and Florida. In addition, the Company actively participates in programs designed to enhance the Company's image on two-year campuses. Some of these programs include:

Industrial Workshop, Career Day, Job Fair, and Workshops for high school guidance counselors on career opportunities for two-year graduates.

The Company is deeply involved in locating, screening, and recruiting experienced personnel. The advertising program includes publications such as Engineering News Record, Power Magazine, Nuclear News, IEEE Spectrum, ASME Publication, Navy Times, and other military newspapers. Local newspapers are used whenever a supply of qualified personnel is made available due to industrial layoffs or job changes. The Company continuously encourages referrals by employees. Particular emphasis is placed on engineering and technical personnel.

The Company is an equal opportunity/affirmative action employer and has a number of special programs designed to attract qualified minority, female, and handicapped personnel.

Staffin the Harris Plant Or anization The present and projected staffing level for the Harris Plant staff 'is shown on Figure 12 and is based on an evaluation of the staff needs in con)unction with the startup schedule. Recruiting of personnel for the plant is already active as previously discussed and will continue on a schedule designed to ensure that personnel can be placed in training classes, training positions, or other positions of responsibility far enough in advance to ensure that sufficient numbers of qualified personnel are available to staff the Harris organization when needed to perform the required startup functions; develop the required operating, maintenance, and administrative procedures; operate and maintain systems following acceptance by startup personnel; and provide

'for safe operation of the various units while other units are being completed or started up.

5-2 7390SNP

PRESIDENT CHIEF EXECUTIVE OFFICER d CHAIRMAN EXECUTIVE VICE PRESIDENT POWER SUPPLY d ENQINEERINQ d CONSTRUCTION GROUPS SENIOR VICE PRESIDENT SENIOR VICE PRESIDENT SENIOR VICE PRESIDENT POVIER SUPPLY FUEL d MATERIALS ENGINEERING d GROUP MANAGEMENT GROUP CONSTRUCTION GROUP VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT VICE PRESIDENT MANAGER CORPORATE NUCLEAR OPERATIONS FUEL DEPARTMENT NUCLEAR PLANT BRUNSWICK NUCLEAR CORPORATE NUCLEAR SAFETY d DEPARTMENT CONSTRUCTION PROJECT QUALITY ASSURANCE RESEACII DEPARTMENT DEPARTMENT DEPARTMENT MANAGER GENERAL MANAGER VICE PRESIDENT VICE PRESIDENT GENERAL MANAGER MANAGER CORPORATE HARRIS PLANT MATERIALS NUCLEAR PLANT BRUNSWICK PLANT QUALITY ASSURANCE/

HEALTH PHYSICS MANAGEMENT ENGINEERING OUAI.ITY CONTROL DEPARTMENT DEPARTMENT HARRIS PLANT MANAGER GENERAL MANAGER MANAGER MANAGER CORPORATE ROBINSON PLANT ENQINEERINQ d QUALITY ASSURANCE/

NUCLEAR SAFETY CONSTRUCTION QUALITY CONTROL BRUNSWICK BRUNSWICK d ROBINSON PLANTS MANAGER MANAGER Rf SEARCH TECHNICAL SERVICES MANAGER DEPARTMENT QUALITY ASSURANCE SERVICES Figure I Carolina Power d LIOhl ComnanY

PLANT OENERAL MANAOER A8SISTANT TO OENERAL MANAOE4 MANAOER DIRECTOR MANAOER MANAOER MANAOER ADMNIISTRATIOH PLANNINO j SCHEDULINO TECHNICAL SUPPORT PLANT OPERATION8 START UP 8 TEST MANAOER DIRECTOR MANAOER ADMINISTRATIVE ENVIRONMENTAL MANAOER START UP REOULATO4Y OPERATION8 SUPERVISOII RADIATION MAINTENANCE SUPERVISORS COMPLIANCE COH'TROL SRO EME4OENCY ENVI4ONMENTAL MAINTENANCE OPERATIONS ENOINEERINO CHEMISTRY SUPERVISOR SUPERVISOR PREPA4EDNESS SUPE4VISORS SUPERVI8OR ELECTRICAL SRO PROJECT P 4 IN C I PA L SPECIALIST LIAINT'EHANCE SECURITY ENOIHEER SUPERVISO4 SIIIFT FOREMAN ENVIRONMENTAL SUPPORT ME C I I A HID A L 8 CI<EMISTRY 8RO SENIOR CONT4OL RADIATIOH PROJECT OPERATORS CONTROL E N CINE E R S SRO SUPERVISOR

~

LEOEND Admlnl ~ Iratlve Or@ants ~ lion

---Lines ol Communlcsllon PROJECT CONTROI.

OPERATORS RO SRO Senior Reactor Operators License SPECIALIST 40 Reaolor Oper ~ lors Lloense 4ADIATION AUXILIARY CONTROL OPERATORS P4INCIPAL RADWASTE ENOIHEER SUPERVISOR OPERATIONS SIIIFT FOREMAN STA'8 Fleur ~ 2 Carolina Power 8 Light Company Harris Plant Stall Orpanlsntlon RW CONTROL OPERATIONS Untie I 8 2 OPERATORS E N 0 Nl E E 4 S AUXILIARY FIRE PROTECTION OPERATORS SPECIAI,ISTS

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ONIITC I I OCNCRAL SUtCRYIOOR I OASOC OENERAL CNOINCCRINO NUCLEAR OASETY UANAOCR I HARRIS tLANT ~ IANAOCR HARRIS tLANT I

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PRESIDENT CHIEF EXECUTIVE OFFICER d CHAIRMAN Sherwood H. Smith,Jr.

EXECUTIVE VICE PRESIDENT POWER SUPPLY d ENGINEERING d CONSTAUCTION GROUPS E. E. Utley SENIOR VICE PRESIDENT SENIOR VICE PRESIDENT SENIOR VICE PRESIDENT POWER SUPPLY GROUP FUEL d MATERIALS ENGINEERINQ d MANAGEMENT GROUP CONSTRUCTION GROUP L. W. fury J. M. Davis,Jr. M. A. McDullle VICE PRESIDENT VICE PRESIDENT MANAGER Plant Operations d Nuclear Fuel Site d Plant Design CORPORATE NUCLEAR BRUNSWICK I Maintenance Procucement Features Review CORPORATE SAFETY d RESEARCH NUCLEAR PROJECT I Training d Retraining Reactor Operations Plant Modlllcatlon QUALITY ASSURANCE DEPARTMENT I I Environmental d Suppoct Support Radiation Control Plant Materials Plant System Problem Dr. T. S. Elleman I P. W. Howe I Procurement Resolution Support H. R. Banks Control Board Human Factors Review Contracts Outside Contractor Plant Chemistry d Supervision Radlochemlslry Nuclear, Thermal Oil-Site Nuclear Plant Opecallons d I Industrial Security hydraulic, Stcuclural, Corporate Quality Solely Review I Maintenance Regulatory CompHance Instrumentation d Assurance Audits On-Site Nuclear Salety I Plant Modlllcatlons Emergency Planning Controls, Process Quality Assurance Review Plant System Pcoblem RolueHng Operations Computer, d Electrical Englneeclng d Senior Management Aesolutlon Technical Support Plant Operating NRC Bullellns d Ordocs d Mechanical Nuclear Saloty Response Engineering Support OualHlcatlon d Audit ol Oversight Experience Review Vendor d Conlcactor Plant Modillcatlons Corporate Radiation Plant Systont Problem Activities Protection Pollcios d Rosolullon OA/OC Training Programs Nuclesc Licensing Envlronmonlsl Monltocing Meteorological d Seismic Monitoring Anslyllcal Chemistry d ON-SITE Metallurgical Laboratory FSAR Prepscstlon Figure 4 CacoHna power d Light Company Oll-Site Nuclear Operations Support

EXECUTIVE VICE PRESIDENT POWER SUPPLY h ENOINEERINO CONSTRUCTION 04OUPS E.E. Ulley SENIOR VICE P4ESIDEHT SENIOR VICE PRESIDENT SENIOR VICE 44ESIDENT ENOINEERIHO S POWER SUPPLY CROUP FUEL S MATERIALS COHSTRUCTION CROUP MANAGEMENT CROUP M. A, McDulll~ L.W. Eury J.M. Daul ~, Jr.

REFE4 TO FIOURE 8 MANAOER VICE PRESIDENT MANAOER VICE PRESIDENT VICE P4ESIDEHT CORPORATE DUALITY NUCLEAR OPERATIOrlS TECHNICAL SERVICES FUEL DEPARTMEHT ATERIALS MANAOEMEN SSURAHCE DEPARTMEN DEPARTMENT DEPARTMENT DEPARTMENT ILR. Danke S.J. Furr W.J. Hurlord R.A, Watson W.B. Itlncald MAHAOER MAHAOER MAINTENANCE NUCLEAR FUEL SUPPORT SECTION SECTION MAHAOER ADMINISTRATIVE MANAOER MAHAOER MANAOER DIRECTOR MAHAOER DUALITY ASSURANCE/ h TECHNICAL RADIOLOOICAL S NUCLEA4 LICENSIHO & EMEROENCY ENVIRONMENTAL DUALITY CONTROL SUPPORT CHEMICAL TRAINIHO PERMITS SECTION PREPAREDNESS TECHNOLOOY HARRIS PI,AN' SECTION SUPPORT SECTION UNIT SECTION SECTION MANAOER DUALITY ASSURANCEI HEALTH NUCLEAR h NUCLEAR OUALI'TY COH'THOL PHYSICS SIMULATOR I.IC E 4 SINO BRUNSWICK g UNIT TRAIHINO UNIT UN I' ROSIHSDN PLANTS CRAFT SPECIAL EHVIRONMEHTAL MANAGER TECHNICAL NUCLEAR UNIT DUALITY ASSURANCE TRAINIHO UNIT PROORAMS UNIT SERVICES CUR4ICULUM CHEMI STR Y DEVELOPMEHT PERMITS UNIT UNIT UNIT DIRECTOR TRAIHIHO HARRIS DIRECTOR Figure S TRAIHIHO Carolina Power h Light Company DRUNSWICK Duality Assurance. Power Supply a Fuel snd Materials Management Oroupe DIRECTOR OH-Sll ~ Support Organlxatlon T4 A IHIrlo ROSINSON

SENIOR VICE PRESIDENT ENGINEERING 4 CONSTRUCTION GROUP M.A. McDuffle VICE PRESIDENT VICE PRESIDENT NUCLEAR PLANT NUCLEAR PLANT CONSTRUCTION ENGINEERING DEPARTMENT DEPARTMENT A.B~.Cutter S.D ~ 8mlt h MANAGER PROJECT GENERAL MANAGER ENGINEERING HARRIS SITE HARRIS PLANT SECTION MANAGEMENT SECTION MANAGER MANAGER ENGINEERING SUPPORT ROBINSON CONSTRUCTION NUCLEAR PLANTS SECTION MANAGEMENT SECTION DIRECTOR MANAGER NUCLEAR ENGINEERING CONSTRUCTION PROCUREMENT SAFETY REVIEW 5 CONTRACTING SECTION Figure S Carolina Power 8 Light Company Englneerlng 5 Construction Group Off-Site Management Resources

PRESIDENT CHIEF EXECUTIVE OFFICER

& CHAIRMAN Sherwood H. Smith,Jr.

EXECUTIVE VICE PRESIDENT POWER SUPPLY L ENGINEERING CONSTRUCTION GROUPS E. E. Ut Icy VICE PRESIDENT CORPORATE NUCLEAR SAFETY RESEARCH DEPARTMENT T. S. Elleman MANAGER MANAGER MANAGER RESEARCH SECTION CORPORATE NUCLEAR SAFETY CORPORATE HEALTH PHYSICS SECTION SECTION SPECIALIST CORPORATE NUCLEAR SAFETY DIRECTOR DIRECTOR DIRECTOR DIRECTOR NUCLEAR SAFETY REVIEW ON-SITE NUCLEAR SAFETY ON-SITE NUCLEAR SAFETY ON-SITE NUCLEAR SAFETY (BSEP) (HBR) (SHNPP)

Figure 7 Carolina Power 8 Light Company Independent Safety Revlow Groups

BOARD OF DIRECTORS I

LEGEND I AEMINETRATIVE I PRESIDENT CHANNELS CHIEF EXECUTIVE OFFICER & CHAIRMAN

~ ~ ~AUTHORIZED DIRECT Sherwood H. Smith,Jr.

C 0 M M UN I C A T I ON CHANNELS EXECUTIVE VICE PRESIDENT POWER SUPPLY & ENGINEERING

& CONSTRUCTION GROUPS E.E. Utley MANAGER VICE PRESIDENT CORPORATE NUCLEAR SAFETY &

QA DEPARTMENT RESEARCH DEPARTMENT H.R. Banke T.S. Elleman fSI I I I

I I I I I I I

MANAGER QA/QC MANAGER MANAGER GENERAL MANAGER BRUNSWICK QA/QC 'NUCLEAR MANAGER I CORPORATE ROBINSON HARRIS PLANT TRAINING t HEALTH

'HARRIS PLANTS I PHYSICS MANAGER MANAGER GENERAL GENERAL MANAGER QA ENVIRONMENTAL MANAGER MANAGER CORPORATE SERVICES RADIATION NUCLEAR ROBINSON BRUNSWICK CONTROL SAFETY Figure 8 Carolina Power & Light Company SENIOR MANAGEMZNT OVERSIGHT FUNCTIONS

P 4 E 8 ID E 4 1 CHIEF EXECUTIVE OFFICER 6 CHAI4MAN Sherwood H. Smith, Jr.

EXECUTIVE VICE PRESIDENT POWER SUPPLY 6 ENOINEERINO 6 CONSTAVCTION OAOVPS E. E. Ulley SEHIOR VICE PRIEIDENT SENIOR VICE PAESIDENT SENIOR VICE PRESIDENT FUEI. 6 MATERIALS EHOINEERINO 6 POWEA SUPPLY CROUP CONSTRUCTION CROUP MANAOEMENT CROUP L. W. Eury J. M. Dsvts Jr. M. A. McDulll~

VICE PRESIDENT MANAOER VICE PRESIDENT VICE PAESIDENT VICE PRESIDENT VICE PAESIDENT NUCLEAR PLAHT CORPORATE NUCLEAR SAFETY 6 SRVNSWICK NUCLEAR OPE4ATIONS FUEL DEPA4TMENT CONSTRUCTION DUALITY ASSUAANCE RESEARCH DEPARTMEHT NUCLEAR PLANT DEPARTMENT DEPARTMENT DEPARTMENT S. J. Furr R A~ Watson S D Smith H. A. Banks Or. T. S. EHeman P. W. Howe NUCLEAR FUEL AUTHORIZED: A VTHORIZED AUTHOAIZEDI AUTHORIZED: AUTHORIZED:

44 Prolesslonsls 362 Prolesslonals 263 Prolesslonals SECTION ONLY: 103 Prolesslonals 04 Pro I ~ s ~ toasts 36 Dsyrsed Engineers 147 Deyreed Engineer ~ 138 Degrsed Enylneer ~ 06 Deyreed Engineers 10 Degreed Engineers AUTHORIZED:

ACTUAL: ACTUAL: ACTUAL: 30 Prol ~ salon ~ Is ACTVALI ACTVALI 44 Prolasslonalx 320 Prolesslonsls 243 Protesslonals 28 Degreed Engineers tyd Prolesslonals 82 Prot ~ sslonals 130 Degresd Engineers SS Degreed Enplneers 31 Dsgreed Engineers 33 Depreed Engineers 120 Depreed Enylneers ACTUAL:

Y ~ srs ol 3088 Total Years ol 2021 Tol ~ I Y ~ ars ot 1436 Total Years ol ddd Total Years ot 4184 Tol ~ I 28 Prolesslonals Experience Experience Experience 24 Degresd Engineers Experience Experience 223 Tol ~ I Years ol Experience VICE PRESIDENT MANAOEA NUCLEAR PLAN'T TECHNICAL SERVICES E NO IN E E 4 INC DEPARTMENT DEPARTMENT W. J. Hurtord A 6 Cutl ~ r AUTHORIZED: A VTHORIZEDI 211 Prot ~ salonals 126 Prot ~ sslonsls 31 Degrsed Engineers 100 Depreed Enplneers ACTVALI ACTVALI 171 Prot ~ s ~ lonsl ~ 100 Prol ~ aston ~ I ~

43 Dagraed Enylneers 04 Degreed Engineers 1666 Total Y ~ srs ol 1268 Total Years ol Experience Ex per lane ~

Flyer ~ 0 Carolina Power 6 Llyhl Company Nuclear Support Slstl Aeaources ss ol June. 'I083

SITE EMERGENCY COORDINATOR DIRECTOR DIRECTOR. DIRECTOR RADIOLOGICAL CONTROLS PLANT OPERATIONS LOG ISTIC S Figure 10 Carolina Power 8 Light Company Harris Technical Support Center Organization

~ Operation

~~~ Information CORPORATE MANAGEMENT REGULATORY MANAGER CORPORATE AGENCIES CORPORATE EOC MEDIA CENTER l l I I I I I I I I I COUNTIES I

I I I I I I l MANAGER CORPORATE STATE EMERGENCY RESP, SPOKESMAN I I

I I

SITE MANAGER MANAGER MANAGER PLANT EMERGENCY TECHNICAL ADMINISTRATIVE RADIOLOGICAL MEDIA CENTER

.COORDINATOR ANALYSIS LOGISTICS CONTROL Figure 11 Carolina Power 5 Light Company Emergency Response Organization

700 800 60G 400 3GO 200 100 1980 1981 1982 1983 1984 1986 1986 1987 1988 YEAR LEGEND ACTUAL PROJECTED Figure 12 Carolina Power 8 Light Company Shearon Harris Plant Staff

CORPORATE NUCLEAR SAFETY PROGRAM POLICY STAT&fENT It is the policy of the Carolina Power & Light Company to design, construct, and operate nuclear power plants without jeopardy to its employees or to the public health and safety. Nuclear safety programs shall be developed, implemented, and updated as necessary to assure that the Company's nuclear generating units will be managed such that all plant systems used to treat, store, or convey waste produced by the generation of nuclear steam will be designed, const ucted, and operated in a safe manner. Deviations from these programs shall be permitted only upon written authority from the corporate management position origiaally approving the program.

The design, construction, and operation of nuclear plants shall be accomplished in accordance with U. S. Nuclear Regulatory Commission (NRC} regulations specified in Title 10 of the U. S. Code of Federal Regulations. All commitments to the NRC Regulatory Guides and to engineering and construction codes shall be carried out.

The operation of the Company's nuclear power plants shall be in accordance with the tems and conditions of the facility operating license issued by the NRC.

Any changes in operating procedures, experiments at the facility, modifications to the plant hardware or systems, shall be made in accordaace with the terms and condit'oas of the facility operatiag license.

The corporate nuclear safety (CNS) section of the nuclear safety & research department shall monitor the Company's auclear programs on a continuous basis to assure they are being carried out ia an effective manner. The CNS section shall implement the nuclear safety functions as defined in ANSI N18.7 and as required by nuclear plant safety aaalysis reports, technical specific "ioas, nd K3C Regulatory Guide 1.33.

The vice president - nuclear safety & research department and the manager - CNS section shall review with the senior operating officer of the Comoany with the ultimate responsibility f'r the operation of all nuclear power plants on a regular periodic basis the overall effectiveness of the Company's nuclear safety programs. They shall be expected to communicate directly with corporate manage-ment up to and including the chief executive officer aad if board of directo s to resolve any nuclear safety-related concerns appropriate with the if the concerns cannot be resolved satisfactorily at a lower management level.

Issued by: Date:

Sherwood H. Smith, Jr.

Chairman/President Chief Executive Officer EXHIBIT 1.

CAROLINA POWER & LIGHT COMPA'K CORPORATE QUALITY ASSURANCE PROGRAM POLICY STATEMENT It is the policy of the Carolina Power & Light Company to design, construct, and operate nuclear power plants without jeopardy to its employees or to the public health and safety. The quality assurance programs shall be developed, implemented, and updated as necessary to assure that the Company's nuclear generating units will be managed such that all plant systems used to produce, convey, or use nuclear generated steam and all plant systems used to treat, store, or convey waste produced by the generation of nuclear steam will be designed, constructed, and operated in a safe manner. Deviations from these programs. shall be permitted only upon written authority from the corporate management position originally approving the program or implementing pxocedures.

The design, construction, and operation of nuclear plants shall be accomplished in accordance with U. S. Nuclear Regulatory Commission (NRC), regulations speci-fied in Title 10 of the U. S. Code of federal Regulations. All commitments to the NRC Regulatory Guides and to engineering and construction codes shall be carried out.

The operation of the Company's nuclear power plants shall be in accordance with the terms and conditions of the facility operating license issued by the NRC.

Any changes in operating procedures, experiments at the facility, modifications to the plant hardware or systems, shall be made in accordance with the terms and conditions of the facility operating license.

The corporate quality assurance department. shall monitor the Company's nuclear programson a continuous basis to 'assure they are being carried out in an effective manner. The corporate quality assurance department shall implement the quality assurance audit functions as defined in 10CFR50, Appendix B, ANSI N18.7, and as required by nuclear plant safety analysis reports and techni-cal specifications, Company approved ASME programs, and NRC Regulatory Guide 1.33.

The manager << corporate quality assurance department shall review with the senior operating officer of the Company with the ultimate responsibility fox the opera-tion of all nuclear power plants on a regular periodic basis the overall effec-tiveness of the Company's quality assurance program. He shall be expected to communicate directly with corporate management up to and including the chief e:cecutive officer and if appxopr'ate with the board of directors to resolve any if quality assurance related concerns the concerns cannot be resolved satis-factorily at a lower management level.

The engineering & construction quality assurance (E&CQA) section of the corporate quality assurance department shall implement the Corporate Quality Assurance Programs and the approved ASME QA Programs assuring effective implementation by surveillance of portions applicable to the engineering & construction group.

EXHIBIT 1A

The operations quality assurance (OQA) section of the corporate quality assurance department shall implement the Corporate Quality Assurance Programs assuring implementation by surveillance of portions applicable to the power supply group. The OQA section shall also implement the applicable portions of the Corpox'ate QA Program as implemented by the engineering & construction group for nuclear plant modifications.

The managers of all functions involving nuclear fuel, eng'neering and construc-tion, and operations shall assure that their personnel are adequately trained for their jobs and they have the experience and education required to carry out their assigned responsibilities. Personnel who habitually or willfullydisre-gard or violate the nuclear safety and quality assurance policies and procedures will be subject to disciplinary action.

Issued by: Date:

Sherwo'od H. Smith, Jr.

Chairman/President Chief Executive Officer EQilBIT 1A

Caroiina Power & Light Company

~ lhlUf~~  %&4a tP%4a&ell~a P. O. Box 1%1 ~ Raleigh, N. C. 27602 J~ A. JONES Senior Executive Vice President Chief Operating Oificer Apzil 15, 1981 HELORAiKUM TO: Hessrs. T. S. Elleman M. A,. HcDuffie E. E. Utley PROM: J. A. Jones Recent organizational changes make it necessary to update the "assigned responsibilities of the Vice President Nuclear Safety & Research to meet per odically with Company personnel concerning nuclear sa ety and quality assurance.

This memorandum confirms our understanding that the Vice Pzesident - Nuclear Safety & Reseaxch will meet periodically with the individuals occupying the positions listed later in this memorandum to review nuclear safety and quality assurance matters. These discussions will encompass matters outside of normal nuclear safety and quality assurance reports and could include (but not necessarily be Umited to) such items as trends in quality assurance which could lead to safety problems, generic quaLLty assurance and safety problems in the nucleax'ndustry, long-tean concerns requiring extended or intensive investigation affecting CP&L's quality assurance or nuclear programs, attitude px'oblems which could adversely affect implementation of CP&L's programs, or any personnel concerns related to potential safety problems.

These meetings will give the individuals whose total responsibility is nuclear safety and quality assurance an additional avenue to egress concerns to the Chief Executive Officer of the Company if necessary to get them satisfactorily resolved.

It will be the responsibility of the Vice President Nuclear Saxety &

Research to init:iate these meetings on a frequency of no less than once each quarter. If, however, a meeting is essential to resolve any quality assurance or safety issues that require immediate attention, individuals occupying the positions U.sted in this memorandum are responsible for contacting the Vice President - Nuclear Safety & Research immediately.

In the event of his absence in these situations; or in the event the Vice President - Nuclear Safety & Research does not respond satisfactorily, the Chic" Operating Officer or the Chief Executive Officer should be contacted.

EXHIBIT 2

This memorandum is applicable to the individuals occupying the following positions:

Manager - Corporate Quality Assurance Manager - Oper'ations Quality Assurance/Quality Control Manager >> Engineering and Construction Quality Assurance/Quality Cont ol Manager - Nuclear Training Manager - Environmental and Radiation Control Manager - H. B. Robinson Plant Manager Brunswick Steam Electric Plant In addition, the Vice President for Nuclear Safety 6 Research may elect to meet with plant personnel having responsibilities for operations, maintenance, training and health physics as he deems necessary.

TSE/phT6 cc: Messrs. H. Banks J. D. E. Jeffries J. R. Bohannon C. H. Moseley A. L. Cutter S. D. Smith C Dietr. S. H. Smith, Jr.

L. W. Eury R. B. Starkey B. J. Furr B. H. Webster P. W. Howe EXHIBIT 2

PRL Caroiicm Power 8 Light Company P. O. BOx I&1 ~ AaknIIn. N. C. 276Q2 SHERWOOD I":, SMITH. JR.

Presa en<

Mar h 17, 1980 RANDU.. TO'- Dr. Thomas S. ~Ieman, Vice President Hucleaz Sa "aty and Research PROP:. Shazwood B. Smith, Jr.

Sb~:-CT: Nuclaa= Sazaty and Quality Assurance Fetters

"'oc" iMS me orandm T 'll conf'u unde standing that you vi3.l mea pe cally with me to discuss nuclear safety and qua>> ty assurance matte s.

These d'scuss'cns may encompass matte s outside of routine nuclear safety and qt"'ity assu=ance aud'nd surveillance repor s, as well as appropriate items i" those reports. These reviews should include, but not be lim.tad to, sucn it~ as trends in qua1ity assu ance which could lead to safety conca m, o.enaric quaXity assurance and safety problems in the nuclear indust~, any t t"a s requ'g extended or intensive investigations affec~g CP&L's quality su ance or nuclear safety prog ams, the att" ides and morale r of pe sonnel involved in the implementation of CP&L's prog ams, or any other personnel conca=s related to safety matte s.

Since you total responsibility includes all aspects of nuc'aar safety from a corporate standpoint with no other assigned duties that might con lier with th's inta est, I am certain that your reviews will provide an objective evaluation oz the effectiveness of nuclear safety, qua'~~ assurance, and ALARA prog ams. I will also expect to be informed about the conca~ of other person-nel working in areas of nuclear sa ety, quality assn=ance, and health phys" cs, there are any safety matt'ers not being satisfactorily addressed by othe" levels of management.

Our conversat"'ons, of cou se, wild. not be intended to rap'aca,or alter, the regular and rourine octan reports on nuclea" safety matters, which a e provided to ma.

Zt wM~ be vour asponsibi~'w to initiate these meetings on a =equency of no.3.ess thaz: semiannually. Of cou se, however, should s" tuations ar"'se that you fee'houlc be brought to my attention immediately for satisfactory esolution, I will expect you to contac" me promptly and make me aware of the urgency o the s" "uatioa;

~ORANG)UM TO: Dr. Thomas Mrch 17, l980 your judgment, Ae response to your e~ressions of concern "s nott satisfactory =rom a safety standpoint, it will be you" respons'b'~ to make the Chai~n of the ~orecasting, System Development and Finance Com='ttee of the Boa"d oz D'=.e tors aware of your concern or in the absence of dividual you may contact any member of the Board.

Since the Dire tors oz our Company a=e tiveness of our nuclear, sa"ety programs, vie it vol h.">>ested in be your the responsibmty e""ec-to appear bezo e the:o ecasting, System Development and Finance Comm'<<=ee oz the Boa-d annua" y to provide a prof essiona'I evaluation of ou- programs. P~ange-ments w&l be made zoz you to meet wi.A this Ccnnmittee periodically dur"ng scheduled meetings, and when appropriate to appear also be=ore the its'egularly ent='e Board. '

Sherwood H'. Smith, Jr.

SBS, Jr.:pt

~

0 CC: Boazd of Directors

~

Senior management Ccmmd.ttee

C C,grolier>.> I'>>vii g c l II)ltd (' ~ e ~

i(>>t>y

~ I, I l'L April 22, '981

)MlOR 4NDU'1 TO: Mz. E. E. Utley Mz. J. M. Davis, Jr.

Dr. T. S. Elleman Mr. L. M. Eury Mr. M. A. McDuffie Mr. 'rl. J. Ridout, Jr.

FROM: J. A. Jones

SUBJECT:

Corporate Health Physics Policy Attached 's a revised copv of ouz Corporate Health Physics Policy. The Policy Statement is appl'caole to all health physics programs which aze establ'shed to engineer, construct, and operate nuclear power plants on our system.

This Policy supersedes the Corporate Health Physics Policy attached to my memorandum dated January 11, 1980. Please take measures which you deem appropriate to disseminate the revised Policy Statement t'o appropriate personnel under your supervision to assure that they are fully aware of the Policy Statement and the commitments contained therein.

. JAJ/dwjW2 Attachment cc: Mz. H. R. Banks Mr. R. L. Mayton, Jr.

Mr. Sherwood H. Smith, Jr.

EXHIBIT 4

Carolina Power B Light Company CORPORATE HEALTH PHYSTCS POLTCY Ia line with the overall policy of Carolina Power & Light Company to eng'neer, construct, and operate nuc?.ear power plants w'thouc jeopardy to public he"'n and safety, it is the. policy of the Company to develop, implement, and maincain sound health physics "programs at each Company faci3.ity where radiac'on produc'ng equipment and/or radioactive materials are used oz stored. The health physics progra s shall be structured to ensure that the exposure to radiation of Company personnel, contractor personnel, and the general'ublic will be maintained at levels which are as low as reasonably achievable (AL~GA) and cons'scent with United States Nuc'ar Regulatory Commission Regulations in Title 10 of the United States Code of Pede a3. Regu3.ations. The heal h pnysics programs associated with activities licensed by state regu'atory agencies shall be developed to comply with applicable state regulations.

The health physics programs developed by the Companv shall ensure tha" personnel, che gene al public and the off-site environs a e proteccec, anc procedures and records syste=s are establ'shed to meet a13. appl'cab'e federa3.

or state regulac'ons.'ach Companv employee and concractor personnel working in a facility where exposure to zad'ation m3.ghc occu- shall make eve y zeasonab3.e eifozt to ma'ntz'n radiat'on exposures and releases of zadioact ve materials to unr'est icted azeas as far below specified limits as reasonably ach.evable.

Personnel who habitually or wil3.fully disregard oz violate health physics procedures and practices will be subject to disciplinary action.

Health. physics programs shall be developed which wi13. be stzictly adhered to by the Company'and its contractors to lim.'t occupational exposuzes ac Company facilities in which exposure to radiation may occur to ALARA levels. The health physics programs shall be documented in writing and shall be reflected in written administrative procedures and inst.uctions for operations involving potentia3. exposures of personnel to radiation and for design activities assoc" ated wi"h each faci3.ity. Instructions to designers, constructors, vendozs, and facil'cy personnel responsible for specifying or zeviewing facility features, systems, or equipment shall reflect the health physics programs goals and obje'ct ves The goals and objectives of the health physics programs shall be to maintain the annual dose to individual facility personne3. to as low as reasonably achievable and to maintain che annual incegraced dose to facility personnel; i.e., the sum of annual doses (expressed in man-zcm) to all facility personnel, as low as reasonab3.y achievable. The health physics programs shall identify the organizations participating in the programs, che positions invo3.ved, and t)te rcspottsibilicies and func" ions of thc various pos'tions in conducting the programs ~

411 Fayettt.vittt. Street ~ P 0 t3oa t'St ~ -Aalt.ion. N C 27602

~ ~ Ir ~

g,

Thc design of nuclear facilitics shall bc consistent with the goa's and object'ves of the hca'h physics programs. Modif ications to ex'sting nuc3c" r fac'litics shall bc designed and implemented in compliance witn the health physics programs to meet i'KARA rcquiremens. Design review shall re cct cons'ceration of the ac" ivi"ies of fac. ity personnel such as opcrat'ons, ma'ntcnancc, refuel'ng, in-service inspection, radioact'vc waste processing, and deconta...'nat'cn.

Adequate trained personnel shall be provided to develop and conduct all necessary health physics programs. The health physics personnel sha3.1 possess the necessary training and expertise to c"rry out the health physics programs in an efficient manner to assure tha" Company and regulatory reaui.rements are met.

Thc dircczor of corporate health physics will make hi...self avai.3.able to all Company personnel for advice and consultation on matters relating to health physics.

A13. health physics programs sh-13. require procedures, job planning, record keeping, special equipment, operating philosophy, and other suppor" conduc've to meet ALW requirements. Proper prepa ation and plann.'ng shall be performed prior to entering radiation areas whe.e significant doses could be rece ved.

Adequate supervision and radiat'on protection surveillance shal'e prov'ded during operations in radiation areas to ensure that the appropriate procecures arc followed, that planned precaut'ons are observed, and that al'otential radiat'on h""ards which m ght deve'op dur'ng the operat'ons are considered in a time'y manner. Results of ac ivities in radiat o.. zones sha2.1 be analyzed to ident' deficiencies in the progr m and to provide the basis for revi.sing procedures, mod'fying facility features, or make other adjus ments which may reduce exposures during subsequent. act'v.'t'es.

Health physics facilities, instrumentation, and protective equipmcnt shall be adequate to permit the staffs to function effectively. Thc selection of in trumcntat'on and equipment and the quantitics provided shall bc adequate to meet the ant'c.'pated needs of the facilities dur"ng nor..~3. operations, major outages, and accident conditions.

Appropriate training programs in thc fundamentals of radiation protection and facility cxposurc control proccdurcs shall bc established to prov dc instructions to all facility personnel including contractors whose dut'es requ're working in rad'ation areas. Training programs for health physics personnel shall bc provided to improve their performance in thc health physics programs.

Appropriate health physics programs shall bc established for all Company operation:

which deal with radiation. The programs shal'c consistent with thc corporate health physics policy and all applicable rcgulat'ons. The director of corporate health physics shall period.'cally evaluate the various health physics

.programs and other Company activ'c" which have impacts on thc programs and rcport to senior managcmcnt regarding thc cifectivcncss and adcquac; of thc programs Thc director of corporate health physics sha3.1 'akc recommend't'ons to senior management as necessary to maintain effect'vc ovcra3.1 health physics programs.

The director of corporate health physics shall review with thc senior operating officer of the Company with the ultimate responsibility for the operation o all nuc3.ear power plants on a regular periodic.basis thc overall effect'veness of the corporate health physics progra-.s. He sha3.1 be expected to co-,.=.unicate directly with corporate a:anagcrent up to and including the ch'ef exccut've off'ccr to resolve any concern in thc area of corporate health physics if the concern cannot bc resolved satisfactorily at a lower nanager.:cnt level.

Issued by& .

~~~a,.~.f'P', h Jlrc "+'i' Date: ( -Cy~o c Sherwood H. Smith, Jr.

Chairman/President Orig.'. 6-17-77 Chief Executive Officer Rev. 1: 1-'-80 Rev. 2: 4~21 Q3 EZHT.BIT 4

LIST OP OPEN ITE fS, REVIEW BRANCH AND REVIEWER Containment Systems Branch/ J. Huang Open Item 63 Core Performance Branch/J. Voglewede Open Items 31, 313 Environmental and Hydrologic Engineering Branch/R. Gonzales Open Item 14

• Instrumentation and Control Systems Branch/H. Li.

Open Items 82, 94 Meteorological and Effluent Treatment Branch/J. Hayes Open Items 160, 163, 179, 18l., 229 Power Systems Branc'n/0. Chopra Open Item 344 Quality Assurance Branch/J. Spraul Open Item 220 (Partial)

Reactor Systems Branch/E. Marinos Open Items 47, 49 Structural Engineering Branch/S. B. Kim Open Item 339 83081.700'30-

Containment Systems Branch/J. Huang Open Item 63

Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER)

Containment Systems Branch 0 en Item 63 (DSER Section 6.2.1.5, a e 6-14)

The staff has reviewed the applicant's input parameters used in the minimum containment pressure analysis including initial containment conditions, containment net free volume, passive heat sinks, heat transfer to passive heat sinks, containment active heat removal, and containment purge system operation, and found them all to be acceptably conservative and in conformance with BTP CSB 6-1, with two exceptions. The first exception is the incomplete listing of the thermophysical properties of the passive heat sinks used in the analysis (see NRC Question 480.25) . The second exception is the lack of demonstrated conservatism in the times assumed for post-LOCA initiation of the containment spray system and fan coolers (see NRC Question 480.26) ~

Based on the above, the staff cannot find acceptable the applicant's minimum containment pressure analysis for ECCS performance studies until the above concerns regarding the thermophysical properties of the passive heat sinks and the containment spray system and fan cooler initiation times have been satisfactorily resolved. This is an open item.

