ML20040D833
| ML20040D833 | |
| Person / Time | |
|---|---|
| Site: | 05000514, 05000515 |
| Issue date: | 01/31/1982 |
| From: | PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20040D827 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-***, TASK-TM NUDOCS 8202020299 | |
| Download: ML20040D833 (300) | |
Text
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File this instruction sheet in the front of Volume 1 as a record of
) changes.
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The following information and checklist are furnished as a guide for the
- insertion of new sheets for Amendment 13 into the Preliminary Safety Analysis Report for the Pebble Springs Nuclear Plant. This material is
- denoted by use of the amendment number and date in the lower outside corner of the page.
New sheets should be inserted as listed below:
Discard Old Sheet Insert New Sheet (Front /Back) (Front /Back)
Transmittal Letter CHAPTER 1 1-1/1-11 1-1/1-11 1-v/1-vi 1-v/1-vi 1.4-1/1.4-2 1.4-1/1.4-2 thru thru 1.4-15/1.4-16 1.4-15/ blank Fig 1.4-1 -
CHAPTER 13 13-i/13-11 13-1/13-11 13-iv/13-v 13-iv/13-v 13.1-1/13.1-2 13.1-1/13.1-2 thru thru 13.1-17/ blank 13.1-11/ blank Tab 13.1-1(Sh 1 of 46)/ Tab 13.1-1(Sh 1 of 41)/
Tab 13.1-1(Sh 2 of 46) Tab 13.1-1(Sh 2 of 41) thru thru Tab 13.1-1(Sh 45 of 46)/ Tab 13.1-1(Sh 41 of 41)/ blank l Tab 13.1-1(Sh 46 of 46) i
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Amendment 13 8202020299 820129 (January 1982)
PDR ADOCK 05000 B
Discard Old Sheet Insert New Sheet (Front /Back) (Front /Back)
Tab 13.1-2(Sh 1 of 39)/ --
Tab 13.1-2(Sh 2 of 39) thru -
Tab 13.1-2(Sh 39 of 39)/ --
Tab 13.1-3 Fig 13.1-1/ Fig 13.1-2 Fig 13.1-1/ blank thru thru Fig 13.1-11/ blank Fig 13.1-4/ blank CIIAPTER 17 Attachment 1 - Bechtel Quality Assurance Program 1/ blank 1/2 thru 7/ blank TAB - A.9 A.9-1/ blank thru A.9-119/A.9-120 Tab A.9-1/ Tab A.9-2 thru Tab A.9-9/ blank Fig A.9-1/ blank thru Fig A.9-10/ blank l
l Amendment 13 2 (January 1982)
CHAPTER
1.0 INTRODUCTION
AND GENERAL DESCRIPTION
/ OF PI. ANT
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CONTENTS Section Title Page 1.1 Introduction . . . . . . . . . . . . . . . . . . 1.1-1 1.1.1 Principal Aspects of Application . . . . . . . . 1.1-1 1.1.2 Report Format . . . . . . . . . . . . . . . . . 3.1-2 1.1.3 Glossary of Terms . . . . . . . . . . . . . . 1.1-3 1.1.3.1 Text Definitions and Abbreviations . . . . . . 1.1-4 1.1.3.2 Drawing Index and Symbols . . . . . . . . . . 1.1-4 1.1.3.3 Valve Identification . . . . . . . . . . . . . 1.1-4 1.1.3.4 Instrumentation Identification . . . . . . . . 1.1-5 1.1.3.5 Electrical Component Identification . . . . . 1.1-5 1.2 General Plant Description . . . . . . . . . . . 1.2-1 1.2.1 Design Criteria . . . . . . . . . . . . . . . . 1.2-1 1.2.2 Operating Characteristics . . . . . . . . . . . 1.2-2
-s 1.2.3 Nuclear Steam System . . . . . . . . . . . . . . 1.2-2 V 1.2.4 Engineered Safety Features . . . . . . . . . . . 1.2-3 1.2.5 Instrumentation and Control Systems . . . . . . 1.2-6 1.2.5.1 Reactor Protection System . . . . . . . . . . 1.2-6 1.2.5.2 Engineered Safety Features Actuation System . 1.2-7 1.2.5.3 Integrated Control System . . . . . . . . . . 1.2-7 1.2.5.4 Nuclear Instrumentation System . . . . . . . . 1.2-7 1.2.5.5 Nonnuclear Instrumentation System . . . . . . 1.2-8 1.2.5.6 Incore Monitoring System . . . . . . . . . . . 1.2-8 1.2.6 Reactor and Plant Control . . . . . . . . . . . 1.2-8 1.2.7 Electrical Systems . . . . . . . . . . . . . . . 1.2-8 1.2.8 Power Conversion . . . . . . . . . . . . . . . . 1.2-9 1.2.9 Fuel Handling and Storage System . . . . . . . . 1.2-9 1.2.10 Cooling Water and Other Auxiliary Systems . . . 1.2-10 1.2.11 Radioactive Waste Treatment Systems . . . . . . 1.2-12 1.3 Comparison Tables . . . . . . . . . . . . . . . 1.3-1
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1.3.1 Comparisons with Similar Facility Designs . . . 1.3-1
%J 1-1
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1 CHAPTER 1.0 l INTRODUCTION AND GENERAL DESCRIPTION OF PLANT CONTENTS Section Title _ _ _ _ _ , _ _ _
Page 1.3.2 Comparison of Final and Preliminary Information 1.3-1 1.4 Identification of Agents and Contractors . . . . . 1.4-1 1.4.1 Architect / Engineer-Constructor . . . . . . . . . . 1.4-1 1.4.1.1 Division of Responsibility for Engineering and Design . . . . . . . . . . . . . . . . . . 1.4-1 gl 1.4.1.2 Engineer's Design Organization and Methods . . . 1.4-3 C
1.4.1.3 Portland General Electric Company's Engineer-ing Organization and Methods . . . . . . . . . 1.4-4 1.4.2 Nuclear Steam System Supplier . . . . . . . . . . 1.4-5 1.4.3 Award of Construction and Procurement Contracts . 1.4-7 1.4.3.1 Division of Responsibility for Contracting and Procurement . . . . . . . . . . . . . . . . 1.4-8 1.4.3.2 Engineer's Contracting and Procurement Organ-ization and Methods . . . . . . . . . . . . . 1.4-9 1.4.3.3 Portland General Electric Company's Contract-ing and Procurement Organization and Methods . 1.4-10 1.4.4 Construction Management . . . . . . . . . . . . . 1.4-10 1.4.4.1 Division of Responsibility for Construction Manager . . . . . . . . . . . . . . . . . . . . 1.4-10
- 1. 4. 4 . 2 Construction Manager's Organization and Methods . . . . . . . . . . . . . . . . . . . . 1.4-14 1.4.4.3 Portland General Electric Company's Organiza-tion and Methods . . . . . . . . . . . . . . . 1.4-15 1.5 Requirements for Further Technical Information . . 1.5-1 1.5.1 Fuel Densification Rehearch and Development Program . . . . . . . . . . . . . . . . . . . . . 1.5-1 1.5.2 Reactor Vessel Model Flow Tests . . . . . . . . . 1.5-1 1.5.3 Mark C Fuel Assemblies (17 x 17 Array) . . . . . . 1.5-2 1.5.3.1 Fuel Assembly Flow Tests . . . . . . . . . . . . 1.5-2 1.5.3.2 Fuel Assembly Mechaniccl Tests . . . . . . . . . 1.5-4 1.5.3.3 Control Rod Tests . . . . . . . . . . . . . . . 1.5-5 1.5.3.4 Component Mechanical Tests . . . . . . . . . . . 1.5-6 1.5.3.5 Critical Heat Flux Tests . . . . . . . . . . . . 1.5-7 1.5.3.6 Manufacturing Feasibility Tests . . . . . . . . 1.5-8 1.5.3.7 Experimental Facilities . . . . . . . . . . . . 1.5-9
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1.5.3.8 Reflood Heat Transfer Coefficient Tests . . . . 1.5-10 O
Amendment 13 1-11 (January 1982)
j CHAPTER
1.0 INTRODUCTION
AND GENERAL DESCRIPTION 0F PLANT FIGURES Number Title a
Section 1.1 i 1.1-1 Regional Map I 1.1-2 Vicinity Map 1 1.1-3 Fluid System Symbols and Instruments, Piping, Valve and Duct Identification and Numbering Systems 1.1-4 Notes 'and Symbols for Electrical Drawings 1.1-5 Logic Legend Section 1.2 1.2-1 Pebble Springs Site Showing Anticipated Boundaries
- 1.2-2 Site and Yard Development 1.2-3 Equipment Location, Fuel Building, El. 740 ft 0 in.
I 1.2-4 Equipment Location, Fuel Building, El. 757 ft 4 in.
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, 1.2-5 Equipment Location, Fuel Building, El. 774 ft 8 in.
1.2-6 Equipment-Location, Fuel Building, El. 792 ft 0 in.
I.2-7 Equipment Location, Containment & Fuel Buildings, ES
, Section A-A 1.2-8 Equipment Location, Auxiliary Building, Section "B-B" i 1.2-9 Equipment Location, Fuel Building, Section D-D [j l' 1.2-10 Equipment Location, Containment & Auxiliary Buildings, [}
- El. 700 ft 0 in.
1.2-11 Equipment Location, Containment & Auxiliary Buildings, }} El. 718 ft 0 in. & 720 ft 0 in. 1.2-12 containment and Auxiliary Buildings, El. 740 f t 0 in. 72 C 1.2-13 Equipment Location, Containment, Auxiliary & Control 72 Buildings, El. 757 ft 4 in. & 760 ft 0 in. t3 l 1-v Amendment 11 (April 1978)
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CIIAPTER
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF PLANT FIGURES Number __ T_itle _ 2 1.2-14 Equipment Location, Containment, Auxiliary & Control Buildings, El. 774 ft 8 in. 7 1.2-15 Equipment Location, Containment & Auxiliary Buildings, El. 792 ft 0 in. 2 1.2-16 Equipment Location, Containment, Auxiliary & Control Buildings, El. 806 ft 0 in. @ 1.2-17 Equipment Location, Containment & Auxiliary Buildings, Section C-C 1.2-18 Equipment Location, Turbine Building, El. 717 ft 0 in. 1.2-19 Equipment Location, Turbine Building, El. 740 ft 0 in. 1.2-20 Equipment Location, Turbine Building, El. 764 ft 0 in. 1.2-21 Equipment Location, Turbine Building, El. 792 ft 0 in. 1.2-22 Equipment Location, Turbine Building, Section A-A 3 1.2-23 Equipment Location, Turbine Building, Section B-B 1.2-24 Equipment Location, Turbine Building, Section C-C 1.2-25 Equipment Location, Turbine Building, Section D-D 1.2.26 Equipment Location, Turbine Building, Section E-E 1.2-27 Equipment Location, Turbine Building, Section F-F 1.2.28 Equipment Location, Turbine Building, Section C-G 3 1.2-29 Equipment Location, Diesel Generator Building, s- El. 740 ft 0 in., 767 ft 0 in., & 792 ft 0 in. @ 1.2-30 Equipment Location, Diesel Generator Building, Section A-A & B-B 3 1.2-31 Piping & Mechanical Area 7, Auxiliary Building, v El. 720 ft 0 in. 2 O O Amendment 13 1-vi (January 1982) .