The applicant's minimum containment presure analysis for performance capability studies on the ECCS required by Appendix K to 10 CFR 50 has also been reviewed and found acceptable except for two items for which additional information has been requested (see NRC Questions 480.25 and 480.26) .

~Res onse'hrough discussions with the NRC-CSB reviewer and the NRC consultant, the response for Question 480.25 was discussed and noted as being acceptable.

Therefore, the response to Question 480.25 resolves one of the two items of concern.

With regards to the second item, please refer to the information provided in the response to Question 480.26.

uestion 480.26 Justify the times assumed in the minimum containment pressure analysis for fastest post-LOCA initiation of the containment spray system and fan coolers (listed as 42.03 sec. and 32.30 sec., respectively, in FSAR Table 6.2.1-62) by giving a conservative account of the delay times associated with the sequence of events leading to full operation of the containment heat removal systems (e.g., signal process time, diesel starting time, sequencer delay time, breaker closing time, pump startup time, time to bring fans to the operational speed, valve opening time, and spray line fill-up time)., Also, verify that these start times are consistent with or conservative with respect to the worst single active failure assumed for the limiting case for the ECCS analysis (DECLG break, CD~0.4) .

(7544PSAlcv)

0 Res onse to uestion 480.26 FSAR Table 6.2.1-62 will be revis'ed to amend the containment spray and fan coolers start times. The fastest post-LOCA initiation of the Containment Spray System with loss of offsite power is 42.94 seconds and the fastest post-LOCA initiation of the fan coolers with loss of offsite power is 21.62 seconds. Table I provides a break down of the times used to calculate the containment spray actuation time and the fan coolers to reach full speed time. These times are calculated by considering a conservative account of minimum delay times associated with the sequence of events leading to full operation of the Containment Heat Removal System.

FSAR Section 6.2.1.5.4 will be revised in a future amendment to reflect the above information.

(7 544PSAlcv)

TABLE I MINIMUM CONTAINMENT SPRAY AND FAN COOLER ACTUATION TIMES CONTAINMENT CONTAINMENT SPRAY FAN COOLER EVENT WITHOUT OFF WITHOUT OFF SITE POWER SITE POWER SIGNAL PROCESS TBiE 1.0 1.0 TIME FOR DIESEL TO REACH FULL 10.0 10.0 SPEED AND POWER AVAILABLE FOR 1st LOAD BLOCK SE UENCER DELAY 5.0 5.0 CLOSE PUMP BREAKER 0.04 0.04 TIME TO REACH FULL SPEED 1.0 5.58 FOR PRIMP or FAN CONTAIPifENT SPRAY 25.9 PIPING FILL-UP TLUTE TOTAL TL"IE IN SEC. 42.94 21.62 (7544PSAlcv)

Core Performance Branch/J. Voglewede Open Items 31, 313

Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER)

Core Performance Branch 0 en Item 31 (DSER Section 4.4.8 a es 4-47 4-49 & 4-50)

Provide the itemized documentation required by Item II.F.2 of NUREG-0737.

~Res ense The following information describes the instrumentation utilized for monitoring ICC and is organized per NUREG-0737, Item II.F.2 "Documentation Required."

Information utilized to give the operator an advance warning of the approach to ICC and to monitor the recovery from ICC, if it occurs, is obtained via a qualified instrumentation package. The information is obtained by the use of the Reactor Vessel Level Indicating System (RVLIS) and core exit thermocouples.

The RVLIS is a fully qualified and redundant system for monitoring water inventory in the reactor vessel. Each of the two channels provide differential pressure cells and transmitters for narrow and wide range monitoring over the full length of the vessel, with the reactor coolant pumps off (natural circulation) and on respectively.

Additionally, narrow range monitoring is provided for each channel of the upper plenum during natural circulation. Each channel's microprocessor utilizes these D/P signals in conjunction with other inputs such as RCS pressure, RCS temperature (loop RTD's or incore thermocouples), and RVLIS reference leg temperature sensors, to compensate for density changes in the system reference legs so as to provide direct water level readings available for operator use.

These water level readings will be displayed by redundant, qualified, alpha/numeric displays in the control room.

Incore thermocouples are utilized to determine core exit temperature. These 51 thermocouples outputs will be data li.nked to the ERFIS computer system for primary display on the SPDS CRT located on the MCB. Additionally, the thermocouple outputs are transmitted to a back up display in the control room. The display has a number of switches for selection based upon thermocouple identification and location core map.

The input to the ERFIS computer will also be used to determine the margin of saturation which can be displayed on demand on the SPDS CRT or continuously on a strip chart recorder.

The operator can also use Hain Control Room display information in conjunction with steam tables to determine the margin to saturation.

Shearon Harris Nuclear Power Plant Draft SER 0 en Item 31 (Cont'd)

~Res onse (cont'd)

(2) Design analysis and an evaluation of instruments to monitor water level, and available test data to support the design described in item 1 above may be found in NUREG CR-2628 regarding the Westinghouse design (RVLIS) ~

(3) A description of test programs conducted for evaluation and qualification of the RVLIS was provided in NUREG CR-2628. For qualification of the thermocouples, see item (4) below.

Although the system sensors and microprocessors are not directly testable at power for calibration, the calculated parameter of margin to saturation can be readily verified at power through use of the steam tables and observation of the independent indications of pressure and temperature. These observations should show higher margin to saturation since the system uses conservatively auctioneered values.

An evaluation on the conformance of ICC instrumentation to item II.F.2 Attachment 1 and NUREG 0737 Appendix 8 is provided in NUREG CR-2628 for the RVLIS. RVLIS meets the intent of Regulatory Guide 1.97.

Technical Specifications will be prepared for the instrumentation specifically installed for the detection of inadequate core cooling. The Technical Specifications will be prepared considering the recommendation of NRC's Standard Technical Specifications (STS's) for Westinghouse PWR's (Rev.4) ~ CPSL is currently reviewing the Technical Specifications in Chapter 16.0 of the FSAR in view of the recommendations of Revision 4 of the Westinghouse STS; a revision of the Technical Specifications will be submitted to the NRC in the second quarter of 1984.

The thermocouples meet the intent of design criteria outlined in XI.F.2 Attachment 1 as indicated below:

utilized for the core exit for each core quadrant

'.l Thermocouples (in conjunction with core inlet temperature data) are sufficient to provide indication of radial distribution of the coolant enthalpy (temperature) rise across representative regions of the core.

A.2 The primary display has the following capabilities:

(a) A spatially oriented core map indicating the temperature or temperature difference across the core (at each thermocouple location) is displayed on the CRT.

Shearon Harris Nuclear Power Plant Draft SER 0 en Item 31 (Cont'd)

~Res onse (Cont'd)

(b) A selective reading of core exit temperature which is consistent with parameters pertinent to operator actions in connection with plant-specific inadequate core cooling procedures, will be displayed on demand.

(c) Direct readout and hard copy capability is available for all thermocouple temperatures. The range extends from 200'F to 2300'F.

Hard copy will be provided by computer printout.

(d) Trend capability showing the temperature-time history of representative core exit temperature values is available on demand.

(e) Alarms are provided in the control room consistent with operator procedure requirements.

(f) The operator display device (CRT) interface will be located in accordance with human-factor design in order to provide rapid access to requested displays.

A.3 A backup display is provided with the capability for selective reading of each of the operable thermocouples.

Backup display is provided in the control room.

A.4 The types and locations of displays and alarms will take into account:

(a) the use of this information by an operator during both normal and abnormal plant conditions; (b) integration into emergency procedures; (c) integration into operator training; and (d) other alarms during emergencies and the need for prioritization of alarms Normal operating and emergency operating procedures are currently being developed and will be submitted at a later date. They will be available for onsite review six months prior to fuel load.

Shearon Harris Nuclear Power Plant Draft SER 0 en Item 31 (Cont'd)

~Res onse (cone'd)

A.5 The RVLIS instrumentation meets the requirements of Appendix B, "Design and Qualification Criteria for Accident Monitoring Instrumentation," as modified by the provisions of items 6 through 9 below.

A.6 The primary and backup display channels are electrically independent, energized from independent station Class 1E power sources, and physically separated in accordance with Regulatory Guide 1.75 up to and including the isolation devices. The primary display and associated hardware beyond the isolation device are energized from a high reliability power source. The backup display and associated hardware is Class 1E.

A.7 Primary and Backup displays are located in. the control room envelope. Backup display will be completely qualified in accordance with IEEE 323 (1974) and 344 (1975) as defined in WCAP 8587/8687. The isolation device is located in an area which is accessible for maintenance following an accident.

A.8 The primary and backup display channels are designed to provide 99% availability for each channel with respect to functional capability to display a minumum of four thermocouples per core quadrant.

A.9 Quality assurance meets the requirements of 10CFRSO as applicable.

(5) For a description of the computer furictions associated with ICC monitoring, refer to item (1) above.

(6) ICC Instrumentation, including associated testing, and calibration, will be installed before plant operation.

(7) Guidelines for the use of instrumentation for monitoring ICC and the basis for these procedures will be provided after their completion.

(8) A description of key operator action instructions in the emergency procedures for ICC will be provided when available.

(9) Additional information to support the acceptability of the ICC monitoring system will be provided by August 30, 1983 in the applicants response to Supplement 1 of NUREG 0737 (item R.G.1.97 Rev. 2)

(7292PSAtda)

Shearon Harris Nuclear Power Plant (SHNPP)

Draft Safety Evaluation Report (DSER)

Core Performance Branch 0 en Item 313 (NRC letter dated March 24 1983)

Table 15.6.5-2 of the SHNPP Final Safety Analysis Report (FSAR) gives a value for the F used in loss-of-coolant accident (LOCA) analysis. Section 4.3 of the FSAR (3hich should discuss all aspects of all power distributions used in Chapter 15 analyses, including in particular the power peaking factors used to satisfy LOCA analysis requirements) does not mention such a value.

Instead Section 4.3 presents only the standard Westinghouse discussions demonstrating that an F~ value of 2.32 can be maintained, using the standard Westinghouse CAOC (with improved load follow) analysis, control and excore (split detector) surveillance. The only conclusion that one can draw from this information, as it stands, is that it will be necessary to derate the reactor to 91 percent power. If there is an alternate power distribution analysis, control scheme or surveillance system to be used with your reactor operations which will demonstrate that an F of 2.11 can be maintained, please modify paction 4.3 (and other indications o pn and peak kp/it such as Table 4.1) to present this new limit, and a discussidn in detail of the modifications involved to hardware, analyses and operations, including a full uncertainty analysis. Topical report's may be submitted and referenced, but modifications to Chapter 4 (and possible Chapter 7) are required.'Res onse Accident analyses for SHNPP are presented in Chapter 15 of the FSAR. The results of these analyses determined a limiting value of total peaking factor, F , of 2.11 under normal operation, including load following maneuvers. This value is derived from the conditions necessary to satisfy the limiting conditions specified in the LOCA analyses of FSAR Section 15.6.5. As noted in FSAR Section 4.3.2, an upper bound envelope of F x Power equal to 2.32 x K(Z), as shown in FSAR Figure 4.3.2-21, results from operation in accordance with Constant Axial Offset Control procedures using ex-core surveillance only. Since SHNPP is limited to an F of 2.11, the Technical Specifications will require extra surveillance in the power regime above 90% RTP (sm ~2'11 This extra surveillance will be obtained by use of the Axial Power 2 32 F

Distribution Honitoring System (APDNS) to be described in a topical report to be issued.

The surveillance of the core hot channel factors, in accordance with the above, is presented in FSAR Section 3.2.6 of the SHNPP Technical Specification.

The APMS is a surveillance tool to verify compliance with limits of F (Z).

This system utilizes a four section excore detector (which has been calibrated and normalized) to provide on-line real-time monitoring of F (Z), This system provides audible and visual alarms when a predetermined APING alarm setpoint and power distribution are exceeded. Core scanning is initiated by exceeding

.a power range setting, by exceeding a control rod step dead band, or manually.

It is anticipated that future analyses will permit a rise in the F limit to 2.32 or greater. This rise in the limit will eliminate the need tP operate the APDMS.

(7215PSAkj r)

Environmental and Hydrologic Engineering Branch/R. Gonzales Open Item 14

2.4. 1

~ ~ Cooling Water Supply pg.~ 2-28

~session:

Because plant site drainage, including overland runoff, flows into the emergency service water intake and discharge channels, there is a potential for sediment to'uild up in the channels and auxiliary reservoir during operation, especially while heavy construction is still in progress. The staff has asked the applicant to describe the program for monitoring sediment buildup in the emergency service water channels and the auxiliary reservoir.

The staff's concurrence in the need for and the design of a sediment-monitoring program will be necessary before an operating license is issued.

~Res onse:

The emergency service water channels are inspected in accordance with the requirements of Reg. Guide 1.127 , Rev. 1. Profile data for the ESW intake and discharge canals and the auxiliary reservoir channel are attached. These profiles show the design and present configuration of the channels. The difference between the design and present profiles in the discharge canal is due to sediment.

The following drawings are attached:

SHNPP W. Aux. Channel Plan and Profile D-3341 SHNPP W. Aux. Channel Cross Sections D-3442 SHNPP Emergency Service Water Discharge Channel Plan and Profile D-3443 SHNPP Emergency Service Water Discharge Channel Cross Sections D-.3444 SHNPP Emergency Service Water Intake Channel Plan and Profile D-3445 SHNPP Emergency Service Water Intake Channel Cross Sections D-3446 (7463PXTtda)

• The full size drawings included in the response to Open Item 14 are being transmitted only to R. Gonzales.

Instrumentation and Control Systems Branch/H. Li.

Open Items 82, 94

Shearon Harris Nuclear Po~er Plant Draft- SER Onen Item No. 82 Provide response to IE Bulletin 80-06 concerns (NRC Question 420.6).

Response

In order to ensure proper functioning of safety related equipment during an emergency mode, an ESF actuation signal is generated to activate associated equipment. The ESF Actuation System (ESFAS) is described in FSAR Section 7.3.1 and is functionally shown on FSAR Figures 7.3.1-1 Sheet 2. The functional diagrams are further detailed on FSAR Figures 7.2.1-.1 Sheets 6 through 8.

The following ESF actuation signals are automatically initiated and latched in during emergency conditions:

The Safety Injection Signal (SI), generated by high containment pressure or low pressurizer pressure or low steam generator pressure. The presence of either one of these conditions will automatically actuate the safety related equipment associated with the Safety Injection outlined in FSAR Table 7.3.1-5.

2. The Control Room Isolation Signal, generated by the Safety Injection Signal or by high radiation in the control. room air intake or by high chlorine in the control room air intake. The presence of either one of these conditions will automatically actuate the Safety equipment associated with the Control Room Isolation listed in FSAR Table 7.3.1-5.

3. The Containment Isolation Phase A signal, generated by a Safety Injection Signal outlined in 1 above. The presence of an SI will automatically actuate the safety equipment associated with the Containment Isolation Phase A listed in FSAR Table 7.3.1-7.

4. The Containment Isolation Phase B signal and Containment Spray Actuation Signal, generated by a high-3 containment pressure.

Tne presence of high-3 containment pressure will automatically actuate the safety equipment associated with the CI Phase B and Containment Spray listed in FSAR Table 7.3.1-8.

5. The Containment Ventilation Isolation Signal, generated by high containment radiation or by a safety injection signal outlined in l.above. The presence of either one of these conditions will automatically actuate the safety related equipment associated with the Containment Ventilation listed in FSAR Table 7.3.1-9

6. The Feedwater Isolation Signal, generated by a High-High level in any Steam Generator or by a SI signal outlined in l above. The presence of either one of these conditions will automatically actuate the safety related equipment associated with the Feedwater Isolation listed in FSAR Table 7.3.1-11.

Shearon Harris Nuclear Power Plant Draft DRR Open Item Ro. 82 Cont.'d~

~Res onae (Cont'd)

7. The Main Steam Isolation Signal (MSIS), generated by a high Containment pressure or by Low steam pressure or by high stear.

pressure rate. The presence of either one of these conditions will automatically actuate the safety related equipment associated with the MSIS listed in FSAR Table 7.3.1-10.

The ESF actuation signals outlined in 'items 1 through 7 above are equipped with ta manual reset capability at the Main Control Board. A review of the control wiring diagrams for this safety related equipment confirmed that after resetting the ESFAS signal, safety-related equipment remains in its associated emergency mode.

FSAR Test Abstract 14.2e12.1.59 will be revised to verify that equipment remains in its associated emergency mode upon the resetting of various ESFAS signals.

Shearon Harris Nuclear Power Plan't Draft SER en Item No. 94 NRC uestion 420.5 Loss of Non-Class lE Instrumentation and Control Power System Bus Durin Power Ooeration (IE Bulletin 79-27 If reactor controls and vital instruments derive power from common electrical distribution systems, the failure of such electrical distribution systems may result in an event requiring operator action concurrent with failure of important instrumentation upon which these operator actions should be based.

This concern was addressed in IE Bulletin No. 79-27. On November 30, 1979 IE Bulletin No. 79-27 was sent to Operating License (OL) holders, the near OL applicants (North Anna 2, Diablo Canyon, NcGuire, Salem 2, Sequoyah, and Zimmer), and other holders of Construction Permits (CP) .

Of these recipients, the CP holders were not given explicit direction for making a submittal as part of the licensing review. However, they were informed that the issue would be addressed later.

Your are requested to address these issues by taking IE Bulletin 79-27 Actions 1 through 3 under "Actions to be taken by Licensees." Complete the review and evaluation required by Actions 1 through 3 and provide a written response describing your review and actions. This report should be in the form of'n amendment to your PSAR and submitted to the NRC office of Nuclear Reactor Regulation as a licensing submittal.

Response

Action Item No. 1<<

Safety"related instrumentation and controls are designed, as described belo~ so that the impact of losing a bus or safety-related instrumentation vill not impact the ability to achieve cold shutdown.

Safety-related instrumentation and systems conform to the criteria of IREE-279-1971, "Criteria for Protection Systems for Nuclear Generating Systems." This means that the safety systems and instrumentation are designed for high functional reliability such that no single failure results in loss of

, safety function and does not result in loss of the required minimum redundancy. The safety systems are designed to assure that the effects of natural phenomena, normal operating maintenance, tasting and postulated accident conditions do not result in loss of the safety function. The safety systems are separated from the control systems to the extent that failure of any single control system component, power supply, or channel which is common to the control and safety system leaves intact a system satisfying the reliability redundancy, and independence requirements for a safety system.

The following identifies the nonsafety-related instrumentation and control systems and corresponding power supplies whose loss could affect the ability to achieve'old shutdown condition. An evaluation of the effect of is the loss of power supply on the ability to achieve cold shutdown condition also provided.

• A list of safety and non-safety power buses is attached as Table Z.

The following items are included on .Table I: whether or not the bus has a dedicated alarm, bus failure detection and redundant buses.

Shearon Harris Nuclear Power Plant Resoonse (Cont'd)

Action No. 1* (Cont'd)

Pover Panel UPP-1 Vital (120 V hC)

Sys tems af fected are: Pressurizer Pressure Control System (PPCS)

Pressurizer Level Control System (PLCS)

Pressurizer Temp Control System (PTCS)

Evaluation for achieving cold shutdovn:

Loss of power vill result ia loss of ehe 26 volt pover supply to the Process Instrumentation Coatrol Panel C5 (PIC-C5), but there is a backup 2S volt supply feed from PP-lE212. If this feed vas ppe , e operator can switch to ehe also lost s , v vhile th e reactor is pressure control which vill not impact redundane control for manual level and the .ability to achieve cold shuedova.

Power Panel PP-lE212 (120 V hC)

Systems affected are same as those of Paver Panel UPP"1 vital (120 V hC).

Evaluation for achieving cold shutdovn:

The evaluation for Pov Pover Panel UPP"1 vital Vital is redundant to Power Panel PP-1E212. applies, since Power Panel UPP-1 Pover Panel PP-ID212 (120 VAC)

Systems affected are: Pressurizer Relief Tank Inst Residual Hest Exchanger Bypass Plov Evaluation for achieving cold shutdovn:

Loss of this pover panel results ia loss of the 26 volt power supply ia the PICM5. But there is a backup 2S V pover supply feed from UPP-1B vital pover panel.

Paver Panel UPP-1B Vital (120 VhC)

Systems affected are same as those of Pover Panel PP-1D212 (120 V hC) ~

Evaluaeioa for achieving cold shutdown:

The evaluation for Pcver Panel PP-ID212 applies, since Power Panel UUP-1B Vital is redundant to Power Panel PP-1D212.

Shearon Harris Nuclear Power Plant Draft SER 0 en Item No. 94 (Cont'd)

~Res onse (Cont'd)

ACTION ITEM 1* (Cont'd)

Power Panel PP-1E211 (120VAC)

Systems affected are: Rod Position Indication Boric Acid Control Batching Tank Agitator Evaluation for achieving cold shutdown:

Loss of this power panel will result in loss of rod position indication but the operator can manually switch to Power Panel PP-1D211. There will also be a loss of boric acid flow indication, however, the operator can manually borate via the emergency boration flow path to achieve cold shutdown.

Power Panel PP-1D211 (120 V AC)

Systems affected are same as those of Power Panel PP"1E2 11 (120 V AC).

Evaluation fox achieving cold shutdown:

The evaluation for Powex'anel PP-1E211 applies, since Power Panel PP-1E211 is redundant to Power Panel PP-1D211.

ACTION ITEM 2 CF&L will prepare abnormal and emergency operating procedures that will be used by control room operators, including procedures required to achieve a cold shutdown condition, upon loss of instrument bus power to safety and nonsafety related instrument and control systems. The procedures will include the diagnostics/alarms/indicators/symptoms resulting from the review and evaluation conducted per IE Bulletin No. 79-27 action item No. 1. These procedures will also describe methods for restoring power to the bus. The procedures will be available six months prior to fuel load.

To ensure subcomponents are operable following repair, administrative controls have been included in various plant programs.

Maintenance administrative controls are specified in the plant's Corrective Maintenance Procedure. Operability of repaired safety system subcomponents is ensured by the Operational Work Permit (OWP).

When a safety system or one of its components is removed from service, an approved OWP is written to guarantee that the appropriate component and system retesting is completed prior to declaring the system operable. In addition, the OWP details any other systems that must be tested prior to taking the desired system out of service. After testing, the system lineup will be returned to normal service with an approved OWP, and it will be independently verified.

Shearon Harris Nuclear Power Plant Draft SER 0 en Item No. 94 Cont'd Resnonse (Cont'd) hctfon Item 3 The non-ssfety Vital h"C Power Supplies are fed from tvo Static Uninterruptible Pover Supplies (SUPS). These are used to power noa-safety instrunentation aad controls circuits, fire detection and radiation monitoring systems. They are both rated 120V, 60 Hs, single phase 60 KVh and 7.5 KVh.

The 60 KVh SQPS is powered from MCC 1D21 for the main feed and MCC 1E21 for the bypass feed. The dc fs supplied froa the 250V dc bus DP-1-250. The 7.5 KVh SUPS is povered frcxa MCC lE21 for the main feed and dc provided from the 250V bus DP-1-250 for backup. Bypass for the 60 KVh SUPS is regulating transformer.

provided'hrough a stepdovn Both SQPS are designed to provide power to their loads normally from the main feed through the rectf ffer/faverter. Tailure of the asia ac feed vfll cause the SUPS to be povered froa the battery.

Circuitry is provided fn the noasafety 60 KVh vital ac SUPS to shut dovn the iaverter portioa and transfer to aa alternate hC source vfa a high speed static switch. Transfer fs initiated oa oae of the folloving conditions: (1) inrush load exceeds 260 KVh for a duration of 1 sec at rated voltage, (2) output voltage drops 10 percent of instantaneous nominal voltage, (3) output frequency devfates more than +0.5 Hs from nominal value. Voltage sensing circuitry is provided vhich vill respond vhea a preset voltage has been reached. The period of detectioa and cmapletfoa of transfer does aot exceed 8.3 mill llisecoads and the power interruptioa interval during transfer from the inverter hC supply to bypass source does not exceed 50 microseconds. Upon restoration of normal conditfons retransfer fs done manually. hfter a time delay, the unit vill alarm whenever a transfer of the bypass source has taken place to prevent alarming on momentary transfers.

The safety related instruments for the Reactor Protection System and the Engfneered Safety Features hctuation System are povered from four safety related SQPS (channels I, II, III arid IV). ~ They are rated 118V hC 60 Hs I s fagle phase ?.5 KVh.

1 Channels I and III are 0 povered froca MCC lh21 Sh and HCC Ih31 Sh respectively for the main feeds and 125V dc bus DP-lh-Sh for backup. Channels II and IV are povered from MCC lB21 SB and MCC1B31 SB respectively for the main feeds and 125V dc bus DP-lB-SB for backup. h manual svftch fs provided to power the 118V hC bus, in case the fnverter fs lost.

These sources are from 120/208V Power Paael lh211 Sh, lh311 Sh for channels I and III, respectively, and Power Panel 1B211 SB, 1B311 SB for channels II and IV, respectively.

Duriag normal operation the four 7.5 KVh faverters for the safety related fnstrumentatfoa is powered from an hC source with the 125V DC source assuming the load upon loss of hC voltage. Blocking diodes are provfded to each input circuit to prevent voltage feedback. Lov hC and dc input voltage, low hC output voltage aad overcurreat are all alarmed. There is ao automatic traasfer to a bypass source.

fifth regard to safety system operability after maintenance, refer to Action Item 2.

Based on the revfev of IE Circular No. 79-02 and evaluatioa discussed above, no design modifications are required.

TABLE I ALARM INDICATION FOR LOSS OF P(WZR TO SAFETY AND NON-SAFETY P(VER BUS DEDICATED REDUNDANT BUS OR DISTR PNL CLASS hLhRM BUS FAILURE DETECI'ION BUS 1 120/208V PP-1D121 HS Ho Failure of IND and Controls None 2 120/208V PP-1D211 NS Ho Failure of IND and Controls PP-IE211 3 120/208V PP-1D343 HS No Failure of IND and Controls None 4 120/208V PP-ID131 NS Ho Failure of IND and Controls None 5 120/208V PP-1D212 HS Ho Failure of IND and Controls UPP-lhhlB 6 120/208V PP-1-4A232 NS Ho Failure of IND and Controls None 7 120/208V PP-lE121 NS Ho Failure of IND and Controls None 8 120/208V PP-lE211 HS Ho Failure of IND and Controls PP-1D211 9 120/208V PP-1E311 NS No Failure of IND and Controls None 10 120/208V PP-lE341 NS No Failure of IND and Controls None 11 120/208V PP-lE231 NS Ho Failure of IND and Controls None 12 120/208V PP-lE212 NS Ho Failure of IND and Controls None 13 120/208V PP-1-4B231 NS No Failure of IND and Controls None 14 120/208V PP-1-.4A111 HS Ho Failure of IND and Controls None 15 120/208V PP-1-4A10211 NS Ho Failure of IND and Controls None 16 120/208V PP-1-4A10221 'NS No Failure of IND and Controls Hone 17 120/208V PP-1-4A241 HS Ho Failure of IND and Controls None 18 120/208V PP-1-4A21-1 HS No Failure of IND and Controls None 19 120/208V PP-1-4A22-1 HS Ho Failure of IND and Controls None 20 120/208V PP-1-4A25-1 NS No Failure of IND and Controls None 21 120/208V PP-1-4A31-1 NS No Failure of IND and Controls None 22 120/208V PP-1-4A32-1 NS No Failure of IND and Controls None 23 120/208V PP-1-4A32-2 HS No Failure of IND and Controls None 24 120/208V PP-1-4A32-3 NS No Failure of IND and Controls None 25 120/208V PP-1-4A33-1 NS No Failure of IND and Controls None 26 120/208V PP-1-4A33-2 NS Ho Failure of IND and Controls None 27 120/208V PP-1-4A33-3 NS No Failure of IND and Controls None 28 120/208V PP-1-4A33-4 HS No Failure of IND and Controls None 29 120/208V PP-1-4A611 NS No Failure of IND and Controls None 30, 120/208V PP-1-4A231 NS Ho Failure of IND and Controls None 3"1 120/208V PP-1-4Blll NS No Failure of IND and Controls None 32 120/208V PP-1-4B112 HS Ho Failure of IND and Controls None 33 120/208V PP-1-4B10211 NS Ho Failure of IND and Controls None 34 120/208V PP-1-4B10212 Ho Failure of IND and Controls None 35 120/208V PP-1-4B241 NS No Failure of IND and Controls None 36 120/208V PP-1-4B31-1 NS Ho Failure of IND and Controls Hone 37 120/208V PP-1-4B32-1 HS Ho Failure of IND and Controls None 38 120/208V PP-1-4B32" 2 NS No Failure of IND and Controls None 39 120/208V PP-1-4B32-3 HS Ho Failure of IND and Controls None 40 120/208V PP-1-4B33-1 HS Ho Failure of IND and Controls None 41 120/208V PP-1-4B34-1 NS Ho Failure of IND and Controls Hone 42 120/208V PP-1-4A10121 NS Ho Failure of IND and Controls None 43 120/208V PP-lE213 HS Ho Failure of IND and Controls None 44 UPP-1 VITAL NS Ho Failure of IND and Controls None 45 120/1PH PP1 ROD POS NS Ho Failure of IND and Controls PP-lD211, PP-1E211

TABLE I Cont'd DEDICATED REDUNDANT BUS OR DISTR PNL CLASS ALARM BUS FAILURE DETECTION BUS 46 120/240 PP-B Failure of and Controls PP-ID211, PP-IE211 120/208 PP-IE321 Failure of IND Controls PP-ID2 ll o PP-IE211 120/208 PP-IE322 Pail ure of Controls PP-ID2 I l, PP-IE211 49 120/208 Pp-lh 321 IE/SA Pailure of IND and Controls PP-IB32 1 50 120/208 PP-IB321 IE/SB Failure of IND and Controls PP-lh321 51 120/208 PP-l&4A33 IE/SA Failure of IND and Controls pp-I &4B33 52 120/208 PP-1&4B33 IE/SB Failure of IND aad Controls PP-1&4A33 53 120/208 PP-IA211 IE/SA Failure of IND and Controls PP-IB211 54 120/208 PP-IA311 IE/SA Failure of IND aad Controls PP-IB311 55 120/208 PP-IB211 IE/SB Failure of IND and Controls PP-lh211 56 I20/208 PP-IB311 IE/SB Failure of IND and Controls PP-IA311

57 120/208 PP-lh231 IE /SA Failure of IND and Controls PP-IB23 }

'58 120/208 PP-IB231 IE/SB Pailure of IND and Controls PP-ih231 59 120/20S PP-ID122 Failure of IND and Coatrols 60 120/IPH IDP-lh-SI SA Ann, Caap and Failure of IND ID P-IB-SII 61 120/IPH IDP-lh-SIII SA Yes han, Camp and Failure of IND IDP-IB-SIV 62 120/IPH IDP-IB-S II Yes ESP, Ann, Canp and Failure of IND ID P-I h-S I 63 120/IPH 120/208 IDP-IB-S IV PP-1E-122 SB Yes ESP, Ann, Coep and Failure of IND II IDP-A-S I 6 4 No Failure of IND and Controls 65 120/208 PP-ID213 NS No Pailure of IND and Controls None 66 120/208 PP-ID341 NS Ho Failure of IND and Controls None 67 125VDC DP-Ih IE/SA Ho Failure of IND and Controls DP-IB 68 125VDC DP-lhl IE/SA No Failure of IND and Controls DP-I Bl 69 120/208 PP-ID344 NS No Failure of IND and Controls Noae 70 120/208 PP-IE312 Ho Failure of IND and Controls None 71 125VDC DP-IB IE/SB Ho Failure of IND and Controls DP-Ih 72 125VDC DP-IBl IE/SB Ho Failure of IND and Controls DP-lhl 73 120VhC UPP-Ih-VIThL HS Ho Failure of IND and Controls PP-ID212 74 120VAC UPP-IB-VITAL NS No Failure of IND and Controls PP-ID212 75 125VDC DP-lh NS No Failure of IND and Controls None 76 125VDC DP-Ih-I NS Ho Pailure of IND and Controls DP-Ih-2 77 125VDC DP-IA-2 NS Ho Pailure of IND and Controls Dp-Ih-I 78 125VDC DP-lhll NS No Pailure of IND and Controls DP-lh21 79 125VDC DP-IA21 NS No Failure of IND and Controls DP-lhl I 80 125VDC DP-104h NS No Failure of IND and Controls DP-I-4B 81 125VDC Dp-I-4AI NS Ho Failure of IND and Controls DP-I-4B I 82 125VDC DP-I-4A2 NS No Failure of IND and Controls DP- I- <82 83 125VDC DP-I-4A3 Ho Pailure of IND and Controls DP-!- q83 84 125VDC DP-I-4B Ho Pailure of IND and Controls DP-I-qf, 85 125VDC DP-I-4BI NS No Pailure of IND and Controls DP-I- qual 86 125VDC DP-I-4B2 No Failure of IND and Controls DP-I-4A2 87 125VDC DP-I-4B3 No Failure of IND and Controls DP-'I-4A 3 88 120/208 PP-ID342 No Failure of IND and Controls None

Meteorological and Effluent Treatment Branch/J. Hayes Open Items 160, 163, 179, 181, 229

t Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER) 0 en Item 160 (DSFR Section 11. 1. 2 Page 11-6) 0 en Item 163 (DSER Section 11.4.2 Page 11-22)

Question 0 en Item 160 The applicant has not presented any detail, on the VR system. In particular, the applicant has not addressed, (1) the volume of waste to be handled by the VR system, (2) the quantity of airborne radioactive effluents released from the VR system, (3) the additional volume of waste to be treated by the liquid waste processing system as a result of operation of the VR system, and (4) the additional radioactive liquid effluents resulting from operation of the VR system.

0 en Item 163 Provide a detailed description of the volume reduction system that addresses:

o how the system operates o expected input streams, their volumes, and associated activities o liquid and gaseous effluents resulting from system operation o volume reduction factors achieved o activity associated with the disposable ash o conformance with RGs 8.8, 1.140, and 1.143, and BTP CMEB 9.5-1 (SRP 9.5.1)

~Res'ese The Volume Reduction System processes liquid wastes to free-flowing, anhydrous, radioactive salts via water evaporation in a heated, fluidized bed. This system is represented by the system described in previously-approved Aero)et Topical Report No. AECC-1-A dated February 21, 1975, with the following exceptions:

1. The fluid bed dryer was incr'eased from 24 to 30 inches to allow for a higher processing rate.

s 2~ The scrubber/condenser units are separated to provide cleaner condensate overflow which is recycled to the Floor Drain Treatment System (refer to FSAR Section 11.2.2.1). Also, the scrubber utilizes feed solution as the scrubbing liquid, therefore creating a worst case efficiency of 99.86%

based on feed input to waste return. The scrubber now is designed to automatically control the solids content of the scrubber liquid to 28%

solids, which is fed to the fluid bed dryer. This eliminates critical control on upstream evaporators, which can increase the VRS feed rate up to 80 gph.

3. The Liquid Heat Exchanger for the condenser was replaced with condenser tubes to allow more efficient use of cooling water since it eliminated the loss of pressure by two-fold indirect heat exchange.

4. The Aerojet Energy Conversion Company (AECC) designed charcoal adsorber was replaced by an 11 /2 inch-thick nuclear grade (HSA design) commercial unit to improve radioiodine entrapment, process life, and reliability.

5. Two additional HEPA filters were added to the fluidizing air line to the fluid bed dryer. This will prevent buildup of dust in the windbox and distributor plate, thus reducing problems with dust buildup and decontamination requirements

6. A detailed breakdown of components and design standards is shown in Tables H through K.

II. S stem Desi n Features The Volume Reduction System as described in FSAR Section 11.4.2.1.2 is designed to the guidelines of NRC Regulatory Guide 1.143, Rev. 1, 1979 with components code requirements and material selection presented in Tables H through K.

The VRS design is consistent with the guidelines set forth in Regulatory Guide 8.8 so as to achieve "As -Low As Reasonably Achievable" (ALARA) occupational exposure. The general arrangement of the VRS hardware utilizes individually shielded cubicles to separate components containing the majority of the radioactive material from mechanical components such as the pumps and blowers, which contain small amounts of radioactive material an'd which are expected to require periodic maintenance. In addition, the components containing the majority of the radioactive material can be decontaminated and the radioactive salts can be flushed from the system components prior to performing maintenance.

Fire protection considerations are detailed in FSAR Section 9.5.1 (including conformance to BTP CHEB 9.5-1) ~ CP&L conformance to Regulatory Guide 1 '40 is stated in FSAR Section 1.8. Conformance to other regulatory bases and guides are included in Topical Report No. AECC-1-A dated February 21, 1975.

III. Influent Waste Streams The VRS will process liquid radioactive wastes from, the following sources:

a. Backwashed sludges from filters on liquid waste process streams.

b. Waste evaporator concentrates.

c. R.O. concentrate evaporator concentrates.

d. Secondary waste evaporator concentrates.

e. Boron recycle evaporator concentrates.

The expected input volumes and associated activities are shown in attached Tables A through E, which are expressed for two-unit operation. Figure 1 shows the process flow diagram and material balance. FSAR Tables 11.4.1-2

through 11.4.1-7 will be revised in a future amendment to the FSAR. All input streams to the VRS are sent to the VRS feed tanks (dual 5,000 gal.

presolidification tanks) where they are pH adjusted.

XV. Li uid Effluent Streams During operation of the VRS, condensate overflow (1 gpm) from the condenser is recycled to the Floor Drain Treatment System (see FSAR Section 11.2.2.1 and FSAR Figure 11.2.2-4). The dissolved solids content in the overflow is 200 ppm and the suspended solids content essentially nil. Since the average dissolved content of the feed stream is expected to be 110,000 ppm, the activity"in the overflow can be calculated by multiplying the feed stream activity by (200/110,000) or 1.8 x 10 . From Figure 1, the condenser overflow volume of 4.24 x 10 kg/yr will be treated by the Floor Drain Treatment System. This system provides for treatment via a reverse osmosis unit followed by a cation demineralizer. This stream is then sent to the Waste Monitor Tanks where it is sampled and monitored through Radiation Monitor REM-1WV-3541 prior to release. Figure 1 also shows 7.8 x 10 kg/yr of water to be processed through the etch disk filter. This water source is the carrying water from the 38 backflushable filters (17 types) throughout the plant when the filters are backflushed and is sent to the Fquipment Drain Treatment System. The activity of this water is a weighted average of the dissolved radionuclide concentration in each of the 17 types of filters as shown in attached Tables D and E. This stream is processed through the Equipment Drain Treatment System and sent to the Waste Monitor Tanks where it is sampled and monitored through Radiation Monitor REM-lWV-3541 prior to release. The annual release of radionuclides to the environment due to operation of the VRS is shown in the attached Table F.

V. Gaseous Effluent Stream Figure 1 indicates that the exhaust gases will come off the condenser and be recycled serving as the fluidizing air to the fluid bed dryer system. A small portion (about 6%) of the recycled stream will be routed to the VRS air cleaning units. These air cleaning units consist of two parallel trains, each consisting of a pre-HEPA filter, charcoal adsorber, and HEPA filter in series. After processing through the VRS air cleaning units, the gas is monitored by Radiation Monitor REM-1WV-3551 (see FSAR Figure 9.4.3-5). The gas then enters the Waste Processing Building exhaust header where it is refiltered by parallel trains, each consisting of a pre-HEPA filter, charcoal adsorber, and HEPA filter before entering the WPB Exhaust Stack 85. Here, is again monitored, this time by Radiation Monitor REM-1WV-3546 before being it exhausted to the atmosphere (see FSAR Figure 9.4.3-8). Total decontamina)ion factors from input to t)e VRS until exhaust to the environment are 1 x 10 for radioiodines and 1 x 10 for all others. Using this, the annual release of radionuclides to the environment due to operation of the VRS is shown in the attached Table G.

VI. Volume Reduction Factors The volume reduction factor achieved will vary with the input stream concentration. The process procedure will ensure that the feed stream will remain above 10% solids; therefore, the expected volume reduction factor will be approximately eight.

VII. Actitit Associated with the Dls osable Ash The normal and design base activity for unsolidified Volume Reduction dry salt product is shown in attached Table L.

TABLE A VRS INFLUENT WASTE STREAMS Sources ~Te Volume (ft3 I r.)

Waste Evaporator Bottoms 12% Na2B~07 325 R.O. Concentrates 12% boric and 2500 Evaporator Bottoms other salts Secondary Waste 22% Na2SO~ 9350 Evaporator Bottoms Boron Recycle Evaporator 4% H3BO 2050 Filter Particulates 3% filter sludge 3600

NUCLIDE ACTIVITY INPUTS TO THE SOLID RADWASTE SYSTEM EVAPORATOR CONCENTRATES (pWi/ ) NORMAL OPERATlONS Waste RO Concentrate S. W. EU.gh Boron Concentrat Evaporator Evaporator Conductivity Evaporator Isotope Bottoms Bottoms Evaporator Bottoms Bottoms Br 83 1.24E-03 4. 69E-05 1.01E-04 Br 84 1.46E-04 1.25E-06 1 21E-05 I 130 2.71E-03 2.73E-04 1.62E-04 I 131 5.04E+00 5.93E-Ol 3.42E-02 6. 7E-01 I 132 2.46E-02 8.91E-04 2.01E-03 8.2E-02 I 133 8.07E-01 9. 18E-02 3.60E-02 6.8E-01*

I 134 4.37E-03 *' 6.11E-05 3 60EW4 I 135 1.34E-01 1.05E-02 1.00E-02 1.5E-01 Rb 86 " : 3.13E-03 2.88E-04 l. 10EW5 Rb 88 6.18E-03 2.98E-05 5.21E-04 Cs 134 1.93E+00 1.25E-01 3.29E-03 5.6E-01 Cs 136 3.69E-01 '3.78E-02 1.68E-03 7.2E-01 Cs 137 1.42E+00 9.27E-02 2.37E-03 3.7E-01 Cr 51 8.67E-03 7. 23E-04 2.47E-04 54 2.31E-03 1.62E-04 4.07E-05 Fe 55 1.24E-01 7.87E-03 2.10E-04 Fe 59 5.54E-03 4.20E-04 1.30E-04 Co 58 1.00E-01 7.18E-03 2.09E-03 Co 60 1.56E-02 ~ " 1.09E-03 2.62E-04 Sr 89 2. 02E-02 1.50E-03 4.57E-05 Sr 90 7.86E-04 4.96E-05 1.31E-06 Sr 91 6.54E-04 6.03E-05 4.37E-05 Y 90 7.66E-07 9.94E-08 1.42E-07 Y 91M 3.22E-06 4.31E-08 2.65E-06 Y 91 3.81E-04 2.78E-05 8.35E-06 Y 93 3.68E-06 3.48E-07 2.39E-06 Zr 95 3.69E-04 4.26E-05 7.84E-06 lib 95 2.53E-04 4. 29E-05 6.51E-06 99 5.54E-03 7.18E-04 9.94E-03 1.6E-01 Tc 99M 3.07E-04 2.27E-05 2.34E-03 1. 5E-01 Ru 103 2.38E-04 2.01E-05 5.86E-06 Ru 106 7.52E-05 3.28E-05 1.31E-06 Rh 103M 4.59E-07 7.01E-09 3.78E-07 Te 125M 1. 73E-03 1.27E-04 3.79E-06 Te 127M 1.89E-02 1.29E-03 3.66E-05 Te 127 8.23E-04 7. 44E-05 5.58E-05 Te 129M 6.90E-02 5.53F;03 1.82E-04 Te 129 1.93E-.04 3.49E-06 1.59E-05 Te 131M 7.24E-03 8.72E-04 2.58E-04 4.8E-03 Te 131 4.82E-05 3.23E-07 4.02E-06 Te 132 2.08E-01 2.71E-02 3.24E-03 6.7E-02 Ba 140 6.12E-03 6; 32E-04 2.82E-05 140 6.02E-05 .7.55E-06 1.66E-05 Ce 141 3.43E-04 2.76E-05 9.11E-06

TABLE >(Continued)

Paste RO Concentrate S. V. High Boron Concentra:

Evaporator Evaporator Conductivity Evaporator Isotope Bottoms Bottoms Evaporator Bottoms Bottoms Ce 143 1.32E-05 1.62E-06 4.26E-06 Ce 144 2.45E-04 7.38E-05 4.33E-06

~

Pr 143 1.46E-04 . 1.48E-05 6.42E-06 Pr 144 9.78E-07 4.52E-09 8.25E-08 Np 239 6.72E-03 8. 68E-04 1.40E-04 TOTAL 1.03E+01'. 01E+00 1.10E-01 3.61E+00

~ ~

TABLE C NUCLIDE ACTIVITY INPUTS TO THE SOLID RADVASTE SYSTEM EVAPORATOR CONCENTRATES (pWi/ ) DESIGN BASIS Paste RO Concentrate S. V. High Boron Concentrat Evaporator Evaporator Conductivity Evaporator Isotope Bottoms Bottoms Evaporator Bottoms Bottoms Br 83 2.19E-02 8. 23E-04 3.57E-03 Br 84 2.30E-03 1.97E-05 3.83E-04 I 129, 3.22E-.06 2.04E-07 1.07E-08 I 130 2.42E-02 ' 2.44E-03 2.90E-03 I 131 5.11E+01 6 '2E+00 6.94E-01 6. 70E+00 I 132 6.31E-01 . 2.29E-02 1.03E-01 8.20E-01 I 133 '.30E+00 9.49E-Ol 7.40E-01 6.80E+00 I 134 4.69E-02 6.54E&4 7.72E-03 I 135 1.42E+00 l. 11E-Ol 2.12E-Ol 1.50E+00 Rb 86 7.65E-Ol 7.05E-02 5.40E-01 Rb 88 1.21E-01 5.83E-04 2.04E-02 Rb 89 4.66E-03 1.87E-05 7.91E-04 Cs 134 1.69E+02 1.08E+01 5.75E-01 5.60E+00 Cs 136 8.02E+01 8.21E+00 7.30E-01 7.20E+00 Cs 137 1.12E+02 7. 10E+00 3.75E-01 3.70E+00 Cs 138 4.76E-02 4.11E-04 7.90E-03 Cr 51, 2. 38E-02 1.99E-03 1.36E-03 54 2.84E-03 2.14E-04 9.99E-05 Mn 56 5.19E-04 2.08E-05 8.48E-04 Fe 55 1.70E-01 1.08E-02 5.75E-04 Fe 59 3.06E-03 2.32E-04 1.44E-04 Co 58 8.94E-02 .6.48E-03 3.73E-03 Co 60 . 1.41E-02 1.17E-03 4.75E-04 Sr 89 2.41E-01 1.79E-02 1.09E-03 Sr 90 8.98E-03 5.68E-04 3.00E-05 Sr 91 ~ 5.71E-03 5.25E-04 7.63E-04 Sr 92 3.23E-04 1.35E-05 5.27E-05 Y 90 2.08E-05 2.70E-06 7.71E-06 Y 91M 3.15E-05 4.22E-07 5.19E-05 Y 91 3.29E-03 2.40E-04 1.44E-04 Y 92 3.97E-05 2.09E-06 6.41E-05 Y.. 93, 3.74E-05 3.54E-06 4.86E-05 Zr 95 3.86E-03 3.20E-04 a.64E-04 hb 95 3.17E-03 3.12E-04 1.63E-04 Mo 99 5.91E 7.69E-03 1.72E-01 l. 60E+00 Tc 99M 3.91E-03 2.89E-04 5.95E-02 .1.50E+00 Ru 103 2.92E-03 2.30E-04 1.44E-04 Rn 106 1.00E-03 1.37E-04 3.50E-05 Rh 103M 4.97E-06 7.60E-08 8.19E-06 1

1@i

~ 'V ~ 9.83E-03 6.56E-04 3.49E-04 Te 12'e 1.59E-02 1.16E-03 6.96E-05 127M 1.86E-01 1.28E-02 7.23E-04 Te 127 1.03E-02 9.32E-04 1.40E-03

C (Continued)

Waste RO Concentrate S. W. High Boron Concentrat Evaporator Evaporator Conductivity Evaporator Isotope Bottoms Bottoms Evaporator Bottoms Bottoms Te 129M 8. 92E-01 7.15E-02 4.70E-03 Te 129 1.83E-03 3.30E-05 3.01E-04 Te 131M 6.92E-02 8.34E-03 4.94E-03 4.8E-02 Te 131 4.75EW4 3.18E-06 7.93E-05 Te 132 2.23E+00 2.92E-01 6.98E-02 6.7E-01 Te 134 2.01E-03 2.26E-05 3.31E-04 Ba 140 1.13E-01 1.17E-02 1.05E-03 La 140 5.20E-04 6.52E-05 2.87E-04 Ce '141 2.98E>>03 2.40E-04 1.58E-04 Ce 143 '.59E-04 1.95E-05 1.02E-04 Ce 144 'r 2.76E-03 3.30E-04 9.74E-05 143 1.81E-03 1.83E-04 1.58E-04 Pr 144 9.63E-06 4.45E-08 1 63E-06 TOTAL 4.28E+02 3. 37E+01 3. 81E+00 3.61E+01

TABLE D FILTER SLUDGE ACTIYITT ChLCUIATIOtlb (HOIUIAL) filter dyeteu Farticulate Activity uCI/cc - '

Dieeolved Activity uCiicc IIaas oi

?-131.

bpeot caela bluice I.ZG(O) 2.06(-l) G.S3(-I) 2.7S(-l) 1.83(0) 1.49(O) 2 86( I) 4.18{-l) I o68(-2) 2mlo{-3) 2+63(-2) 1.90(-2) 3.179(7)

Filter Fuai Pool bkiaser 2. 16( I) 2. 74 (0) 1.90(-5) 2.70(-5) 1.10(-6) 1.40(-7) 1,70(-6) 1.20(-6) 9.536(7)

Filter Ilaiualing Iiater 6.99 (0) 8.90(- I) I . 90(-5) 2. 70(-5) I . 10( 6) I+40(-7) I. 70("6) l. 20(-6) 1.590(8)

I'liter Iia e C ~ r F v a po a C 0 r 1.33("2) 2.35(-3) 6 ~ 73(-5), 3.29(-7) L.34(-6) '2.08(r7) 2.57(-6) 4.92(-7) 1.059(6)

Condenaate Filter baaL Mater Ileturn 2.60(2) 3.40{1) 2. 86 {-I) 4. 18 (-I) 1.68(-2) 2 F 10(-3) 2.63(-2) 1.90(-2) 6.352(6)

Filter Seal Mater 1.90(2) 2.54(I) 2.86(-I) 4.18(-1) le68(-2) 2 ~ 10(-3) Zo63("2) 1.90(-2) 4o533(6)

InJrction filter recycle KvaporaCO'r I 14( 2) ~ 5,80(-Z) 2.86(;2) 4.18(-2) 1.68( 3) 2.10('-4) 2.63("3) 1.90{-3) 1.276{6)

Feed Filter Recycle Kvaporator 3.86(-4) 6.58(-5) 1<<059(6)

Condenaata Filter Ilecycie FvaporaCor 8 ~ 48(Q) 3.02(0) 9.68(-1) 7e 10(- I) 6.62 (- I) 6.61(-I) 6.39(-2) 4.68(-2) 7.953(6)

Concentrate filter boric Acid FII ter 2 '5{Q) I 81(Q) 3.O8(-l) 3.41(-I) G.GZ(-I) 6 ~ 61(-l) 6.39(-2) 4.68(-2) 6.357(7)

Caa Decay Tark 3 ~ 99(-I) 7o953(6)

Filter Ileactor Coolan.t Filter 6.71(I) 8 ~ 40(o) 2.86("I) 4 'a(-I) lo68(-2) Zo 10(-3) 2.63(-2) I 90(-2) 6 'S7(7) fuel Pool filter 2.45(-i) I.4i(-Z) I . 79(-3) I.Z7(-2) 8.99(-3) 1.90(-5) 2. 70(-S) 1. I'0(-6) 1.40(-7) 1.70(-6) 1. 20(-6) I, 574(a)

Secondary Maa:e I ~ 14 (-4) I .43{-5) 2.43(-G) 3. 18(-6) 1.42(-7) I o77(-8) 2.23(-7) 1.61(-7) 6.357(7)

Filter Maate Evaporator 4.03(I) 9.32(0) 6. 7 3(-2) I. 08(-2) l.. 34 (-2) 2,08(-3) 2.5I(-2) 1.9o(-z) 1,059(7)

Filter Laundry 4 llot 9.46(-3) Z. 19(-2) 9. I 7(-7) 6.30(-6) 1.45(-5) 2.10("5) 3.86(-6] 3. I 79(7)

SIlouer I'iter Floor Drain 1.59(0) Z.ZO(-I) 8 97( 3) I 66( 3) I 06( 3) Io46( 4) Ieaz( 3) 1+33( 3) I 589{8)

Filter

H FILTFR SI.UDGK ACIIVITf CAI.CIII (DKS 108>

Filter Systeu Particulate Activity nCf /cc Dissolved Activity uCI/cc IIsrs ol

~l-I I I-l33 8 -38 C -68 Ce-l34 C -l33 I-l3l 1-133 Co-58 Co-60 Cs-134 C -III ~LI I ~l ~

Spent laein Sluice 1.26(1) 2.06(p> -3. 54i(-I) I Ai9(-I>

~ 7,79(P) 4 ~ 33(P> 2 '0(0) 4 '0(0) 1.5(-2> f,9Q( 3) 2.30(0) I ~ %0{0> 3.179(7)

Filter FWL POOl Skfsirer 2.01(1) 2.4S(0) 1.90(-4) 2.80(-4) 1.00(-6) 1830(-6) 1.00("3) 9.536(7)

I,SQ(-4) filter SefualfnS Water 6.53(0) 7.95(-I) 1.90(-4) 2.80(-4) 1.00(-6) 1.30(-6) 1.50(-4) 1.00{-3) L.S90(8)

Filter Waste tvsporator I. 18(-2) 2. 13(-3) I'll(-4) L. 19(-6) 1888(-7) 2.25(-4) 1.50(-4) L.OS9(6)

Condensate Filter '.82(-4)

Seal Water Return 2.32(2) 3.07( I) 2. 90(0) 4. 30(0) 1.50(-2) 1.90(-3) 2. 30(0) I . 50(0) 6. 352(6) filter Seal Water L.70(2) 2.30(1) 2. 90(0) 4. 30(0) I, 50(-2) I. 90(<<3) 2,30(0) I. 50(0) 4.533(6)

In)ection Pilter Sacycle Kvaporator I l6(-I) 5.07(0) 2.90(-1) 4.30(-1) 1.50(-3) 1.90(-4) 2.30(-1) I.SO(-l) 1.276(6)

Feed Filter

'Recycle Evaporator 3.91(-3) 6'78(-4)

Condenssts FILter Sacycle'vaporator 8.58(l) 3.10(l) 8.48(1) 5.62(L) 6.70(0) 6.80(0) 5.60(0) 3. 70(0) 1.953(6) 1.059(6)'.60(0)

Concentrate Fl ltar Soric Acid filter 2. $ 8( I) 2 ~ 03(I) 2.69(l) '.69{ I) 6.70(0) 6.80{0) 3.70(0) 6. 351(7)

Css Decay Tank 3.48(1> 7.953(6) filter Iteactor Coolant 6.00{ I) . 7,63{0) 2.90(0) 4 ~ 30(0) L.SO("2) I ~ 9Q( 3) 2 '0(0) LE 50(0) 6 '57(7)

Filter .

fuel Pool Filter 7,45(0) 3.35(p) I. I3(-2) 1.6 I (-3) I, I 1(P> 7.37(-I),1.90(-4) 2.80(-4) 1.00(-6) 1.30(-6) 1.50(-4) 1.00(-3) 1.$ 74(8)

Secondary Waste 1.02(-4) I. 29 (-5) 2.47 ("5) 3.28(-5) 1.27(-7) 1.61(-8) 1.9$ (-5) 1.27(-5) 6.3$ 1(7)

FLLter Waste Kvaporator 3.64(l) 44(0) 6.82{-l) "I) 8 1.11{ 1819(-2) 1.88(-3) 2.2S(0) 1.50(0) I 059(1)

Filter

~

Laundry 4 Ilot 2.47(-2) 5.65("2) 2.39(-6) 1.65{"5) 3.78(-5) $ .48("5) 1.01(-4) 3. 179{7)

Shove r Pi 1 to r floor Drain 1.42(0) 2.00{-L) 9.09(-2) 1.71(-2) 9.44(-4) 1.32(-4) I.S9(-L) 1.71(-2) 1.589(8)

Filter

TABLE F ANNUAL RELEASE OF RADIONUCLIDES TO THE ENVIRONMENT (Ci/yr per Unit)

LI UID RELEASES FSAR NUCLIDE DESIGN BASIS* NOWL OPERATION* TABLE 11.2.3-1 Co-58 1. 87(-6) 1.75(-6) 2. 3(-3)

Co-60 2. 66(-7) 2.30(-7) 5.9(-4)

I-131 1. 16(-2) 1. 07(-3) '1. 8(-1 )

I-133 6. 25(-3) 5.35(-4) 2.5(-2)

Ce-134 3. 34(-3) 4.45(-5) 4.6(-3)

CE-137 1. 42(-3) 3.10(-5) 3.9(-3)

• Releases due to operation of the VRS.

TABL1'. G GASEOUS RELEASES (Ci/yr per Unit)

FSAR NUCLIDE DESIGN BASIS* NOlQlAL OPERATION* TABLE 11.3.3-1 Co-58 1.30(-8) 1.20(-8) 1. 6(-2)

Co-60 1.85(-9) 1.60(-9) 7.6(-3)

I-131 8. 10(-4) 7.45(-5) 4.6(-2)

I-133 4 ~ 34 (-4 ) 3.70(-5) 6.0(-2)

Ce-134 2. 32(-5) 3.10(-7) 4.9(-3)

CE-137 9.85(-6) 2.15(-7) 8.2(-3)

• Releases due to operation of the VRS.

E H VOLUME REDUCTION SYSTEM TANKS AND PRESSURE VESSELS DESIGN DESIGN COMPONENT NAME CODES/STANDARDS MATERIALS VOLUME PRESSURE(PSI) TEMPERATURE; ( F)

VRS Feed Tank ASME VIII, Div. 1 Incoloy 825 5000 gal 15 160 Caustic Tank API 620 304 SS 1130 gal 5 200 Fluid Bed Dryer Vessel ASME VIII, Div. 1 321 SS Freeboard 12 1200 Inconel Inconel 625 1100 321 Free-Shell/Bottom board 1200 321 Windbox Product Conveyor ASME VIII, Div. 1 316 SS 10 200 Gas/Solids Separator ASME VIII, Div. 1 321 SS 12 1000 Scrubber-Preconcentrator ASME VIII, Div. 1 '316 LSS/Body 200 gal 12 1000 Venturi Inconel 625 inlet pipe Venturi 200 Pre-concentrator Bod Condenser ASME VIII, Div. 1 316 SS 200 gal Full vacuum 200 TEMA Class "R" to 12-gas side 150-CW side Fines Hopper ASME VIII, Div. 1 304 SS 120 Ft 12 200 CS water

)acket Bed Storage Hopper ASME VIII. Div. 1 304 SS =20 ft 12 700

TABLE I VOLUME REDUCTION SYSTEM PUMPS AND BLOWERS COMPONENT NAME TYPE DESIGN MATERIALS (SEALS)

CAPACITY Dryer Feed Pump Prog. Cavity 0.5 gpm 316SS (Tungsten Carbide Double Mech.)

Scrubber Recirculation Centrifugal 20 gpm 316SS (Tungsten Pump Carbide Double Mech.)+

Liquid Waste Feed Pump Prog. Cavity 2 gpm 316SS (Tungsten Carbide Double Mech.)*

Condensate Pump Centrifugal 25 gpm 316SS (Tungsten" Carbide Double Mech.)*

Caustic Feed Pump Gear 10 gpm 316SS (Tungsten Carbide Double Mech.)*

Air Blower Centrifugal 1200 scfm CS (Packing Gland)

• Seal water required

TABLE J VOLUME REDUCTION SYSTEM HEATERS COMPONENT NAME TYPE POWER DESIGN RATED OTHER Max. (Oper) Voltage Dryer Air Heater Tubular 130 (100)KW 460 Vac 3g 321SS body Incoloy Gas Heater Tubular 45(25)KW 460 Vac 39 321SS body Incoloy Bed Heater Tubular 124 (60) KW 460 Vac 3g Incoloy

TABLE x.

VOLUME REDUCTION SYSTEM FILTERS AND ADSORBERS COMPONENT NAME DESIGN CRITERIA MATERIALS HEPA Filter ASME VIII, Div. 1 Housing Permanent Part (No stamp) 304SS Charcoal Filter Design pressure~12 psig Charcoal Adsorber Design Temperature~210'F i

FIGURE I

'LOCK FLOW DIACIRAN) OF VRS (2 UNIT OPERATIO)ij PACKWAQ FILTE,R YE)ITORI GCRUBBE'R

'F9'. SCII f SAC K%AD

JJOAA60 SCRu513ER, paELocE)llhAIE COND E)IGE DISv, FIOLRSL TAHITI Z4rr0 CsAI FEED TANK PRE-FILTER ff Y TO PLANT TO WASTE IIOLDOP IANK R.TCII ISO g ,

I STENOEPAI OVERFLOW TO CHARCOAL FOR DISK FILTER FLooa DRAI)I PhOOES'hINQ t4 6AI . TREATMENT

. O'ISTQNI YR5 FEED FILTC$L T.'I l. K TANK PART ICMI.*TI. a COHCCtt IAATK

~000 GAL <32 IS) F4/TR 1440 QAI Q TANK5)

C)IARCOAL WASTE FILTCRS EVAPORATOR

'l.6Tl3) x ya SOTTOr45 TO SOLIDS PACKAGING FLUID SEO one<

QASISOL19 STOM,GE HOPPER R/o T L1(A K SEPARATOR Ff NES Hol Pgg C4}l(ENTAAT6 AIR HEATER EVAPORATOTL SOT T Alg QLONEg PRODUCT CO)II/ENVOI)

SEC WaetE EVAPORATOR 8OTTOgS TO S,OLIE)S PACKAQINQ .

<< H) ~91?'R..

I3RS EVAPORATOR 13OTTOIrIS VRS aaAeco Naav)ere INcoaroaArxo CAROLINa R)VIE.IC g CIc))IT FI QUIP (X) DENoTES IO an~' AtTAOTrrr NI'P, S)IEARaV HAaats Qa TANK HAS SAMPLING CANBILILITY OAT% ST DAM OT AttrtOYID ~ra.r r a. n~ Lr. BLOCK FCOW DIAGRAM OF VRS

TABLE L VOLR1E REDUCTION ASH PRODUCT TO SOLIDS PACKAGING Normal 0 eration Specific Total Activity Activity*

Radionuclide ~(pc(/ m) ~(c(/ r)

Co-58 1.30 (+2) (+4)

Co-60 1.67 (+1) 1.41 (+3)

I-131 5.42 4.58 (+2)

I-133 2.59 2.19 (+2)

Cs-134 1.66 1.40 (+2)

Cs-137 1.35 1.14 (+2)

Total 1.577 (+2) 1.334 (+4)

Design Basis Co-58 1.14 (+2) 9.64 (+3)

Co-60 1.50 (+1) 1.27 (+3)

I-131 5.59 (+1) 4.72 (+3)

I-133 3.55 (+1) 3.0 (+3)

Cs-134 7.20 (+1) 6.08 (+3)

Cs-137 5.26 (+1) 4 '4 (+3)

Total 3.45 (+2) 2.915 (+4)

• Based on annual product 8.5 x 10 kg/yr (7321NL Vk] r)

Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER) 0 en Item 179 (DSER Section 11.5-2 Page 11-27)

Question Provide an effluent monitor for the Turbine Building vents (release points 3A and 3B) as required by Table 1 of SRP 11.5 and provide for sampling.

~Res oese The plant airborne effluent release points are shown on Figure 9.4.0-2 and listed on Table 9.4.0-2.

As shown on FSAR Table 12.2.2-2, the Turbine Building Areas shall have negligible airborne concentrations and consequently do not significantly enhance 'airborne radioactive releases. The ventilation of the various turbine building areas and process systems is shown on Figure 9.4.4-1. The ventilation exhaust for these areas and systems is to the turbine building vent, release points 3A and 38. The only radioactive system discharging to this common vent stack is the condenser vacuum pump.

The condenser evacuation system is described in Section 10.4.2 and shown on Figure 10.1.0-4. The exhaust and the effluent treatment system is shown on Figure 9.4.4-1, which will be revised to show radiation monitoring system RE-1TV-3534-1, RE-1TV-3534-2, and RM-1TV-3534-1, for Unit 1, monitoring the exhaust prior to release to release point 3A and monitoring system RE-2TV-3534-1, RE-2TV-3534-2, and RM-2TV-3534-1,'for Unit 2, monitoring the exhaust prior to release to release point 38. The condenser vacuum pump effluent treatment system is described in Section 9.4.4.2.4 and the condenser vacuum pump effluent monitors will be described in revised Section 11 5.2. 7.2.9.

~

The gland seal condenser system is described in Section 10.4.3 and shown on Figures 10.1.0-4 and 1.2.2-71, and includes indication of the system's effluent exhaust location, release point 3E for Unit 1 and release point 3F for Unit 2. These effluent exhausts are monitored by radiation monitors REM-1AE-3536 (Uni.t 1) and REM-2AE-3536 (Uni,t 2) which will be shown on revised Figure 10.1.0-4, and will be described in new Section 11.5.2.7.2.18. (These are existing monitors REM-1TV-3534 and REM"2TV-3534, shown on Figure 9.4.4-1, whose designations will change to REM-1AE-3536 and REM-2AE-3536, respectively.) During hogging, the condenser vacuum pumps discharge to the atmosphere via a short pipe vent, line number 7AE12-19<<1 with a valve 7AE-B9"1, which during hogging provides another potential release point identified as release points 3C and 3D.

Sampling capability for the effluent exhaust streams of the condenser vacuum pump and gland steam condenser is provided on the effluent stream off-line monitors as described in Section 11.5.2. These off-line monitors provide sample points for obtaining grab samples for radiological analysis in the radiochemistry facility described in Section 12.5.

FSAR Figure 9.4.0-2 will be revised to indicate the release points for the gland seal condenser effluent exhaust and the hogging vent for the condenser vacuum pumps discharge.

(7318NLUkj r)

Shearon Harris Nuclear Power Plant Draft Safety Evaluation Report (DSER) 0 en Item 1S1 (DSER Section 11.5.2 Pa e 11-27) 0 en Item 229 (DSER Section 11.5.2 Pa e 11-30)

Question 0 en Item No. 181 The service water system does not possess the capability to obtain a continuous sample as required by Table 2 of SRP 1 1.5.

0 en Item No. 229 Revise Table 9.3.2-2, Figure 9.3.2-2 and Section 9.3.2.2.2 to reflect grab sampling of the service water system.

~Res esse The service water system is described in Section 9.2.1 and shown on Figures 9.2.1-1 and 9.2. 1-2. Normal service water is used to cool both the safety and non-safety grade containment fan coolers during normal operation. The non-safety grade containment fan coolers are not used and are isolated during accident conditions where radiation may be present. The discharged service water from the emergency fan coolers are monitored by radiation monitors discussed in Section 11.5.2.7.2. 1 during normal and emergency operation as shown on Figure 9.2. 1-1. The recommended sampling of the service water is provided as required by SRP 11.5. As required by SRP 11.5 Table 2, the service water is grab sampled at the liquid monitor location as shown on CAR-2165-G-105 and discussed in Section 11.5.2.6.2. The radiation monitor location is shown on Figure 9.2. 1-1. The liquid monitors have sample connections for obtaining representative grab samples of the service water for analysis in the radiochemistry facility. The service water return headers A and B, lines 3SW30-25SA-1, 3SW30-146SB-1 shown on drawing CAR-2165-G-047, Figure 9.2. 1-1, will be provided with sample taps for representative periodic automatic grab samples via two collection vessels. The sampling system will be designed to .include the appropriate components to collect a fixed volume of sample at a rate proportional to the flow in the sample stream to be discharged. Sample vessels will be analyzed in the radiochemistry facility. This sample system meets the requirements of SRP 11.5 for continuous sampling as required by SRP 11.5 Table 2.

The service water is to be sampled at a point downstream of all potentially radioactive inputs and prior to discharge.

(7392NLU cfr)

Power Systems Branch/0. Chopra Open Item 344

OPEN ITEN 344 ORIGINAL QUESTION 430.90 It is not apparent from the information presented in Section 8.2 of

,the FSAR that the design of the on-site power systems and the switchyaxd components meets the requirements of GDC-18, "Inspection and Testing of Electric Power Systems." Describe in more detail this aspect of the design. In particular, describe the capability for testing transfer of power from the unit auxiliary transformers to the start-up transformers (and vice versa) during plant operation.

RESPONSE

The generator lockout relay trip initiates the transfer of power from the auxiliary to the start-up transformers. Westinghouse flexitest switches have been incorporated into each lockout relay's circuitry to provide the capability for testing the relay during plant operation. The flexitest can be opened for the required trip blocks, and the generator lockout relay can be manually tripped.

This action will initiate the automatic fast transfer between the auxiliary and the start-up transformers. The generator lockout relay can then be reset and the flexitest switches closed, after which manual transfer from start-up to auxiliary transformers may be initiated. There is no automatic transfer from the start-up to the auxiliary transformers.

Quality Assurance Branch/J. Spraul Open Item 220 (Partial)

Shearon Harris Nuclear Power Plant Draft SER en Item No. 220 The NRC requested the inclusion of the following listed information in the SHNPP FSAR:

260.0 - Quality Assurance Branch 260.67 Section 17.1.2.2 of the standard format (Regulatory Guide 1.70) requires the identification of safety-related structures, systems, and components controlled by the QA program. You are requested to supplement and clarify the Harris FSAR in accordance with the following:

a. The following items do not appear on FSAR Table 3.2.1-1.

Add the appropriate items to the table or justify not doing so.

1. Containment emergency sumps

2. Equipment hatch and personnel air locks

3. Guard pipes and leak-tight compartments for contain-ment emergency sump recirculation lines and valves

4. Engineered safety fea'tures actuation system

5. Charcoal filters in the control room HVAC system

6. Fuel building radiation monitor

7. Roof and site drainage system including drains, parapets, grading, culverts, and channels

8. AC Onsite Power System (Class 1E)

Diesel generator governor, voltage regulator, and excitation system Instrumentation, control, and power cable splices, connectors, and terminal blocks

- Conduit and cable trays and their supports which do not contain Class lE cables but whose failure may damage other safety-related items Load sequencer AC control power inverters AC vital bus distribution equipment

Shearon Harris Nuclear Power Plant Draft SER en Item No. 220 Cont'd

9. DC Power System (Class lE)

- Conduit and cable trays and their supports ~hich do not contain Class 1E cables but whose failure may damage other safety-related items

- Battery racks

- Protective relays and control panels

10. Radioactivity sampling (air, surfaces, .liquids) ll. Radioactivity contamination measurement and analysis equipment

12. Personnel monitoring equipment (internal, e.g. whole body counts, and external, e.g., TLD system)

13. Instrument storage, calibration, and maintenance program

14. Decontamination facilities, personnel, and equipment

15. Respiratory protection equipment (including testing)

16. Contamination control

b. Clarify Table 3.2.1-1 as noted below or justify not doing so.

l. Provide a commitment that all safety-related instrumentation and controls (I&C) described in FSAR Section 7.1 through 7.6 and other safety-related I&C for safety-related fluid systems will be subject to the pertinent requirements of the FSAR Appendix B QA program. This can be done by a footnote to FSAR Table 3.2.1-1.

2. Under "Electrical Systems and Components" on FSAR Amendment 3 page 3.2.1>>43 is "Safety-related power control and instrument cables and raceways." It appears that a comma should be inserted after "power" and one after "control."

3. Notes 26 and 27 to FSAR Table 3.2.1-1 state that some required information can be found in Table 11.4.2-1..

Table 11.4.2-1 does not have this information, and clari-fication is needed.

c. FSAR Amendment 3, on page 3.2.1-42, includes "Post Accident Monitoring Instrumentation NUREG 0737." However, Enclosure 2 of NUREG-0737 identified numerous other items that are also safety-related or of such importance to safety that they should have the pertinent requirements of the FSAR Operational QA program applied. These items are listed below. Provide such a commitment in Table 3.2.1-1 of the FSAR or justify not doing so.