1.4 IDENTIFICATION OF AGENTS AND CONTRACTORS
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This section identifies the contractors and priiicipal consultants employed or to be employed in the design, construction and operation of the plant. The responsibilities of each contractor, consultant and service organiza-tion are stated and the division of responsibility between them and PGE is indicated. 1.4.1 ARCHITECT / ENGINEER-CONSTRUCTOR Eechtel Power Corporation (Bechtel), San Francisco, California, has been selected by PGE as the architect-engineer, and will provide the design and the construction management for the plant. The design and construction management of the reservoir and reservoir dams, the intake / pumping facility on the Columbia River, the intake pipelines to the plant site, i the roads and the railroad spur will be provided by Bechtel Hydro and Community Facilities Division, Inc., San Francisco, California. Bechtel will furnish preoperational testing and startup operation services, and will provide a quality rssurance/ quality control program consistent with the requirements of 10 CFR 50( ). Bechtel has been continuously engage'd in construction or engineering activities since 1898. For more than 20 years, Bechtel has been active j in the fields of petroleum, power generation and distribution, harbor l development, mining and metallurgy, and chemical and ind 7 trial processing. i For over 20 years, Bechtel has been engaged in the study, design and cor.struction of nuclear installations. Its experience includes design and construction of such facilities as accelerators, nuclear research labcratories, hot cells, experimental reactors and nuclear fuel processing plants, as well as nuclear power plants. 1.4.1.1 Division of Responsibility for Engineering and Design l Responsibility for engineering and designing the plant is divided between p Bechtel and PGE as shown below: O 1.4-1 [ l l
(1) Functions to be performed by Bechtel: (a) Site feasibility report (b) Engineering data collection and analysis (c) Engineering design and definitive design development (d) Implementation of quality assurance program (e) Preparation of construction and equipment bid document specifications and drawings (f) Preparation of contract cost estimates (g) Scheduling of construction (h) Assistance in PSAR and FSAR preparation (1) Other engineering support for PGE as requested. (2) Functions to be performed by PCE: (a) Site investigations (b) Obtaining of licenses and permits ( (c) PSAR and FSAR preparation and presentation (d) Quality asstrance supervision and auditing (e) Review of engineering design criteria and selected g engineering design documents C (f) Procurement of material and equipment for construction. O Amendment 13 1.4-2 (January 1982)
1.4.1.2 Engineer's Design Organization and Methods O h All of Bechtel's engineering and design are performed under the direction of the Project Manager. A Project Engineer is assigned to direct and supervise the detailed design with a design group consisting of architectural, civil, mechanical, electrical and layout supervisors, engineers and draftsmen. Bechtel's management regularly reviews this work in order to assure that it will conform to the highest professional standards. The Project Engineer coordinates all matters related to this work with PGE. Other Bechtel departments perform supporting services as required under the direction of the Bechtel Power Corporation management. The Estima-ting Department prepares the definitive estimate, monitors cost trends, periodically reviews costs and prepares cost forecasts. The Legal and Insurance Department reviews and approves construction and equipment specifications. The Purchasing Department conducts factory inspection and performs expediting functions, preparing related reports as required.
/N Quality assurance duties are performed as described in Section 17.1.
In performing its engineering and design functions, Bechtel first develops conceptual designs and layouts and discusses them with PGE. Upon approval in principle, Bechtel prepares technical specifications covering construc-tion contracts and equipment procurement contracts or purchase orders. i These specifications are then submitted for review in draft form to PGE. After thorough review, PCE sends written comments to Bechtel. Bechtel then either accepts the comments or indicates the reasons why they cannot. After further communication, specifications are prepared which meet the requirements of both Bechtel and PCE. Drawings are administered in a manner similar to that of engineering and design. PCE reviews the drawings prepared as part of the bid docu-ments and major revisions thereto in the following categories: (1) Architectural plans and elevations, layouts,
-g building services, materials, finishes and colors (b
1.4-3 Amendment 13 (January 1982) 4
(2) Civil plot plans and structures j (3) Mechanical - general arrangemeat of mechanical and O1l electrical equipment, piping and instrument diagrams (4) Electrical - elementary schematic diagrams, single line drawings. Bechtel is authorized to employ consultants in special fields when specifically approved by PGE. 1.4.1.3 Portland General Electric Company's Engineering Organization and Methods PGE is organized to handle its primary responsibility for the design of the plant. Figure 13.1-3 shows the PGE nuclear-related organization. The Vice President, Nuclear, is in charge of both engineering and con-struction activities. His division is divided into the following departments: Nuclear Prajects Engineering, Generation Licensing and p Analysis, Nuclear Projects Administration, Nuclear Projects Quality O Assurance, and Nuclear Projects Construction. Support in operational aspects of nuclear plant design is obtained from those groups reporting to the Vice President, Nuclear. Environmental support comes from those groups reporting to the Vice Chairman of the Board. Purchasing is under the direction of the General Manager, Purchasing and Materials Management, who reports to the Vice President, Operating Services. Additional technical support is obtained from consultants PCE retains to provide particular expertise not available within the company. Those retained to date and their areas of expertise are listed in Table 1.4-1. When technical specifications are received from Bechtel for review, they are assigned to an " action party" in whose field of primary responsibility the specification lies. Copies are also distributed to all interested l l Amendment 13 1.4-4 (January 1982)
parties in PGE for comment. The action party reviews and consolidates [ all comments and organizes meetings as necessary within PGE to consider the proposed revisions. When required, PGE comments are discussed with
.Bechtel by telephone or in joint meetings. Final PCE comments are incorporated in a letter requesting Bechtel to effect the changes.
Drawings are reviewed by the cognizant technical branch. Where PGE review comments follow the issuance of a drawing, design change notices are issued as required and followed by a contract change order. 1.4.2 NUCLEAR STEAM SYSTEM SUPPLIER Babcock & Wilcox Company (B&W) has been selected to supply the nuclear steam system. B&W was founded as a partnership in 1867 and organized as a corporation in 1881. B&W is a widely diversified company serving the electric utility, transportation, steel, petroleum, chemical, pulp and paper, and machinery industries, as well as one of the world's leading suppliers of specialty steels. B&W is one of the leading suppliers of components for the nuclear Navy. B&W employs over 36,000 persons in facilities throughout the United States, Australia, Canada, Great Britain and Sweden. Of this total, over 1000 are technical and scientific personnel working in nuclear power-related activities. As the world's largest manufacturer of steam generating equipment, B&W is a recognized leader in American industry. The company has contributed materially to the development of fundamental materials data, heat transfer data, manufacturing and erection processes and inspection techniques used in the steam generating equipment industry. B&W's participation in the development of nuclear power dates from the Manhattan Project. The company's broad activities include applied research to develop fundamental data, the design and manufacture of nuclear systems components, and design and manufacture of complete nuclear steam generating systems. 1.4-5 Amendment 13 (January 1982)
Major activities in the water-cooled and -moderated reactor field include the following: (1) Indian Point 1 (Consolidated Edison Company of New York, Inc) (2) NS Savannah (USAEC) (3) Advanced Test Reactor (USAEC) (4) Otto Hahn Power Plant Design (5) Oconee Nuclear Station Units 1, 2 and 3 (Duke Power Company) p (6) Three-Mile Island Nuclear Station Units 1 and 2 (CPU Nuclear) (7) Crystal River Station Unit 3 (Florida Power Corporation) (8) Arkansas Nuclear One, Unit 1 (Arkansas Power and Light Company) (9) Rancho Seco Unit 1 (Sacramento Municipal Utility District) 2 (10) Midland Plant Units 1 and 2 (Consumers Power Company) 2 O (11) Davis-Besse Nuclear Power Station Unit 1 (Toledo Power Company) 2 O (12) Bellefonte Nuclear Power Plant Units 1 and 2 (Tennessee Valley Authority) O Amendment 13 1.4-6 (January 1982)
(13) North Anna Power Station, Unit 3 (Virginia Electric and Power Com9any)
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(14) WPPSS Nuclear Project Nos. I and 4 (Washington d Public Power Supply System) 1.4.3 AWARD OF CONSTRUCTION AND PROCUREMENT CONTRACTS Construction will be accomplished by employing a number of prime contractors, each responsible for a separable portion of the work, such as civil-structural, mechanical, electrical, piping, instrumentation, etc. Major equipment, materials required in large quantities (such as structural steel), and long-lead-time items will be owner-furnished. Other materials and minor equipment will be furnished by the construction contractors. Concrete will be furnished to all contractors from a central mixing plant which will be operated by a supplier under direct contract with PGE. (% Upon preparation of a contract package consisting of a satisfactory set V; of technical specifications and applicable Irawings, Bechtel will send it to PGE with its recommended list of qualified bidders. PGE will then either approve or amend the list, referring any changes to Bechtel for comment. PGE will advertise for competi'.tve bid by the approved qualified bidders. Bidder technical questions will be referred directly to Bechtel, whereas commercial inquiries will be sent to PGE. For major construction contracts, prebid conferences will be held with PGE, Bechtel, and prospective bidders attending. The answers to bidders questions will be confirmed by addenda to the specifications, which then will be sent l to all prospective bidders. Bids will be received by PGE and will be referred to Bechtel for analysis
- and recoma.endation. After receipt of this analysis and recommendation, awards will be made by PGE to the lowest bidder fully meeting the terms of the specifications, or if there are minor deviations, the lowest bidder that meets the specifications with exceptions acceptable to k
1.4-7 Amendment 13 (January 1982) l I t
Bechtel and PGE. A principal contract objective will be to obtain firm lump sum or unit price bids. 1.4.3.1 Division of Res<onsibility for Contracting and Procurement The responsibility for contracting for construction and owner-furnished materials and equipment, and for nerforming quality assurance functions and expediting such equipment supply contracts will be divided between Bechtel and PGE as follows: (1) Functions to be performed by Bechtel: (a) Supplying to PGE a list of qualified bidders (b) Participating in prebidding conferences with prospective construction contractors and PGE, and preparing addenda to the specifications considered necessary as a result of the conferences (c) Analyzing bids received through PGE and making recommendations for award (M Conducting quality assurance surveillance of suppliers' manufacturing plants (e) Expediting the manufacture, shipment and delivery of materials and equipment when requested by PGE (2) Functions to be performed by PGE: (a) Advertising for bids among montractors or suppliers considered qualified by PGE and Bechtel (b) Conducting prebidding conferences O Amendment 13 1.4-8 (January 1982)
(c) 2eceiving bids O (d) Furnishing acceptable bids received to Bechtel for analysis and recommendation (e) Reviewing Bechtel's analysis and recommendation for award in the light of PCE's analysis (f) Negotiating with prospective contractors as required to reach agreement on contract terms (g) Making contract awards (h) Overseeing Bechtel's quality assurance and expediting functiocs on materials and equipment procurement , (i) Overseeing Bechtel's contract management
/N activities. \~.
1.4.3.2 Engineer's Contracting and Procurement Organization and Methods Bechtel's design organization, described in Section 1.4.1.2, has the primary responsibility for ensuring that the functions assigned to . Bechtel are performed satisfactorily. In such duties as preparing a list of qualified bidders and analyzing bids, the design organization is assisted by the Construction Department of Bechtel. Quality assurance and inspection functions at the various suppliers' factories will be performed by the Bechtel Quality Assurance organization and the Bechtel Procurement Department, respectively. Section 17.1 describes the quality assurance program in detail. Bechtel's Procurement Department will supervise materials and equipment delivery, expediting when necessary 1
.to maintain the constructicn schedule.
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a 4 1.4-9 Amendment 13 (January 1982) i
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1.4.3.3 Portland General Electric Company's Contracting and Procurement Organization and Methods The Purchasing and Material Mana8ement Department, under the Vice Presi-dent, Operating Services, is responsible for PCE's contracting and 2 procurement activities. The Purchasing Department acts on the advice and recommendation of the Nuclear Projects Engineering Department under the General Manager, Technical Functions. The Nuclear Projects Engineering Department has primary responsibility for procurement technical matters, whereas the Purchasing Department is principally responsible for commer-cial aspects. Both departments participate in all meetings and decisions. 1.4.4 CONSTRUCTION MANAGEMENT The Bechtel Power Corporation is assigned the responsibility for con-struction management of the majority of the constrw cion contracts and for providing overall jobsite coordination and supervision. 1.4.4.1 Division of Responsibility for Construction Manager O The responsibility for managing the construction of those parts of the plant on which Bechtel will be the construction manager is divided between Bechtel and PGE as follows: (1) Functions to be performed by Bechtel: (a) Assisting in developing the scope and composition of the construction contract packages. i
, s-(b) Developing the overall construction plan.
l (c) Conducting site tours for prospective bidders. O , Amendment 13 1.4-10 (January 1982) l
(d) Supervising the site operations of con-struction contractors. (e) Ensuring that construction contractors meet their contractual obligations by surveillance of their work. (f) Performing tests, surveys, measurements and inspections necessary to insure that the work performed meets contract speci-
'fications.
(g) Verifying contractors' payment estimates and certifying to PGE for payment. (h) Interpreting drawings and specifications for contractors and providing information not given in the contract documents, such as details of the portions of drawings marked " hold". (i) Evaluating changes, additions or deletions l in the scope of the contract resulting from design changes; evaluating the effect on cost and schedule of' unforeseen site and weather conditions, strikes, interface conflicts with other contractors, delays in receipt of owner-furnished equipment, and other problems beyond'the control of the contractor; receiving contractors' claims, estimating their cost, negotiating equitable settlements, and recommending to PGE appropriate contract modifications and just payments to be made. l 1.4-11 Amendment 13 (January 1982)
(j) Coordinating the efforts of contractors and allocating facilities and space equitably among them. (k) In cost-reimbursable portions of the work, establishing a detailed cost reporting system and exercising cost control procedures to ensure that the work is accomplished in an economical manner. (1) Maintaining a current project schedule and recommending to PGE measures to be taken to ensure that target dates are met. (m) Maintaining cost records and forecasts of phased future costs (cash flows) to keep PGE continually informed of the financial outlook of the project. (n) Monitoring schedules for fabrication and delivery of owner-furnished materials and equipment and, with PGE's concurrence, providing expediting assistance when con-struction progress is apt to be affected adversely by delays. (o) Keeping the startup organization informed of aspects of construction progress that might affect that program. (p) Maintaining construction records, including quality assurance and quality control documents. (q) Reviewing and approving construction contractors' quality assurance programs. O Amendment 13 1.4-12 (January 1982)
(r) Receiving and inspecting owner-furnished materials and equipment and issuing them to
- _ appropriate contractors.
(s) Overseeing project safety matters, operating a site first-aid station and ambulance service, and ensuring conformance with appli-cable health and safety regulations. (t) Implementing the field quality assurance pro-gram as outlined in the Bechtel Quality Assurance and Field Inspection Manuals. (2) Functions to be performed by PCE: (a) Providing overall supervision of Bechtel's construction management activities. (b) Maintaining liaison between PGE office I {"N) staff and Bechtel's field force. (c) Handling contacts with the public in the
! vicinity and conducting visitors through the site.
(d) Handling all aspects of the construction contracts prior to establishment of Bechtel's field office. 4 , (e) Supervising certain housekeeping and supporting activities such as road main-tenance, providing construction water for concrete curing, providing potable water, etc; providing expedient sewage disposal, i electric power, etc. O\ (__/ l.4-13 Amendment 13 (January 1982)
(f) Keeping PCE management informed of con-struction progress, planned near-term construction activities, anticipated con-struction problems, labor problems and other cignificant matters. 1.4.4.2 Construction Hanager's Organization and Methods Bechtel, ne Construction Manager, functions as PCE's agent in planning, organizing, scheduling and administering the construction program for the plant. All of Bechtel's wo'rk on the plant will be under the general supervision of PCE; PGE will exercise the right of approval at all stages of the construction process. Bechtel is charged with keeping PGE informed of its plans and actions and of construction progress and the activities of contractors. While Bechtel's organization for performing construction management funtions may vary somewhat with time, the organization described herein is anticipated to have the major part of the construction responsibility. The highest level in the Bechtel organization responsible solely for construction canagement is the Division Manager of Consruction. Reporting to him is the San Francisco Manager of Construction, who supervises the home office-based Construction Manager responsible for this project. The home office-based Construction Manager is respon-sible for: (1) Providing liaison between the Project Construction Manager (in charge at the site) and the Bechtel home office design group (2) Assisting in the development of contract packages, l especially from the viewpoint of construction feasibility, interfaces with other contracts and other aspects of construction management O Amendment 13 1.4-14 l (January 1 32) i t I
(3) Developing with PGE a broad construction plan N (4) Guiding and assisting the Project Construction Manager (5) Coordinating the use of Bechtel's corporate specialists and support personnel. The Bechtel field construction management organization will be headed by a Project Construction Manager. The field organization necessarily will be varied from time to time depending upon the work in progress. 1.4.4.3 P;rtland General Electric Company's Organization and Methods The Manager of Nuclear Projects Construction, located in Portland, will perform those construction management duties retained by PGE and will exercise general supervision over the Bechtel construction management p] ( organization. Operating through the Resident Engineer, the Nuclear Projects Construction Manager will continually monitor and evaluate
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Bechtel's site construction management program and its implementation. The Resident Engineer, whose office will be at the site, and the Nuclear Projects Administration Department in Portland will support the Manager
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of Nuclear Projects Construction in this activity. The Resident Engineer and his staff, consisting of the Assistant Resident Engineer, Mechanical and Electrical Engineers, Project Accountanc and the Plant Analyst, will provide the day-by-day interface at the construction site between Bechtel's construction management organization and PGE. The PGE quality assurance site representative will provide the day-by-day interface with the Bechtel quality assurance organization. The Nuclear Projects Administration Department will maintain the permanent l file on all contracts, correspondence and drawings. All supply and construction contracts and correspondence will be processed through the Nuclear Projects Administration Department, which is also responsible for l () providing accounting distribution, project budgeting and cash flows. 1.4-15 Amendment 13 (January 1982)
CHAPTER 13.0 CONDUCT OF OPERATIONS , , CONTENTS. Section _ Title _Page l 13.1- Organizational Structure of Applicant . . . ... 13.1-1 l 13.1.1 Corporate Organization'. . . . .. . .. . . . . ... 13.1-1
- -13.1.1.1 Corporate Functions, Responsibilities and Authorities . . . . . . . . . . . . . . . ... 13.1-1 i
13.1.1.2 PCE's Nuclear-Related Organization . . .. . .. -13.1-2 l{ 13.1.1.3 Interrelationships and Organizational v Interfaces . . . . . . . . . . . . . . . ... 13.1-3 13 .1 '.1' . 4' PGE Staff's Duties, Responsibilities,' 1_ Authorities and Qualifications . . . . . ... .13.1-4 13.1.2 Operating Organization . . . . . . . . . . ... 13.1-5 ! 13.1.2.1 Plant Organization . . . . . . . ~ . . . . ... 13.1-5 13.1.2.2 Personnel Functions, Responsibilities and Authorities . . . . . . . . . . . . . . ... 13.1-6 13.1.2.3 Shift Crew composition . . . . . . . . . ...
. 13.1-9 13.1.3 Qualification Requirements for Nuclear Facility Personnel . . . . . . . . . . . . ... 13.1-11 C
g 13.2 Training Program . . . . . . . . . .. . . . ... 13.2-1 v-l 13.2.1 Program Description . . . . . ... . . . .-. ... 13.2-1 13.2.1.1 Program Content . . . . . . . . . . . .. . . -. 13.2-2 13.2.1.2 . Coordination with Preoperational Tests and Fuel Loading . . . . . . . . . . . . . . ... .13.2-4a- 3
- 13.2.1.3 Practical Reactor Operation . . . . . . . .. .. 13.2-5 13.2.1.4 Reactor Simulation Training . . . . . . . ... 13.2 t 13.2.1.5 Previous Nuclear Training . . . . . - . . . ... 13.2-6 ^
! 13.2.1.6 Other Scheduled Training .. . . . . . . ... 13.2-6a O I 13.2.1.7 Training Programs for Nonlicensed Personnel . . 13.2-6a. @ 13.2.1.8 General Employee Training . . . . . . . . ... 13.2-8
- j. 13.2.1.9 Responsible. Individual . . . . . . . . . ... 13.2-8 4
13.2.2 Retraining Program . . . . . .. . . . . . ... . 13.2-9 l 13.2.3 Replacement Training . . . . . . . . . . . ... 13.2-10 ! e 13.2.4 Trainee Progress Evaluations . . . . . . . ... 13.2-13 3 13.3 Emergency Planning . . . . . . . . . . . . ... 13.3-1 13.3.1 Introduction . . . . . . . . . . . . . . . ... 13.3-1 '{ 13.3.1.1 Scope . . . . . . . . . . . . . . . . . . ... 13.3-2 ) 13.3.1.2 Definitions . . . . . . . . . . . . . . . ... 13.3-5 4 ! l
\
13-1 Amendment 13
.(January 1982)
- - . , , . . . . . - . ~ . - - - - - ,, . . - -- - -. , - -. ~ _ . . - . -
CilAPTER 13.0 CONDUCT OF OPERATIONS CONTENTS Section Title Page 13.3.2 Emergency Organization'. . . . . . . . . . . . 13.3-7 13.3.2.1 Emergency Operating Staff. . . . . . . . . . 13.3-7 13.3.2.2 Emergency Control Center Staff . . . . . . . 13.3-9 13.3.2.3 Company Control Center Staff . . . . . . . . 13.3-12 13.3.3 Emergency Communications Network . . . . . . . 13.3-15 13.3.4 Emergency Supplies . . . . . . . . . . . . . . 13.3-15 13.3.5 Notification of Offsite Groups . . . . . . . . 13.3-19 13.3.5.1 Direct PCE Notifications . . . . . . . . . . 13.3-19 13.3.5.2 Indirect Notifications . . . . . . . . . . . 13.3-23 13.3.6 Accident Classification. . . . . . . . . . . . 13.3-24 ,, 13.3.6.1 Local Onsite Emergencies . . . . . . . . . . 13.3-24b C3 13.3.6.2 General Onsite Emergencies . . . . . . . . . 13.3-24c 13.3.6.3 Offsite Emergencies. . . . . . . . . . . . . 13.3-25 13.3.7 Emergency Exposures. . . . . . . . . . . . . . 13.3-25 . 13.3.7.1 Plant Personnel. . . . . . . . . . . . . . . 13.3-26 13.3.7.2 General Public . . . . . . . . . . . . . . . 13.3-26 13.3.8 Review and Updating of Emergency Plan. . . . . 13.3-26 13.3.8.1 Review Frequency . . . . . . . . . . . . . . 13.3-26 13.3.8.2 Provision for Changes. . . . . . . . . . . . 13.3-26 13.3.9 Medical Support. . . . . . . . . . . . . . . . 13.3-27 13.3.9.1 Decontamination Facilities and Supplies. . . 13.3-27 13.3.9.2 First-Aid Facilities and Medical Supplies. . 13.3-27 13.3.9.3 Physicians . . . . . . . . . . . . . . . . . 13.3-27 13.3.9.4 Ambulance Service. . . . . . . . . . . . . . 13.3-27 13.3.9.5 Hospital Service . . . . . . . . . . . . . . 13.3-28 13.3.10 Drills . . . . . . . . . . . . . . . . . . . . 13.3-28 13.3.10.1 Frequency and Participants . . . . . . . . . 13.3-28 13.3.10.2 Evaluation and Recommendations . . . . . . . 13.3-28
- 13.3.11 Training . . . . .. . . . . . . . . . . . . . 13.3-29 13.3.12 Recovery and Reentry . . . . . . . . . . . . . 13.3-29 13.4 Review and Audit . . . . . . . . . . . . . . . 13.4-1 13.4.1 Review and Audit - Construction. . . . . . . . 13.4-1 13.4.2 Review and Audit - Test and Operation. . . . . 13.4-2 Amendment 3 13-11 (November 1974)
, . . . - - . . . - = .- - .. . . - . ~ . . - . - . . . - _ . - - . - . - . . . ~ - - . ~ . . - - . - - - _ . ~ . . _ . . . . - -
l l i
; CHAPTER 13.0 I ]*~ CONDUCT OF OPERATIONS !