Shearon Harris Nuclear Power Plant Draft SER en Item No. 220 Cont'd NUREG-0737 (Enclosure 2)

Clarification Item

1. Plant-safety-parameter display console I.D.2

2. Reactor coolant system vents II.B.1

3. Plant shielding II.B.2

4. Post accident sampling capabilities II.B.3

5. Valve position indication II.D.3

6. Auxiliary Feedwater system II.E.1.1

7. Auxiliary feedwater system initiation and flow II.E.1.2

8. Emergency power for pressurizer heaters II.E.3.1

9. Dedicated hydrogen penetrations II.E.4.1

10. Containment isolation dependability II.E.4.2 ll. Instrumentation for detection of inadequate II.F.2 core-cooling

12. Power supplies for pressurizer relief valves, II .G. 1 block valves, and level indicators

13. Automatic PORV isolation II.K.3.1

14. PID controller II.K.3.9

15. Anticipatory reactor trip on turbine trip II.K.3.12

16. Power on pump seals II.K.3.25

17. Emergency plans III.A1.1/III.A.2

18. Emergency support facilities III.A.1.2

19. Inplant I2 radiation monitoring III.D.3.3

20. Control room habitability III.D.3.4

Shearon Harris Nuclear Power Plant Draft SER en item No. 220 Cont'd Res~ense a.1 The containment emergency sump is included under containment building and containment liner which appears on Page 3.2.1-4 of FSAR Table 3.2.1-1. See FSAR Section 3.8.1 for a discussion of the concrete containment. The containment sump screens, which are seismically designed, are listed as part of the Containment Spray System. Page 3.2.1-19 of FSAR Table 3.2.1-1 wi11be amended to reflect the seismic design of the CS sump screens.

a.2 The equipment hatch and personal airlocks appear on Page 3.2.1-5 of FSAR Table 3.2.1-1.

a.3 The guard pipes and leak-tight compartments for containment emergency sump recirculation lines and valves are included as part of the containment building and containment liner which appears on Page 3.2.1-4 of FSAR Table 3.2.1-1. See FSAR Section 3.8.1 for a discussion of the concrete containment.

a.4 The Engineered Safety Features Actuation System appears on Page 3.2.1-44 of Table 3.2.1-1.

a.5 The charcoal filters in the Control Room HVAC System is part of the Control Room HVAC System Emergency Filtration System which appears on Page 3;2.1-32 of Table 3.2.1-1. See FSAR Section 9.4.1 for a discussion of the Control Room HVAC System.

a.6 The fuel building emergency safety effluent radiation monitors are included as part of the Radiation Monitoring System Effluent Monitors which appear on Page 3.2.1-42 of Table 3.2.1-1. See FSAR Section 11.5.2.7.2.3 for a discussion of the Fuel Handling Building Emergency Exhaust moni.tors.

a.7 The Roof and Site Drainage System including drains, parapets, grading, culverts, and channels are non-safety related and therefore are not included as part of Table 3.2.1-1.

a.8 AC On-Site Power Systems (Class lE)

1. The Diesel Generator Governor, Voltage Regulator and Excitation system are part of the Standby Diesel Generator System which appears on Page 3.2.1-36 of the table.

See FSAR Section 8.3.1.1.15, Standby AC Power Supply, for a discussion of the Diesel Generator.

2. Table 3.2.1-1 will be amended to include the instrumentation, control, and power cable solices, connectors, and terminal blocks, for safety related circuits.

Shearon Harris Nuclear Power Plant Draft SER 0 en Item No. 220 Cont'd

~Res oese (Csee'd) a.8 AC On-Site Power Systems (Cont'd) 3.

s Table 3.2.1-1 will be amended to include this item.

4. The load sequencer is included as part of the instrumentation for the Standby Diesel Generator System which appears on Page 3.2.1-36 of Table 3.2.1-1. See FSAR Section 7.3.1.5.1 for a discussion of the Diesel Generator Instrumentation.

5&6. The AC control power inverters and AC vital bus distribution equipment are included as part of the ESF 120V Uninterruptible AC System which appears on Page 3.2.1-43 of Table 3.2.1-1..

See FSAR Section 8.3.1.1.1.4 for a discussion of the 120V Uninterruptible AC System.

ae9 DC Power Systems (Class lE)

1. Table 3.2.1-1 will be amended to include this item.

2. Table 3.2.1-1 will be .amended to include battery racks on Page 3.2.1-43.

3. Protective relays and control panels are considered as part of the instrumentation listed'nder each system.

a.lO-a.16 These items are controlled by CP&L Procedures and will not be added to Table 3.2.1-1. The implementation of procedures is subject to audit by the CP&L Quality Assurance Program.

b.l Note 15 to Table 3.2.1-1 will be amended by adding the following statement:

All IE instrumentation is subject to the pertinent requirements of the Appendix B QA program.

b.2 Page 3.2.1-43. will be amended as noted.

b.3 Notes 26 and 27 to FSAR Table 3.2.1-1 will be amended'to reference FSAR Table 11.4.2-4.

c.l The Plant-safety-parameter display console is non-safety related and will not be added to FSAR Table 3.2.1-1. Interfaces between the SPDS and safety systems will have isolation that meets all the required class IE requirements. The isolation will ensure the integrity of the safety systems. Calibration of safety-related instrument loops comprising the SPDS is included in plant operating procedures which are subject to audit by corporate QA.

Shearon Harris Nuclear Power Plant Draft-SER en Item No. 220 Cont'd

~Res onse (Conc'd) c.2 Table 3.2.1-1 vill be amended to include the Reactor Coolant System Vents.

SHNPP utilizes no portable shielding except for personnel protection during maintenance activities. Use of portable shielding for maintenance activities is governed by plant operating procedures which are subject to audit by Corporate QA.

c.4 The Post Accident Sampling System is not safety-related, and therefore, is not included in Table 3.2.1-1. The use of such equipment is included in plant operating procedures which are subject to audit by Corporate QA. See FSAR Section 9.3.2 for a description of this system.

c.5 Safety related position indication switches have been provided on the pressurizer safety relief valves.

Valve position has been indicated in the main control room. FSAR Section 7.4.1.7 will be amended to reflect this design modification.

c.6 The Auxiliary Feedwater System appears on Page 3.2.1-40 of the table. See FSAR Section 10.4.9 for a description of the Auxiliary Feedwater System.

c.7 The Auxiliary Feedwater System initiation and flow is a part of the Auxiliary Feedwater System instrumentation which appears on Page 3.2.1-41 of the Table. Compliance with this item is described in FSAR Sections 7.2.2, 7.3.1, 7.3.2 and 7.5. Compliance with IEEE-279-1971 is detailed in FSAR Section 7.3.2.2.

c.8 Provi,sions have been provided to manually connect two groups of pressurizers heaters to the ESF power bus. No change is required to Table 3.2.1-1. Compliance with this item is discussed in FSAR Section 8.3.1.2.35.

c.9 This item does not apply to SHNPP since qualified, redundant hydrogen recombiners are provided inside the containment.

See FSAR Section 6.2.5 for a description of the hydrogen re-combiners in containment.

c.10 SHNPP as designed complies with containment isolation dependability requirements. The appropriate equipment and instrumentation is included in FSAR Table 3.2.1-1. Compliance with this item is discussed in FSAR Sections 6.2.4, 7.3. 1.1 and 7.3. 1.3.2.

'C Shearon Harris Nuclear Power Plant Draft SER 0 en Item No. 220 Cont'd

~Res ense (Cont'd)

c. 11 Table 3.2.1-1 will be amended to include the Inadequate Core Cooling (ICC) System. All components of the ICC System inside containment will be class 1E seismic Category I. The inputs from these components are processed by the class 1E, seismic Category I Reactor Vessel Level Indicating System microprocessors.

An isolated non-class lE output from the qualified microprocessor will be data linked to the ERFIS Computer System for primary display on th'e SPDS CRT located on the main control board.

FSAR Section 7.5 will be amended to include a description of the ICC System.

c.12 On-site sources of power are provided for the pressurizer relief valves, block valves and level indicators. Compliance with this item is discussed in FSAR Section 8.3.1.2 '5.

As the on-site power supplies are included in FSAR Table 3.2.l.l, no additional information need be provided.

c.13 Based on the results of the Westinghouse Report, WCAP-9804, dated February 1981, no modifications are necessary. No change will be made to FSAR Table 3.2.1-1.

c. 14 The proportional Integral Derivative controller has the derivative action setting set to zero, thereby eliminating Controller calibration is covered in Plant Calibration it from consideration.

Procedures.'o addition to Table 3.2.1-1 i.s required.

c.15 Compliance with this item is discussed in FSAR Section 7.2.1.

On FSAR Table 3.2.1-1, it is included as part of the Reactor instrumentation which Coolant System appears on Page 3.2.1-5.

c. 16 During normal operation, seal injection flow from the Chemical and Volume Control System is provided to cool the RCP seals and the Component Cooling Water System provides flow to the thermal barrier heat exchanger to limit the heat transfer from the reactor coolant to the RCP internals. In the event of a loss of offsite power the RCP is deenergized and both of these cooling supplies are terminated; however, the diesel generators are automatically started and either seal injection flow or component cooling water to the thermal barrier heat exchanger is automatically restored within seconds. Either of thes cooling supplies is adequate to provide seal cooling and prevent seal failure due to a loss of seal cooling during a loss of offsite power for at least two hours.

Shearon Harris Nuclear Power Plant

~Res onse (Cont'd) c.17 Emergency plans are subject to audit by QA. This information will not be included in Table 3.2.1-1.

c.18 Emergency support facilities are not considered safety-related.

This information will not be included in Table 3.2.1-1.

c.19 Calibration of portable and fixed iodine monitors is covered by plant operating procedures which are subject to audit by QA. This information will not be included in Table 3.2.1-1.

c.20 The components required for control room habitability appear on Pages 3.2.1-31 and 32 of Table 3.2.1-1. See FSAR Section 6.4 as augumented by FSAR Section 9.4.1 for a description of this item.

A marked copy of Table 3.2.1-1 is attached to show revisions that will be made in a future FSAR amendment.

TABLE 3,2.l-l (Contlnuod)

CLASS I F I CAT ION OF STRUCTOlES SYSTEMS ANO COMPONENTS 0esl n and Construction 0 eratlons Raosrks Qua I I ty Qua I I ty Safety Codo Sel sml c Quality Group Assuranco S stems and Com onunts Class ill Codo Class ~Cate or l2l Assuranco (3) (23) (24) c) Norma I ly or automat I ca I ly NNS ANSI (5l ~ I 0 R Isolated f rom parts of systole covorod by a or b d) Operators tor Safety-Related IE IE Q Soe Note (3l)

Actlvo Valvos Instrumentat Ion ( In part) IE IE Q See Note ()5)

Contalnmont S ra S stem Rofuellnq Mater Storaqe Tank 2 ASME I I I 2 Spray Addltlve Tank ASME I I I Contalnmont Spray Pumps 2 ASME III Containment Spray Pump Motors IE IE Eductors 2 ASME III Contalnmont Spray Nozzlos 2 ASME I I I Contalnmont Sump Screens NA System Piping and Valves a) Required for Inltlal InJectlon 2 ASME I I I or long term recirculation of sump ~a ler b) Required for spray add(tive 3 ASME I I I 3 c) Normal ly or automatlcal ly IINS ANSI (5l.l I so lated f rom parts of sys tom covered by a 4 b

TABLF. 3.2.1-1 (Continued)

CLASSIFICATION OF SIRIICTI YSTE HS ANO CO)4 ONENTS Oosl n and Construction 0 eratlons Romarks Quality (4ality Safety Codo So I sml c Qjallty Group Assurance S stems and Com ononts Class (I) Codo Class ~Cats or (2) Assurance (3) (23) (24) lllgh Rango Contalnmont IE Q Soo Note (15)

Radlat Ion Honltorlng Electrical S stems and Co ononts ESF 6,9 kv Bus IE IE ESF 6,9 kv S<<ltchqoar IE ESF 480V S<<ltchgear and IE IE Transformers fSF 480V Notor Control IE ting)

Contors, Including 120V Transformors and Po<<or Panels 480-208/120V Transformers IE (emergency I lgh n.

ESF 120V Unltorruptlble AC System IE IE Q FSF Station Batter lesvand Chargors IE IE Q T3cdLcrg (bc&

Safety related motors 8 IE Q Safety rolalod po<<or control IE IE Q and Instrument cablos and race<<ays orcl a.-"o t~trcl cni2lc or(:.cst io~ncctctr nod tcrhs<~( 12(oaf s r~s <2f ttt "-'~m ~

Contalnmont electrical 2 IE Q Ponetratlons 120 volt D.C Systotn a

'\'u Qr)ntrJ r~cciv~>s

~L S<,)a)or(S (<so<'En<<<<<<<I hara fcis cn(tfcs ugose 4~21<,rc r a3 da sc r2\4rsg Q Qcl tc

(,21,2<< f<...,)S 2l 5 1

2'. l.<r S...r,v-, c a

IN ~

~

TABLE 3.2. I -I (Continued)

CLASS IF I CAT ION OF STRUCTIIRES SYSTEMS ANO COI42ONENTS Oesl n and Construction 0 eratlons Rsusrks Quality ()uallty Safety Code Se I sml c (4a Ilty Group Assurance S ste<<<s and Components Glass lit Coda Cl ss CstorooCr l2l Assurance (3) (23) (24)

Pr l<<<ary and backu p protect I ve IE IE 0 devices associated 2< I th class IE containment elect( Ice I penotrations and 6,9 kv nonwlass IE systera IE Outott~ro Houtroo~Monttorln

~ds too IE IE ESF Protection S ste<<<

(T, P, or S signals) IE IE Q 2

pep Jor CGR~I<ktt (st c'>> 'oc<2YG P pu. Gs ~ Q vJsum 2 fk I'> 'l>>sf ssiC< ~ ~ rats ~ '1 Sm.>>s'd2 d(r21 ~ I w oc4C ,. ~ ( <- I<I 2 g(i okapi~( p.q NIIS

( ee )k<2ld.

("t4 <a.

s T

). n(k i(c)ssR(C Qr<

r C L<<<tl<d']

at(( %2 <<t i < tw 'PdkssA L(= c +(( . rt(

/] ~ s o <~ f. C (t j<<f<2<)<kd d<)'1 I I s</" O'(

Notes to Table 3.2.1-1 (Continued)

(8) Classified on basis that flow restriction is provided in the piping.

(9) " The reactor coolant system supports are not stamped. They are also not

~

part of th'e reactor coolant pressure boundary, and therefore, are not included in any Quality Group.

(10) Failure could cause releases of radioactivity.

r (11) Portions of containment boundary.

(12) Protects fuel from damage during transportation.

(13) Supports for the gas decay tanks are designed to withstand the Operating Basis Earthquake loads.

(14) There is no specific code which classifies the diesel generator as Safety Class 3. IEEE-323 is the governing design code.

(15) Instrumentation Instrumentation required to actuate, maintain operation of, or detect failure of equipment needed to safe shutdown, isolate, and maintain the reactor in a safe condition, and prevent uncontrolled release of radioactivity from the plant is Class IE/Seismic Category I. Instrumentation designated as Class IE/Seismic Category I includes as Seismic Category I all sensing lines, instrument valves and instrument racks. Instrument racks containing Class IE equipment are also considered Class IE.-

Systems noted as Class IE may also contain non-Class IE equipment.

Refer to Chapter 7 for specific identification of Class IE ~ tA C'I equipment P'L$ f L + ff'fp+r4 AP Nw 'Is '4l )cc t% t~Q Q f't A ~

cg, ~4m A gq~"~ v. 8 o~ P<~g<>~'16)

This piping is designed to the Class 3 requirements of ASlK B6PV Code, Section III and fabri. cared in accordance with ANSI Power Piping Code B31.1.

(17) Piping which serves hose stations and standpipes required to protect safe shutdown equipment is designed to ANSI B31.1 requirements and is 'seismically qualified.

(18) Those portions of this system whose failure may have an adverse effect on a nearby safety related component are seismically supported.

(19) The reinforced concrete mat and walls of the Unit 1 Turbine Building between column line 42 (approx.) and 43 (approx.) are designed and constructed to Seismic Category I requirements due to the presence of the diesel generator service water pipe tunnel and Class 1 electrical cable .area above the pipe tunnel (see Figure 1.2.2-60). This area is designed and constructed to withstand the collapse of the Turbine Building concurrent with a SSE.

3.2.1-46 Amendm(nt No. 3

Notes to Table 3.2.1-1 (20) Provides mechanical support for Safety Class 1 component.

(21) Will be designed and fabricated to the applicable portions of ASME III, although it is not classified as ANS Safety Class 1, 2, or 3.

(22) Provides support to the Safety Class 1 pressure boundary conduit.

(23) Quality Group classification in accordance with Regulatory Guide 1.26, Rev. 3, as defined in Section 3.2.2; Class IE is defined in IEEE 308.

(24) Quality Assurance Requirements (Operations Phase)

Q

- QA requirements will meet 10CFR50 Appendix B criteria.

R QA requirements will meet ETSB 11-1 QA requirements as a minimum.

Optionally "Q" requirements may be imposed.

F QA requirements will meet "Q"

Fire Protection OA requirements as a minimum. Optionally requirements may be imposed.

QA requirements of 10CFR50 Appendix B are not mandatory.

(25) The code and code class for individual components in the Liquid Waste Processing System can be found on Table 11.2.1-7.

(26) The code and code class for individual components in the Solid Waste Processing System can be found on Table 11.4.2-f,.

i (27) The ETSB 11-1 QA applies to components listed in Table 11.4.2-P except those listed as manufacturer's standard.

(28) ASME III Code applies to oil cooler and trip/throttle valve only.

(29) Not Stamped.

(30) Structure is designed to withstand design wind/tornado loadings and missile impacts.

(31) Valve operators for active valves, as listed in Tables 3.9.3-13 and 3.9.3-14, are motor, solenoid, or electrohydraulic and qualified to meet IEEE 344 and 323 standards.

(32) Containment water level and containment pressure indication are redunda,.t and class IE up to the isolation device at the SPDS cabinet.

(33) Radiation Monitor is not redundant.

(34) The fuel oil storage tanks are reinforced concrete tanks with steel liners. Alth'ough these tanks are not classified as Safety Class 3, they are designed and constructed commensurate with their intended safety function.

(35) See Figure 9.5.5-1 for ASME III/non-ASME III boundary.

Qow+aawr c~ Sump scree~~ n.~ <g~~lq Le"l>sro ~

3. 2. 1-47 Amendment No. 5

Reactor Systems Branch/E. Harinos Open Items 47, 49

Shearon Harris Nuclear Power Plant (SHNPP)

Draft Safety Evaluation Report (DSER)

Reactor Systems Branch 0 en Item 47 (DSER Section 15.9, a e 15-36)

Provide detailed design information for the Reactor Coolant System (RCS) High Point Vent System.

~Res oese Figure l-l is the diagram of the RCS High Point Vents indicating the valve numbers, power sources and the seismic Category.

2~ The vent system utilizes a 3/8 inch diameter orifice. The size was selected to limit the flow to less than the make-up capacity of one charging pump. Thus, the limiting mass loss is below the definition of the LOCA in 10CFR50, Appendix A. The 3/8 inch orfice in conjunction with one (1) inch (nominal size) piping will provide adequate venting capacity for the anticipated operating conditions.

The system is capable of venting half of the RCS volume (gas) in approximately one (1) hour.

3~ The design temperature and pressure of the piping, valves and components in the vent system are as follows:

Design Pressure 2485 psig Design Temperature 650'F and 680'F The piping and component material used up to and including the second isolation valves are as follows:

Piping ASNE SA-376 TP304 or TP316 Seamless Flange AS'A-182 Gr. F316 Valve Body ASNE SA-182 Type F316 L Bonnet ASHE SA-479 Type 316 Disc ~ ASt4E SA-564 Gr. 630 All materials selected are compatible with reactor coolant chemistry and will be fabricated and tested in accordance with SRP Section 5.2.3 "Reactor Coolant Pressure Boundary Materials."

4~ The RCS vent system is designed to preclude failures that may prevent the essential operation of safety-related systems required for safe reactor shutdown or mitigation of the consequences of a design basis accident:

a) All essential portions of the RCS vent system are seismic Category I.

b) All essential systems required for safe shutdown or mitigation of the consequences of a design basis accident are protected against postulated missiles as described in FSAR Section 3. 5.

The effect from RCS Vent system component failures is considered.

0 en Item 47 Res onse (Continued) c) The RCS vent'ystem does not have piping greater than one-inch nominal size.

d) Fluid sprays from RCS vent system component failures in the normally pressurized portion will not affect tne essential operation of safety related systems required for safe shutdown or mitigation of the consequences of a design basis accident.

5. Table I shows a failure mode and effects analysis for the RCS vent system.

The RCS vent path to containment is loc'ated in-the lower elevation of the containment building. There is no safety related system and/or component located in the vicinity of the discharge point. The area is well ventilated and communicates with the upper space of the containment building, thus, providing good mixing with containment air and prevent accumulation or pocketing of high concentration of hydrogen.

7. The RCS vent system valves will be tested in accordance with ASNE Section XI, Subs IWV, for Category B valves ( reference NUREG 0737, Item .1, Clarification A.(II)). The testing frequency may be educed during cold shutdown.

8. SHNPP is currently revising its Technical Specifications (FSAR Chapter 16). This revision will consider the recommendations of Revision 4 to the NRC's Standard Technical Specifications for Westinghouse Plants. It will also address the appropriate specifications for the RCS High Point Vent System. SHNPP's revised Technical Specifications will be submitted to the NRC during the second quarter of 1984.

9. SHNPP's Fmergency Operating Procedures (EOPs) will be based on the Westinghouse Owner's Group (WOG) Emergency Response Guidelines (ERGs) . In plant procedure FRP-I.3, entitled "Response to Voids in Reactor Vessel," guidance will be provided for using the RCS vent system, and the following information will be included:

a) Entry conditions (through the use-of Critical Safety Function Status trees);

b) Consideration of containment hydrogen concentration in determining when and how long to vent the Reactor Vessel Head; c) Reference to instrumentation used to detect and control upper head voiding; d) Guidance for the Control Operator on the use of the Reactor Vessel Head Vent System.

0 en Item 47 Res onse (Continued)

In addition, the EOPs for Inadequate Core Cooling (ICC) will provide methods other than the RCS venting to assure that decay heat is removed from the core (i.e., RCP operation and primary feed/bleed).

The EOP network will be reviewed to see if additional operator guidance is needed in detailing the methods for determining the size and location of a noncondensible gas bubble.

10. The location of the control instrumentation in the,lfain Control Room required for system operation or testing of the RCS vent system have been reviewed in accordance with the requirements of NUREG-0737 Item

1. D.l, "Control Room Design Review."

The SHNPP FSAR will be amended to reflect this information.

TABLF. I RCS VENT SYSTEH FAIl URE HOPES AND EFFFCT ANALYSIS Effect on S stem Hethod of Detection* Remarks Solenoid a. Fail Failure results in loss Valve position indication Parallel redundant isolatlon Operated Closed of ability to vent the in the Hain Control Room valve 2RC-V281SA-1 (2RC-V28088-1)

Valves reactor vessel allows ventinp of the reactor 2RC-V280SB-1 vessel.

(2RC-V201SA-1 analogous)

b. Fail No impact on normal Valve position indication Redundant isolation valves to PRT Open operation. in the Hain Control Room (2RC-V285SB-1) and containment Inability to vent the (2RC-V284SA-1) preclude uncontrolled pressurizer without also venting of the reactor vessel.

venting the reactor vessel

2. Solenoid a. Fail Lose the ability to vent Valve position indication Parallel redundant isolation valve operated Closed the pressurizer in the Hain Control Room 2RC-V283SA-1 (2RC-V284SA-1) allows valves venting of the pressurizer.

2RC-V282SB-1 (2RC-V283SA-1 b. Fail No impact on normal Valve position indication Redundant isolation valves to PRT analogous) Open operation. in the Hain Control Room. (2RC-V285SB-1) and containment Inability to vent the (2RC-V284SA-1) preclude uncontrolled reactor vessel without also venting of the pressurizer.

venting the pressurizer.

3. Solenoid a. Fail No impact on normal Valve position indication Al terna t e vent inp of reac tor vessel Operated Closed operation. in the Hain Control Room or pressurizer to the containment is Valve Lose the ability to vent available via 5solation valves 2RC-V285SB-1 the reacor vessel or 2RC-V284SA-I pressurizer to pressurizer relief tank (PRT)-

b. Fail Lose the ability to Valve position indication Redundant isolation valves For the Open isolate tne pressurizer in the Hain Control Room reactor vessel (2RC-280SB-1, relief tank (PRT) from 2RC-V283SA-1) and pres. urizer the RCS vent system (2RC-V282SB01, 2RC-V283SA-1) prevent uncontrolled ventinp to the pressurizer rel.ief tank (PRT).

TABf.F. I (Continued)

RCS VENT SYSTEH FAII.IJRI.'HODFS AND FFFECTS ANALYSIS Failure Node Effect on S stem Ifethod of Detection* Remarks

4. Solenoid 'a. Fail No impact on normal Valve position indication Alternate venting of reactor vessel Operated Closed operation. in the )fain Control Room or pressurizer to the pressurizer Valve Lose the abilf.ty to vent relief tank (PRT) is available via 2RC-V284SA-1 the pressurizer to the isolation valve 2RC-V285SB-l.

containment

b. Fail Lose the ability to Valve position indication Redundant isolati.on valves for the Open isolate RCS vent system in the Ifain Control Room reactor vessel (2RC-V280SB-1) and

.from containment. pressurizer (2RC-V282SR-1, 2RC-V283SA-1) prevent uncontrolled venting to the containment.

5. Flow switch a. Fail to No impact on normal Valve position indication Valves RC-V284SA-1 and 2RC-V285SB-1 FS-5752-S alarm on operation. in the Ifain Control Room provide redundant isolation to the pre- Loss of ability to detect prevent uncontrolled venting of RCS.

sence of leakage into the vent condensate/ system piping steam

b. Spurious No impact on normal Valve position indication alarm operation. in the Hain Control Room Loss of ability to detect leakage into the vent system piping.

6. Position False indication Loss of ability to deter- Flow switch (FS-5752-S) indicator of valve mine valve position in alarm indicates valve is for position reactor vessel vent line. opened 2RC-V280SB-1 and 2RC-V281SA-1

7. Position False indication Loss of ability to deter- Flow switch (FS-5752-S) indicator of valve mine valve position in alarm indicates valve is for position pressurizer vent line. opened.

2RC-V282SB-1 and 2RC-V283SA-1

TABLE I (Continued)

RCS VENT SYSTEM FAILURE MODES AND EFFECTS ANALYSIS Failure Mode Effect on S stem Method of Detection" Remarks

8. Position False indication Loss of ability to deter- Flow switch FS-5752-S and indicator of valve mine valve position in the pressurizer relief tank for position vent line to pressurizer pressure and temperature 2RC-V285SB-1 relief tank (PRT). can verify valve position.

9. Position False indication Loss of ability to deter- Flow switch FS-5752-S and indicator of valve mine valve position in the containment humidity/rad-for position vent line to the contain- iation levels can verify 2RC-V284SA-1 ment. valve position.

• As part of plant operation, periodic test, surveillance inspections and instrument calibrations are made to monitor equipment & performance. Failures may be detected during such monitoring of equipment in addition to detection methods noted.

(7 53 7P SAc f r)

Shearon Harris Nuclear Power Plant (SHNPP)

Draft Safety Evaluation Report (DSER)

Reactor Systems Branch 0 en Item 49 (DSER Section 5.4.7.5, page 5-26)

Address the requirements of BTP 5-1 and demonstrate that suitable plant systems and procedures are available to place the plant in a cold shutdown condition, with only offsite or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.

~Res oose The SHNPP's cold shutdown methodology per the requirements of SRP 5.4.7 is as follows:

I ~ ACHIEVING COLD SHUTDOWN The plant has been evaluated to demonstrate actions needed to achieve cold shutdown conditions following a safe shutdown earthquake, loss-of-offsite power, and the most limiting single failure. To achieve and maintain cold shutdown, the following key functions must be performed:

1) residual heat removal, 2) boration and inventory control, and

3) depressurization.

1. Residual Heat Removal The function of residual heat removal is performed in two stages to accomplish the cooldown to cold shutdown.

The first stage is cooldown from operating temperature to 350'F.

During this stage, circulation of the reactor coolant is provided by natural circulation with the reactor core as the heat source and the steam generators as the heat sink. Steam is initially released via the 1fain Steam relief and safety valves. This occurs automatically as a result of turbine and reactor trip. Steam release for cooldown occurs via the Hain Steam power-operated atmospheric relief valves. As the cooldown proceeds, the operator adjusts the amount of steam dump to permit a reasonable cooldown rate. Feedwater makeup may be provided by the auxiliary feedwater system during loss of power or accident conditions.

The Hain Steam power operated relief valves (PORV's) are Safety Class 2, seismic Category I and seismically and environmentally qualified. The Applicant has reviewed the arrangement and has determined the most limiting single failure to be the loss of a division of electrical power which renders the components of that division inoperable. However, even with such a failure, two Steam Generators are available for cooldown of the plant to the conditions needed to allow initiation of the Residual Heat Removal System (RHRS) (i.e., approximately 350'F and 425 psig).

0 en Item 49 Res onse (Continued)

For more information on the ifain Steam System see FSAR Section 10.3.

The Auxiliary Feedwater System (AFWS) is Safety Class 2, seismic Category I and seismically and environmentally qualified. The AFWS is comprised of two separate subsystems with sufficient alignment capability and flow capacity to ensure that sufficient water can always be provided to any combination of the three steam generators. The first subsystem is comprised of two motor-driven pumps, one is controlled and powered from the Emergency A Bus and the other is controlled and powered from the Fmergency B Bus. The second system is comprised of a turbine-driven pump controlled from the Emergency DC B Bus and is powered by steam supplied from Steam Generators B and/or C. Any one of the three AFW pumps can supply sufficient flow for cooldown of the plant. For more information on the AFWS see Section 10.4.9.

The second stage is from 350'F to cold shutdown. During this stage, the RHRS is brought into operation. Circulation of the reactor coolant is provided by the RHR pumps, and the heat exchangers in the RHRS act as the means of heat removal from the Reactor Coolant System (RCS). In the RHR heat exchangers, the residual heat is transferred to the Component Cooling Water System (CCWS) which then transfers the heat to the Service Water System (SWS) with final heat rejection to the Ultimate Heat Sink (UHS) ~

The RHRS, CCWS and the SWS are safety grade, seismic Category I with redundant trains powered from the Emergency A Bus and the Emergency B Bus. Further information on these systems can be found in FSAR Sections 5.4.7 (RHRS), 9.2.1 (SWS), 9.2.2 (CCWS) and 9.2.5 (UHS).

2~ Boration and Inventor Control Boration is accomplished using portions of the Chemical and Volume Control System (CVCS). The boric acid transfer pumps supply four weight percent boric acid from the boric acid tanks to the suction of the centrifugal charging pumps which inject the borated water into the RCS via the normal charging and/or reactor coolant pump seal injection flow paths. Hakeup in excess of that required for boration can be provided from the Refueling Water Storage Tank (RWST) using the centrifugal charging pumps and the same injection flow paths as described for boration. Two motor-operated valves, one powered from Emergency A Bus, the other from Emergency B Bus and connected in parallel, transfer the suction of the charging pumps to the RWST. The CVCS is Safety Class 2 and 3, seismic Category I and seismically and environmentally qualified. For additional information refer to FSAR Section 9.3.4.

3~ De ressurization Depressurization is accomplished using portions of the CVCS. Either four weight percent boric acid or refueling water can be used as desired for depressurization with the flow path heing from the centrifugal charging pumps to the auxiliary spray valve to the pressurizer.

0 en Item 49 Res onse (Continued)

As an alternative, depressurization could be accomplished, by discharging RCS inventory from the pressurizer to containment via the pressurizer power-operated relief valves. This operation can be integrated with the cooldown function near the end of the cooldown to 350'F. As RCS inventory is relieved to the containment, the pressurizer temperature and pressure is reduced, thus reducing the pressure in the RCS. Makeup is provided as necessary to maintain a minimum level in the pressurizer.

II. SINGLE FAILURE EVALUATION

1. Residual Heat Removal A. From Hot Standby to 350'F Reactor coolant loops and steam generator Three reactor coolant loops and steam generators are provided, any one of which can provide natural circulation flow for adequate core cooling. Even with the most limiting single failure, two of the reactor coolant loops and steam generators remain available.

Main Steam PORV's Three valves are provided (one per generator) .

Auxiliary feedwater pumps Three (two motor-driven and one steam-driven) auxiliary feedwater pumps are provided. Each pump can provide sufficient cooling water to any combination of steam generators.

Auxiliary feedwater flow control valves - Electro hydraulic operated, fail open valves are provided. In the event of a single failure of one flow control valve (which affects flow to one steam generator from either a motor-driven pump or the steam-driven pump), auxiliary feedwater can stall be provided to any combination of steam generators from the other pumps.

Condensate storage tank Upon depletion of the primary source of auxiliary feedwater in the seismic Category I condensate storage tank, a backup source of auxiliary feedwater is the UHS via either train of the SWS.

B. From 350'F to Cold Shutdown RHR Pumps 1 and 2 Two RHR pumps are provided, either one of which can provide adequate circulation of the reactor coolant.

In the event of a single failure, either pump can provide sufficient RHR flow.

0 en Item 49 Res onse (Continued)

RHR Suction Isolation Valves 8701A and 8702A (to RHR Pump 1) and 8701B and 8702B (to RHR Pump 2) The two valves in each RHR

,subsystem are each powered from different emergency busses (8701A and 8701B from A bus; and 8702A and 8702B from B bus).

Failure of either emergency bus can prevent initiation of RHR cooling in the normal manner from the control room. In the event of such a failure, the affected valve(s) can be deenergized and opened with its handwheel or can be opened using alternate power via operator action outside of the control room. Therefore, any single failure can be tolerated since it would only affect one of the two redundant RHR subsystems.

RHR Heat Exchangers 1 and 2 If either heat exchanger is unavailable for any reason, the remaining heat exchanger can provide sufficient heat removal capability.

RHR Flow Control Valves HCV-603A and HCV-6038 If either of these normally open, fail open valves closes spuriously, sufficient RHR cooling can be provided by the unaffected RHR subsystem.

RHR/Safety Injection System Cold Leg Isolation Valves 8888A and 8888B - If either of these normally open, motor-operated valves (each is powered from a different emergency bus) closes spuriously, sufficient RHR cooling can be provided by the unaffected RHR subsystem.

Component Cooling Water System Two redundant subsystems are provided for safety-related loads. Either subsystem can provide sufficient heat removal via one of the RHR heat exchangers.

Service Water System Two redundant subsystems are provided for safety-related loads. Either subsystem can provide sufficient heat removal via one of the component cooling water system heat exchangers.

2. Boration and Inventor Control Boric Acid Tanks 1 and 2 Two boric acid tanks are provided. Each tank contains sufficient four weight percent boric acid to borate the RCS for cold shutdown.

Boric Acid Transfer Pumps 1 and 2 Each pump is powered from a different emergency bus. In the event of a single failure, either pump can provide sufficient boric acid flow.

Isolation Valve 8104 If valve 8104 (located between the boric acid transfer pump discharge and the CVCS centrifugal charging pump suction), which is supplied from the Emergency B Bus and is normally closed, cannot be opened due to bus or operator failure, it can be

0 en Item 49 Res onse (Continued) opened locally with its handwheel. If valve 8104 cannot be opened with its handwheel, an alternate flow path is available via air-operated, fail open valve FCV-113 and normally closed manual valve 8439.

Refueling (later Storage Tank Isolation Valves LCV-11SA and LCV-11SB Each valve is powered from a different emergency bus; only one of these normally closed motor-operated valves needs to be opened to provide a makeup flow path from the RUST to the CVCS centrifugal charging pumps.

Centrifugal Charging Pumps A, B, and C Pump A is powered from the Emergency A Bus, pump B is powered from the Emergency B Bus, and pump C can be powered from either emergency bus (see FSAR Subsection 8.3.1.1.2.4). In the event of a single failure, any one pump can provide sufficient boration or makeup flow.

Flow Control Valve FCV-122 This valve (located on the CVCS centrifugal charging pump discharge) is normally open and fails open on loss of air. If FCV-122 closes spuriously, the CVCS centrifugal charging pumps can safely operate on their miniflow circuits.

Efforts would be made to open it or its manual bypass valve (8403) can be opened. Boration can be accomplished by realigning the CVCS centrifugal charging pumps discharge in order to use the boron injection or SIS high-head injection in flow path.

informal Charging Isolation Valves 8107 and 8108 If either of these normally open, motor-operated valves, each of which is powered from a different emergency bus, closes spuriously, operator action can he used to deenergize the valve operator and reopen the valve with its handwheel. Boration can be accomplished by realigning the CVCS centrifugal cha1ging pumps discharge in order to use the boron injection tank on the SIS nigh-head injection flow path.