t i TABLES l r i : Numbe_r _ _ Title .I - i h Section 13.1 : 3, , t i 13.1-1 Qualifications of Portland General Electric Company Technical i Staff Engaged on Nuclear Plants m i - 1 W
--Section 13.3 : ',
13.'3-1 Emergency Alert Conditions and Action Levels f i 4 ! 1 J
-1 i
l 1
- I g x I
- - 4-it ,
r pa s 4 1 t i j , ;
- i I
i , J 2-i i s 1 I ' i 13-iv Amendment 13 (January 1982) f a I
' -,___-..-.I.... _ . , . . - . _ _ _ . _ _ _ _ _ . . - _ , _ , . _ . . . _ . _ _ . - -
, I ~f ,
.s J ,e QIAPTER 13.0
/ ; CONDUCT OF OPERATIONS
,- FIGURES
,, Number _
Title _
';- .Se.ction 13.1
.13.1-1 Pebble Springs Project Organization
+
,, 13.1-2 PGE Corporate Organization
- s >
.13.1-3 PGE Nuclear-Related Organization b
v 13.1-4 Pebble Springs Plant Organization j <- s Sect.lon 13.2 13.2-1 Plant Staff Training Program Schedule O A a
/
f 6.
^
I:. Amendment 13 13-v (January 1982) s
13.1 -ORG/.NIZATIONAL STRUCTURE OF APPLICANT i
. (-
13.1.1 CORPORATE ORGANIZATION The structure and qualifications of Portland General Electric . Company's (POE) corporate organization and of .those Bechtel Corporation (Bechtel) organizations, which. will serve as the Engineer for the plant, are i described in this section. 4 L
- s Bechtel Power Corporation will provide the design and construction man-agement for the nuclear plant proper. Bechtel Hydro and Community
; Facilities Division, Inc. , _ will provide the design and construction
. management for the reservoir, reservoir _ dams, the intake / pumping facility on the Columbia River and :the intake pipelines to the plant site, roads and railroad spur. -s v
C i 13.1.1.1 Corporate Functions, Responsibilities and Authorities l l Bechtel is. responsible for tha design and construction management of the ! plant. However,. PGE has a controlling role in the design by making design decisions during the conceptual stage, participating with Bechtel in consideration of the definitive design determinations, and by thoroughly reviewing all technical specifications and design layouts. PGE also ;
~
participates in.the design change process after construction contracts are awarded. I The design of equipment, including the Nuclear Steam System (NSS), will f be accomplished by the supplier, in accordance with the requirements and design criteria stated in the specifications on which they bid. B&W is furnishing the NSS, as well as some odme 7uipment, and supplying Bechtel and PGE with pertinent interface data and information. Bechtel, in turn, will furnish B&W with design information~ supplementing the speci-fications as necessary to ensure that the B&W equipment will be compatible with the structural, seismic, mechanical, electrical, instrumentation - ; and insulation design of the plant. B&W will participate in all meetings, 1 - 13 1-1 Amendment 13 (January 1982) I l-.~.. . - . ~ . . - -. . . . - . - . - - . . - . . - . . - . . . -- - ----. - . - . - . - .
telephone conference calls and hearings when the subject relates to the interface between the NSS or fuel and the balance of plant. The construction of the plant will be accomplished as described in Sect ion 1.4.4. PCE will do none of the construction with its own crews, except installing temporary power lines and lighting both for the site and access road. Construction contractors will be responsible for meet-ing their own power and lighting requirements, using power sources dis-tributed to a few convenient central points. Bechtel will serve as Construction Manager on all major elements of the work but will not perform auy of the actual construction. PGE will act as Construction Manager for the preliminary construction contracts and for the specialized construction as discussed in Section 1.4.4. Quality assurance duties will be performed as described in detail in Chapter 17. PGE is responsible for ensuring that quality assurance and quality control duties are executed properly by Bechtel and by all material and equipment suppliers and construction contractors. PGE's Manager of Nuclear Projecte Quality Assurance reports directly to the d Vice President, Nuclear. Testing is the responsibility of Bechtel under the supervision of PGE. Bechtel's construction management organization includes specialists in each required area of testing. Bechtel will use its San Francisco laboratories during the design stage and as required during the construc-tion period to check the work of other testing agencies. PGE will exercise general surveillance over all testing operations and will perform tests in its own laboratory when required for checking purposes or in order to obtain an independent report. 13.1.1.2 PGE's Nuclear-Related Organization g The Pebble Springs project organization is shown in Figure 13.1-1. PGE's U corporate organization is shown in Figure 13.1-2. PGE's nuclear-related organization for handling matters pertaining to the plant. including design, construction, quality assurance, operation and testing, is shown Amendment 13 13.1-2 (January 1982)
in Figure 13.1-3. Organization for quality assurance is discussed in
,O Chapter 17. The responsibilities, authority, qualifications and experi-ence of PGE personnel who have a major role in the design and construction of the plant are shown in Table 13.1-1. O S
,13.1.1.3 Interrelationships and Organizational Interfaces The working interrelationships and organizational interf aces among PGE, Bechtel, B&W, and other suppliers and contractors are discussed in Section 1.4.
Bechtel is the center of the design effort, incorporating in technical specifications all the requirements and design criteria for plant equipment. Bechtel's final designs frequently will be based upon the specific dimensions, performance characteristics and other data furnished by equipment manufacturers, who will generally be selected after competitive bidding among prequalified suppliers. The Bechtel supplier-design contract will continue until equipment meeting gFa specifications is received in satisf actory condition at the site. PGE L will participate in important discussions and will review Bechtel's design work. Approval is the responsibility of PGE. B&W, as the supplier of the NSS, has a major role in the coordination of the design effort. For the construction stage, Bechtel, as Construction Manager, will act l l as coordinator among the construction contractors who must; work simul-taneously in a very limited space, of ten using the same items of equip-l ment. At times multishif t and overtime work may be necessary in order j to resolve a conflict of work space requirements or equipment use. PGE l reserves approval rights on all such solutions because of the substant.ial f added costs. Weeklfschedulingmeetingswillbeheldatthesite. Equipment suppliers, such as B&W and oth.tra who have technicians at the site, will be represented when considered desirable. l I O V 13.1-3 Amendment 13 (January 1982)
13.1.1.4 PCE Staff's Duties, Responsibilities, Authorities and Qualifications PCE's " Engineer in Charge" of the design, construction and performance of quality assurance functions for the plant is the Vice President, Nuclear. He reports directly to the President of PGE. His leading assistant in this work will be the General Manager of Technical Functions, who has overall responsibility for activities related to plant scope, regulatory requirements, procedures, design schedule, licensing schedule, construction schedule, cost, major scope changes, contacts with partners, and general contacts with Bechtel, B&W and other supply and construction contractors. The Pebble Springs Project Manager and his staff, under the General Manager of Technical Functions, will direct and coordinate these { activities among the various PCE support groups and external organiza-tional entitie s. The Manager of Nuclear Projects Quality Assurance, who has responsibility for all quality assurance activities related to design and construction, reports directly to the Vice President, Nuclear. Table 13.1-1 gives the duties, responsibilities, authority, qualifica-tions, educational background and technical experience of key and tech-nical PGE personnel in the home office and at the site having a role in the design, c:onstruction and quality assurance surveillance of the project. This duties portion of this table provides specific information as to how duties are assigned within the headquarters staff. Outside consultants will be employed to handle highly specialized work. All l outside consultants will report to the PGE departn;ent responsible for their particular type of work, except that the geological and geophysical bl consulting boards report directly to the Vice President, Nuclear. Th.e responsibility for staff recruiting and training is centralized
}l 'chrough the Human Resources Department. However, identification of
{ specific recruiting needs or training programs is the responsibility of individual supervisors. When the plant becomes operational, the two divisions most directly
}
engaged in activities supporting the operation are Power Operations and O l Amendment 13 13.1-4 l (January 1982) l l l
Nuclear. The Nuclear Division has the responsibility for construction, [] operations and engineering support of operation in the areas of design, licensing, technical support and quality assurance activities The Power y Operations Division, shown in Figure 13.1-2, provides scheduling of plant generation and planned outages and fuel management services. Table 13.1-1 lists the current duties, responsibilities, authority, qualifications, education and technical experience of key incumbent PGE personnel in the staff and illustrates the areas they are responsible for during design and construction and will be responsible for during opera-tion. This staff will be increased, depending upon needs and require- { ments, as the project proceeds during the design and construction phase. Such additions also include considerations of operational needs following [ completion of construction. 13.1.2 OPERATING ORGANIZATION The plant will consist of two identical nuclear generating units. The plant organization, applicable when both units are operational, is shown / in Figure 13.1-4. The staff will consist of three main groups: Opera- n t'k]M tions & Maintenance, Technical Services and Plant Services. An initial y staff size of approximately 200 will be maintained during single-unit operation; additional personnel will be added as required while the second unit is being placed in service. Plant staff positions will be filled in accordance with the schedule criteria of ANSI /ANS 3.1-1981,
" Selection, Qualification and Training of Personnel for Nuclear Power Plants".