Normal Charging Valve 8146 If this normally open, fail open valve closes spuriously, alternate charging valve 8147, which fails open, can be used. Boration can be accomplished by realigning the CVCS centrifugal charging pumps discharge in order to use the boron injection tank on the SIS high-head injection flow path.'

~ De ressurization Pressurizer Auxiliary Spray Valve 8145 This normally closed valve fails closed on loss of air. In this case, 8145 can be opened by using a portable nitrogen bottle. If 8145 is stuck closed as a result of a single failure, the pressurizer power-operated relief valves can be used to depressurize the RCS by discharging the pressurizer inventory to the pressurizer relief tank.

0 en Item 49 Res onse (Continued)

Charging Valves 8146 and 8147 - These valves fail open on loss of air. In this case, isolation is provided by two check valves in series with 8347 and 8378 or 8346 and 8379 respectively.

RHR Suction Valves 8701A, 8702A and 87028 See Section II.I.B above.

(7536PSAccc)

Structural Engineering Branch/S. B. Kim Open Item 339

e Shearon Harris Nuclear Power Plant Draft SER 0 en Item No. 339 Tabulation of loadings at the anchor point of mainsteam and feedwater pipes on containment wall.

~Res oese A tabulation of loadings is attached to provide the above information.

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Old New Old New Old New Old New Old New Old New R-HMS-2-1B i56 +63 i537 %483 +1326 +812 %36623 +41803 +5533 +5841 +2595 41478 R-HMS-2-2B +110 i37 +405 i454 %1022 +852 +36456 +42656 410128 +3520 17025 =

+948 R-HMS-2-3B +538 %318 +1043 +1039 i36737 J39075 R-HMS-2-5B +78 i101 +1446 i1176 i11844 +9235 %34 +12 NOTE: Pipe rupture loads are the resultant loads due to piping inside and outside containment.

(7325NLllkjr)

PIPE RUPTURE LOADS AT FEEIXfATER PENETRATION MAXIMIM REACTXONS Break No. Maximum Time of and Run No. Reaction Max. Reacts Fx lbs) F (lbs) F (lbs) H(in-lbs) H (in-lbs) M (in-lbs)

%.06984 +55601 %8712 %84279 %769639 %5014101 +360504 i.04670 j4989 260150 %45748 %2391272 22017240 %152465 RIIFW  %.00323 i372 +2702 2329502 %8258 +1397 +216125 15-1A +.05639 i22082 %58603 %356502 +3466431 +2966556 +442416

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(7325NLUkjr)

r ~

0 l

TABLE OF CONTENTS Page

" INTRODUCTION+

DI SCU S SION o

~ ~...... ~ ~ ~ . ~ ~ i o ~ ~ ~ ~ ~ ~ o ~ o.... ~ ~ . ~ ~ a ~ .

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ o. ~ ~ ~ ~ ~

~ ~ ~ ~ ~

2 2

l. De fense-x.n-Depth............. . ~ ~ ~ ~ ~ ~ 3 2~ Use of Water on Electrical Cable Fires............, ~ ~ ~ ~ ~ 4 3~ Establishment and Use of Fire Areas.. ............. ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ 5 4~ De fxnztXOnS............................,...........

f ~ ~

~ ~ ~ ~ ~ 5 C. PO SIT ION o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

Fire Protection Program Requirements. ~.............

ao Fire Protection Program........................ ~ ~ ~ 8

b. Fare Hazards Analysis........ ~ ~ ~ . ~ ~ ~ .~.~.~~ ~~~ ~ ~ ~ ~ 12 C ~ Fire Suppression System Design Basis........... ~ ~ ~ ~ ~ ~ ~ ~ 14. ~

d~ Alternative or Dedicated Shutdown.............. ~ ~ ~ ~ ~ ~ ~ 14 ~ ~

e. Implementation of Fire Protection Programs..... ~ ~ ~ ~ ~ ~ ~ ~ 15 ~

2~ I

~ ~

Admxnxstratxve Controls............................ ~ ~ ~ 15 3~ F ire Brigade....................................... ~ ~ ~ ~ ~ ~ ~ ~ ~ 18 4~ Quality Assurance Program.......................... ~ ~ ~ 22

a. Design and Procurement Document Control........ 0 ~ ~ 22

b. Instructions, Procedures, and Drawings......... ~ ~ ~ 22 C ~ Control of Purchased Material, Equipment, and S ervt.ces. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 23

d. Inspection ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ 23

e. Test and Test Control.......................... ~ ~ ~ 23

f. Inspection, Test, and Operating Status. ~ ~ ~ ~ ~ . .~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 23 go Nonconf orming Items............... ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 23 h Corrective Act@on... ~ . ~ .. ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 23 lo R e cord s ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ 23 3 ~ Audlts d ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 23 5 ~ General Plant Gul.delxnes........................... 24 a~ Buxldxng Desi.gn...woo.o...woo.....:.-.o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 24

b. Safe Shutdown Capability.......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 29 C ~ Alternative or Dedicated Shutdown Cap abxlxtyo ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ 30

d. Control of Combustibles.............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 32

e. Electrical Cable Construction, Cable Trays, and Cable- Penetratxons.......... ~ . ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 34

f. Ventxlatxon.......................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 37 g ~ Lighting and Communication........ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 38

6. Fire Detection and Suppression....... ~ . ~ .. ~ ~ ~ ~ ~ 39

a. F are Detection....................... ~ ~ ~ ~ ~ ~ ~ t~~~ ~ ~ 39

b. Fire Protection Water Supply Systems... ~ ~ ~ ~ ~ ~ ~ 40 C ~ Water Sprinkler and Hose Standpipe Systems. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 45 dt Halon Suppression Systems. ~ ~ ~ ~ ~ .. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 48

e. Carbon Dioxide Suppression Systemic . ~ ..... ~ ~ ~ 40

f. Portable Extxnguxshers..................... ~ ~ ~ 49

$ 30'Q01 6

TABLE OF CONTENTS (Cont'd)

Page

7. Guidelines for Specific Plant Areas................... ~..... 49 a~ Primary and Secondary Containmen't.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 49

b. Control Room Complexes.-...ooooo..o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 52 c~ Cable Spreading Room............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 55 d~ Plant Computer Rooms............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 56

e. Swxtchgear Rooms................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 57 Remote Safety-Related Panels....... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 57 go Safety-Related Battery Rooms. ~..... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 58

h. Turbine Building................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 58 lo Diesel Generator Areas............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 59 Q ~ Diesel Fuel Oil Storage Areas...... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o 61

k. Safety-Related Pumps............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 61

l. New Fuel Area...................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ 62

m. Spent Fuel Pool Area............... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 62 Radwaste and Decontamination Areas. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 63 0~ Safety-Related Water Tanks......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ 63 po Records Storage Areas.............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 63 Cooling Towers ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ 64

r. Miscellaneous Areas................ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ 64

8. Special Protect>on Guxdelxnes............................... 64

a. Storage of Acetylene-Oxygen Fuel Gases. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 64

b. Storage Areas for Xon Exchange Resins.. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 64 C ~ Hazardous Chemicals........ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 65

d. Materials Containing Radioactivity..... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 65

BRANCH TECHNICAL POSITION CMEB 9.5-,1 (Formerly BTP ASB 9.5-1)

GUIDELINES FOR FIRE PROTECTION FOR NUCLEAR POWER PLANTS

Your fire protection program will be reviewed to the guidelines of BTP CHES 9.5"1 (NUSEG-0800), July 1981 Provide a comparison that shows conformance: of the plant fire protection program to these guidelines. Deviations from tne guidelines should be specifically identified. A technical basis should be provided for each deviation.

NUREC-0800 CUIDFLINFS CONFO'RNA'NCR A. INTRODUCTION General Desfgn Criterion 3, "Fire Protection" of Appendix A, "General Followed ln design Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Wccnsfng of Production and Utllizatlon Facilities," rcqufres that structures, systems, and components important to safety be designed und located to mlnlmlre, consistent with other safety requirements, tne probability and effect of fires end exploslons. Noncombustible and heat-resistant materials src required to be used wherever practical throughout the unit, particularly in locations such as the containment end control room ~

Criterion 3 also requires that fire detection and suppression systems of appropriate capacity and capability be provided and designed to minimize thc adverse effect of fires on structures, systems, and components important to safety end that fireflghtlng systems be designed to ensure that their Eailure, rupture or inadvertent operation docs not significantly impair the safety capability of these structures, systems'nd components.

This Branch Tcchnical Position (BTP) presents guidelines acceptable to the Followed in design NRC staff Eor implementing this criterion ln the development oE e Eire protection program Eor nuclear power plants. These revised guidelines include the acceptance criteria listed ln e number of documents, including Appendfx R to 10 CFR Part 50 and 10 CFR Part 50, end S0.48. The purpose of the fire protection program ls to ensure the capability to shut down the reactor and maintain ft In a safe shutdown condition und to minimize radioactive releases to the environment ln the event of e fire. It implements the philosophy oE defense-ln-depth prvtcctlon agufnit'he hazards of fire snd its associated effects on safety-related equipment.

If designs or methods different from tNe guldellncs recommended herein are used, they must provide equivalent Eire protection. Suitable bases and gustfffcatfon should be provided for alternatlvc approaches to establfeh acceptable Implementation of General Design Criterion 3.

This BTP addresses fire protection programs for safety-related systems and Followed in design equfpmcnt and for other plant areas containing fire hazards that could adversely affect safety-related systems. It does not give guidance for protecting the life or safety of the site personnel or for protection against economic or property loss. This document supplements Regulatory Cu! tc 1.7S, "Physical Independence of Electrical Systems," fn determining the Efre protection for redundant cubic systems.

B. DISCUSSION There have been numerous fires in operating U.S. nuclear power. plants through December 1975 of which 32 were important enough to rcport. Of these, the Eire on Harch 22, 1975 et Browne Ferry nuclear plant was the most severe. With approximately 250 operating reactor years of experience, one may infer a frequency on the order of one fire per 10 reactor years. Thus, on the average, a nuclear power plant may experience one or more ffrcs of varying scvcrlty durinF ltu operating liEe. Alihouph MASN-1400, "Reactor Safety Study An ussessmcnt of Accident Risks ln U.S.

Commercial Nuclear Power Plants," dated October 1975 concluded that thc

~

Drowns Ferry fire did not af [ect the vaf ldlty nf flu. overall risk asuessmcnt, the stuff concluded that cost-oflectlvc fire protection meu,:urea should bc instituted tu algol fluently decrease thc frequency und

NURKC-OBUU CUIDKLINFS COHFORHANCK severity of fires and consequently initiated the development of this.BEP.

In this development, the staff made use of many national standards and other publications related to fire protection. Thc documents discussed belou vere particularly useful.

A document entitled "The International Guidelines for the Fire Protection of Nuclear Pover Plants" (ICL), I974 Edition, Second Reprint ~ pubitshed on behalf of the National Nuclear Risks Insurance Pools and Association, provides a step-by-step approach to assessing the Eire risk in a nuclear pover plant and describes protective measures to be taken as a part of thc fire protection of these plants. It provides useful guidance in this important area. The Huclear Knergy Liability and Property Insurance Association (Hgl.PIA) and the Hutual Atomic Energy Reinsurance Pool (HAKRP) have prepared a document titled "Specifications for Fire Protection of Hcu Plants," uhich gives general conditions and valuable criteria. A special revieu group organised by HRC under Dr Stephen H Hansuer, Technical Advisor to the Kxecutive Director for Operations, to study the Brouns Ferry Fire, issued a report, NURKC-0050, "Recommendations Related to Brouns Ferry Fire," in February 1976, vhich contains recommendations applicable to all nuclear pouer plants. This BTP uses the applicable information contained in these documents.

The fire protection program for a nuclear pouer plant presented in this BTP consists oE design features, personnel, equipment, and procedures that provide the defense-in-depth protection of the public health and safety.

The purpose of the program is to prevent significant fires, to ensure the capability to shut dovn thc reactor and maintain it in a safe shutdovn condition, and to minimiae radioactive releases to the environment in the event of a significant Eire. To meet these objectives, it is essential that management participation in the program begin uith early design concepts and plant layout, uork and continue through plant operation and that, a qualified staff be responsible for engineering and design of fire protection features that provide Eire detection annunciation, conEinemcnt, snd suppression for the plant. The staff should also be responsible for fire prevention activities, maintenance oE fire protection systems, training, and manual firefighting activities't is the combination of all these that provides the ncedcd defense"in-depth protection of the public health and safety-Some of the major conclusions that emerged from the Brovns Ferry fire investigations uarrant emphasis and are disccussed below.

l ~ Defense-in-De th Nuclear PoMer plants use the concept of deEense"in-depth to achieve the Defence in depth wss considered in the design.

required high degree of safety by using echelons of safety systems. This concept is also applicable to Eire safety in nuclear poser plants'ith respect to the fire protection program, the dc(ense-in"depth principle is aimed at achieving an adequate balance in:

a. Preventing fires from starting;

b. Detecting fires quickly, suppressing those fires that occur, putting thi m out quickly, and limiting their damage; and

HUREC&800 CUIDKI.IHES CONFORHA NCE

c. Designing plant safety systems so that a fire that starts in spite of the fire prevention program and burns for a considerable time in spite of Eire protection activities vill not prevent essential plant safety functions from being performed.

No one oE these echelons can be perfect or complete by itself. Each echelon should meet certain minimum requirements, however, strengthening

.any one can compensate in some measure for weaknesses, knovn or unknovn, in the others.

'Ihe primary ob)ective of the fire protection program is to miniml'ie both Considered in the design.

the probability and consequences of postulated fires. In spite oE steps taken to reduce the probabi.lity oE fire, fires are expected to occur.

'Iherefore, means are needed to detect and suppress fires with particular emphasis on providing passive and active fire protection of appropriate capability and adequate capacity for the systems necessary to achieve and maintain safe plant shutdovn vlth ot'ithout oifsite power, For other safcty-related systems, the Eire protection should ensure that a fire vill not cause the loss of function of such systems, even though loss of redunda'ncy within a system may occur as a result of the fire. Generally, in plant areas where the potential fire damage may )eopardixe safe plant shutdown, the primary means of fire protection should consist of fire barriers and fixed automatic fire detection and suppression systems.

Also, a backup manual firefighting capability should be provided .-

throughout the plant to limit the extent o! fire damage. Portable equipment consisting of hoses, noxxles, portable extlnguishers, complete personnel protective equipment, and air breathing equipment should be provided for use by properly trained firefighting personnel. Access Eor effective manual application of fire extinguishing agents to combustibles should be provided. The adequacy of fire protection for sny particular plant safety system or area should be determined by analysis of the:

effects of the postulated fire relative to, maintaining the ability to safely shutdovn the plant and minimize radioactive releases to the environment in the event of a Eire.

Fire protection starts vith design and must be carried through sll phases Complied vith the intent. As stated on page 9.5.1"48 of Subsection 9.5-1 of construction and operation. A quality assurance (QA) program is needed of the FSAR a quality assurance program has been developed for fire to identify and rectify errors in design, construction, and operation and protection. %be Design Construction QA program ls described in thc PSAR is an essential pert of defense"in<epth. and uas approved by the HRC. However, for components of the fire protection system designed, specified, procured, manufactured, fabricated or installed prior to institution of the Fire Protection IIA program (February 8, 1977) the program was folloved to the extent practicable.

'ihe Engineering snd Construction fire protection quality assurance program uas approved by the NRC during construction permit review, 'Ihe Operational Quality Assurance Program is described in Section 17.2 oE the FSAR,

2. Use of Mater on Electrical Cable Fires Experience uith wa]or electrical cable fires shous that vater vill Complied with.

promptly extinguish such fires, Since prompt extinguishing of the fire is vital to reactor safety, fire and water damage to safety systems is reduced by the morc efficient application of vater from fixed systems spraying directly on the fire rather than by manual application vith fire hoses. Appropriate flrefightlng proccdurca and fire training should provide thc techniques, equipwcnt, and skills for thc usc of water In fighting electrical cable fires in nuclear plants, part fcularly in areas l

containing a h lgh concentration of electr c cables ul th plastic insulation.

NUREC-0800 GUIDELINES 'ONFORYJlNCE This is not to ssy that fixed vater systems should be installed Complied vith.

everyvhere. Equipment that may be damaged by voter should be shielded or relocated avay from the fire haxard and the vster. Drains should'be provided to remove any vater used for Eire suppression and extinguishment to ensure that voter accumulation does not incapacitate safety-related equipment.

3. Establishment and Use of Fire Areas Separate fire areas for each division of safety-related systems vill Complied vith the intent. As stated in the FSAR on page 9.5.1-1, separate reduce the possibility oE fire-related damage to tedundant safety-related fire areas, vhlch reduce thc possibility of fire related damage to equipment. Fire areas should be established to s>>parat>> redundant safety redundant safety-related trains, vere established to separate redundant divisions and isolat~ safety"related systems from fire hazards in safety divisions and to isolate safety related systems from haxatds in nnnsafety-related areas. Particular design attention to the use of non-safety related areas to the extent possible in the previously isolated fire areas fot redundant cables vill help to avoid loss

'eparate established plant design established priot to issuance of NUREC-0800 (See oE redundant safety-related cables. Separate fire ateas should also be first parsgtaph on FSAR page 9.5.1-1 and Response to Comment on Page 4 employed to limit the spread of fires bctveen components that sre maJor Item I), vhere fire barriets could not be installed to separate redundant fire ha>>atda vithin a safety division. @here redundant systems cannot be systems, alternate means, as permitted by Appendix A to Branch Technical separated by Eire barriers, ss in containment and the control toom, it is Position APCSB 9.5-1 Rev 0 Cuidelines for Fire Protection Eor Pover Plants necessary to employ other measures to prevent a fire from causing the loss Docketed Prior to July I, 1976, such ss limitation of the amount of of function of safety-related systems. combustible materials through administrative procedures utilixstion of Eire tesistive construction, provision of fire breaks snd/or fire Mlthfn fire areas containing components of a safety-related system, . teterdant. coatings In cable trays, snd installation oE Eire detection special attention should bc given to detecting and suppressing fires that systems snd automatic Eire extinguishing systems ot combined thetof.

may adversely effect the system. Heasures that may be taken to reduce the effects of s postulated fire in a given fire area include limiting the amount of combustible materials, installing fire-resistant construction>

providing fire rated barriers for cable trays, installing fire detection systems snd fixed fire suppression systems> or providing other ptotection suitable to the installation. The fire hsxatd analysis vill'e the, mechanism to determine thc Eire areas have been properly aelected.

Suitable design of the ventilation systems can limit the consequences of a fire by preventing the spread of the products oE <<ombustlon to other fire areas. It is important that means be provided to ventilate, exhaust, or isolate the fire area as required and that consideration be given to the consequences of failure of ventilation systems due to fire causing lose of control for ventilating, exhausting, or isolating a given fire area. Ihc capability to ventilate, exhaust, or isolate is particularly important to ensure the habitability of rooms or spaces that must be attended ln an emergency. In the design, provision should be made for personnel access to and escape routes from each fire area.

4. Definitions For the user's convenience, some of the terms related to fire protection ste presented belov vith their definitions as used in this BTP. J Approved - tested and acc>>pted for a specific purpose or application by a net~one ly recognised tcstlng laboratory.

NURKC-UBUU CUIDCLINBS CONFORNANCK Automatic - self-acting, operating by its ovn mechanism >>hen actuated by some impersonal influence such as change in current, pressure, temperature, or mechanical configuration.

Combustible Haterisl - material that does not meet the definition of noncombustible.

Control Room Com lex " the rona served by the control room emergency ventilation system (see SRP Section 6.4> "Nabitability Systems" ):

Exposure Fire - An exposure fire is a fire in a given area that involves either in situ or transient combustibles and 'is external to any structures, systems> or components located in or adjacent to that same area. The eEfects of such Eire (e.ge> smoke, heat, or ignition) can adversely efEect those structures, systems, or components important to safety. Thus, a fire involving one train of safe shutdo>fn equipment may constitute an exposure fire for the redundant train located in the same area, and a fire involving combustibles other than either redundant train may constitute an exposure fire to both redundant trains located .in the same a'rea Fire Area - that portion of a building or plant that is separated from other areas by boundary fire barriers.

Fire Barrier " those components of construction (>>alla> floors, and their supports), including beams, joists, columns, penetration seals or closures, fire doors, and fire dampers that sro rated by approving laboratories in hours of resistance to fire snd are used to prevent the spread of fire.

~pf ~ >t - a feat re of eonstreetion that prese ts fire props>ation al n>

the length oE cables or prevents spreading of fire to nearby combustibles Mithin a given fire area or fire xone.

~Fi e nr de - the te of plant personnel s sltned to fir fidhtid> snd Mho are equipped for and trained in the fighting of fireso Fire Detectors - a device designed to automatically detect the presence of fire and initiate an alarm system and other appropriate action (see NFPA 72C> "Automatic Fire Detectors" ). Some typical fire detectors are classified as folio>>s:

Neat Detector - a device that detects a predetermined (Eixed) temperature or rate oE temperature rise.

Smoke Detector - a device that detects the visible or invisible products ol combustion.

Flame Detector - a device that detects the infrared, ultraviolet, or visible radiation prf>duccd by a fire.

NUREC&800 GUIDEI INES CONFORMANCE a path> e.ga > fixed-temperature, heatmensitivc cable and rate-of-rise pneumatic tubing detectors.

Fire Protection Pro ram - the integrated effort involving components, procedures, and personnel utillxed ln carrying out all actlvltles of fire protection. It includes system and facility design, fire prevention, fire detection, annunciation, confinement, suppression, administrative controls, fire brigade organlratlon, inspection and maintenance> training, quality assurance, and testing.

Fire Resistance Ratln - The time that materials ot'ssemblies have withstood a fire exposure as established in accordance with the test procedures of "Standard Methods of Fire Teats or Building Construction and Materials" (NFPA) 251).

Manual fire suppression is the use of hoses, portable extingulshers, or manually-actuated fixed systems by plant personnel. Automatic fire suppression is the use of automatically actuated fixed systems such as water> Melon, or carbon dioxide systems.

Fire Zones - the subdivision of fire areas in which the fire suppression systems are designed to combat particular types of fires.

Noncombustible Material

a. A material which in the form in which it is used and under the conditions anticipated, will not ignite, burn, support combustion, or release flammablc vapors when sub)ected to fire or heat.

b. Material having a structural base of noncombustible materiel, es defined in aa above, with a surfacing not over I/8-Inch thick that has e flame spread rating not higher than 50 when measured using ASIM E-84 Test "Surface Burning Characteristics of Building Materials."

~Race s r fer ta R Ralstarf Garde 1.7>.

Restricted Area any area to which access ls controlled by the licensee for purposes of protecting individuals from exposure to radiation and radiobctlvc materials.

Safet Related S stems and Com onents systems and components required to shutdown the reactor, mitigate the consequences of postulated accidents, or maintain the reactor in a safe shutdown condition.

Secondar Containment - a structure that completely encloses primary conte nmcnt> use or controlling contalnmcnt Ienkape.

~Srlnkl r R st

- eet ark ef plplar. c ect d t s ref leal ster supply tl>at will dlstrlt>f>te thc water throughout thc area protected and will dlschatgc the water through sprinklcrs ln sufflclf nt quantity either

HUREG&000 OUIDELIMFS COWFORIAHCE to extinguish the fire entirely or to prevent its spread. Thc system, usually activated by heat includes a controlling valve and a device for

~

actuating an alarm when the system ls in operations Thc following categories of sprinkler systems are defined in MFPA 13, "Standard for the Installation of Sprinkler Systems" ~

Met Pipe System Dr y- Pi pc System Preaction System Deluge System Combined Dry-Pipe and Preaction System Oi~if System Stand I e and llose S stems - a fixed piping system with hose outlets, hose, and nozzles connected to a reliable water supply to provide effective fire hose streams to specific areas inside the buildings Mater S a S stem - a network of piping similar to a sprinkler system except that it utilizes open-head spray nozzles. NFPA 15, "Mater Spray Fixed Systems," provides guidance on these systems Co POSITIOM li Fire Protection Pro ram Re uirements ai Fire Protection Pro ram A fire protection program should be established at each nuclear power planti 'Ihe program should establish the fire protection policy for the protection of structures, systems, and components Important to safety at each plaut and the procedures, equipment ~ and personnel required to implement the program at thc plant site~

(I) The fire protection program should be under the direction of an CP&L will comply individual who has been delegated authority commensurate with the responsibilities of the position and who has available staff personnel knowledgeable in.both fire protection and nuclear safety.

(2) The fire protection program should extend the concept of a. The fire protection program utilizes the concept of defcnsc-in-depth for defense-In&cpth to fire protection in fire areas Important to fire protection in fire areas Important to safety by safety, with the following ob]cctivcs:

to prevent fires from starting; Preventing fires from starting to detect rapidly, control, and extinguish promptly those fires Rapid detection, control and extinguishing of those fires that do occur~

that do occur; to provide protection for structures, systems, and components Providing protection for structures, systems and components important 'to Important to safety so that a fire that is nut promptly safety so that a fire which Is not promptly extinguished by fire extinguished by thc fire suppression activities will not prevent suppression actlvltlcs will not prevent safe shutdown of the plant~

thc sale shutdown of thc plant ~

UUREG&8VO UUIDELIIIRS CONFORMANCE (3) Responsibility Eor the overall fire protection program should be CP&L will comply assigned to a person who has management control over all organixations involved in fire protection activitiesi Formulation and assurance of program implementation may be delegated to a staff composed of personnel prepared by training and experience in Eire protection and personnel prepared by training and experience in nuclear plant 'safety to provide a balanced approach in directing the fire protection program for the nuclear power plant>

The staff should be responsible Eor:

a, Fire protection program requirements> including consideration of potential hazards associated with postulated fires, with knowledge of building layout and systems design+

b, Post-fire shutdown capabilityi c, Design, maintenance, surveillance, and quality assurance of all fire protection features (e.g., detection systems, suppression systems, barriers, dampers, doors, penetration seals, and fire brigade equipment) ~

d~ Fire prevention activities (administrative controls and training) ~

e> Fire brigade organization and training.

E, Pre fire planning>

(4) The organlxational responsibilities and lines of communication CP&L will comply pertaining to fire protection should be defined between the various positions through, the use of organixational charts and functional descriptions of each position's responsibilities> The following positions/organizations should be designated:

a> 'Ihe upper level offslte management position which has management responsibility for the f ormulation> implcmcntation> and assessment of the effectiveness of the nuclear plant fire protection program.

bi The offslte management position(s) directly responsible for formulating, implementing, and periodically assessing the effectiveness of thc fire protection program for the licensee's nuclear power plant including fire drills and training conducted by the fire brigade and plant personnel. The results of'hese assessments should be reported to the upper level management position responsible for f Ire protection w I.th recommendations for improvements or corrective actions ea deemed necessary ci The onslte management position responsible for the ove'rail administration of the plant operations and emergency plans which Include the f irc protection and prevention pr>>gram and which provide a single point of control and contact for all conLIngcncIcs ~

tNREG-0000 GUIDEEUlES C00 FOR N ECE di The onsite position(s) which:

Implements periodic inspections tot minimize the amount of combustibles in safety-related areas; determine the effectiveness of housekeeping practices; assure the availability and acceptable condition of all fire protection systems/equipment, emergency breathing apparatus, emergency lighting, communlcatlon equipment, fire stops, penetration seals, and fire retardant coatings; and assures the prompt and effective corrective actions are taken to correct conditions adverse to fire protection and prccludc their recurrence.

Is responsible for the fire fighting training for operating plant personnel and the plant's fire brigade; design and selection of equipment; periodic inspection and testing of fire protection systems and equipment in accordane with established procedures, and evaluate test results and determine the acceptability of the systems under test iii Assists in the critique of all f ire drills to determine how well thc training obJectives have been meti iv. Reviews and evaluates proposed work activities to identify potential transient fire loads'o Implements a program for indoctrination of all plant contractor personnel in appropriate administrative procedures which implement the fire protection program, and the emergency procedures relative to fire protection.

vi. Implements a program for instruction of personnel on the proper handling of accidental events such as leaks or spills of flammablc materials that are related to fire protection+

Co The onslte position responsible for fire protection quslity- Refer to FSAR Section 17.2 ~ 19 assurancei This position should be responsible for assuring the effective implementation of the fire protection program by planned inspections, scheduled audits, and verification that the results of these inspections of audits are promptly reported to cognizant management personncli The positions which are part of the plant fire brigadeo The plant fire brigade positions should be responsible for flghtlng f ires. Thc authority and responsibility of each f irc brigade position relative to fire protection should be clearly defined.

The responsibilities of each fire brigade position should correspond with thc actions required by thc fixe fighting procedurcsi

HUREC&000 CUIDELIHES III COHFORtlhHCE iii. The responsibilities of the fire brigade members under normal plant condl,tions should not conflict with their responsibilities during a fire emergency.

ivo The minimum number of trained fire brigade members available onsite for each operating shift should be consistent with the activities required to combat the most significant f irci The size of the f ire brigade should be based upon the functions required to fight fires with adequate allowance for inJurieso ve Thc recommendations for organization, training, and equipment of "Private Fire Drigades" as specified in HFPA Hoi 27-1975, including the applicable HFPA publications listed in the appendix to HFPA Ho.

27, are considered appropriate criteria for organizing, training and operating a plant fire brigadco (5) Personnel Qualifications ao The position responsible for formulation and implementation of the CP&L will comply fire protection program should have within his organization or as a consultant a f ire protection engineer who is a graduate of an engineering curriculum of accepted standing and shall have completed not less than 6 years of engineering attainment indicative of growth in engineering competency and achievement, 3 years of which shall have been in responsible charge of fire protection engineering works These requirements are the eligibility requirements as a Member in the Society of Fire Protection Englnecrs ~

bo The fire brigade members'ualifications should include satisfactory completion of a physical examination for performing strenuous activity, and of the fire brigade training described in Position C.3odo C~ Thc personnel responsible for the maintenance and testing of the fire protection systems should be qualified by training and experience for such works d, The personnel responsible for the training of the fire brigade should be qualified by training and experience for such work+

(6) The following HFPA publications should be used for guidance to CP&L will comply develop the fire protection program.

Ho 1201-1977 - "Organization for Fire Services" Ho. l2U2"1976 - "Organiration of a Fire Department Hoi 27 -1975 "Private Fire Drigadcs"

NURFG&800 GUII)BLINtB CONFORHANCE (7) On sites where there is an opctating reactor nnd construction ot CPSL will comply vhen it becomes applicable.

modfffcntfon of other units is underway, the superintendent of the operating plant should have. the lend responsibility for site fire protection.

b. Fire Hazards Anal sls

&e Eire hazards analysis should demonstrate that the plant vill maintain Design in compliance the ability to perform safe shutdovn functions nnd minimize radioactive releases to. the environment in the event of n fire.

1he fire hazards snalysfs should be performed by qualified fire protection Design fn compliance and reactor systems engineers to (1) consider potential in situ and transient fire hazards; (2) determine the consequences of fire in any location in the plant on the ability to safely shutdown the reactor or on.

the ability to minimize and control the release of radioactivity to the environment; nnd (3) specify measures for fire prevention, fire detection, ffre suppression, snd fire containment and alternative shutdown capability as requfred for each fire area containing structures> systems, and components important to safety that are fn conformance with NRC guidelines and regulations.

"Morat case" Eires need not be postulated to be simultaneous vith Design fn compliance nonfire"related failures fn safety systems, plant accidents, ot the most severe natural phenomena.

On multiple-reactor sites, unrelated fires in tvo ot more units need not Unrelated fires fn tvo units vere not postulated to occur simultaneously, be postulated to occur simultaneously. Fires involving facilities shared Fires involving shared facilities between units vould involve the Paste betvcen units and fires due to man~ada site-related events that have a processing Building, Fuel Dandling Building, Emergency Setvfce Mater reasonable probability of occurring and affecting more than one reactor Screening Structure snd Recactor Auxilfary Building, Common to Units 1 and unit (such as nn aircraft crash). should be considered. 2 vhich sre separated from each othct and othet plant bufldfngs by three hour fire rated barriers snd doors. For details of Unit 1 nnd 2 Safety Because ffre may affect safe shutdown systems and because the loss oE Related equfpment located vithin the same Eire area, refer to the Safe function of systems used to mitigate the consequences of design basis Shutdovn Analysis fn Case of Fire. See Section 2.2.3 of the FSAR.

accidents under postffre conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve nnd maintain safe shutdown conditions is greater than the need to limit fire damage to those systems required to mftfgnte the consequences of design basis nccfdents. '%ree levels of Eire damage limits nre estnblfnhed according to the safety function of the structure, system, or component.

Safet Function Fire Damn e Limits Not shutdovn One train of equipment necessary to achieve hot shutdovn from either the control toom or emergency control station(s) must be mafntnined free of Eire damage by n sfngle fire, including nn exposure fire.

Cold shutdown Both trains of equipment necessary to achieve cold shutdown mny bc damaged by n sfngle fire, fncludfnp an exposure fire, but damage must be limited so that nt least one trnin cnn be repaired or made operable vfthfn 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> usfng onsltc cnpnbllity.

NUREC-0800 CUIDFLINES CONFORttANCF.

Safety Function >~>re Dans e L>afts Design basis Both trains of equipment necessary for mftlgatfon accident of consequences following design basfs accidents may be damaged by s single exposure fire.

The most stringent fire damage limit should apply for those systems that fall into more than one category. Redundant systems used to mftfgate the consequences of other desfgn basis accidents but not necessary for safe shutdown may be lost to a single exposure ffre. lfowever, protection shall bc provided so that a fire wfthfn only one such system will not damage the redundant system.

'lbe ffre harards analysis should sepaiately identify hazards and provfdc Tbe Eire hazard analysis separately identifies hazards and appropriate approprfate protection fn locations where safety-related losses can occur protection fn locations where safety-related losses can result from the as a result of: eight situations described.

(I) Concentrations oE combustible contents, fncludlng transient fire load due to combustibles expected to be used fn normal operations such as refueling, maintenance, and modfflcatfons; (2) Contlnulty of combustfble contents, furnfshlngs, bufldfng materials, or combinations thereof fn configurations conducive to fire spread; (3) Exposure ffre, heat, smoke, or water exposure, including those that may necessitate evacuation from areas that are required to be attended for safe shutdown; (4) Fire fn control rooms or other locations having critical safety-related Eunctlons; (5) Lack of adequate access or smo'ke removal facilities that impede fire extinguishment fn safety-related areas; (6) Lack of explosion"prevention measures; Loss of electric po~er or control ',cfrcufts; I

(8) Inadvertent operation of ffre suppression systems.

The fire hazards analysis should verify that the NRC fire protection 'lbe ffre hazards analysis wss performed to meet tbe intent of Appendfx A to program guidelines have been met. The analysis should list applicable BTP-APCSB 9.5-1 and identifies the locatfon, type of system snd design elements of the program, with explanatory statements as needed to fdentfEy crfterfa. It describes the fEre hazards for each fire area and the location, type of system, and design criteria. Tbe analysis should )ustfffcatfons, for the protection provided. Alternate protection measures fdcntffy and justify any deviations from the repulatory guldelfn<s. are provided where plant: design and arrangement indicate the alternate Justfflcatfon for deviations from the regulatory guidelines should show offers, at a minfmum, equivalent protcctfon.

that an equivalent level of protectfon will be achieved. Deletion of a protective feature without compensating alternative protection measures As descrfbcd in FSAR Appendix 9.58, Safe Shutdown hnalysfs in Case of Ffre, will not be acceptable, unless ft ls clearly d< mnnstrated that the Section I, CI>ronnlogy of Fire Protection Submlttals thc Sbearon Barris fire protective m<asurc is not needed because ol'he dcslpn nnd arrangement of protection program was based on the guidelines of Appendix A to USNRC Branch the particular plant. To<1>nlcal Posltlon (BTP) APCSB 9.5-1 dated August 23, 1976 for plants dncLetcd prfnr to July I, 1976, and tl e PSAR Fire Naznrds Analysis (FNA) was performed fn acnrdancc with NRC Iett< r of September 30, 1976 on the l>asfs of the above listed criteria.

On Junc ZG, 1980 CML sub<>>ftted th< SIINPP FSAR to tbe NPC,'t which time the FNA was cxpan<lcd (of lowfnp RC 1.70 Revision 3 - November 1978, Subsection 9.5.1..'> Sa(<ty Fvalnntlnn (Fire Naxardn Analysis) how<vcr, tl>e dcsfpn crl-tcrla and <Icslpn l>asfs were the same ns established ln thc 1'SAR Amen>tncnt 58>.

NURFC-0800 CUIDEBINES 8 CONFORHAV('E

'Ihe guidelines of an NRC BTP may be followed, or acceptable alternative n'y be provided by a Utility. Appendix A to BTP-APCSB 9.5-1 was not complied with as written by NRC. Alternatives to it were included in the SUNPP design and accepted by the NRC at the time when the Plant Construction I'crmlt was issued.

The FNA presently filed with the NRC, FSAR Section 4.5-1 Appendix 9.-5A did not address Appendix R or NURFC-0800 Criteria, because Appendix R was issued on November 19, 1980, becoming effective on February 17D 198l an<<

NURFC-0800 was issued on July 1, 1981. Both these documents were issued after the Plant FSAR Docleting.

The SUNPP FNA presently filed with the NRC did not verify that their NRC guidelines have been met'. Further, we know that the intent was met'.

SNNPP Safe Shutdown Analysis in Case of Fire (Appendix R), FSAR Appendix 9.5B, wss performed in response to NRC Questions 280.13 and 280.14 and was Submitted tO the NRC. In this analysis either conformance to Appendix R

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was indicated or sn exemption request was made for which Justification is contained in the SSA.

NUREC&800 was addressed in response to NRC question 280.1 and this was Submitted tO the NRC either..

e. Fire Be ressien B seen Design Basks (1) Total reliance should not'e placed on a single fire suppression Provision of fire suppression for the various plant areas avoid total system. Appropriate backup fire suppression capability should be reliance on any single system, automatic or manual.

provided. extinguishers src installed A full complement of appropriate hand fire throughout the plant to provide either initial fire-fighting capability or backup to any automatic or manual suppression systems. As a backup to hand fire extinguishers and/or automatic suppression systepsD a system of 1"1/2 inch small hose connections are installed throughout so that all areas within each building will be reached with 100 feet of this hose, attached to a standpipe connection. As a final backup to all of the protection outlined above, outside hydrants snd hose houses are also provided, (2) h single active failure or a crack in s moderate-energy line (pipe) Redundancy of equipment and systems are used as required to avoid in the fire suppression system should not impair both thc primary and impairment of both primary and secondary .fire suppression capability.

backup fire suppression capability. For example, neither the failure of.a fire pump, its power supply or controls nor a crack in a Niere feasible, fire protection, detection and suppression system control modcratemnergy line in the fire suppression system, should result in circuitry are routed through areas not served by the systems and thus not loss of function of both sprinkler and hose standpipe systems in an exposed to failure by thc fire incident.

area protected by such primary backup systems.

es '3) hs a minimum, the delivering water to fire suppression system should be capable of manual hose stations located within hose rerich of The fire suppression system is capable of delivering water to manual hose stations located within reach of areas containing equip<<<ent required for sal'c plant shutdown following the safe shutdown earthquake.

areas <nntaining equip<Dent required for safe plant shutdown following

\hr safe shutdown earthquake (SSE). In areas of high seismic activity, the staff will consider on n c<<se-hy-case basis the need to design thc fire d< tcction nnd suppr< salon syst<ms to be functional fol low/up tt<c SSE.

'I I I I NUREC&800 ClllDELINES CONFORHANCE (4) The fire protection systems should retain thefr original design Piping and supply Eor those portions of thc fire protection system not capability Eor (a) natural phenomena of lees severity and gtcater desfgnated to be operable post-SSE are designed and installed;to vlthstand frequency then the most severe natural phenomena (approximately once natural phenomena of less severity and greater frequency than the most in 10 years) such as tornadoes, hurricanes, floods, fce storms, or severe natural phenomena. See Section 2.2.3 oE the FSAR for the small-intensity earthquakes that ate characteristic of the geographic discussion on manmade sfte-related events.

region, and (b) potential man~de site-related events such as oil barge collisions or aircraft crashes that have a reasonable probability of occurring at a specific plant site. 'Ihe effects of lightning strikes should be included in the overall plant fire protection.

(S) The consequences of inadvertent operation of or a crack in a moderate The consequences of inadvertent operation or a crack on a moderate energy energy line in the fire suppression system should meet the guidelines line in the fire suppression system has been evaluated and meets the specified for moderate-energy systems outside containment in SRP guidelines specified in SRP 3.6.1 as stated in FSAR Scctlon 3.6.1 snd Section 3.6.1i 3e6.2o

'd ~ Altetnative or Dedicated Shutdown Aitetnative or dedicated shutdown capability should be provided where the Alternative or dedicated shutdown ls not contemplated at this time.

protection of systems vhose functions are requfted fot safe shutdown is not provided by established fire suppression methods or by Position C.S.

e~ Im lementation of Fire Protection Pro rags

'lhe fire protection program (plans, personnel, and equipment) for CP&L vill comply buildings storing new reactor fuel and for adjacent fire atcas that could affect the fuel storage area should be fully operational before fuel is received at the site. Such adjacent areas include those vhose flames, hot gases, and fire-generated toxic and corrosfve products may jeopardize safety and surveillance of the stored Euel (2) The fire protection program for an entire reactor unft should be CP&L vill comply fully operational prior to fnitial fuel loading in that reactor unft.

(3) On reactor sites vhere there is an operating reactor and constructfon CP&L vfll comply when it becomes applicable or modification of other unfts is under way, the fire protection program should provide for contfnuing evaluatfon of ffre hazards.

Additional fire barriers, fire protection capability, and administrative controls should be provided as necessary to protect the operating unit from construction fire hazards.

2~ Adminfstrative Controls Administrative controls should be used to maintain the performance of the fire protection system and personnel. 'lhese controls should establish procedures to:

a. Prohibit bulk storage of combustible materials inside or adjacent to CP&L vill comply safety-related buildfngs or systems during operation or maintenance periods. Regulatory Guide 1.39 provides guidance on housekeeping, fncluding the disposal of combustible materials.

b. Covern the handling and limitation oE thc use of ordfnnty combustible CP&L vill comply materials, cnmbustfblc and flammable gases and liquids, high ef[iclcncy pnrtfculntc air and charcoal filters, dry fon cxchanpc rcalns, or other combustfblc supplies ln safety-related areas.

I HUREC&800 CUlDEL1HES COHFORMAHCH Govern the handling of and limit transient fire loads such as CP&L vill comply combustible and flammable liquids, wood and plastic products, or other combustible materials in buildings containing safety-related systems or equipment during all phase of operating, snd especially during maintenance, modification, or refueling operations.

Designate the onsite staff member responsible for the inplant fire CP&L vill comply protection review of proposed work activities to identify potential transient fire hasards and specify required additional fire protection vork activity procedure.

Govern the use of ignition sources by use of a Elame permit system CP&L vill comply control welding, Elsme cutting, bracing, or soldering operations. h separate permit should be issued Eor each area vhere vork is to be done. IE vork continues over more than one shift, the permit should be valid for not more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the plant is operating or for the duration of a particular )ob during plant shutdown.

Gontrol the removal from the area oE all waste, debris, scrap, oil CP&L vill comply spills'r other combustibles resulting from the vork activity immediately Eollowing completion of the activity, or at the end of each work shift, vhichever comes first.

Covern leak testing; similar procedur'es such as airflow determination CP&L vill comply should use one of the commercially available techniques'pen flames or combustion-generated smoke should not be permitted.

Maintain the periodic housekeeping inspections to ensure continued CP&L vill comply compliance with these administrative controls.

Gontrol the use oE specific combustibles in safety-related areas. CP&L vill comply All wood used in safety-related areas during maintenance; modification, or refueling operation (such ss lay&own blocks or scaffolding) should be treated with a Elsmc retardant. 'Equipment or supplies (such as nev fuel) shipped in untreated combustible packing containers may be unpacked in safety-related areas if requited for valid operating reasons. However, sll combustible materials should be removed from the area immediately folloving unpacking. Such transient combustible material, unless stored in approved containers, should not be left unattended during lunch breaks, shift changes, or other similar periods. Loose combustible packing material such as vood or paper excelsior, or polyethylene sheeting should be placed in metal containers vith tight-fitting self-closing metal covers.

Disarming fire detection or Eire suppression systems should be CP&L vill comply controlled by a permit system. Fire watches should be established in areas where systems are so disarmed.

Successful fire protection requires testing and maintenance of thc CP&L will comply fire protection equipment and the emergency lighting and communication. A test plan that lists thc individuals and their rcsponsibiliteis in connection with routine tests and inspections of the fire detection and protection systems should bc dcvclopcd. Ibc test plan sl<nuld contain the types, frequency, ond detailed procc<lures for testing. Proccdurcs should also contain, instructions on m<<lntainlng fire protection during those periods when thc fire

1 NUREC&800 CUIOEI IN FS ptotection system is impaired or during periods of plant maintenance, CONFORH4NCE c.g., fire Matches or temporary hose connections to uatet systems.

Control actions to be taken by an individual discoveting a fire, for CP6L vill comply example, notification of control toom, attempt to extinguish Eire, and actuation of local fire suppression systems.

mo Conttol actions to be taken by the control room operaton to determine CP6L vill comply the need for brigade assistance upon report oE a fire or receipt of alarm on control room annunciator panel, fot example, announcing location of fire over PA system, sounding fire alarms, and notifying the shift supervisor and the fire brigade leader oE the types'ize ~

and location of the fire.

i l n. .Control actions to be taken by the Eire brigade after notification by CP6L uiil comply the control room operator of a Eire, for example, assembling in a designated location, receiving directions frow the fire brigade leader, and discharging specific Eire fighting responsibilities, including selection and ttsnspottation of Eire Eighting equipmcnt to fire location, selection oE protective equipment, operating instructions fot use of fire suppression systems, and use of preplanned strategies fot fighting fires in specfic areas, oo Define thc strategies fot fighting fires in all safety-related areas CP&L vill comply and areas presenting a hazard to safety-related equipment 'Ihese strategies should designate.

(I) Fire hazards in each area covered by the specific prefire plans (2) Fire extinguishants best suited for controlling the fires associated with the fite hazards in that area and the nearest location of these extingui aha nt s.

Nost favorable direction frow which to attack a Eire in each area in viev of the ventilation direction, access hallvays, stairs, and doors that are most likely to be Eree of Eire, and the best station or elevation for fighting the Eire. hll access and egress routes that i,nvolve locked doors should be specifically identified in the procedure uith the appropriate precautions snd methods fot access specified.

(4) Plant systems that should bc managed to reduce the damage potential during a local fire and the location of local and remote conttols for such management (eg, any hydraulic or electrical systems in the zone covered by the specific fire Eighting procedure that could increase the hazards in the area because of ovetpressurization or electrical hazards).

Vital heat-sensitive system components that need to be kept cool uhilc Eight ing a local f ire. Particularly hazardous combustibles that need cooling should bc designated.

Organization of fire fiphting brigades and the aasignmcnt of special duties according to Job title so that all fire fighting functions nre covered hy any complete shift personnel complcmcnt. These Qutics include command control of the brigade, transporting fire supprcsslon nnd support cquipmcnt to the fire sccllcsi applying thc extinguisher';

NUREG&800 GUIDELINES o'ONPORHANCE r

'I to the fire, communication vith the control room, and coordination

't I

v it h outside firc department s.

B Potential radiological and toxic hazards in fire zones.'

(8) Ventilation system operation that ensures desired plant air distribution vhen the ventilation flow is modified for fire containment or smoke clearing operation.

".;-.. l Operations requiring control room and shift engineer coordination or a

authorization.

(10) Instructions for plant operators and general plant pc rsonncl during Eire.

3~ ~Pare Brt ade

t]

a~ The need for good organization, training, and equipping of fire CPSL vill comply brigades at nuclear pover plant sites requires that effective measures ba implemented to ensure proper discharge oE those functions. Tbe guidance in Regulatory Aide 1.101, Emergency zr,.'ly Planning for Nuclear Povcr Planta," should be Eolloved as applicable.

b. A site Eire brigade trained and equipped for fire fighting should be CPSL vill comply established to ensure adequate manual fire fighting capability for all areas of the plant containing structures, systems, or components important to safety. The fire brigade should bc at least five members on each shift. The brigade leader and at least tvo brigade members should have sufficient training in or knovlcdgi of plant safety-related systems to understand the effects of fire and fire suppressants on safe shutdovn capability. The qualification of fire brigade members should include an annual physical examination to determine their ability to perform strenuous fire fighting

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activities. The shift supervisor should not bc a member of the fire brigade The'brigade leader shall be competent to assess the potential safety consequences of a fire and advise control room personnel. Such competence by tha brigade leader msy be evidenced by .,B..

possession of an operator'a license or equivalent knovlcdge of plant safety-related systems.

C ~ The minimum equipment provided for the brigade should consist of CPSL vill comply (Refer to NLU Reviev 81-527) personal protective equipment such as turnout coats, boots, gloves, hard hats, emergency communications equipment, portable lights, portable ventilation equipment, and portable extinguishers.

Self-contained breathing apparatus using full-Eace positive-pressure masks approved by NIOSN (National Institute for Occupational Safety and llcaltb--approval formerly given by tbe U.S. Bureau'f Hines) should be provided for Eire brigade, damage control, snd control room personnel. At least 10 masks shall be available for fire brigade personnel. Control room personnel may be furnished breathing air by a manifold system piped from a storage reservoir if practical.

B 'B Service or rated operating life shall be a minimum of onc-ltalf hour for the self-contained'units.

't

'1a At least tuo extra air bottlcu sltoul.l lB located onsitc for each If one hour rated SCBAPs are used, only one spare bottle per unit vill be

P.'l self-contained breathing unit. In addition, an onsite 6-ltour supply supplied.

of reserve sir should be provided and arranged to permit quick snd rnprdpdtd t n rnntnpttnlpBdanl nF ndpturrdntnal ~rarerrtar nlr lv plna ~n tie er nrn B

NUREG&800 GUIDELINES CONFORMANCE returned. If compressors are used as a source of breathing air, only units approved for breathing air shall be used; compressors shall bc operable assuming a loss of ofEsite power. Special care must be taken to locate the compressor in areas free of dust and contaminants.

Thc fire brigade training program shall ensure that the capability to See response to questLon 630.8 as transmitted on July 20, 1983 fight potential fires is established and maLntsined. The piogram shall consist of an initial classroom instruction program followed by periodic classroom instruction, fire fighting practice, and fire drills; (1) The initial classroom instruction should include:

(s) Indoctrination of the plant fire Eighting plan with specific identification of each individual's responsibilities.

(b) Identification of the type and location of fire hazards and

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associated types of fires that could occur in the plant.

(c) The toxic and corrosive characteristics of expected products of combustion.

(d) Identification of the location of- Eire Eighting equipme'nt for each fire area and familiarization with the layout of the plant, including access and egress routes to each ares, (e) The proper use of available fire fighting equipment and the corrective method of fighting each type of fire. The types of fires covered should include fires in energized electrical equipment, fires in cables and cable trays, hydrogen fires, fires involving flammable and combustible liquids or hazardous process chemicals, fires resulting from construction or modification (welding), snd record file Eires.

(f) The proper use of communication, lighting, ventilation, snd emergency breathing equipment.

(g) The proper method for fighting fires inside buildings and confined spaces.

(h) The direction and coordination oE the fire fighting activities (fire brigade leaders only).

(i) Detailed review of fire fighting strategies and procedures.

(g) Review of the latest plant modifications and corresponding changes in Eire fighting plans.

(k) Training of the plant fire brigade should be coordinated with the local fire department s'o that responsibilities and duties are delineated in advance. This coordination should be part of thc training course and should be included in the training of thc local fire department staff.

NUREC&800 GUIDELINES CONFORMANCE Local fire departments should be provided training in operational See response to question 630.8 as transmitted. on July 20, 1983 precautions when fighting fires on nuclear power plant sites and should be made aware of the need f or radiological protection of personnel and the special haxards associated with a nuclear power plant site.

Note: Items (i) and (j) may be deleted Erom the training oE no more than two of the nonoperations personnel who msy be assigned to the fire brigade.

(2) The instruction should be provided by qualified individuals who are See response to question 630.8 as transmitted on July 20, 1983 knowledgeable experienced, and suitably trained in fighting the types of fires that could occur in the plant and in using the types of equipment available in the nuclear power plant.

(3) Instruction should be provided to all fire brigade members and fire See response to question 630.8 as transmitted on July 20, 1983 brigade leaders.

(4) Regular planned meetings should be held at least every 3 months for See response to question 630.8 as transmitted on July 20, 1983 all brigade members to review changes in the fire protection program and other sub)acts as necessary.

(5) Periodic refresher traLning sessions shall be held to repeat the See response to question 630.8 as transmitted on July 20, 1983 classroom instruction program for all brigade members over a 2~ear period. These sessions may be concurrent with the regular planned meetings.

(6) Practice Sce response to question 630.8 as transmitted on July 20, 1983 (a) Practice sessions should be held for each shift fire brigade on the proper method of fighting the various types of fires that could occur in a nuclear power plant. 'Ihese sessions shall provide brigade members with experience Ln actual fire extinguishment and the use of emergency breathing apparatus under strenuous conditions encountered in fire fighting.

~ (b) These practice sessions should be provided at least once per year for I each Eire brigade member.

i (>) Drills See response to question 630.8 as transmitted on July 20, 1983 (a) Fire brigade drills should be performed in the plant so that the fire See response to question 630.8 as transmitted on July 20, 1983 brigade can practice as a team.

(b) Drills should be performed at regular intervals not to exceed 3 See response to question 630.8 as transmitted on July 20, 1983 months for each shift fire brigade. Each fire brigade member should participate in each drill, but must participate in at least two drills per year.

21 CONPORMhNCE NUREG&800 GUIDELINES I h sufficient number of these drills, but not lese than one for each Sce response to question 630.8 as transmitted on July 20, 1983 shift fire brigade per year, should be unannounced to determine the fire fighting readiness of the plant fire brigade, brigade leader, l and fire protection systems and equipment. Persons planning and

~ $ .i authorizing an unannounced drill should ensure that the responding l shift fire brigade members are not aware that a drill Ls being planned until it is begun. Unannounced drills should not be scheduled closer than 4 weeks, ht least one drill per year should be performed on a "back shift" for each shift fire brigade.

(c) The drills should be preplanned to establish the training objectives See response to question 630.8 as transmitted on July 20, 1983 of the drill and should be critiqued to determine how well the training objectives have been met. Unannounced drills should be planned and critiqued by members of the management stafE responsible for plant safety and fire protection. Performance deficiencies of a fire brigade or of individual fire brigade members should be remedied by schedulLng additional trainLng for the brigade'r members.

Unsatisfactory drill performance should be folio~ed by a repeat drill within 30 days.

(d) These drLlls should .provLde for local Eire department participation See response to question 630.8 ss transmitted on July 20, 1983 periodically (at least annually).

(e) ht 3~ear intervals> a randomly selected unannounced drill should be See response to question 630.8 as transmitted on July 20, 1983 critiqued by qualified individuals Lndependent of the licensee's staff. A copy of the written report from such individuals should be available for NRC review.

(E) Drills should as a minimum include the following: See response to question 630.8 as transmitted on July 20, 1983

i. Assessment of fire alarm effectiveness, time required to notify and assembly fire brigade, and selectLon, placement, and use of equipment and fire fighting strategies ii. assessment of each brigade member's knowledge of,his or her role in the fire fighting strategy for the area assumed to contain the fire. hssessment of the brigade members'onformance with e

J J

established plant fire fightLng procedures and use of fire fighting equipment, including self-contained emergency breathing apparatus, communication equipment, and ventilation equipmcnt, to I thc extent practicable.

iii. The simulated use of fire fighting equipment required to cope

/ with the situation and type of Eire selected for the drill. The area and type of fire chosen for the drill should differ from those used in the previous drills so that brigade members are trained in Eightiog fires in various plant areas. The situation selected should simulate thc size and arrangcmcnt of a Eire that could reasonably occur Ln thc area selected, allowing for Eire development duc to thc time required to respond, to oltain

~,

equipment, and organize for thc Eire, assuming loss of automntic suppression capability.

HUREC&800 GUIDELINES II CONFORMANCE I

iv. Assessment of brigade leaders direction of the fire fighting efforting as to thoroughness, accuracy, and effectiveness.

(8) Records Individual records of training provLded to each fire brigade member, See response to question 630.8 as transmitted on July 20, 1983 including drill critiques, should be maintained for.at least 3 years to ensure that each member receives training in all parts of the training program. These records of training should be available for NRC review.

Retraining or broadened training for Eire fighting within buildings should be scheduled for all those brigade members whose pcrEormance records show deficiencies.

(9) Guidance Documents NFPA 27I "Private Pire Brigade," should be followed in organization, CP&L will comply training, and fire drills, This standard also is applicable for the inspection and maintenance of Eire fighting equLpment. Among the standards referenced in this document NPPA 197, "Training Standard on InitLal Fire Attacks," should be utilized as applicable. NPPA booklets and pamphlets listed in NPPA 27 may be used as applicable for training references. In addition, courses in fire prevention and fire suppression that are recognized or sponsored by the fire protection industry should be utilized.

I

4. alit Assurance Pro ran The quality assurance (QA) programs oE applicants and contractors should The Eire protection quality assurance program is described in PSAR ensure that the guidelines for design, procurement, installation, and Section 17.2.19. As Stated on page 9.5.1%8 of Subsection 9.5-1 of the testing and the administrative controls for the Eire protection systems FSAR a quality assurance program has been developed for fire protection.

for safetymelated areas are satisfied. The QA program should be under The Design Construction QA program is described in the PSAR and vas thc management control of the QA organization. This control consists of approved by the NRC. NoveverI for components of the fire protection (1) formulating a fire protection QA program that incorporates suitable system designed, 'specified, procured> manufactured, fabricated or requirements and is acceptable to the management responsible for fire installed prior to institutILOn OE the Pire Protection QA program (February protection or vcrifyLng that the program incorporates suitable 8I 1977) the program was folloved to the extent practicable. The requirements and is acceptable to the management responsible for fire Engineering and Constructon fire protection quality assurance program vas protection, and (2) verifyLng the effectiveness of the QA program Eor fire approved by the NRC during constructon permit review. The Operational protection through reviev, surveillinceI and audits. Performance of other Quality Assurance Frogrsm is described in Section 17.2 of the FSAR.

QA program functions for meeting the, fire protection program requirements may be performed by personnel outside of the QA organization. The QA program for Eire protection should be part oE the overall plant QA program. It should satisfy the specific criteria listed belov.

a. Desi n and Procurement Document Control Measures should be established to ensure that the guidelines of the Described in PSAR Section 17.2.19 regulatory position of this guide are include in design and procutemen't documents and that deviations therefrom are controlled.

b. Instructions Procedures and Drawings Inspecting, tests, administrative controls, fire drills and training that

~ Described in FSAR Section 17.2.19 govern the fire protection program should be prescribed by documented instructions, procedures, or drawings,and should be accomplished Ln accordance with these documents.

l23 NUREG&000 GUIDELINES COHPORHANCE

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C~ Control of Purchased Haterial ui ment and Services Heasures should be established to ensure that purchased material, Described ln FSAR Section 17.2.19 equipment, and services conform to the procurement documcntsn

d. ~Inn ntdnn A program for independent inspection of activities affecting. Eire Described in FSAR Section 17 2 19 protection should be established and executed by or for the organization performing the activity to verify conformance with documented installation drawings and test procedures for accomplishing the activitiesn dd

'V e Test and Test Control A test program should be established and implemented to ensure that Described in FSAR Section 17 ' '9 testing is performed and verified by inspection and audit to demonstrate conformance with design and system readiness requirementsn The tests should be performed in accordance with wri.tten test procedures; test results should be properly evaluated and acted onn fn Ins ection Test and 0 eratin Status Heasures should be established to provide for the identification of items Described in FSAR Section 17.2n19 that have satisfactorily passed required tests and inspectionsi gn Nonconformi Items Heasures should be established to control items that do not conform to Described in PSAR Section 17 2nl9 specified requirements to prevent inadvertent use or installationnI h, Corrective Action.

Heasures should be established to ensure that conditions adverse to Eire Described in FSAR Section 17n2nl9 protection, such as failuresd malfunctions, deficiencies, deviatlonsd defective components, uncontrolled combustible material and nonconfvrmances, are promptly identified, reported, and corrected.

i. Records Records should be prepared and maintained to furnish evidence that the -Described in FSAR Section 17n2nl9 criteria enumerated above are being met for activities affecting the fire protection programn Audits Audits should be conducted to verify compliance with the Eire protection Described in FSAR Section 17n2 ~ 19 program, including design and procuremcnt documents> instructionsd procedures and drawings, and inspection and test activitiesn

24 NUREG&800 GUIDELINES CONFORHANCE

5. General Plant Guidelines (1) Fire barriers with a minimum fire resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> should be provided to:

(a) Separate safety-related systems from any potential fires in a) Complied with. Safety related systems werc separated from non nonsafety-related areas that could affect their ability to safety areas such as the Turbine Building, Maste Processing perform their safety function; Building, Mater Treatment Building and Administration Building by thrcc hour rated fire barriers. As stated in the FSAR, pape 9.5,1-1, separate fire areas were established to separate redundant safety divisions and to isolate safety related systems from hazards in non safety related areas to the extent possible in the previously established plant design established prior to issuance of NUREG"0800.

( b) Separate redundant divisions or trains of safety-related systems b) "Complied with 'the intent for separation of redundant safety from each other so that both are'ot subject to damage from a related systems required for safe shutdown (refer to FSAR single fire; Appendix 9.5B.3 for conformance or exemption request.) For redundant safety related systems not required for Safe Shutdown, Regulatory Chide 1.75 separation criteria werc followed, as described in the PSAR Section 8.3.

(c) Separate individual units on a multiple-unit site unless the c) Complied with requirements of General Design Criterion 5 are met with respect to fires.

(2) Appropriate fire barriers should be provided within a single safety (2) Complied with

'ivision to separate components that present a fire hazard to other safety-related components or high concentrations of safety-related cables within that division.

(25 HUREG&000 GUIDELIHES COHFORMAHCE (3) Openings through fire barriers for pipe, conduit, and cable trays CphL will comply which separate fire areas should be sealed or closed to provide a fire resistance rating at least equal to that requlrcd of the barrier itself Openings inside conduit larger than 4 inches In diameter should be sealed at the fire barrier penctratlono Openings Inside conduit 4 inches or less in diameter should be sealed at the fire barrier unless the conduit extends at least 5 feet on each side of the fire barrier and is sealed either at both ends or at the fire barrier with noncombustible material to prevent the passage of smoke and hot gascsi Fire barrier penetrations that must maintain environmental isolation or'ressure differentials should bc qualified by test to maintain the barrier integrity under such conditions Penetration designs should utilize only noncombustible materials and Complied with should be qualified by tents The penctrations qualification tests should use the time-temperature exposure curve specified by ASTM E-119, "Fire Test of Building Construction and Materials." The acceptance criteria for the test should require thatt (a) The fire barrier penetration has withstood thc fire endurance (a) Complied with test without passage of flame or Ignition of cables on the unexposed side for a period of time equivalent to the fire resistance rati,ng required of the barrier.

(b) The temperature levels recorded for the unexposed side are (b) Complied with analyzed and demonstrate that the maximum temperature does not exceed 325'F ~

(c) The fire barrier penetration remains Intact and does not allow (c) The hose stream test is in accordance with ASTM E 119 and Huclear pro)ection of water beyond the unexposed surface during thc hose Mutual Limited A-14 and meets the intent of this position.

stream test. The stream shall be delivered through a 1"1/2 inch nozzle set at a discharge angle of 301 with a nozzle pressure of 75 psi and a minimum discharge of 75 gpm with the tip of thc nozzle a maximum of 6 ft from the exposed face; or the stream shall be delivered through a l-l/2 inch nozzle set at a discharge angle of 152 with a nozzle pressure of 75 psi and a minimum discharge of 75 gpm with the tip of 'the nozzle a maximum of 10 f t from the exposed face; or the stream shall be delivered through a 2-1/2 inch national standard playpipc equipped <<Ith 1-1/2 Inch tip, nozzle pressure of 30 psi, located 20 ft from the exposed face.

~ ~

NUREGWUUU UUIDFI.INES 0)lIFOR I11'NCV I><><>tw<>rk whl< h p<'nctrat<< fire I arriere <<II I I e pr<>vl led with f Irc dampern I<avlng a fire r<<ululant rating nt I<<aut <<

al to that nf the I'Ire harrl<<r with the following cxrcpt iona: (4) Penetration openings for ventilation systems should be protected by I) Exhaust and Intakes nt <'xterl<>r walls, stacks and roofs. fire dampere l>avlng a rating equivalent to that requited of the Because these walls are not contiguous with fire oreas barrier (eec NFPA-90A, "Air Conditioning and Ventilating Systems" ) ~ It was not necessary to provide fire da><>pere. Flexl.ble ait duct coupling in ventilation and filter systems should I I) Transfer air fro<<> RAB, NVAC cqulpacnt roo>> to thc tank be noncombustiblei area Elevation 286 hecrn>sc the tank arcs has ncpllglhlc combustibles. III) IA>cal cooler ductwork pcn<<tratlng floors ln RAB which are designated fire xone boundaries wlthln I'lre areas I-A-BAI. nnd l2-A-BAI. and w<re upgraded to the equivalent of mini<<>u<>> 3-hour fire rcslstanr<<rating. The>><a)or part of flexible air duct coupling In vcntllntl<>n and filter syste>>>s are metallic flexible connectors, which are nonconlx>stlhlc. The remainder are one of thc following con<<<crcinlly available types, ns <<>anufactured by DuPont "Fairprene NN-0003 or DX-0002", or equal. (5) Door openings in fire bartiers should be protected with equivalently Door openings in fire barriers have fire resistant ratings equivalent to rated doors, ftamcs, and hardware that have been tested and approved that of the fire barrier and ate certified and Fuatanteed by the by a nationally recognized laboratoryi Such doors should be manufacturer to have fire resistant construction. The doors are either self"closing or provided with closing mechanisms and should be self closing or automatic closing or normally secured closed inspected semiannually to verify that automatic hold-open> release,, and closing mechanisms and latches are operable~ (See NFPA 80> "Fire Doors and Hindows.") Onc of the folio~leg measures should be provided to ensure they will CP&L will comply protect the openinF as required in case of firer (a) Fire doors should be kept closed and electronically supervised at a continuously manned location; (b) Fire doots should be locked closed and inspe'cted weekly to verify that the doors are in the closed position; (c) Fire doors should be provided with automatic hold-open and release mechanisms and inspected daily to verify that doorways are free of obstructions; or (d) Fire doors should be kept closed and inspected daily to verify that they are in the closed position. 1lie fire brigade leader should have ready access to keys for any CP&L will comply locked fire dootso Areas protected by automatic total flooding gas suppression systems CP&L will comply should have electrically supervised selfmlosing fire doors or should satisfy option (a) above~ J NUREC"0800 CUII>FLINFS CONFORHANCE (6) Personnel access routes and escape routes should be provided fot each ln most cases, more than one means of access or egress are provfded, es fire aren. Stafrvelle outside primary contnfnment serving as escape detnfled in the fire hazards analysis, for each fire ares. Stnirwnys routes, access routes for ffreffghting, or access routes to areas outside the primary containment nre enclosed by fire barriers vfth a containing cqufpment necessaty for safe shutdown should be enclosed minimum fire resistive rating of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and provided with self-closing fn masonry or concrete towers with a minimum fire rating of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Class B type fire doors. and seffmlosfng Class B fire doors. (7) Fire exit routes should be clearly marked. Fire exit routes sre clearly marked. (8) Each cable spreading room should contain only one redundant safety Complied vfth except for Cable Spreading Room A, where ae detefled in FSAR division. Cable spreading rooms should not be shared betveen Appendix 9.5B-3 Sketch CAR-SN-SK-668SI8> redundant B cables vhfch tun fn reactors. Cable spreading rooms should be separated from each other the cable tray CC0078-SB and Condufts 16020C-SR2-2, 160207-SR4-2, I1nd from other areas of the plant by'arriers having a minfmum ffre 109SSB SR4 2> 16106K SR4 I> 1063211 SR4 I ~ 16020R-SR2-1, They will be tesfstnnce of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. enclosed in one hour fire resistnnce rated enclosure due to sprinkler system already existing in this ares. (9) Intetfor.vali and sttuctural components> thermal insulation Complied with. Interior ffnfshes including thetmal insulation radfston materials, radiation shielding materfals, and sou'ndprooffng should be shielding and sound proofing have a flnme spread, smoke and fuel noncombustible. Interior finishes should be non-combustible. contribution of 50 or less ss defined in AS'-84, "Surface Burning Characteristics of Building Materfals". Hatetfals that are acceptable for use ss interior ffnfsh vfthout evfdence of test and listing by a nationally recognised laboratory are the folloving: Plaster, acoustic plaster, gypsum plasterboard (gypsum wallboard), either plain, vallpapered, or painted vlth oil- or vater-base pafnt; Ceramfc tile> ceramic panels; Class, glass blocks; Brick, stone, concrete blocks, plain or painted; Steel and aluminum panels, plain, painted, or enameled; Vinyl tile, vinyl-asbestos tile, linoleum, or asphalt tile on concrete floors ~ HUREG&800 CUIDELIHFS COHFORHAHCE (10) Hetal deck roof construction should be noncombustible snd listed as Hetal deck roof construction is not used on saEety-related buildings. "acceptable for fire" tn the UL Building Hatetials Directory; or listed as Class I in tbe Factor Mutuel System Approval Guldens (11) Suspended ceiling and their supports should be of noncombustible Suspended ceilings and their supports are of non-combustible construction. Concealed spaces should be devoid of combustibles 'construction. Electrical viring to lighting fixtures and HVAC systems in except as noted in Position C.l.b. these spaces is in condutts to reduce the combustible loading. (12) Transformers installed inside fire areas containing safety-related Transformers installed inside phe buildings are dry type systems should be of the dry type ot insulated and cooled vith noncombustible liquid. Transfotmers filled with combustible fluid only. that are located tndoore should be enclosed tn a transfotmet vault (see Section 450(c) of NFPA 70, "National Electrical Code" ). (13) Outdoor oil-Eilled transformers should bart oil spill confinement Outdoor oil Eilled transformers are mote than 50 Eeet from safety-related features or drainage avay from tbe buildings. Such ttansfotmers buildings and separated from the Turbine Building by tvo hour fire rated should be located at least 50 feet distant from the building, or by valls. Each transformer is installed over a gravel filled pit for ensuring that such building veils vitbtn 50 feet of otl-filled drainage. transformers are vtthout openings and have a ftte resistance rating of at least 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. I (14) Floot drains sized to remove expected ftrefighting waterflov vitbout Floor drain s sre provided in areas in which fixed water suppression flooding safety-related equipment should be provided in those areas systems amf hoses are installed. Drainage requirements and impact vhere fixed water Etre suppression systems are tnstalledo Floor on safety-related equipment will be considered before extending or drains should also be provided in other eteas vhere hend hose lines adding water suppression systems. See Section 9.3.3 of the FSAR for may be used if such fiteftgbting vatet could cause unacceptable damage to ssfetywelated equipment t.n tbe area (see NFPA<2, a description of the drainage systems. Gas suppression systems are not used in safety-related areas. Equipment containing quantities of "Waterproofing and Draining of Floors" ). Where gas suppression oil which may be of concern regarding backElow, such as the reactor systems sre installed, the drains should be provided vith adequate coolant pumps, RHR pumps, chillers, the diesel generator, DG day tank seals ot the gas suppression system should be sized to compensate for the lose of the suppression agent through the drains. Drains in and fuel oil stotage, storage tank pumps, are remote from each other areas containing combustible liquids should have provisions fot and are serviced by different drainage systems. In the containment preventing the backflov of combustible liquids to safety-related any leakage would flow by gravity into the Containment Building sump areas through the interconnected drain systems. Mater drainage Erom where it would be pumped to the waste holdup tanks in the Waste areas that may contain radioactivity should be collected'ampled, Processing Building. In the RAB leakage would drain to the floor and analyzed before discharge to the environment. drain tanks, except at elevation 190 ft where it goes to the sump. From these points it would be pumped to the waste holdup tanks in the Waste Processing Building. In the Diesel Generator Building any leakage from the diesel generator would be retained by curbs and flow into the Diesel Generator sump for transEer to the oil separator. In the DG Day Tank Room the spill would be retained by a 3 foot high dike before manually opening the drain valve to the Diesel Generator Building sump for transfer to the oil separator. The Diesel Generator F.O. transfer pumps have individual sumps ftom which oil is transferred to the oil separator. Water drained from areas having a potential for radioactive contamination is collected, sampled and analyzed before discharge to the environment. Areas with equipment coatnth(IIy 'significant amounts oE combustible liquids will have containment curbing to control inadvertant oil flows to surrounding arc-o cnd the drainage system. Where feasible, drains Eor these areas will be designed to minimize the possibility of combustible liquid fires spreading to other areas through the drains. 29 HUREC-0800 CUIDELINES CONFORMANCE

b. Safe Shutdown Ca abilit (l) Fire protection features should be provided for structures systems Complied with as detailed in the Safe Shutdown Analysis.

and components important to safe shutdowns These features should be capable of limiting fire damage so that: (a) One train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station(s) is free of fire damage; and (b) Systems necessary to achieve and maintain cold shutdown from either the control room or emergency control station(s) can be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. (2) To meet the guidelines of Position C5.b.l, one of the following means Generally complied with, as detailed in the safe shutdown of ensuring that one of the redundant trains is free of fire damage analysis, FSAR Appendix 9.58, with noted exceptions. should be provided; Additional evaluation and identification of exceptions (a) Separation of cables and equipment and associated circuits- of resulting from recent clarifications are ongoing. redundant trains by a fire barrier having a 3-hour rating. Structural steel forming a part of or supporting such fire barriers should be protected to provide fire resistance equivalent to that required of the barrier; (b) Separation of cables and equipment and associated circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazardsi In addition, fire detectors and an automatic fire suppression system should bc installed in the fire area; or (c) Enclosure of cable and equipment and associated circuits of one redundant train in a fire barrier having a I-hour rating~ In addition, fire detectors and an automatic fire suppression system should be installed in the fire area. (3) If the guidelines of Positions C5.baal and C5.b.2 cannot be met, then Alternate or dedicated shutdown is not contemplated at this time, FSAR alternative or dedicated shutdown capability and its associated Appendix 9,5Bi circuits, independent of cables, systems or components in the area/ room, or zone under consideration should be providedi .- 30 NUREG-0800 GUIDEI INCS CONFORHANCK

c. Alternative or Dedicated Shutdown Ca abilit Alternate or Dedicated Shutdown Capability is not contemplated at this time ~

(I) hlternative or dedicated shutdown capability provided for a speciEic f ire ares should be able to achieve and maintain subcritical reactivity conditions in the reactor, maintain reactor coolant inventory, achieve and maintain hot standby* conditions for a PNR (hot shutdown~ for a BMR) and achieve cold shutdown+ conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and maintain cold shutdown conditions thereafter. During the post fire shutdown, the reactor coolant system process variables shall be maintained within those predicted for a loss of normal ac power, and the fission product boundary integrity shall not be affected; i.e., there shell bc no fuel clad damage> rupture, or any primary coolant boundary, or rupture of the containmcnt boundary. (2) The performance goals for the shutdown functions should be: (a) The reactivity control function should be capable of achieving and maintaining cold shutdown reactivity conditions (b) The reactor coolant makeup function should be capable oE maintaining the reactor coolant level above the top of the core for BNRs and bc within the level indication in the prcssuriser for WRs. ~ (c) The reactor hest removal function should be capable of achieving and maintaining decay heat removal. (d) The process monitoring function should be capable of providing direct readings of the process variables necessary to perform snd control the above functions. (e) The supporting functions should be capable of providing the process cooling, lubrication, etc., necessary to permit the operation of the equipment used for safe shutdown Eunctions. (3) the shutdown capability for specific Eire areas may be unique for each such area, or it may-be one unique combination of systems for all such areas. In either case, the alternative shutdown capability shall be independent of the specif'ic fire area(s) and shall accommodate postfire conditions where of fsite power is available and where offsite power is not available for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Procedures shall be in effect to implement this capability ~ (4) If the capability to achieve and maintain cold shutdown will not be availablc because of fire damage> the equipment and systems comprising the means to achieve and maintain the hot standby or hot shutdown conditions shall be capable of maintaining such conditions until cold shutdown can be achieved. If such equipment and systems will not be capable of being powered by both onsitc and of fsitc electric power systems because of fire damage, an independent onsite power system shall be provided. The number of operating shift personnel, exclusive of fire brigade members, required to operate such equipmcnt and systems shall hc onsitc at all times.