13.1.2.1 Plant Organization The plant staff will be under the general administration of the General Manager and is divided by function into three major departments: Opera-tions & Maintenance, Technical Services and Plant Services. O The Operations & Maintenance Department, directed by the Manager, Opera-tions & Maintenance, is responsible for the general operation and mainte-A nance of plant systems and equipment on a continuous basis and for ! i Q) 13.1-5 Amendment 13 (January 1982)
maintaining plant cleanliness. This department is comprised of Opera-tions, Maintenance, and Planning & Scheduling. The Operations & Mainte-nance Manager also coordinates the onsite industrial safety program with the Safety Coordinator. The Technical Services Department, directed by the Manager, Technical Services, provides technical support to plant Operations & Maintenance. This department is comprised of the following groups: Engineering, Chemistry, Radiation Protection and Training. The Plant Services Department is directed by the Manager, Plant Services. This department includes the Office, Quality Assurance, and Material Control activities. The Manager, Plant Services also coordinates the onsite security activities with the Security Supervisor. 13.1.2.2 Personnel Functions, Responsibilities and Authorities The General Manager is responsible for the management of the plant. In the performance of his duties he administers and enforces applicable policies, regulations, codes and practices to ensure safety, security, efficiency and a continuity of service in the operation and maintenance of the plant. The Manager, Operations & Maintenance is responsible to the General Manager for the continued operation and proper maintenance of the plant and for coordinating the activities of the Operations & Maintenance groups to provide the most efficiency in the efforts of these two activities. { The Operations & Maintenance Manager also coordinates the onsite indus-trial safety program with the Safety Coordinator assigned to the plant. The Operations Supervisor is responsible for the operation and functional maintenance of the plant. He directs the Assistant Operations Supervisor, Shif t Supervisors, Assistant Shif t Supervisors and the necessary Control Operators, Assistant Control Operators, Auxiliary Operators and Helper Operators to man the operating shifts. Refueling operations are performed O Amendment 13 13.1-6 (January 1982)
by the Operations Group and assistance from Plant Maintenance and Plant Engineering. V The Maintenance Supervisor directs the Control & Electrical Supervisor, Mechanical Supervisor, Instrument & Control Supervisor, Mechanical Foremen, Electrical Foreman, Maintenance Engineers, Maintenance Planner / Scheduler, and the necessary technicians, mechanics, welders and helpers to maintain and repair the mechanical, electrical, and instrument and control equipment in the plant, and also maintain plant cleanliness. The plant Planner / Scheduler supervises plant outages and coordinates and schedules plant evaluations during normal operation. The Manager, Technical Services is responsible to the General Manager for engineering assistance, chemistry control, radiation protection, training and computer programming. He is responsible for providing the necessary i I 1 technical support for safe and efficient operation and maintenance of the plant. m i O" The Engineering Supervisor directs the Reactor Engineer, Reliability Engineer, Supervising Engineers and Nuclear Plant Engineers. This group provides procedures for plant testing, engineering for plant modifica-l tions, safety analyses for procedural changes and plant modifications, and reactivity monitoring services and verifies performance of the plant and related equipment. The Chemistry Supervisor directs the Plant Chemists, the Effluent Analyst, and the Chemistry & Radiation Protection Technicians. The Chemistry Group performs chemical analyses on various plant systems to verify operating parameters, and ensures that operating specifications are met and that plant chemistry is monitored to maximize the operating life and reliability of plant systems and equipment. The Chemistry Group also monitors and reports liquid and gaseous radioactive effluents. The Radiation Protection Supervisor is responsible for development and r implementation of the radiological control program and radiological
; J 13.1-7 Amendment 13 (January 1982)
emergency plans. The Radiation Protection staff includes the Radiation Protection Supervisor, Assistant Radiation Protection Supervisor, Radiation Protection Engineers, Radiation Protection Specialist, Records Coordinator, Radioactive Waste Supervisor, Utility Workers, and Chemistry & Radiation Protection Technicians. The Training Supervisor administers the plant training programs in accor-dance with plant and NRC policies and license regulations to ensure that plant personnel are properly trained to perform their assigned responsi-bilities and tasks. The Training Group is comprised of the Training Supervisor and Training Assistants. The Manager, Plaat Services is responsible for: (a) coordinating the various administrative functions at the plant, (b) ensuring that the corporate quality assurance programs and requirements of those programs are fully implemented at the plant and (c) ensuring a suitable inventory of spare parts and supplies is maintained as necessary for operation and b maintenance. The Manager, Plant Services also coordinates the onsite security activities with the Security Supervisor. The Office Supervisor provides the clerical support for the plant including the establishment and maintenance of records, typing, filing and correspondence related to plant activities. This position also provides the administration and direction of the custodian contracts. The Administrative Group is comprised of the Office Supervisor and the plant secretarial and clerical personnel. The Office Analyst provides plant budgeting, accounting, bookkeeping, l and time-keeping controls as well as management of cesh funds and job cost control. l l The Quality Assurance Supervisor oversees the administration and imple-mentation of the plant operating quality assurance program. The Quality Assurance Group is comprised of the Quality Assurance Supervisor and a staff of QA/QC Inspectors. O Amendment 13 13.1- 8 (January 1982)
m The Material Control Supervisor is responsible for operating and main-( k / t taining the plant warehouse and ensuring that the required materials are available to support plant operations. The Material Control Group handles all procurement actions for the plant and verifies that all materials received are in accordance with established requirements. The staff includes the Material Control Supervisor, Material Coordinator, Head Storekeeper, and the necessary clerks and warehousemen. In the absence of the Plant General Manager, the overall plant responsi- ;; bility will be designated in writing by the Plant General Manager and IS will be the Manager, Operations & Maintenance or, in his absence, the Manager, Technical Services. Each Shif t Supervisor is responsible for the actual operation of his unit during his assigned shif t and as the designated senior Shif t Supervisor, has the additional responsibility for site facilities common to both units. In the event either is incapacitated, the applicable Control Operator will assume the unit responsibilities and the remaining Shift Supervisor will assume the responsibility for the site until a qualified (~'}
\'
replacement is available. 13.1.2.3 Shift Crew Composition The twelve-man operating crew, which includes six licensed operators, provides adequate supervision of plant operation during normal operation and suf ficient manning to handle emergency situations. Each unit is designed for one-man operation from the control room. Systems essential to power generation that require continuous operation are automatic and have control room operator supervision. Process systems, such as the ;; radioactive waste systems, Boron ' Recovery System, and makeup water d ' treatment systems, operate automatically following manual starting. Such systems have alarms to indicate malfunctions and remote shutdown capability to minimize operator action beyond initial lineup and super-vision of operation. All of these control system design features, which have been proven in application in large central station steam plants, i 4
\ ,/
- 13.1-9 Amendment 13 (January 1982) l
demonstrate the adequacy of the above-described operating crew for normal ope ra tion. Each unit has all required safety system indication and control in its control room so that in an emergency situation, it can be shue dowe safely from the control room by any one of the licensed operators on shift. Additienally, automatic and redundant reactor protection and Engineered Safety Features (ESP) assure a safe plant shutdown without operator action under the most severe accident conditions. Shif t operating personnel are trained and qualified to implement radia-tion protection procedures including routine and special radiation surveys using portable radiation detectors, use of protective barriers and signs, use of protective clothing and breathing apparatus, performance of con-tamination surveys, checks on radiation monitors and limits of exposure g rates and accumulated dose. O The initial operations personnel have extensive training and/or experi-ence in nuclear and conventional power operation, and will be engaged in a continual retraining program to assure the continued safe and effi - l cient operation of the plant. Before assuming their responsibilities, replacement personnel will have the same capaailities as the initial operations personnel. i During periods of increased activity such as refueling, startup after a shutdown of long duration, or special testing, additional operating per-sonnel will be assigned to augment the normal crew as required. Simi-larly, the shif t organization during initial testing and startup is basically the same as the normal operating organization but is supple-mented as necessary during this period. B&W and Bechtel will be available to provide technical direction and assistance during this period. O Amendment 13 13.1-10 (January 1982)
._ _ _ , . .. - __ ~ .- . .. . . . . - . - . - . . . .- -. .. - - -
k 13.1.3 QUALIFICATION REQUIREMENTS FOR NUCLEAR FACILITY PERSONNEL 4 The qualification requirements used for personnel assignments on Pebble i Sprin3s will meet or exceed the minimum requirements set forth in 4' i ANSI /ANS 3.1-1981, " Selection, Qualification and Training of Personnel O ! u } for Nuclear Power Plants". " i h i 1 i i i
- l 1
i. 1 1 4 1 i i i l 13.1-11 Amendment 13 , (January 1982) . l 4
, _ _ , . . . . . . _ . . . . . . , . _ . _ . . . - -. _ -.....-,...-.,,, _m... .
m., , _ , , , . - , , . . . . . , , , . . , . _ , , , ,, ,,-,... . , ,,,, ._ - -
t TABLE 13.1-1 Sheet 1 of 41 QUALIFICATIONS OF PORTLAND GENERAL ELECTRIC COMPANY
; .TECilNICAL STAFF ENGAGED ON NUCLEAR PLANTS NAME: Richard.L. Ackerman TITLE AND DUTIES: Supervisor, Contract Services - responsible for providing an early and controlled review of contracts to ensure
.that appropriate attention is given to legal and commercial
. aspects and that contracts are prepared in a uniformly auditable manner to minimize corporate liability and risk. Also respon-sible for providing data on scheduling and cash flows for major
- generating projects.
4 EDUCATION: U of Idaho - BS in CE-(1957); JD - Northwestern School of Law, Lewis & Clark College (1974). ,s
. Q
_ PROFESSIONAL. AFFILIATIONS: PE - Oregon; Member, Oregon State Bar and ' ' - , American Bar Association. EXPERIENCE: U.S. Army Biological Lab - Research on physical defenses against biological warfare (1958-1960);-Chicago Bridge & Iron
. Company - Engineer, Field Engineer, Shop Foreman and Shop Superin-tendent for plate metal designer,. fabricator and erector (1957-1969);.
- g' .
Various engineering and shop positions while attending night-law
- school (1970-1973); Swan Wooster Engineering, Inc. - Project Engi-neer in charge of structural design for waste wood boiler (1973-1975);
i Portland General Electric Company - Schedule Analyst to Supervisor,- Cost / Schedule Section (1975-1977); Supervisor, Cost and Schedule (1977-1981); Supervisor, Contract Services (1981- ).. i h i 1 i i i i 1 i e Amendment 13 (January 1982) ! 5. _-- ~ _ . - , . _ _ . _ - , , _ . _ - , _ . . . _ _ _ _ - _ _ . _ _ _ _ _ _ _ _ _ . . . . _ _ . _ ._. _ . .~ - - -- J
TABLE 13.1-1 Sheet 2 of 41 O NAME: Jacob K. Aldersebaes TITLE AND DUTIES: Manager, Nuclear Maintenance and Construction - R responsible for all construction and major maintenance and C modification work at the Trojan Nuclear Plant and for storage and maintenance of the Pebble Springs equipment. EDUCATION: llogere Technische School, Amsterdam - BS in CE (1949); NF - OSU (1968); INP - NUS (1969); Non-destructive Testing Course - Convair (1970). PROFESSIONAL AFFILIATIONS: PE - Oregon; ASCE. EXPERIENCE: Municipal Water System, Amsterdam (1951-1952); Arabian-American Oil Company - construction and project engineer on civil works in Saudi Arabia (1952-1958); Portland General Electric Company - design and construction supervision of civil-structural projects (1959-1970); Trojan assistant resident engineer (1970-1972); Trojan 2 resident engineer (1972-1981); Manager, Nuclear Maintenance and O Construction (1981- ). [ l l l l l l O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 3 of 41 O NAME: Mark A. Bell TITLE AND DUTIES: Supervising Engineer, Mechanical Engineering Branch - responsible for mechanical system design of nuclear projects. EDUCATION: BYU - BS in ChE (1072). , C PROFESSIONAL AFFILIATIONS: PE - Oregon; AIChE; ANS. EXPERIENCE: Bettis Atomic Power Laboratory qualified engineering officer of the watch and shift supervisor (1972-1975); Portland General Electric Company - Chemist, Trojan Nuclear Plant (1975-1978); Plant Engineer, Trojan Nuclear Plant (1978-1979); Chemistry Supervisor, Trojan Nuclear Plant (1979-1981); Supervising Engineer, Mechanical Engineering Branch (1981- ). O O Amendment 13 (January 1982)
1 TABLE 13.1-1 Sheet 4 of 41 O NAME: Glen E. Bredemeier R C TITLE AND DUTIES: Vice President, Power Operations - responsible for system load dispatching, scheduling of power plant operations and power interchanges, power planning activities, pooling and O coordination, power contracts, fuels planning and procurement,
- hydrogeneration and communication.