• As Jclincd in the Standard Tcchnical Specifications.

4 NUREG 08UO GUIDELINES CONFORHhNCE (5) Equipment and systems comprising the means to achieve and maintain cold shutdown conditions should not be damaged by fire; or the fire damage to such equipment and systems should be limited so that the systems can be made operable and cold shutdown achieved within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Naterials for such repairs shall be readily available onsite and procedures shall be in effect to implement such repairs. If such equipment and systems used prior to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the fire will not be capable of being powered by both onsite and offsite electric power systems because of fire damage, an independent onsite power system should be provided. Equipment and systems used after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> may be powered by offsite po~er only. (6) Shutdown systems installed to ensure postfire shutdown capability need not be designed to meet seismic Category I criteria, single failure criteria, or other design basis accident criteria, except ~here required for other reasonsy e g y because of interface with or impact on existing safety systems, or because of adverse valve actions due to fire damage. (7) The safe shutdown equipment and systems for each fire ares should be known to be isolated from associated circuits in the fire area so that hot shorts, open circuits, or shorts to ground in the associated circuits will not prevent operation of the safe shutdown equipment. The separation and barriers between trays and conduits containing associated circuits of one safe shutdown division and trays and conduits containing associated circuits or safe shutdown cables from the redundant division, or the isolation of these associated circuits from the safe shutdown equipment, should be such that a postulated fire involving associated circuits will not prevent safe shutdown. NUREG-0800 GUIDELINES CONFORHANCE d Control of Combustibles Safety-relsted systems should be isolated or separated from SaEety-related equipment and systems are isolated- or protected against combustible materials. Mhen this is not possible because of the exposure from ignition sources or high combustible loading. This nature of the safety system or the combustible material, special separation and protection consists of spatial separation, physical fire protection should be provided to prevent a fire from defeating the rated barriers, fire suppression, fire control or damage limitation safety system function. Such protection may involve a combination of systems or any combination of these which provides the degree oE automatic fire suppression, snd construction capable of withstanding separation required by the fire harard analysisi and containing a Eire that consumes all combustibles present. Examples of such combustible materials that msy not be separ'able from Examples of isolation and protection of typical combustible materials the remainder of its system are: using this concept sre: (a) Emergency diesel generator fuel oil day tanks. The diesel fuel oil day tanks are located within concrete vaults, which sre separated Erom other plant areas by three hour fire barriers. (b) Turbine-generator oil and hydraulic control fluid systems. Automatic, multi-cycle sprinkler systems are provided in these areas, with bose station and yard hoseline equipment and portable extinguishers as (c) Reactor coolant pump lube oil system. backup. The turbine-generator lubricating oil system is located within the turbine building, away Erom all safety-related equipmenti This area is provided with sn automatic multi-cycle sprinkler system for equipment and property protection. A Eire in this area will not pose any haxard to safety-related equipment. The reactor coolant pump lube oil system is located within containment ~ nest the reactor coolant pumps'nd will be equipped with in oil collection system. The oil collection system will be designed, engineered and installed so that failure will not lead to fire during design basis accident conditions and that there will be reasonable assurance that the system will withstand the Safe -Shutdown Earthquake. HUREC-0800 CUIUELIHES COHFORHAHCE

2) Sulk gas storage (either compressed or cryopenfc), should not bc Bulk storape of compressed or cryogenic gases is not permitted within permitted inside structuree housing safety-related equipment. structures housfng safety-related equipment. Flammable goses are stored Storage of flammable gas such as hydrogen should be located outdoors outdoors ond will not expose safety-related equipment, systems or or in separate detached buildings so chat o fire or explosion will structures. Systems are designed to applicablc code.

not adversely affect any safety-related systems or equipment. (Refer to NFPA 50A, " Coscous Hydrogen Systems. ) Care should be token to locate hfgh pressure goe storage contofners CPAL will comply with the long axis parallel to building walls. 'Ibis will mfnimfzc the possibility of wall penctratfon in the event of a contafner faflure. Use of compressed gases (especially flammable and fuel "gases) fnsfde buildings should be controlled. (Refer to NFPA 6, Industrial Fire Loss Prevention.")

3) The use of plastic materials should be minimized. In pnrtfcular, Plastics are used only where requfred se essential equipment and to the hologenoted plastics such as polyvinyl chloride (PVC) and neoprene minimum extent possible, ss detailed in the fire hazards analysis. A should be used only when substftute noncombustible materials are not small quantity of vinyl is used for trimming of non-seismic available. All plastic materials, including flame ond fire retardant instrumentation cable tray cover cutouts for cable exits from those trays materials, will burn with an intensfty and BTU production fn a range which are solid bottom wfth cover construction. Stondard Products similar to that of ordinary hydrocarbons. Mhen burnings they produce qufckedge Hfnitrfm Port Ho. 75000341, which fs a vinyl wos the only heavy smoke that obscures visibility and can plug air filters, material available to meet the fnstallotion requirements, and therefore especially charcoal and HEPA. The hslogenotcd plastics also release wos selected for this application. It's use is limited to a minimum 6 fn.

free chlorine and hydrogen chloride when burning which ore toxfc to radius to a maximum of 12 fn. by 13 fn. rectangular cutout. The humans and corrosfve to equfpment. "Qslckedge" vinyl fe self-extfnpuishfng passing Federal Specfficetfon FSS"302. NUREG&800 GUIDELINE CONFORMANCE ) Storage of flammable liqufds should, ss a minimum, comply vfth the Storage and use of flammable and combustible liquids follovs tbe intent requirements of NFPA 30, "Flammable and Combustfble Liqufds Code." and basic criteria of NFPA 30, Flammable and Combustible Liquid Code" except that the requirements of NFPA-37 "Installntfon and Use of Stationary Combustion Engines and Css Turbines" apply to the installation of the Diesel Generator. Day Tank. Specific standard requirements are met vhere compatible vith other design requfrements. ) Nydrogen lines fn safety-related areas should be either designed to Nydrogen lines do not pass through areas bousfng safety-related equipment. seismic Class I requfrements, or sleeve such that tbe vater pipe is directly vented to tbe outside, or should be equipped vith excess flov valves so that fn case of a line break, the hydrogen concentration in the affected areas vill not exceed 2X. Electrical Cable Construction Cable Tra s and Cable Penetrstions ) Only metal should be used for cable trays. Only metallic.tubfng Cable trays and other racevays are constructed of nonmombustfble should be used for conduit. Thin~all metallic tubing should not be material. Metallic tubing is used for conduit and thin vali tubing is not used. Flexible metallic tubing should only be used fn short lengths used. Short lengths of flexible metallic tubing are used to connect to connect components to equipment. Other racevays should be made of components to equipment. h small quantfty of vinyl is used for cable noncombustible materfal. trays as described.fn Positfon C.5.d(3). PVC is used only for imbedded fn concrete or underground applfcations. HUREG&800 GUIDEI.INES COHFORMA HCE Redundant safety-related cable systems outside the cable spreading room are separated from each other and from potential fire exposure hazards (2) Redundant safety-related cable systems outside the cable spreading ln non-safety related areas by separation criteria given in Regulatory room should be separated from each other and from potential Eire Guide 1.75 plus automatic fire detection and/or preaction or multicyclc exposure hazards in nonsafety-related areas by fire barriers with a sprinkler systems. Redundant cable systems required for safe shutdown minimum Eire rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Tlmse cable trays should be provided in case of fire were separated in accordance with Section III.G.2 of with continuous line-type heat detectors and should be accessible for Appendix R to 10CFR50, as detailed in FSAR Appendix 9.58. Generally manual Eirefighting. Cables should be designed to allow wetting down complied with, as detailed in the Safe Shutdovn Analysis, FSAR Appendix 9.58, with fire suppression vater vithout electrical faulting. Manual bose with noted exceptions. Additional evaluation and identification of exceptions stations and portable lmnd extinguishers should be provided . resulting from recent clarifications are ongoing. Spot type ionization smoke detectors and thermal detectors located above the cable trays are provided instead of line type thermal detection. These detectors are sensitive to products of combustion and provide early varning in the first stapes of a fire. Safety-related cable trays of a single division tlat are separated Safety-rclbted cable trays separated by fire barriers with a mininum from redundant divisions by a fire barrier with a minimum rating oE 3 rating of three hours and accessible for manual fire fighting are hours 'and are normally accessible for manual firefighting should be protected by automatic suppression systems from thc effects of an exposure protected from the effects of a potential exposure fire by providing Eire. Hhere provided, automatic area protection considers cable tray automatic vater suppression in the area whcrc such a fire could arrangements and transient combustibles to assure adequate protection occur. Automatic area protection, where provided, should consider against exposure fires. Manual hose stations are not relied upon for cable tray arrangements and possible transient combustibles to ensure primary Eire suppression for redundant safety cables in lieu of automatic adequate water coverage for areas that could present an exposure water suppression. hazard to the cable system. Manual hose standpipe systems may be relied upon to provide the primary fire suppression (in lieu of automatic water suppression systems) for safety-related cable trays of a single division that are separated from redundant safety division by a fire barrier with a minimum rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and are normally accessible for manual fircfighting conditions are met: if all of the following (a) Thc number of equivalents standard 24-inc~ide cable trays (both safety-related and nonsafcty-related) in a given fire area is six or less; (b) The cabling does not provide instrumentation, control or power to systems required to achieve and maintain hot shutdown; and (c) Smoke detectors are provided in the area of these cable routings, and continuous line-type heat detectors are provided in the cable trays. Safety-related cable trays that are not accessible for manual fire Hot applicable, except for Cable Spreading brea where SB cable trays arc

• fighting should be protected by a zoned automatic water system with enclosed in three hour fire resistive barriers along the outside wall of open"head deluge or open directional spray nozzles arranged so that the area north from column line 418 to 438 and west from 438 to 43E.

adequate water coverage is provided for each cable tray. Such cable trays should also be protected from the effects of a potential exposure Eire by providing automatic water suppression in the area where such a fire could occur.

• Trays exceeding 24 inches should be counted as tvo trays; trays exceeding 48 inches should be counted as three trays, regardless of tray fill.

r Cl NUREC&800 GUIDFI.ItlES COttFOIttt&tlCE In the other areas where it mey not be possible because of other In these areas cf ther pre-action'r multfcycle sprinkler systems equipped with closed sprinkler tu.ads nre provided. Since the tvo step operation of overriding design features necessary for reasons of nuclear enfcty to separate redundant safety-related cable systems by 3-hour-rated fire closed sprinkler'ead systems requires activatfon of thc sprinkler flow barriers, cable trays should bc protected by an automatic water control valve by automatic detectors or manual fire alarm etntfons and system vith open-head deluge or open directional spray nozzlee fusfng of the spr'inkier bends tjy beat from tbe fire before water is arranged so that adequate water coverage is provfded for each cable dischprged, unnecessary water damage resulting from premature dfsctmrge or-tray. Such cable trays should also be protected from tbe effects of inadvertent operation is avoided. Mater is discharged only from sprinkler a potentfal exposure fire by providing automatic water suppression in heads Iin the immediate area of tbe fire. The capability to achieve and tbe arcs where such a fire could occur. The capability to acbfeve mnfntafn safe shutdown fn ense of fire is evaluated in the Safe Shutdovn and maintain safe shutdown considering the effects of a fire Analysis. involving fixed and potential transient combustibles should bc For thc Safe Shutdown Analysis ln Case oE Fire, vhcrevcr 3-hour rated Eire cva'lusted with and without actuation of the automatic suppression barriers could not be provided onc oi the separation criteria oE Section system and steuld be Justified on a suitably defined basis. III.C oE Appendix R to 10CFR50 were provided or sn exception request and thc technical basis for it identiiled, as detailed in the FSAR Appendix 9.5B.3. Cencrally cowplied vith, ss detallcd in thc Sais Shutdown Analysis, FSAR Appendix 3.5B, with noted exceptions. Additional evaluation and identiflcbrion oE exceptions resultine frow recent clarlEl-cations are oneoine. The rest ls cowplied vfth. Exposure fires vere considered only for Safety-Related Systews dcsfenated as required Eor SaEc Shutdown Analysis in Case oE rlrc. (3) Electric cable construction should, as a minimum, pass the flame test Thc electric cable construction conforms to IEEE 383 except communication in the current IEEE Std 383. (This does not fmply that cables cable vhich runs in conduit or underground. passing this test vill not require<fire protection.) (4) Cable raceways should be used only for cables. Cable raceways are only used for cables (5) ttfecellaneoue storage and piping for flammable or combustible liquids CP&L vill comply. Piping is in compliance. or gases should not create e potential exposure hazard to safety-related systems. I NDREG&800 GUIDEDINFS CONFORMANCE Ventilation l) The products of combustion and the means by which they will be Methods of removing the products of combustion from each fire area is removed from each fire area should bc established during the initial specifically delineated in FSAR Appendix 9.5A. Snoke from the stages of plant design. Consideration should be given to the Containmcnt, Reactor Auxiliary, Fuel Handling, Haste Processing and installation of automatic suppression systems as a means oE limiting turbine Buildings will be discharged through main plant stacks whose smoke and end heat generation. Smoke snd corrosive gases should discharge points are well removed from safety-related plant areas. generally be discharged directly outside to an area that will not The Diesel Generator Suilding, Diesel Fuel Oil Tank Building and affect safety-related plant areas. The normal plant ventilation Baergcncy Service Mater Inta'ke Structure are remote Erom the main system msy be used for this purpose if capable and available. To plant area and smoke discharged from these structures will not effect facilitate manual fircfighting, separate smoke and heat vents should other safety-related plant areas. In all cases the outside air be provided in specific areas such ss cable spreading rooms, diesel intakes were located in consideration of the pdssibility of short fuel oil storage areas, switchgear rooms, and other areas where the circuiting of exhaust air and in turn any smoke discharge from potential exists for heavy smoke conditions (see NFPA 204 for reentering the building. Sce Section 9.4.0 of PSAR for discussion of additional guidance on smoke control). plant effluent release points. As indicated in Appendix 9.5A the normal plant ventilation systems vill be used for smoke venting. The Control Room and Electric Equipment Protection Rooms sre provided with specially designed smoke purge systems, Por the Cable Spreading Rooms snd Switchgear Rooms specially designed smoke purge systems will be available for smoke venting in lieu of separate smoke and heat vents; architectural limitations precluded the utilization of smoke and heat vents. qr

2) Release oE smoke and gases containing radioactive materials to the (2) 'Ihc release of smoke that csn"potentially carry radioactive material environment should be monitored in accordance with emergency plans as is monitored at the stack discharge points. In addition, area described in the guidelines oE Regulatory Oside 1.101, Baergency monitors are provided in areas containing radioactive material. See Planning for Nuclear Power Plants." Any ventilation system designed Section 9.5.1.2.2 of FSAR for Eurther discussion of this subject.

to exhaust potentially radioactive smoke or gases should be evaluated Normal ventilation systems sre used to exhaust smoke and products of to ensure that inadvertent operation or single failures will not combustion Eor most of the plant areas. section 9.5.1.2.2 describes violate the radiologically controlled areas of the plant design. these systems operations. 'Ibis requirement includes containmcnt functions for protecting the public an maintaining habitability Eor operations personnel.

3) Special protection for ventilation power and control cables msy be (3) Special protection was provided for cables and ventilation systems rcquiredi The power supply and controls Eor mechanical ventilation which were designated as being required for the safe shutdown of the systems should be run outside the fire area scrvcd by the system plant in ease of a Eire, Eor the balance of thc plant, complied with where practical. to thc extent practical.

4) Engineered safety feature filters should be protected in accordance (4) Complied with, for details see Section 9.5A; with the guidelines of Regulatory Oside 1.52. Any filter that includes combustible materials snd is a potential exposure fire haxard that msy affect safety-related components should be protected as determined by the Eire haxards analysis.

5) 'Ihe fresh air supply intskes to areas containing safety-related (5) See response to item (1) for discussion on this subject.

equipment or systems should be located remote from the exhaust air outlets and smoke vents of other fire areas to minimixe the ir possibility of contaminating the intake a with the products of combustion.

6) Stairwells should be designed to minimixe smoke infiltration during a (6) Complied with.

fire. NURFC&800 GUIDELINES OINFORHANCE ) Where total flooding gas extinguishing systems are used, area intake (7) COL vill comply and exhaust ventilation dampers should be controlled in accordance vfth NFPA 12, "Carbon Dioxide Systems'nd NFPA 12A, "Nalon 1301 Systems," to maintain the necessary gas concentration. Lighting and Communfcation ghting and tv<nosy voice communication ere vital to safe ehutdovn and. ergency response fn the event of fire. Suitable Etxed and portable ergcncy lighting and communication devices should be provided as follows: ) Fixed selfwontafned lfghtfng consfstfng of fluorescent or Except for the control, auxiliary control and computer rooms, fixed sealed-beam unfts with individual 8&our minimum battery pover selfmonteincd seal beam units vith individual 8&our mfntmum battery supplies should be provided in areas that must be manned for safe pouer supplies will be provided fn areas that must be manned for safe shutdown and Eor access and egress routes to and from nll fire ebutdovn snd Eor access and egress routes to and from sll fire areas. 'Ihe areas. Safe shutdovn areas include those required to be manned tf DC Fmergency Lighting System usfng tbe plant 125V battery provides thc control room must be evacuated. emergency lighting for the control room, auxiliary control room, and computer room. See FSAR Section 9.5.3.2.

) Suitable sealed+earn battery-povered portable band lights should be CPSL util comply provided for emergency use by the Eire brigade and other operations personnel required to achieve safe plant shutdoun.

) Ffxed emergency communications independent of tbe normal plant CP6L vill comply communication system should be installed at prcselected stations. ) A portable radio communications system should be provfded Eor use by CPAL vill comply tbe-fire brigade and other operations personnel requfred to achieve safe plant shutdovn. 'Ibis system should not interfere uith the . communications capabilfties of the plant security force. Fixed repeaters installed to permit use oE portable radio communfcstfon units should be protected from exposure fire damage. Preoperational and periodic testing should demonstrate that the frequencies used for portable radio communication will not affect the actuatton of protective relays. NUREC&800 GUIDELINES CONFORHANCE Fire Detection and Su ression Fire Detection Detection systems should be provided for all areas that contain or Detection systems will be provided for all areas that contain or present a present a fire exposure to safety-related equipment. .fire exposure to safety-'related equipment. I Fire detection systems should comply with the requirements of Class A As stated In Section 9.5.1.2.3 of the FSAR the fire detection systems will systems as defined in NFPA 72D, "Standard for the Installation, bc Class A systems in accordance with NFPA 72D and have Class 1 circuits Haintenance, and Use of Proprietary Protective Signaling Systems'" as defined In NFPA 70, "National Electric Code." and Class 1 circuits as defined In NFPA 70, "National Electrical Code." ) Fire detectors should be selected and Installed In accordance with Fire detectors wilL be selected and installed in accordance with NFPA 72E, NFPA 72E, "Automatic Fire Detectors." Preopcrational and periodic "Automatic Fire Detectors." ionization detectors Installed on an area testing of pulsed line-type heat detectors should demonstrate that basis for early warning and rate compensated thermal detectors are used to the frequencies used will not affect the actuation of protective provide supplementary detection capability in lieu of relying solely on relays in other plant systems. linc type detectors. ) Fire detection systems should give audible and visual alarm and Complied with FSAR Section 9.5.1.2.3. annunciation In thc control room. Where zoned dctcction systems are used in a given fire arcs, local means should be provided to identify which detector conc has actuated. Local audible alarms should sound in the fire area. I 40 NURCG-0800 GUIDFl lttES COHFORHAttCE Fire alarms should be distinctive and unique so they will not be Design in compliance confused with any other plant system alarms ~ Primary and secondary power supplies should be provided for the Eire Design in compliance detection system snd Eor electrically operated control valves for, automatic suppression systems Such primary and secondary power supplies should satisfy provisions of SEction 2220 of NFPA 720 can be accomplished by using normal offsite power as the primary 'his supply with a 4-hour battery supply as secondlary supply; and by providing capability Eor manual connection to the Class lg emergency power bus within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of loss of offsite po~er. Such connection should follow the applicable guidelines in Regulatory Guides 1.6, 1.32, and 1.75. Fire Protection Mater Su I S stems An underground yard fire main loop should be installed to furnish Design in compliance anticipated water requirements. HFPA 24, "Standard for Outside Protection,".gives necessary guidance Eor such installation. It references other design codes snd standards developed by such orgsnisations ss the American Hational Standards Institute (AHSI) and the American Mater Marks Association (AMMA)~ Type of pipe and water treatment should be design considerations with tuberculation as on of the parameters. Henna for inspecting and flushing the systems should be provided. Approved visually indicating sectional control valves such as Design in compliance post-indicator valves should be provided to isolated portions of .the main for maintenance or repair without shutting off the supply to primary snd backup fire suppression systems serving areas that contain or expose safety-related equipment. 4l NUREC-0800 CUIOELIHES CONFORHAHCE ) Valves should be installed to permit isolation of outside hydrants Complied with from the fire main for maintenance or repair without interrupting the water supply to automatic or manual fire suppression systems in any area containing or presenting a fire haxard to safety-related or safe shutdown equipment. ) The fire main system piping should be separate from service or Complied with sanitary water system piping, except as described in Position C.b.c.(4) ~ ) h common yard fire main loop may serve multiunit nuclear power plant Complied with sites if cross-connected between units. Sectional control valves should permit maintaining independence of the individual loop around each unit. For such installations, cocoon ~ster supplies may also be utilised. For multiple-reactor sites with widely separated plants (approaching l mile or more). Separate yard fire main loops should be used. 42 NUREG-OBOU GU IDEI INES COHFORNANCE ~ ) If pumps are required to meet system pressure or flow requirements. Two 100Z El.re pumps, one electric drivtn and one diesel driven, installed A suEficient number of pumps should be provided to t nsure that 100Z in accordance with NFPA 20, art providtd. The electric driven fire pump capacity will be available assuming failurt of the largest pump 'or is UL listtd and the diest1 driven fire pump is FH approved. Individual loss of offsite power (e.g., three 5UZ pumps or two lUUZ pumps) ~ fire pump connections to the yard fire main are separated by sectional This can be accomplished, for exemple, by providing t ither: valves betwetn connections. The pumps are installed at opposite ends of the emergency service water intake structure which provides spacial (a) Electric motor-driven fire pump(s) and diesel-driven fire separation in lieu of a fire wall. Tht. diesel fit'e pump fuel supply is pump(s); or located about 12 feet away from the emergency strvice water intake l structure within one foot high curbs which direct the oil to the sump (b) Two or more seismic Category I Glass IE electric motor-driven within the curbs. Alarms indicating pump running, driver availability and fire pumps connected to redundant Class IE emergency power buses failure to start are provided in the control room. A low fire-main (see Regulatory Guides l.b, 1.32, snd 1.75). pressure alarm is not providt d since a jockey pump maintains system pressure at about 100 psig. If system pressure is not maintained by the Individual fire pump connections to the yard Eire main loop should be jockey pump the electric driven pump will start when the system pre.ssure separated with sectionalixing valves between connections. Each pump drops to about 90 psig. The cause of the start"up of the electric driven and its driver and controls should be located in a room sepirated pump will be investigatt d by the operators from the remaining fire pumps by ~ Eire wall with s minimum rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The fuel for the diesel fire pump(s) should be separated so that it does not provide a fire source exposing saEety-related equipment. Alarms indicating pump running, drivtr availability, failure to start, and low fire-main pressure should be provided in the control room. The fire pump installation should conform to HFPA 20, ugtsndard Eor the Installation of Centrifugal Fire Pumps." Outside manual hose. installation should be sufficient to provide an Hydrants are installed t very 250 feet on the yard fire main loop. Each effective hose stream to any onsite location where fixed or transient hydrant is furnished with hose, combination noxxle and other auxiliary combustibles could jeopardixe safety-related equipment. Hydrants equipment as recommended by NFPA-24, "Outside Protection". should be installed approximately evt ry 250 ft on the yard main system. h hose house equipped with hose and combination noxxle'and other auxiliary equipment recommended in HFPA 24, "Outside Protection," should be provided as needed, but at lt ast every 1,000 ft. Alternatively, mobile mtans of providing hose and associated equipment, such as hose carts or trucks, may be used" when provided, such mobile equipment should be equivaltnt to supplied by three hose houses. the'quipment NUREG-0800 GUIDELINES CONFORNANCE J (8) Threads compatible with those used by local f ire departments should Threads for fire hose and equipment are, NSf in accordance with NFPA 1963. bc provided on all hydrants, hose couplings, nnd standpipe risers. Each hose house is equipped with two adapters..tagged "Raleigh Fire Department Adapter" and "Sanford Fire Department Adapter" which fit local fire d<<partmcnt threads. (9) Two separate, reliable fresh~ster supplies should be provided. The water supply is taken from the fresh water supply impounded by the Saltwater or brackish water should not be used unless all freshwater Auxiliary Reservoir. Tanks are not used to store fire protection water supplies have been exhausted. If tanks are used, two IOOZ (minimum supply e of 300,000 gallons each) system capacity tanks should be Installed. They should bc so interconnected that pumps can take suction from either or both. Nowever, a failure in one tank or Its piping should not cause both tanks to drain. Mater supply capacity should be capable of refilling either tank in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or lese. (10) Common tanks arc permitted for fire and sanitary or service water Not applicable. storage. Mhen this ls done, however, minimum fire water storage requirements should be dedicated by passive means, for example> use of a vertical standpipe for other water services. Administrative controls> including locks for tank outlet valves, are unacceptable as the only means to ensure minimum water vol>m>e. (ll) The fire water supply should be calculated on the basis of the The fire water supply, 360,000 gallons, is calculated on the basis of thc largest expected flow rate for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, but not less than greatest system demand, 2,000 gpm plus a maximum hose stream demand of 300>000 gallons. This flow rate should be based (conservatively) on 1000 gpm for 'a duration of two hours. The fire pumps are sixed in 5VO gpm for manual hose streams pius the largest design demand of any accordance with NFPA-20 and are rated at 2500 gpm at 125 psig. Thcsc sprinkler or deluge system as determined ln accordance with NFPA 13 pumps are capable of delivering 150X of rated capacity at not less than or NFPA 15. The fire water supply should be capable of delivering 652 of rated head over the longest route of the water supply system. this design demand over the longest route of the water supply system. NUREG-OBOO GUIDEl IN'ONFORHANCB .2) Freshvater lakes or ponds of sufficient site may qualify as sole,. Thc Auxiliary Reservoir supplies fresh voter to the yard fire main. The source of vater for fire protection but require separate redundant electric and diesel vertical fire pumps are veil separated by being auctions in one or more intake structures. 'lhese supplies should be installed at opposite ends of the emergency service vater intake separated so that a failure of one supply will not result ln a structure. Each pump takes suction from a separate vct pit and has failure of the other supply. independent discharge connections, about 40 feet apart, to the main fire protection loop. .3) Nmn a common vater supply is permitted for fire protection and the ultlmatc heat sink, the following conditions should also be satisfied: (a) The additional fire protection vater requirements are designed (1) The additional fire protection voter requirements are designed into into the total storage capacity, and the total storage capacity and (b) Failure of the fire protection system should not degrade the (2) Failure of the Eire protection system will not degrade the function function oE the ultimate hest sink. of the ultimate heat sink (see Section 9.2.5)

4) Other water systems that may be used as one of the two fire voter Not applicable supplies should be permanently connected to the fire main system.and should be capable of automatic alignment to the fire main system.

Pumps, controls, and power supplies ln these systems should satisfy the requirements fro the main fire pumps. 'The use of other water systems for fire protection should not be incompatible vith their functions required for safe plant shutdown. Failure of the other system should not degrade the Eire main system. NUREC-0800 GUIDELINES CONFORHANCE 'Mater S rinkler ond Nose Stand I e S stems I) Sprinkler systems snd manual hose station standpipes should hove Ikeders, fed from both ends fn the Mnete ond Fuel Nandf fng Buildings, the connections to the plant underground water main so that a single Turbine snd Reactor Auxfliary Buildings ere fabricated of carbon steel ective failure or o crock fn a moderote-energy line cannot Impair piping end fittings meeting the requirements of ANSI B31.1 "Power both the primary nnd backup fire suppression systems. Alternatfvcly> Pfpfng" . Fach sprinkler snd standpipe system Is equfpped vfth sn OS6Y headers fed from each end are permitted Inside buildings to supply gate valve ond voter flov alarm except thnt In the RAB the header both sprinkler ond standpipe systems, provided steel piping and supplying the hose stond pipes Is arrnnged so that the OSSY gnte valves fn fittings meeting the requirements of ANSI B31.1, "Power Pipfng," ere the header on coch side of e stnndpfpc must be closed to isolate the used for the headers up to and fncluding the first valve supplying standpipe. Since this header Is fed from both ends the vatcr supply to the sprinkler systems where such headers ere part of the se)smicnlly ,.other stondpfpes served by thfs heqder fs not interrupted. 'Ihe Ffre analyzed hose standpfpe system. Mhen provfded, such headers are Itvzards nnnlysis descrfbes the methods used to protect safety-related considered on extension of the yard main system. Each sprinkler and equipment fn each fire area from vnter dnmage. standpipe system should be equipped vlth OSSY (outside acre~ and yoke) gate valve or other approved shutoff valve and vaterElow alarm. Safety-related equipment thnt does not Itself require sprinkler vater Eire protection but is sub)ect to unacceptable damage ff wet by sprfnkler water discharge should be protected by water shields or baffles.

2) Control ond sectfonalirfng valves In the fire water systems should be All sectional valves and any control valve upstream from deluge, pre" electrically supervfsed or admfnistrat~fvely controlled. The action and multimycle valves sre electrfcally supervised snd indicate on electrical supervision sfgnol should indicate In the control room. the Hnin Fire Detection Control Panel. CPIL will comply.

All valves in the fire protection system should be perfodfcnlly" checked to verify position (see NFPA 26, "Supcrvfsfon of Valves"). NUREG&800 GUIDELINES CONFORHANCE (3) Systems'omplied Ffxed voter extinguishing systems should conform to requirements of appropriate standards such as NFPA 13, "Standard for the Installation of Sprinkler Systems," and NFPA 15, "Standard for Mater Spray Fixed with (4) Interior manual hose installation should be able to reach sny Ibse stations are provfded In each ares so that all portions of tbe plant ~ location that contains, or could present a fire exposure hazard tof except the tank building vhich is protected by hydrants, can be reached safety-related equfpment with nt least one effective hose stream. To from at least tvo hose stations. Each station is ptovfded with 100 feet accomplish this, standpfpes vfth hose connectfons equipped vfth n of'"I/2 inches Angus "Red Chfef"'userubber-lined, rubber coated hose snd maxfmum of 100 feet of 1"1/2 Inch woven-)acket, lined fire hose and adjustable norzles suitable for on electr feel equipment. Individual suitable nozzles should be provided in all bufldfngs on all floors. snndpfpes ote 4 Inches fn diameter fot multiple hose stations and 2-1/2 Individual'tandpfpes should be at least 4 inches in diameter for Inches for single hose stations. 'Ibe stations follov the requirements of multiple hose connectfons and 2-1/2 inches fn dfameter for single NFPA 14 "Standpf pe and Nose Systems." hose connections. These systems should follow the requirements of NFPA 14, "Standpipe and Ibse Systems," for siring, spacing, and pipe support requirements. Nose stations should be located as dictated by tbe fire hazard Complied vith analysis to facilitate access and use for ffreffghtfng operations. Alternntfve hose stations should be provided for an area if the fire hazard could block access to a single hose station serving that area. Provisions should be made to supply voter at least to standpfpes and Piping for hose ststfons serving areas containing equipment required for hose connections for manual firefigbting in areas containing safe plant shutdovn after a safe shutdown earthquake hss been analyzed for equipment required for safe plant shutdown in the event of a safe SSE loading and nte provided with seismic supports. The piping and valves shutdovn earthquake. 'Ihe pfplng system servfng such hose stations fn these systems satisfy, as o minimum, ANSI B31.1, "Power Piping." should be analyzed for SSE loading and should bc provided with Following nn SSE, the vater supply Is obtained by manual operator supports to ensure system pressure integrity. 'Ibe piping and valves nctuatfon of valves to connect to the seismic Category I Emergency Service for tbe portion of hose standpipe system affected by this functional Mater System. Tbe system cross connections are capable of supplying tvo " 75 gpm hose stations. Tbe cross connections vere analyzed for seismic requirement should, ns a mfnimum, satisfy ANSI B31.1, Pover Piping" . The vater supply for tlifs condition may be obtained by landings nnd sefsmfcnlly supported to assure system integrity. It will manual operator actuation of valves fn a connection to the hose not degrade the performance of the seismic Category I voter system. standpipe header from a normal scfomfc Category I water system such as the essential service voter system. Tbe cross connection should be (a) capable of providing flow to at least tvo hose stations " (opproxfmately 75 gpm per hose station), and (b) designed to the same standards as the seismic Category I ~ster system; It should not degrade the performance of the sefsmfc Category I vater system. al I NURECW800 CUIDEllNES CONFORHANCF. (5) 'Ihe proper type of hose nozzle to be supplied to each area should be Adjustable spray nozzles, approved for use on energized electrical based on the fire hazard analysis. Ihe usual combination spray/ equipment, are provided on standpfpe hoselfnes available for discharge on straight-stream nozzle should not be used in areas uhere the straight electrical equfpment and cabling. stream can cause unacceptable mechanical damage. Ffxcd fog nozzles should be provided at locations ubere high-voltage shock hazards exist. All hose nozzles should have shutoff capability (Cufdance on safe distances for uater application to live electrical equfpment may be found fn the "NFPA Fire Protection llandbook." ) (6) Fire hose should be hydrostetfcally tested in accordance ufth tbe Fire hose are hydroststfcslly tested in accordance uftb the recommendations of NFPA 1962, "Fire Nose - Core, Usc, Haintenance." recommendatfons of NFPA 1962, "Fire llose - Cere> Use, Hafntenance" . CP6I llose stored in outsfde hose houses should be tested annually. ~ ulll comply. Interior standpipe hose should be tested every 3 years. (7) Certafn fires, such es those fnvolvfng flammable liquids, respond Fixed foam systems ufll not be used to protect safety-related systems. ucll to foam suppression. Consideration should be given to use of Portable foam equipment uill be available, if required. mechanical lou-expansion foam systems, high-expansfon foam generators, or aqueous film-forming foam (AFFF) systems, including the AFFF deluge system. Ihese systems should comply uith the requirements of NFPA 11, NFPA IIA, NFPA IIB, and NFPA 16, as epplfcable. NDREG&800 CVIDELINES CONFORHANCE

d. Halon Su ression S stems Halon fire extinguishing systems should comply <<1th the requirements of CPSL will comply.