EDUCATION: OSU - BS in EE (1942); Stanford U - Executive Development Program (1966); NF - OSU (1968); INP - NUS (1969). PROFESSIONAL AFFILIATIONS: PE - Oregon; PE0; NSPE; IEEE - Senior Member; NELPA; ECO. EXPERIENCE: General Electric Company - test engineer (1942-1943); U. S. Navy - radar countermeasures officer (1943-1946); Portland General Electric Company - electrical draftsman (1946-1948); assistant to superintendent of production (1948-1953); Intercompany pool engineer - Power Pool Operations (1953-1961); Manager, Power g Operations Department (1961-19'iS); Vice President, Power Operations O (1975- ). O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 5 of 41 e u NAME: Donald J. Broehl TITLE AND DUTIES: Assistant Vice President, Nuclear - responsible for ;; providing general management direction for operation, maintenance, 13 and construction / modification work of the Trojan Nuclear Plant. EDUCATION: OSU - BS in EE (1948); GE Power System Engineering Course (1956); GE Nuclear Training Course (1967); GGA Comprehensive Nuclear Course (1967); W Reactor Operators Training Program (1969); , NF - OSU (1968); INP - NUS (1969); Stanford U (1944); N Carolina State (1945). PROFESSIONAL AFFILIATIONS: PE - Oregon; IEEE - Senior Member; ANS; ICOLD; PE0; NSPE; AAAS; Member of ANS Power Division Executive ,, Committee; AIF. g EXPERIENCE: Portland General Electric Company - Underground Department, i electrical engineer (1948-1949); Valuation Department, electrical engineer (1949-1951); Engineering Department, electrical design engineer (1951-1952); Engineering Department, system planning engineer (1952-1957); resident engineer, Pelton and Round Butte Hydroelectric Projects (1957-1966); Construction Coordinating Department, engineer (1966-1968); Nuclear projects coordinator l ,,,) (1968-1970); Nuclear projects engineer (1970-1972); Manager of ! -/ nuclear projects (1972-1977); Assistant Vice President, Generation ;; Engineering-Construction (1977-1980); Assistant Vice President, 13 Nuclear (1980- ). l i l i Amendment 13 (January 1982)
TABLE 13.1-1 Shee.t 6 of 41 O NAME: Theodore E. Bushnell TITLE AND DUTIES: Manager, Civil Engineering Branch - responsible for g the direction and coordination of civil and structural engi-v neering aspects of nuclear projects. EDUCATION: California State U - BS in CE (1963); Stanford U - MS in CE (1964). Rn 31 PROFESSIONAL AFFILITATIONS: ASCE; PE, SE - California, Oregon. EXPERIENCE: Aetron-Blume-Atkinson structural engineer on design of O structures and facilities during construction of the Stanford v Linear Accelerator Center (1964-1965); John A. Blume & Assoc. - structural engineer on design, review and safety evaluation of numerous projects, including nuclear power plants and nuclear waste handling systems (1965-1974); Portland General Electric Company (1974- ). l l O i l O Amendment.13 (January 1982) l
TABLE 13.1-1 Sheet 7 of 41 O . NAME: Leif W. Erickson , TITLE AND DUTIES: Manager, Systems and Analysis Branch - responsible for managing engineering systems analyses for thermal power plants.
- EDUCATION
- OSU - BS in ChE (1969); U. S. Navy Nuclear Power Training j (1969-1970); OSU Reactor Operator Training (1975). 02 3
PROFESSIONAL AFFILIATIONS: PE - Oregon; ANS; AIChE. EXPERIENCE: U. S. Navy - division officer aboard nuclear powered submarine (!'59-1974); Portland General Electric Company - Nuclear Plant Engineer, Trojan Nuclear Plant (1974-1976); Mechanical Engineer, Generation Engineering (1976-1978); Project Licensing Supervisor, Generation Licensing and Analysis (1978-1980); Manager, Systems and Analysis Branch (1980- ). O f l j i O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 8 of 41 O NAME: Frank C. Gaidos TITLE AND DUTIES: Manager, Nuclear Projects Quality Assurance - responsi-ble for the Quality Assurance program. 7.wplementation of QA acti-vities for nuclear projects including those of the consultants, engineers, vendors and contractors. AE, NSSS, vendor and construc-tion site audits. EDUCATION: PITT - BS in Physics (1957); W Nuclear Components School (1962); NDE Qualification to NAVSHIPS 250-1500; PT, MT, UT and RT (1962); NDE Qualification to ASME Code PT, MT, and RT (1968). PROFESSIONAL AFFILIATIONS: ANS; EEI QA Task Force. EXPERIENCE: Westinghouse Electric Corporation - nuclear technician with Naval Reactors Program at Bettis Atomic Power Division (1946-1957); scientist in experimental radiation physics programs (1957-1960); National Reactor Testing Station, Idaho - test engineer (1960-1962); Bettis Atumic Power Laboratory - QC, rising to senior engineer, responsible for QC in Navy nuclear submarine and surface ship projects (1962-1967); Westinghouse Electric Corporation - senior supervisory service engineer responsible for QA in commercial nuclear power (1967-1969); Portland General Electric Company - (1969- ). O l Amendment 13 O (January 1982) l
1 TABLE 13.1-1 Sheet 9 of 41 NAME: Mike R. Candert TITLE AND DUTIES: Supervising Engineer, Civil Engineering Branch - responsible for the design review of civil and structural engi-neering for nuclear projects. EDUCATION: Tri-State College - BS in CE (1963). PROFESSIONAL' AFFILIATIONS: PE - Ohio, Oregon. ; EXPERIENCE: Ohio Department of Transportation.- Assistant Project Engineer and Project Engineer responsible for administration of contract and supervision of inspection of urban interstate i highway construction (1963-1972); Kaiser Engineers - Quality j Assurance Civil Engineer, responsible for QA on all civil / structural items during construction of Zimmer Nuclear Power Plant (1972-1975); Portland General Electric Company -- l (1975 - ). 1 i i i f F C Amendment 13 , (January 1982) r M _,,...__,,.-m _...,_,_,__._,,,.,,....,.y___ _ , , . _ _ ,, . , , _ . , _ _ , _ , , . _ . . - - . . . _ . _ _ . - , _ . . , _ , _ _ _ _ , _
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TABLE 13.1-1 Sheet 10 of 41 O NAME: Scott G. Gillespie TITLE AND DUTIES: Senior Nuclear Engineer, Radiological Engineering Branch performs design review, analysis, assessment and tech-nical documentation of radiological aspects of nuclear power plants for regulatory, design, and operational purposas. 2 EDUCATION: U of California, Berkeley - BS in Engineering Physics (1970); O MS in NE (1972); MS in ME (1978). PROFESSIONAL AFFILIATIONS: ANS; PE - California, Oregon. EXPERIENCE: Pacific Gas & Electric Company (1972-1978) - Engineer in Special Projects Group of Department of Mechanicci and Nuclear Engineering. Licensing and radiation analysis of nuclear power plants; Portland General Electric Company (1978- ). O 1 l O Amendment 13 l (January 1982) l l
TABLE 13.1-1 Sheet 11 of 41
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NAME: R. E. Gillmor TITLE AND DUTIES: Superintendent Communications - supervise PGE-owned communications system design, replacement and maintenance; including telephone, radio, microwave, carrier and telemetering. Supervise leasing communication facilities from common carriers. Supervise PGE Company applications to federal Communications Commission. Supervised design of Trojan communication facilities and will supervise their maintenance and replacement. EDUCATION: OSU - BS in EE (1942); NF - OSU (1968).. PROFESSIONAL AFFILIATIONS: PE - Oregon, Arizona; IEEE; Eti. f EXPERIENCE: General Electric Company - installation and testing power ] control systems and equipment. Installation and testing com-munication systems and equipment; Portland General Electric Company electrical engineer, design power substation and controls. Coordinate installation and maintenance in generating stations and substations (1948-1955); superintendent communications (1955- ). i
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l t i i l I l 1 ( i i Amendment 13 (January 1982).
TABLE 13.1-1 Sheet 12 of 41 O NAME: D. L. Glivinski TITLE AND DUTIES: Quality Assurance Engineer, Nuclear Projects C
- Quality Assurance - responsible for QA auditing of off site and PGE support groups; auditing and inspection of procurecent activities by plant operating organization; QA functions related to onsite new fuel handling and spent fuel disposal activities; QA duties in connection with NRC audits of operating practices.
EDUCATION: Benson Polytechnic School (1956-1960). l PROFESSIONAL AFFILIATIONS: PE - California; NSPE; ASQC. EXPERIENCE: Stevenson Engineering & Mfg. - designer and draftsman (1965-1966); Stevens Thompson & Runyan, Inc. electrical design (1966-1968); Skidmore, Owings & Merrill electrical designer and inspector (1968-1971); Portland General Electric Company - (1971- ). O f 9 Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 13 of 41 ? NAME: Stephen E. Hoag TITLE AND DUTIES: Manager, Onsite Engineering Branch - responsible for ensuring that onsite design activities at the Trojan Nuclear ;; Plant are performed in compliance with applicable federal-state 53 agency regulations in a cost-effective and technologically sound manner. EDUCATION: OSU - BS in CE (1962); U. S. Naval Nuclear Power School, Vallejo, Cailfornia (1962); U. S. Naval Nuclear Prototype Training AlW, Idaho Falls, Idaho (1963). PROFESSIONAL AFFILIATIONS: PE - California; ANS. I EXPERIENCE: U. S. Navy - engineering officer of the watch, responsible for operation and maintenance of naval nuclear propulsion plants on USS Sam Houston (SSBN 609) and USS James Madison (SSBN 627). Qualified to be an engineering department head of a nuclear-powered ship (1964-1969); Crown Zellerbach Corporation - senior engineer, responsible for design changes related to maintenance and operation of chip and pulp mills (1969-1971); Portland General Electric
- Company - (1971- ).