NFPA 12A and NFPA 12B, "Halogenated Fire Fatingufshfng Agent Systems-Halon 1301 and Halon 1211." Only UL-listed or FM-approved agents should Halon suppression system is used fn the Records Storage areas located fn be used. Provisions for locally disarmfng automatic Halon systems should the Administration. Bui fdfng. be key locked and under strict administrative control. systems should not be disarmed unless controls ao described Automatfc'xtfnguishfng in Position C.2.) are provided. In addition to the guidelines oE NFPA 12A and 12B, preventive maintenance Preventive maintenance and testfng will be performed semfannually in and testing of the systems, including checkmefghfng of the Halon accordance with NFPA-12A. cylinders, should be done at least quarterlyo Particular consideration should also be gfven to: (1) Minimum required Halon consideration, distribution, soak time, and ventilation control; (2) Toxicity of Halon; (3) Toxicity and corrosive characteristics of the thermal decomposition products of Halon; and (4) Locatfon and selection of the actfvating detectors.

e. Carbon Dioxide Su resslon S stems Carbon dioxide extinguishing systems should comply with the requirements Carbon dioxide systems will not be used.

of NFPA 12, "Carbon Dioxide Extinguishing Systems." Nhere automatic carbon dioxide systems are used, they should be equipped with a prcdfscharge alarm system snd a discharge delay to permit personriel egress. Provisions for. locally disarming automatic carbon dioxide systems should be key locked and under strict administrative control. Automatic carbon dioxide extinguishing systems should not*be disarmed unless controls as described in Positfon C.2.c are provided . Particular consideration should also be gfven to: / (1) Minimum required C02 concentration, distribution, soak time, and ventf lotion control; (2) Anoxfa and toxicity of C02', (3) Possibility o! secondary thermal shock (cooling) damage; (4) ConElictfng requirements for ventfng during C02 fn)ection to prevent ovcrpressurfzatfon versus sealing to prevent loss of agent; and (5) Location and selection of the activatfng detectors. 49 NVREC&800 CV IDFEINES CONFORHANCE I. Portable Exti uishers Fire extinguishers should be provided in areas that contain> or could Complied with present a fire exposure hazard to, safety"related equipmcnt in accordance with guidelines of NFPA 10, "Portable Fire Extinguishers, Installation,

Maintenance nnd Vse." Dry chemical extinguishers should be installed with due consideration given to possible adverse effects on safety-relnted equipment installed in the area.

7. Guidelines for S ecific Plant Areas

a. Prima and Seconds Containment (1) Normal 0 eration - Pire protection requirements for the primary and Fire protection systems and equipment are provided in the containmcnt secondary containment areas should be provided for hazards identified areas as required for most effective fire control recognizing the by the fire hazards analysis. different types of operations in the area, accessibility and available personnel usage.

Examples of such hazards include lubricating oil or hy'draulic.'luid system for the primary coolant pumps, cable trey arrangements and The following hazards have been identified and protection is installed in cable pcnetrations, and charcoal filters. Because of the general each containment as follows: inaccessibility of primary containmcnt during normnl plant operation, protection should be provided by automatic fixed systems. The Cable penetrations, and external surfaces of effects of postulated fires within the primary containment should be charcoal filter equipment sre protected by an automatic multlmycle evaluated to ensure that the integrity of the primary coolant system sprinkler system, activated by rate-of-rise detectors. System will and the containment is not )eopardized assuming no action is taken to automatically shut off upon drop in temperature (i.e., fire extinguished) fight the fire. to reduce quantity of unborated water introduced into the containments. Closed sprinkler heads end supervisory air pressure provide adequate snfeguards against inadvertent activation. Valving and electrical equipment associated with the system arc located outside the containment building for availability. A lube oil collection system will be provided for the reactor coolant pumps as detailed in Section (;.5.d, and will be capable of collecting lube oil from all potential pressurized and unpressurized leakage sites in the reactor coolant pump lube oil systems. The effects of postulated fire within the primary containment are discussed in the Fire Hazards Analysis, Section 9.5A, of the FSAR. 50 I NURKC&800 CUIDFI.INKS CONFOR NNCK a) Operation of the fire protection systems should not compromise the Section 9.5.1.2.4 of the FSAR and Appendix 9.5A"1 of the Fire Nazard integrity of the containment or other safety-related systems. Fire Analysis describe the fire protection functions in (he contafnment. protection activities fn the containment areas should function fn con)unction with total containment requirements such as ventilation and control of contaminated liquid and gaseous release. l within the contafnmenty separation of cables and equipment and associated b) Inside noninerted containment one of the fire protectfons means" stated in Positions C.5.b.l snd C.5.b.2 or the Eollowlng fire nonsafety circuits oE redundant trains is achieved by spatial separation protection means should be provided: separation of cables and and/or structural barriers. Automatfc suppression systems are provided fn equipment and associated nonsafety circuits of redundant trains by a the cable penetration areas and over the charcoal filter housings and noncombustible radiant energy shield having a minimum f fre rating of reactor coolant pumps. onediaff hour. c) In primary containment, fire detection. systems should be provided for Complied with each Eire hazard. The type of detectfon used and the location oE the detectors should be the most suitable for the particular type of fire hazard identified by the fire hazard analysis. A general area fire detection capability should be provided ln the Backup detection on a general area basis is provfded by ionization primary containment as backup Eor the above described hazard detectors installed in the NVAC recirculation system upstream of system detection. To accomplish this, suitable smoke or heat detectors filters. compatible with the radiatfon environment should be Installed ~ d) Standpfpe and hose stations should be inside FUR containments and BWR Standpipe and hose stations are provided inside the contafnment and are containments that are not inerted. Standpipe and hose stations supplied Erom the yard fire main during normal operations. After a SSK, inside containment may be connected to a high quality water supply oE the stnndpfpes are supplied from the Emergency Service Mater System. See sufficient quantity and pressure other than the fire main loop if the response to Position 6.c(4) above. plant-specific features prevent extending the fire main supply inside containment. For BMR drywells, standpipe and hose stations should be placed outside the drywell with adequate I'engths of hose, no longer than 100 ft, to reach any location inside the drywell with an l ef fective hose stream. The containment penetration of the standpipe system should meet the Complied with isolation requfrements of Ceneral Design Criterion 56 and should be seismic Category I and Quality Croup B. CUNFORMANCE NUREC&800 CUIDFLINES (e) The reactor coolant pumps should he equIpped with an oil collection CP6L will comply. system The if the containment oil collection system is not inertcd during normal operation. should be so designed, engineered, and installed that failure will not lead to fire during normal or design basis accident conditions and that there is reasonable assurance that thc system will withstand the safe shutdown earthquake. Such collection systems should be capable of collecting lube oil from all potential pressurized and unpressurized leakage sites in the reactor coolant pump lube oil systems. Leakage should bc collected and drained to a vented closed container that can hold the entire lube oil system inventory. A flame arrester is required in the vent if the flash point characteristics of thc oil present the hazard of lift fire flashback. Leakage points to be protected should include pump and piping overflow lines, lube oil cooler, oil fill and drain lines and plugs, flanged connections on oil lines, and lube oil reservoirs where such features exist on the reactor coolant pumps. The drain line should be large enough to accommodate the largest potential oil leak. (f) For secondary containmcnt area, cable fire hazards that could affect Cable fire hazards within the Secondary Containment Area are separated by safety should be protected as described in Position C.5.e(2). The structural barriers and physical separation. Automatic sprinkler systems type of detection system for other fire hazards identified by the are provided in the electrioal cable trays, the electrical penetration fire hazards analysis should be the most suitable for the particular areas and over the charcoal filter housings, actuated by thermal i type o f f re ha zard. detectors. Ionization type smoke detectors are also provided over the electrical cable trays on el. 261.00 ft. NUREC&800 GUIDELINE CONFORHANCE h (2) Refuelin and Maintenance - Refueling and maintenance operations in CP&L will comply containment may introduce additional hazards such as contamination control materials, decontamination supplies, wood planking, temporary wiring, welding, and Elame cutting (with portable compressed~as fuel supply). Possible Eires would not necessarily be in the vicinity of fixed detection and suppression systems. Hanagement procedures and controls ncccssary to ensure adequate fire protection for"transient fire loads are discussed in Position C.2. Adequate self-contained breathing apparatus should be provided near CP&L will comply the containment entrances for f irefighting and damage control personnel. These units should be independent of any breathing apparatus or air supply systems provided for general plant activities and should be clearly marked as emcrgcncy equipment.

b. Control Room Complex The control room co~plex (including galleys, office spaces, etc.) should The Control Room Fire Arcs is separated from other areas of the plant by be protected against disabling fire damage and should be separated from walls, Eloors and ceiling having minimum Eire resistance ratings of other areas of the plant by floors, walls, and roof having minimum fire 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. (The Terminal Cabinet Room is a part of thc Control Room Fire resistance ratings of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Peripheral rooms in the control room Area. Refer to FSAR Appendix 9.5B-3 for an exemption request and a detail complex should have automatic water suppression and should be separated )ustification Eor not providing automatic suppression or fire wall in the from the control room by noncombustible construction with a fire Terminal Roon). A kitchen, dfficc and the component cooling water tank resistance rating of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Ventilation system openings between the are located within the Terminal Cabinet Room. Combustibles in the kitchen control room and peripheral rooms should have automatic smoke dampers that consists of limited amounts of ordinary Class A combustibles such as paper close on operation of the fire detection or suppression system.. IE a towels and napkins. The combustibles in the Terminal Cabinet Room are halon flooding system is used Eor fire suppression, these dampers should limited to cables within the panels which are considered negligible. A bc strong enough to support the pressure risc accompanying halon discharge Eire hose station and portable extinguisher are located in this room.. Thc and seal tightly against infiltration of halon into the control'oom. Control Room Fire Area is served by AN-15 (lA-SA) backed up by Carbon dioxide flooding systems are not acceptable Eor these areas. AN-15(lB"SB) cool the Control Room Fire Area. In case of fire, the ventilation to the Control Room Fire Zone is lost and ventilation to the Hain Terminal Cabinet Fire Zone is provided by AN-97(1&2A-NNS) backed up by AN&7(1&2B-NNS) which are both loaded on thc diesel as described in FSAR Section 9.4. This arrangement precludes thc need for automatic smoke dampers in ducts penetrating the walls between the Control Room complex and peripheral rooms.

I NU REC&800 CU IDELINFq CONFOR NANCE Manual firefighting capability should be provided for both: Manual fire fighting capability is provided for fires (1) uithin cabinets. consoles or connecting cables and (2) exposure fires involving (1) Fire originating vithin a cabinet, console, or connecting cables; and combustibles in the general room area by means of Class A and Class C fire extinguishers inside the control room and a hose iocated )ust outside the (2) Exposure fires involving combustibles in the general room ares. control room. portable Class A and Class C fire extinguishers should be located in the control room. A hose station should be installed immediately outside the control room. J Noxxles that are compatible uith the haxards and equipment in the control noxxles, approved for use on electrical fires, wi11 be provided impingement'djustable room should be provided for the manual hose station. %he noxxles chosen should satisfy actual firefighting needs, satisfy electrical safety, and minimixe physical damage to electrical equipment from hose stream for the damage hose station. %he noxxle selected mill also minimixe physical to electrical equipment from hose stream impingement. Smoke detectors should be provided in the control room, cabinets, and Ionization detectors ui11 be provided in the control and peripheral consoles. If redundant safe shutdoun equipment is located in the same rooms. The use of this sensitive type of detector mitigates the need for control room cabinet or console, additional fire protection measures detectors uithin control room cabinets or consoles. Alarm and local should be provided. Alarm and local indication should bc provided in the indication vill be provided in the control room. control room. Breathing apparatus for control room operators should be readily available. CPSL vill comply &e outside air intake(s) for the control room ventilation system should See Section 9.5.1.2.4 of the FSAK for treatment of the smoke detection be provided uith smoke detection capability to alarm in the control room capability in the Control Room f ire area. to enable manual isolation of the control room ventilation system and thus prevent smoke from entering the control room. NUREG-0800 GUIDELINES CONFORMANCE Venting of smoke produced by fire in the control room by means of the See terna Section 9.5.1.2.4 of the FSAR for discussion of this point. Isolatgon normal ventilaton system is acceptable; however, provision should be made dampers arc provided in the recirculation portion of the ventilating to permit isolatfon of thc recirculating portion of the normal v'entilation 8ys system. Manually operated venting of the control room should be, available to the operators. All cables that enter the control room should terminate in the control In Section 9.5.1.2.4 of the FSAR it is stated all cables entering tbc room. That is, no cabling should be routed through the control room from control room terminate there. No cables are routed through the control one area to another. Cables in underfloor and ceiling spaces should meat room from one area to another. There are no raised floors in the control the separation criteria necessary for Eire protection. room. There is a trench under the llVAC Control vhicb is about ll long x 2 feet vide x 8 inches deep vbich contafns only Train B cable, Eeet safety hnd nonsafety. The fire loading is low> less than 2000 BTU/sq. Et. No suppression system is provided. There are redundant safety related radiation monitoring cables, installed in conduits and in accordance wfth Regulatory Guide 1 .75, located above tbe suspended ceiling. As stated in the Fire Ihzards Analysis, Section 9.5A of the PSAR, the combustible loadinglin the Control Room is considered negligible. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> occupancy of the Control Room combined vith thc availabl.lity of fire extinguishers and hose stations mitigate tbe effects of an exposure fire. Air-handling functions should be ducted separately from cable runs in such All air handling functions for the Control Room system are provfded from spaces; i.e., if cables are routed inlunderfloor or ceiling spaces, these spaces should not be used as air plenums for ventilation of the control above the hung ceiling. Cables serving the Control Room come from the floor below and tcrmfnate in tbh control boards. Conduit 4" and smaller room. Fully enclosed electrical racevays located in such underfloor and if vill be sealed in accordance vith NUREG-0800 criteria. Internal smoke ceiling spaces, over 1 square foot in crossmectional arcs, should have seals vill be provided at both terminations of the conduit, alleviating automatic fire suppression inside. Area automatic fire suppressfon should the need for automatic fire suppression systems. be provided for underfloor and ceiling spaces if used Eor cable runs unless all cable is run in 4-inch or smaller steel conduit or the cables I are in fully enclosed raceways internally'rotected by automatic, Eire suppression. I There should be no carpeting in the control room. There is no carpeting in the control room. NVREG&800 GUIDELINES CONFORHANCE

c. Cable S reading Room The primary Eire suppression in the cable spreading room should be an An automatic pre-action water suppression system employing closed automatic water system such as closed-lead sprinklers, open-head deluge sprinkler heads is installed in the cable spreading room. Cable tray system, or open directional water spray system. Deluge and open'spray arrangements were'onsidered in the location of sprinkler heads to insure systems should have provisions formanual operation at a remote station; adequate water coverage. Since there are only cables in this room, thc however, there should be provisions to preclude inadvertent operation. Fire llazards Analysis postulates that transients such as oil, grease, rags Location of sprinkler heads or spray nozzles should consider cable tray or solvents normally associated with equipment maintenance or repair will arrangmcnts and possible transient combustibles to ensure adequate water not be brought into the area. The pre-action valve can be tripped coverage for areas that could present exposure hazards to the cable mechanically at the valve or operation of pull stations located inside system. Ca bles should be designed to allow wetting down with water oroutside the room located at elevation 286'nd 305'. Inadvertent supplied by the fire suppression system without electrical faulting. operation is precluded by the two step discharge cycle of the pre-action system which requires both the operation of the pre-action valve and fusing of the sprinkler head. Cables are designed to allow wetting down by water from the fire protection system.

Open-head deluge and open directional spray systems should be zoned ~ Closed sprinkler heads are used. The use of foam is acceptable. Foam is not being used. Cable spreading rooms should have: The cable spreading rooms have: (I) At least two remote and separate entrances for access by flic brigade (1) Two or more remote and separate entrances; personnel; (2) An aisle separation between tray stacks at least 3 feet wide and 8 (2) Aisles to facilitate access in the cable spreading Eeet high; rooms have been provided, however, duc to redesign to provide redundant cable spreading'rooms, the aisles have been reduced in dimensions. Depending upon their location, they vary from 3 feet wide by 8 feet high to a minimum of l-l/2 to 2 feet wide by 5 feet high. A number of access doors exist. A trained fire fighter can access thc area with his equipment, provided that he is familiar with thc layout through training. A visual display of smoke detectors is provided at the local control panel. Thc Eire fighter will be cognizant of the location of the fire and will use the proper aisle to facilitate fire attack strategy. (3) . Ibse stations and portable extinguishers installed immediately (3) Portable extinguishers located inside and outside the room and hoses outside the room; located immediately outside the room; (4) Arcs smoke detection; and (4) Arcs smoke detection; and (5) Continuous line-type heat detectors for cable trays inside the cable (5) Ionization detectors are used to provide early warning of incipient spreading room. fires and permit early attack by manual means. Thermal detectors located at the ceiling actuate the automatic suppression system. The dual detection system provides supplementary means of fire detection in lieu of solely depending upon line-type temperature detection. $6 NUREC&800 GUIDELINES CONFORMANCE ( Drains to remove firefighting uatcr should be provided. llhen gas sytcms The f lou drainage system is designed to handle the design sprinkler are installed, drains should have adequate seals or thc gas extinguishing discharge. There is no gas system. systems should be sized to compensate for losses through the drains. A separate cable spreading room should be provided for each redundant Neu response addresses it specifically on Page 28. division. Cable'spreading rooms should not bc shared bctuccn rcacfors. Each cable spreading room should be separated from the others and from other areas of the plant by barriers uith a minimum fire rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. If this is not possible, a dedicated system should be provided. The ventilation system to each cable spreading room should be designed to There is no gas extinguishing system for the cable spreading rooms, isolate the area upon actuation of any gas extinguishing system in the therefore, ventilation system isolation is not required. Smoke venting is area. Separate manually actuated smoke venting that is operable from accomplished from the Control Room upon activation of a ionization outside the room should be provided for thc cable spreading room. detector varning of the presence of smoke. Sce Section 9 5 ~ 1 ~ 2.4 of the FSAR,

d. Plant Com uter Rooms Computer rooms for computers performing safety-related functions that are The SHNPP computer is non-safety related and does not perform any safety not part of the control room complex should be separated from other areas function. Thc nonmafety computer is located outside the Control Room and of thc plant by barriers having a minimum fire resistance rating of 3 is provided Mith 3-hr fire resistance rated barriers on all sides except hours and should bc protected by automatic detection and fixed automatic those separating it from the Auxiliary Relay Panels Room, Unit I, snd the

l. Automatic ionization suppression. Computers that are part of the control room complex but not Process Instruments and Control Racks Room, Unit in the control room should be separated and protected as dcscribcd in type smoke detectors ar'e provided in the room, portable extinguisher and a Position C.l. b. Computer cabinets located in the control room should be hose station ad)scent to the room is available.

protected as other control room equipment and cable runs therein. Nonsafety-related computers outside the control room complex..should be separated from safety-related afeas by fire barriers uith a minimum rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and should be prtected as needed to prevent fire and smoke damage to safety-related equipment. I NUREC&800 GUIDEI.INES CONFORHANCE Switchgear rooms contaning safety"related equipment should be separated Redundant switchgear rooms containing safety-related equipment. are from the remainder of the plant by barriers with a minimum fire rating of separated from each other and other plant areas by walls having, as a 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Redundant switchgear safety divisions should be separated from minimum, a fire resistance rating of three hours. Automatic detectors in each other by barriers with a 3*our fire rating. Automatic fire each room alarm locally and in the control room. Cables passing through detectors should alarm and annunciate in the control room and alarm the switchgear rooms are held to a minimum. 'Ihe rooms sre used only for locally. Cables entering the switchgear room that do not terminate. or switchgear and battery chargers which connot be located in battery rooms-prform a function there should be kept at a minimum to minimixe the . for safety reasons. Carbon dioxide portable extinguishers sre located in combustible loading. 'Ihese rooms should not be used for any other and ad)scent to the rooms. Nose stations sre ad)scent to the rooms. purpose. Fire hose stations and portable fire extinguishers should be readily available outside the area. Equipment should be located to facilitate access for manual firefighting. Equipment is arranged to facilitate access by firefighters. Equipment is Drains should be provided to prevent water accumulation fromdamaging mounted on 4 inch pcdestals and floor drains are provided to handle wats safety-related equipment (see NFPA 92H, "Materproofing snd Draining of discharged by hoses. Smoke is removed by the normal ventilation system Floors" ). Remote manually actuated ventilation should be provided for for this area which is switched remote~nually to once through purge venting smoke when manual fire suppression effort is needed (see Position operation. 'Ihe system used is AN-12 snd 13 for supply snd valved roof C.S.f). vents for exhaust. Refer to Section 9.5A in the FSAR for additional information.

f. Remote Safet -Related Panels Redundant safety related panels remote from the control room complex Areas remote from the control room, containing safety related panels, are should be separated from each other by barriers having a minimum fire provided with detectors which alarm locally nnd alarm and annunciate in rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Panels providing remote shutdown capability should be the Control Room. Panels providing remote shutdown capability are located separated from the control room complex by barriers having a minimum fire in the Aurilary Control Panel Room which is remote from the Control Room rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Panels providing remote shutdown capability should be and separated from other plant areas by barriers having a fire resistance electrically isolated from the control room complex so that s fire in rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, Panels providing remote shutdown in the Auxiliary either area will not affect shutdown capability from the other area. 'Ihe Control Panel Room snd those in the Control Room are electrically isolated general area housing remote safety-related panels should be provided with and sre connected to redundant transfer nanels each of which are located automatic fire detectors that alarm locally 'and alarm and annunciate in in separate fire areas.

the control room. Combustible materials should be controlled snd limited Ionixation detectors to those required for operation. Portable extinguishers snd manualhose in the Auxiliary Control Panel Room alarm locally and alarm and annunciate stations should be readily available in the general area. in the Control Room. Portable extinguishers and manual hose stations are available in the area. Redundant safety related panels required for Safe Shutdown are separated as described in FSAR Appendix 9.SB. I 5 HURSG-0800 GUIDgl.lHES COHFORHAHCE 4 '. Safet -Related Hatter Rooms Safety-related battery rooms should ba protected against fires and The battery rooms are cut oEf from each other and other plant areas by I explosions. Sattary rooms should be separated from each other and other barriers having a minimum fire resistance rating of three hours DC ~ ! arras of the plant by barriars having a minimum Eire rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> switchgear and inverters are not located in the battery rooms. Refer to ~ inclusive of all penetrations and openings. DC switchgear and invrrters RSAR Sections 9.4 5.2.3, 9.4.5.2.4 and 9.5-1 for the description of the ~ lshould not be located in these battery rooms. Automatic fire detection vrnilation and detection systems. should be provided to alarm and annunciate in the control room and alarm 'ocally. Ventilation systems in the battery rooms should be capable of maintaining the hydrogen concentration well b~low 2 vol-Z. Loss of ventilation should be alarmed in the control room. Standpipe and hose and portable. extinguishers should be readily available outside the room. I ! The turbine building should ba separatad from adjacent. structures The design of the turbine building compiles with this position. < containing safety-related equipment by a Eire barrier with a minimum

rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The fire barriers should be designed so as to maintain

structural integrity even in the event of a complete collapsa of the "turbine structure. Openings and penetrations in the fire barrier should ba minimized and should not be located where the turbine oil system or generator hydrogen cooling system creates a direct fire exposure haiard to

'he barrier. Considering the severity oE the Eire hazards> defense, in depth may dictate additional protection to ensure. barrier integrity. ~ ~ NUREC&800 GUIDELINES Q)NFORNANCE Diesel Generator Areas Diesel generators should be separated from each other and from other. areas The diesel gcncratois are located within the diesel generator building of the plant by fire barriers having a minimum fire resistance rating of 3 separated,from each other by barriers having a minimum fire resistance hours rating of three hours and thc diesel generator building is about 175 feet from the other plant buildings. Automatic fire suppression should be installed to combat any diesel Automatic multlcycle suppression systems are installed in the diesel generator or lubricating oil fires; such system should be designed for generator rooms to protect against diesel generator or lubricating oil operation when the diesel is running without affecting the diesel. fires. These systems will not affect the diesel when it is running since fire detection should be provided to alarm and 'annunciate in the 'utomatic combustion air intake is located outside thc room. Automatic dctectlon is control room and alarm locally. Ibse stations and portable extlnguishers provided to alarm locally and annunciate in the control room. lhse should be readily available outside thc area. Drainage for f ireflghting stations and portable extingulshcrs arc available outside the rooms. water and means for local manual venting of smoke should be provided. Drainage for flrcfighting water is 'provided. Thc continuous use of the normal ventilation exhaust system provides smoke purging. The electrical equipment room employs a recirculating system which upon activation of >'moke detectors in the space can be switched to a once through purge system. The system used for this function is described in FSAR Section 9.4. NUREC-0800 CUIDEL1NFS COMFORHANCE Iby tanks ufth total capacity up to 1100 gallons are permitted fn the Ilie dsy tank, capacity 3,000 Fallons, for each diesel generator la diesel generator area under the iolloulng conditions: contained ulthin fts individual enclosure having a minimum ffre resistance rating of three hours. All penetration barriers and the door hove the same fire resistance rating. There is a three foot high dike fn the dooruay vhfch is able to contain 1102 of the tank contents. The tank is also equipped vfth an automatic fill shutoff. 'The floor drain is equipped ufth a normally closed valve located outside the room uhich uhen opened. drains to the diesel generator sump. The installatfon of the day tank ls fn accordance ufth BUFFA 37- "Installation and Use of Stationary Comhastfon-Engines and Css Turbfnes". 'Ibe tonk and associated pfping nre deaf pned to safety Class 3 and seismic Category 1 requirements mfnfmirfng the chance oi oil spflls and fires fn the room. The ateas fs also provided ufth an automatic multicycle water suppression system activated by thermal detectors> backed up by hoses snd portable extfngufshers outside the room. (1) '1he day tank is located in a separate enclosure with a minimum ffre resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, including doors or penetrations. These enclosures should be capable of containing the entire contents of the day tanks and should be protected by an automatic f fre suppression system, or (2) The day tank is located inside the diesel generator room fn a dfked enclosure that has sufffcfent capacity to hold 1102 fo the contents of the day tank or is drained to a safe location. I NUREG-0800 GUI DELlttFS CONFORttAttCE Diesel Fuel Oil Store e Areas Diesel fuel oil tanks with a capacity greater than 1,100 gallons should Thc diesel fuel oil storage tanks sre burled outside st s distance of not be located inside buildings contalnlng safety"related equipment. If about 150 feet from pririclpal plant buildings. Redundant fuel oil above-ground tanks are used, they should be located at least 50 feet from transfer pumps are protected by automatic multi,-cycle suppression systems any building containing safety-related equlpmcnt or, lf located within 50 and separated by barriers having a fire resistance rating of at least three hours. Carbon dioxide and dry chemical extlnguishers as well as feet, they should be housed ln a separate building vith construction havalng a minimum fire resistance raCing of 3 bours. Potential oil spills fire hydrants are available. should be conElned or directed away from buildings containing safety-related equipment. Totally burled tanks are acceptable outside or under buildings (sce NFPA 30, "Flammable and Combustible Liquids Code," Eor additional guidance). Above-ground tanks should be protected by an automatic fire suppression systcmo

k. Safet Related Pum s Pump houses snd rooms housing redundant safety-related pump trains should Safety-related pumps are generally located within areas provided vlth be separated Erom each other and from other areas oE the plant by Eire automatic detection and suppression systcmso Separation between redundant barriers having at least 3&our ratings. 'Ihese rooms should be protected pumps ls achieved by three hour rated barriers partial height barriers, by automatic fire detection nd suppression unless a fire haxards analysis special distance or a combination of these elements. Portable Eire can demonstrate that a fire vill not endanger other safety-related extlngulshers snd hoses are also available. Refer to FSAR Section 9.5A safety related equipment required for safe plant shutdown. Fire detection should alarm for details on pumps.

and annunciate in the control room snd alarm locally. ltose stations and portable extlngulshers should be readily accessible. Floor drains should be provided to prevent water accumulation from Floor drains are designed to e'ccbmmodate any vster discharged from fire damaging safety-related equipment (see Position C.5.a(14)). suppression equipment. Provisions should be made for manual control of the ventilation,system to Smoke removal is assured by continuous operation of the once through facilitate smoke removal if required Eor manual Elrefightlng operation normal used ventilation system for the safety"related pump areas. The system Eor this Eunctlon is described ln FSAR Section 9.4. (sce Position C.5.f). I NUREC-0800 CUIDKllNFS CONFORNANCB

1. Nev Fuel Area Rand portable extinguishers should be located within this area. Also, Portable fire extinguishers snd hose stations are available in the area.

hose stations should be located outside but uithin hose reach of this The Fire Ihzards Analysis reports the amount of combustibles present ia. area. Automatic fire detection should alarm and annunciate in the control negligible and no detection is required. room and alarm locally. Combustibles should be limited to a minimum in thc neu fuel area. The storage area should bc provided with a drainage system to preclude accumulation of Mater. 'Ihe storage configuration of neu fuel should aluays be so maintained as to Complied vith. Described in FSAR Section 9.1.I. preclude criticality for any uater density that might occur during fire ua ter a ppl i ca t ion.

m. S nt Fuel Pool Area Protection for the spent fuel pool area should be provided by local hose Portable fire extinguishers and hose stations are provided for this ares.

stations and portable extinguishers. Automatic fire detection should be Automatic detection is not provided because of the insignificant fire provided to alarm and annunciate in the control room and to alarm locally. loading, as determined by the fire hazards analysis, in the area. Cl I ~ ~ NDREC&800 GUIDELINE CONFORMANCE

n. Radwaste and Decontamination Areas Fire barriers, automatic fire suppression and detection, and ventilation The radwastc and decontamination areas are separated from other plant controls should be provided. areas by reinforced concrete walls constituting three%our barrier ratings with similarly rated doors and penetration barrieis. Ionization detectors arc provided in the drumming area which alarm locally and annunciate in.

the control room. Portable fire extinguishers and hose stations are provided in the area. Smoke removal is assured by the continuous operation of thc once through normal ventilation system for the radwaste and decontamination areas. The .system used for this function is described in FSAR Section 9.4. Based on in-situ and transient combustible loading found in each fire zone/fire ares during normal plant operation, automatic suppression snd/or detection were provided, as detailed in the fire hazards analysis.

o. Safet -Related 'Pater Tanks Thc Condensate Storage Tank that supplies water for safe shutdown is located in a separate tank building which is separated from other plant Storage tanks that supply water for safe shutdown should be protected from areas having barriers with e minimum fire resistance rating of three the effects of an exposure fire, Combustible materials should not be hours. Combustibles will not be stored next to these tanks.

stored next to outdoor tanks.

p. Records Store e Areas Records storage areas should be so located and protected that a -fire in CPAL will comply.

these areas does not expose safety-related systems or equipment (see Regulatory Guide 1.88, "Collection, Storage, and Haintenancc of Nuclear Power Quality Assurance Records"). q. -Coolin Towers NUREC&800 CUIDELINES CO UFO RHANCE Cooling towers should be of noncombustible construction or-so located and The cooling towers will be of noncombustible construction and vill not bc protected that a fire will not adversely affect any safety-related systems used ns the ultimate heat sink or for the primary. source of fire or equipment. Cooling towers should be of noncombustible construction protection water supplies. when the basins are used for the ultimate heat sink or for the fire protection ~ster supply.

r. Hiscellaneous Areas Hiscellaneous areas such as shops, varehouses, auxiliary boiler rooms, Miscellaneous areas such as shops, varehouses, and administration areas fuel oil tanks, and flammable and combustible liquid storage tanks should vill be so located and protected~such that a fire or the effects of a be so located and protected that a fire or effects of a fire, including fire, including smoke, vill not adversely affect safety-related systems or smoke, will not adversely affect any safety-related systems or equipment . equipment.

8. Special Protection Guidelines

a. Storage of Acct lene-Ox en Fuel Cases Cas cylinder storage locations should not be in areas that contain or Complied vith expose safety-related equipment or the fire protection systems that serve those safety-related areas. A permit system should be required to use this equipment in safety-related areas of the plant (also see Position C.2).

b. Stora e Areas for Ion Excha e Resins Unused ion exchange resins should not be stored in areas that contain or CP&L vill comply expose safety-related equipment.

NVREG&800 GUIDEL1NES C0NF0 It MANGE I

c. Hazardous Chemicals l)axardous chemicals should not be stored in areas that contain or expose COL vill comply.

I safety related equipmcnt.

d. Materials Containi Radioactivit Materials that collect and contain radioactivity such as spent ion = CpgL vill comply.

exchange resins, charcoal filters, snd l)EPA filters should be stored in closed metal tanks or containers that sre located in areas free from ignition sources or combustibles. These materials should be protected from exposure to fires in ad)scent areas as veil. Consideration should be given to requirements for removal of decay hest from entrained radioactive 4 materials. Shearon Harris Nuclear Power Plant Open Item 351 estion 280.11 (SR) Verify that the closing of fire doors will be supervised by one of the measures stated in BTP CMEB Section C.5.a. ~Res esse: Fire doors will be supervised by one or more of the methods identified in CMEB 9.5-1 C.5.a.5:

a. Fire doors should be kept closed and electrically supervised at a continuously manned location;

b. Fire doors should be locked closed and inspected weekly to verify that the doors are in the closed position; Ce Fire doors should be provided with automatic hold-open and release mechanisms and inspected daily to verify that doorways are free of obstructions; or d~ Fire doors should be kept closed and inspected daily to verify that they are in the closed position.

CP&L will meet the intent of NUREG-0800 CMEB 9.5-1 C.5.a.5 concerning supervision of fire doors. Presently we are performing an evaluation to determine the most cost effective means of performing this task. s (7459 JDKk) r) Shearon Harris Nuclear Power Plant Open Item 353 estion 280.30 (SR) It is our position that the reactor coolant pumps be equipped with an oil collection system in conformance with Section C.7.a of BTP CMEB 9.5.1. Provide the design description of this system. ~Res esse The response to question 280.30 is addressed in the CPGL response to question 280.1 (Open Item 350, Comparison of the Shearon Harris Plant Fire Protection Program to BTP CMEB 9.5-1). Shearon Harris Nuclear Power Plant Acce tance Review uestion 280.18 Fire Area 14 (Primary and Secondary Shield Mall location): It is our position that you provide an engineered oil containment and collection system for the reactor coolant pump lube oil system to comply with Section D of Appendix R to 10 CFR Part 50. ~Res onse Refer to paragraph 7.a.l.e of the response to SHNPP NRC Question 280.1, the comparison of SHNPP Fire Protection against Branch Technical Position CMEB 9.5-1. Shearon Harris Nuclear Power Plant Acce tance Review Question 280.24 Appendix R to 10 CFR Part 50 will also be used as guidance for our review of your fire protection program. Your compliance with the requirements set forth in Appendix R as modified by accepted exceptions your program takes to the requirements of Appendix R as well as BTP ASB 9.5-1, and describe your alternative for providing an equivalent level of fire protection. ~Res ense Refer to the "Safe Shutdown Analysis in Case of Fire," and Shearon Harris Question 280. 1, the comparison of SHNPP Fire Protection against Branch Technical Position CMEB 9.5-1. (7677FXT cfr) r