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TABLE 13.1-1 Sheet 14 of 41 O NAME: Francisco K. Irlandez TITLE AND DUTIES: Engineer, Civil Engineering Branch - responsible for review of civil engineering and architectural designs and specifications in nuclear power plants. Included in the review responsibilities are site preparation, foundation engineering, cooling water earth retaining structures, and other civil-structural designs. EDUCATION: MAPUA Institute of Technology, Philippines - BS in Archi-tecture (1956); Carnegie-Mellon U - BS in CE (1959). PROFESSIONAL AFFILIATIONS: PE - Pennsylvania, Ohio; SE - Illinois; - ASCE. C EXPERIENCE: E. D'Appolonia Consulting Engineers - Civil engineer, responsible for geotechnical engineering design, including foundation of steel mills, petro-chemical refineries, dams and reservoirs, and other civil engineering works (1957-1958) (1959-1965); The Boeing Company - Stress analyst, responsible for stress analysis of aircraft structures (1965-1971); Gilbert-Commonwealth Associates - Senior engineer, responsible for design and review of utility-related facilities on both nuclear and fossil fuel power plants. Specific responsibilities include siting studies, design and specification preparation of dnma for cooling reservoirs, plant foundation design, and site preparation (1971-1975); Portland General Electric Company - (1975- ). O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 15 of 41 O NAME: Jack W. Lentsch TITLE AND DUTIES: Manager, Generation Licensing and Analysis Depart-ment - responsible for managing and supervising all licensing, '
'2 analysis and radiological engineering activities associated with E$
thermal power projects. EDUCATION: OSU - BS in Chemistry (1965); OSU - MS in Radiological Physics (1966); AEC Health Physics summer program at National Reactor Test- -s . ing Station (1967); New York U - PhD in NE (1975). c. d3 PROFESSIONAL AFFILIATIONS: ANS; HPS; IRPA; Phi Lambda Upsilon; t CHP (American Board of Health Physics). EXPERIENCE: New York U Medical Center, Institute of Environmental -s Medicine assistant research scientist on environmental impact of lC nuclear plants and inplant radiation and process control (1969-1971); Bechtel Corporation - nuclear engineer. Analysis of environmental impact of nuclear plants, including dose assessments for normal and accidental releases of radioactivity (1971-1972); Portland General Electric Company nuclear engineer (1972-1976); Supervisor, Radio- -s logical Engineering (1976-1979); Manager, Generation Licensing & {}
! Analysis (1979 - ). '
i i 4 i 1 k 4 Amendment 13 (January 1982) ] 3
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TABLE 13.1-1 Sheet 16 of 41 l O NAME: William J. Lindblad TITLE AND DUTIES: President - overall responsibility for management of engineering and construction, nuclear and thermal plant operations, corporate planning, and internal audit functions. m O EDUCATION: U of California, Berkeley - BS in EE (1951). PROFESSIONAL AFFILIATIONS: ASME; IEEE. EXPERIENCE: Pacific Gas & Electric Company - increasing levels of res-ponsibility for design and operation of fossil and nuclear generation plants (1954-1976); Portland General Electric Company - (1976- ). O 1 O Amendment 13 (January 1982)
__ _ -. - . _ . _ = . .. . _ . _ _ _ _ _ _ _ _ . .._. _ ..___..._ _ . ._ _ . TABLE 13.1-1 Sheet 17 of 41 s P NAME: Joseph M. Hihelich TITLE AND DUTIES: Engineer, Civil Engineering Branch perform l, engineering analyses and studies including computer modeling for environmentally related nuclear work. Studies include cooling , systems, water and waste treatment systems, meteorological O correlations, hydraulics and hydrology, and effluent diffusien. 3 j Prepare or edit sections of Environmental Reports, SAR, and
; waste discharge permits. General plant engineering design and 4
review, including structural and seismic design and review. EDUCATION: OSU - BS in CE (1971). PROFESSIONAL AFFILIATIONS; None. i EXPERIENCE: Portland General Electric Company - (1970- ). .i i O s i I i i .1 i n v Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 18 of 41 O NAME: Gary E. Mitchell TITLE AND DUTIES: Engineer, Mechanical Engineering Branch - respon-sible for coordinating engineering review of vendor documents and AE designs and administering equipment procurement contracts and ongoing AE support for Pebble Springs Nuclear Plant. EDUCATION: Rose Hulman Institute of Technology - BS in ME (1968). PROFESSIONAL AFFILIATIONS: ANS. EXPERIENCE: Babcock & Wilcox Company - stress analyst in nuclear components engineering (1968-1970); Resident Startup Engineer, Nuclear Field Services Department. Provided technical direction to owner during startup and initial commercial operation at Oconee 1, TMI 1, and Rancho Seco (1970-1976); engineering depart-ment representative covering reactor coolant pump seal develop-ment program at Bingham Willamette Pump Company (1976-1978); Portland General Electric Company - (1978- ). O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 19 of 41
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NAME: William S. Orser TITLE AND DUTIES: General Manager, Technical Functions - responsible for engineering and other technical efforts necessary to support PGE's nuclear projects. EDUCATION: USNA - BS in General Engineering (1966); Naval Postgraduate School - MS in Computer Systems Management (1971). FROJESSIONAL AFFILIATIONS: PE - Oregon; Senior Reactor Operator ;; Licence, Trojan Nuclear Plant; ANS. 13 EXPERIENCE: U. S. Navy - Naval Nuclear Propulsion Training (1966-1968); service as commissioned officer aboard three nuclear powered submarines (1968-1975); Southern California Edison Company - Nuclear Engineer, San Onofre Unit 1 (1975-1976); Portland General Electric Company - Nuclear Plant Engineer, Trojan Nuclear Plant (1976); Material Control Supervisor, Trojan Nuclear Plant (1976-1977); Branch Manager, Engineering, Trojan Nuclear Plant (1977-1979); Manager of Operations and Maintenance, Trojan Nuclear Plant (1979-1980); General Manager, Technical Functions (1980- ).
\
v l r s. Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 20 of 41 O NMIE: James T. Owens TITLE AND DUTIES: Manager, Fuel Operations Department - responsible for planning, procurement, and coordination of the nuclear fuel supply and core physics analysis. 2 EDUCATION: USNA - BS in Mathematics (1966); US Naval Nuclear Power Schoci O and Prototype (1967); Purdue University - MS in NE (1972). PROFESSIONAL AFFILIATIONS: PE - California; ANS; EEI Nuclear Fuels Committee; INMM. EXPERIENCE: U. S. Navy Nuclear Power Program - engineering officer of the wa'ch on USS Long Beach, CGN9 (1966-1970); GE - nuclear core design (1972-1974); Portland General Electric Company - (1974- ). O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 21 of 41 O NAME: William L. Peregoy TITLE AND DUTIES: Engineer, Civil Engineering Bran::h - responsible for the design review of civil and structural engineering for nuclear projects. EDUCATION: U of California, Berkeley - BS in CE (1967); U of California, Berkeley - MS in CE (1972). PROFESSIONAL AFFILIATIONS: PE - California. EXPERIENCE: City of Los Angeles - civil engineer, responsible for design and review of highway and street improvement projects and structural design and review of flood control projects (1967-1969); Dravo Corporation - civil engineer, responsible for engineer-support for several large hydroelectric construction projects (1969-1971); General Atomic Company structural engi-neer for the design and aaalysis of prestressed concrete reactor vessels for high-temperature gas-cooled reactors (1973-1975); Portland General Electric Company - (1975- ).
~
I' O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 22 of 41 O NAME: Carl J. Piluso TITLE AND DUTIES: Supervising Engineer, Electrical Engineering Branch 2 - responsible for electrical system design review and special O projects related to nuclear projects electrical and control systems. EDUCATION: U of California, Berkeley - BA in Physics (1963); U of Oregon - PhD in Nuclear Physics (1968); Westinghouse Nuclear Power School (1972). 2 O PROFESSIONAL AFFILIATIONS: PE - Oregon. EXPERIENCE; Australian National U - research Fellow engaged in nuclear structure research (1968-1971); Westinghouse Electric Corporation - nuclear plant engineer / shift supervisor at the S5G naval propulsion plant prototype. Responsible for the overall operation of the nuclear plant and for determining the most effective methods of dealing with plant incidents and for handling routine maintenance and periodic testing. Supervised operation of all off-hull support systems and development of maintenance and operating procedures (1972-1974); Portland General Electric Company - (1974- ). O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 23 of 41 O NAME: Francis Rogan TITLE AND DUTIES: Manager, Mechanical Engineering Branch - responsible a for the direction and coordination of mechanical engineering
- C efforts, including In-Service Inspection, for nuclear projects.
EDUCATION: Institution of Mechanical Engineers ONC HNC (1957); Durham U - BS in Applied Science ME (1960); Nuclear Advance Course Royal Naval College - Creenwich (1962). PROFESSIONAL AFFILIATIONS: PE - California; ANS. EXPERIENCE: Vickers Corporation - apprentice (1952-1957); Ceneral Electric Company of England - gas-cooled reactor, research and development engineer (1960-1961); Vickers Corporation - design engineer on nuclear submarine design and commissioning (1961-1966); Westinghouse Electric Corporation - design engineer on reactor fluid systems (PWR) (1966-1969); project engineer on large PWR plants (1969-1970); Portland General Electric Company - (1970- ). O l Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 24 of 41 O NAME: Alexander N. Roller TITLE AND DUTIES: Supervising Engineer, Electrical Engineering Branch - responsible for supervising electrical design engineering for design modifications at Trojan Nuclear Plant. EDUCATION: OSU - BS in NE (1970); US Naval Nuclear Power School (1961); US Naval Nuclear Prototype Training SIC, Windsor, Connecticut _ (1962); various industry-sponsored nuclear training courses. C PROFESSIONAL AFFILIATIONS: PE - Cregon; NSPE; ANS. EXPERIENCE: US Navy - Reactor Operator on USS Triton (SSN 586) and USS Haddo (SSN 604) (1959-1964); Bechtel Power Corporation - Nuclear Engineer, responsible for design input on new construction nuclear plant projects (1970-1971); Portland General Electric Company - Nuclear Plant Engineer and Lead Test Engineer, Trojan Nuclear Plant (1971-1976); Electrical Engineer and Supervising Engineer, Electrical Engineering Branch (1976- ). O O Amendment 13 (January 1982)
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4 l j- TABLE 13.1-1 Sheet 25 of 41 ; n v
- NAME
- Ariel De J. Sanchez n
W 1 TITLE AND DUTIES: Senior Engineer, Mechanical Engineering Branch - $$ responsible for design review and engineering of fire protection,. , _HVAC and other mechanical systems for nuclear projects. i . EDUCATION: Havana U, Havana, Cuba (1959); Gonzaga U, Spokane, Washington ] (1961); Columbia Basin College (1963); PSU (1967). PROFESSIONAL AFFILIATIONS: ASHRAE; SFPE. EXPERIENCE: Long Lake Lumber Company - assistant engineer, general j ~ plant maintenance engineering (1960-1961); Boise Cascade Corp; Pulp & Paper Mill - engineer (1961-1966); Boise Cascade Corp..
- Central Engineering - engineer, project handling from conception
; .through design and fabrication of fluid storage and handling i systems, railroad car. bulk unloading systems, sewage treatment plants (1966-1967); Georgia-Pacific Corp. (Gypsum Division) -
project engineer, project handling from conception'through design and fabrication of Gypsum board plants, ready-six joint cement plants, rock handling systems, railroad car bulk unloading system, ' Styrofoam plants, Gypsum specialty machines (1967-1972); Portland j General Electric Company - (1972- ). h I f i i l- _1 t i .
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i t Amendment-13
-(January 1982)
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l 1 l l TABLE 13.1-1 Sheet 26 of 41 O NAME: James G. Schweitzer TITLE AND DUTIES: Supervising Engineer, Nuclear Projects Quality gl Assurance - responsible for supervising activities of Quality v Assurance Engineers. Assist in the development and maintenance of PGE QA Programs and procedures. Conduct internal and external audits, surveillances and reviews of quality-related activities to determine compliance with the Quality Assurance Program. EDUCATION: Marquette U - BS in ME (1972); US Army Nuclear Power School (1961); Lead Auditor Training (1978). PROJESSIONAL AFFILITATIONS: PE - Wisconsin; ASME. EXPERIENCE: US Army Corps of Engineers - Shift Supervisor, Health Physicist, Senior Reactor Operator, PM-1 Nuclear Power Plant (1962-1965); Field Test Engineer, Yuma Proving Grounds (1967-1969); Shift Supervisor, Health Physics Supervisor, Senior Reactor Operator, SM-1A Nuclear Power Plant (1969-1971); Project Officer for Decommissioning the SM-1 and SM-1A Nuclear Power Plants (1972-1974); Chief, Floating Power Plant Branch, Opera-tions Division (1974-1977); Project Officer for Decommissioning the MH-1A Floating Nuclear Power Plant (1977-1978); Portland General Electric Company - Nuclear Projects Quality Assurance Engineer (1978-1980); Supervising Engineer, Nuclear Projects Quality Assurance Department (1980- ). O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 27 of 41 O __ NAME: R. Patrick Sheppard TITLE AND DUTIES: Quality Assurance Engineer, Nuclear Projects Quality Assurance - responsible for reviewing quality-related procedures and assisting in the QA audit program. EDUCATION: Iona College - BS in Physics (1969); U of Notre Dame - MS in Physics (1973). m C PROFESSIONAL AFFILIATIONS: ASME.
- EXPERIENCE: Crane Company - Quality Assurance' Specialist, responsible for QA of Navy nuclear, commercial nuclear, and non-nuclear valves (1973-1977); Sargent & Lundy - Quality control Coordinator, assis-tant in charge of quality control for Byron / Braidwood nuclear plants Units 1 and 2 (1977-1980); Portland General Electric Company - (1980- ).
l Amendment 13
* (January 1982) l m __. ..._.---- = - , - - = . . . - - - - - ' * ~ ' " ' " "
i TABLE 13.1-1 Sheet 28 of 41 O NAME: Robert L. Steele TITLE AND DUTIES: Manager, Nuclear Projects Engineering Department - responsible for the direction and coordination of all engineer-ing efforts for nuclear projects. EDUCATION: OSU - BS in Mathematics (1963); U. S. Naval Power School, Vallejo, California (1964); U. S. Naval Nuclear Prototype Training, AlW, Idaho Falls, Idaho (1965); UW - NE Seminar (1963). PROFESSIONAL AFFILIATIONS: PE - Oregon; ANS; NSPE. EXPERIENCE: U. S. Navy - Engineering Duty Officer, AlW prototype, respon-sible for operation and training during 8-hr shifts at a prototype reactor plant (1965); Enginee ing Officer of the Watch, USS Grant (SSBN 631), responsible for the operation and maintenance of a naval nuclear propulsion plant (1966-1967); Engineer Officer, USS 2 Puffer (SSN 652), responsible for all operations, testing, and O maintenance of a naval nuclear propulsion plant during construction, initial sea trials, and shakedown operations. Officer in charge of ship's engineering department (1968-1970); Westinghouse Electric Corporation - Project Engineer (two large PWR plants), responsible for reactor systems modifications, ECCS studies and refueling outages (1970-1972); Senior Project Engineer, FFTF Program Manage-ment, FFTF Operations, responsible for startup and testing proce-dures for LMFBR mechanical systems at the FFTF/HTSF complex (1972-1973); EDS Nuclear, Inc. - Project Engineer, responsible for preparing quality assurance programs and conducting design reviews of reactor fluid systems for utility clients (1973-1974); Pacific Power & Light - Nuclear Systems Design Engineer, assigned to Portland General Electric to work as an engineer in the Mechani-cal Engineering Section of Generation Engineering (1974-1975); Portland General Electric Company - Senior Mechanical Engineer (1975-1977); Supervising Mechanical Engineer (1977-1980); Manager, Nuclear Projects Engineering (1980- ). Amendment 13 O (January 1982)
j TABLE 13.1-1 Sheet 29 of 41
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NAME: Richard L.' Sullivan d TITLE'AND DUTIES: Branch Manager, Nuclear Projects Administration - ;; responsible for document control and related functions for 33 nuclear generating facilities in the Nuclear Division. l l EDUCATION: USNA - BS in EE (1942); Stanford U - MA in Educational ] Psychology (1954); NF - OSU (1968);,INP - NUS (1969). i
' PROFESSIONAL AFFILIATIONS: None.
EXPERIENCE: Portland General Electric Company - Underground Department, assistant underground electrical engineer (1963-1964); Portland
, Distribution Department, electrical engineer (1964-1965); Executive Department, service methods engineer (1965-1967); assistant to Vice ,
President, Operations (1967-1970); project administrative engineer (1970-1979); Manager, Generation Construction Administration ,, (1979-1981); Branch Manager, Nuclear Projects Administration g (1981- ). - 1 !O i i d 1 i I l i i s I !. Amendment 13 (January 1982)
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TABLE L3.1-1 Sheet 30 of 41 O NAME: Don R. Swanson TITLE AND DUTIES: Senior Licensing Engineer, Licensing Branch - respon-sible for the review, coordination and implementation of Federal and State regulatory requirements related to the design, construc-tion and operation of the Pebble Springs Nuclear Plant. Respon-sible for preparation and maintenance of Pebble Springs licensing documents. 2 EDUCATION: Clark College - Associate of Arts (1968); UW - BS in ChE O (1971); UW - MS in NE (1973). PROFESSIONAL AFFILIATIONS: PE - Oregon. EXPERIENCE: University of Washington - research assistant engaged in nuclear engineering research (1971-1973); Bechtel Power Corporation - licensing and NSS coordinator, safety systems designer for Pebble Springs nuclear project (1973-77); Portland General Electric Company - (1977- ). O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 31 of 41 (l b NAME: Jerry L. Thale TITLE AND DUTIES: Senior Nuclear Engineer, Radiological Engineering Branch - responsible for initiation and review of radiological safety and engineering evaluations. Development and documenta-tion of calculational methods and computer programs used in radiological and safety evaluation, preparation and review of ,, licensing documents, specifications, drawings, etc, to ensure g compliance with regulatory requirements and industry standards. EDUCATION: UW - BS in AA (1970); UW - MS in NE (1971). PROFESSIONAL AFFILIATIONS: PE - California, Oregon. EXPERIENCE: Bechtel Power Corporation - nuclear project licensing and design (1971-1977); Portland General Electric Company - (1977- ). i 1 r . Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 32 of 41 O NAME: Bruce R. Turbitt TITLE AND DUTIES: Quality Assurance Engineer, Nuclear Projects Quality Assurance - responsible for all procurement QA activities. EDUCATION: Olympic College - Associate in Science (1959); UW - BS in Metallurgical Engineering (1962); NDE Qualification to ASME g Code, PT, MT, UT, RT; Seattle U consulting engineering course 3 (1972); Stat-A-Matrix, Inc. - lead auditor qualification and certification course (1975). PROFESSIONAL AFFILIATIONS: PE - California. EXPERIENCE: Boeing Company engineering aide (1961-1962); research engineer (1962-1969); B & K Construction Company - owner (1969-1972); Gilbert Associates, Inc. quality assurance inspector (1972-1974); Portland General Electric Company - (1974- ). O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 33 of 41 O NAME: Thomas D. Walt TITLE AND DUTIES: Manager, Radiological Engineering Branch - respon-sible for managing radiological' engineering, environmental monitoring, and radiological services for thermal power plants. EDUCATION: 'OSU - BS in ChE (1969); Bettis Reactor Engineering School (January 1971).: PROFESSIONAL AFFILIATIONS: Certified in Health Physics by the - American Board of Health Physics; PE - California; HPS.
. C EXPERIENCE: Division of Naval Reactors, USAEC - nuclear engineer.
responsible in Naval Reactors for radiological controls at Naval Reactor facilities and nuclear powered' ships.(1969-1974); Bechtel Power Corporation - nuclear engineer responsible for radiological aspects of nuclear plant design (1974-1976); Portland General Electric Company - nuclear engineer (1977-1978); Radiation Protection' Super-visor, Trojan Nuclear Plant (1978-1979); Radiological Section
- l. Supervisor (1979-1980); Manager, Radiological Engineering Branch (1980- ).
l O l l l Amendaec.t 13 (January 1982) t.
TABLE 13.1-1 Sheet 34 of 41 0 NAME: Rodger J. Wehage TITLE AND DUTIES: Supervising Engineer, Mechanical Engineering Branch
- responsible for supervising mechanical plant design group and engineering activities related to nuclear plant inservice inspection and tese.ing; mechanical equipment material, selec-tion, and qualification; piping stresa analysis; and design of piping and pipe supports.
2 O EDUCATION: OSU - BS in NE (1971); GE BWR Training Center; Conditional Certification as a Senior Reactor Operator (1971). PROFESSIONAL AFFILIATIONS: PE - California, Oregon; ASME. EXPERIENCE: Bechtel Power Corporation - systems engineer responsible for the mechanical design and coordination of electrical, civil, instrument and plant layout design of the Auxiliary Liquid Metal System in the Fast Flux Test Facility (1971-1974); Portland General Electric Company - (1974- ). O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 35 of 41 O NAME: M. E. Willian TITLE AND DUTIES: Manager, Electrical Engineering Branch - responsible for the direction and coordination of electrical and controls engineering efforts for nuclear projects. EDUCATION: WSU - BS in EE (1969); WSU - MS in EE (1974); GE Power Systems Engineering Course (1979). , C PROFESSIONAL AFFILIATIONS: PE - Oregon, Washington; ANS; IEEE. EXPERIENCE: Bechtel Power Corporation - Electrical Field Engineer, Peach Bottom Nuclear Plant (1969-1970); Electrical Field Engineer, Centralia Units 1 and 2 (1970-1972); Electrical Design Engineer, Peach Bottom Nuclear Plant project (1973-1974); Field Electrical Engineer, Centralia Steam Plant Precipitator Project (1974); Portland General Electric Company - (1974- ). O O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 36 of 41 O NAME: Joseph L. Williams g TITLE AND DUTIES: Vice Chairman of the Board - overall responsibility O for management of operating services, division operations, power operations, corporate security and environmental services. EDUCATION: U of Omaha, OSU - BS in EE (1948); Stanford U - Executive Development Program (1963); NF - OSU (1968); INP - NUS (1969). PROFESSIONAL AFFILIATIONS: PE - Oregon; NSPE; ANS; EC0; NELPA; NMA. EXPERIENCE: Portland General Electric Company estimator and chief estimator, Tualatin Valley Division (1948-1954); electrical engineer, Portland Division (1954-1956); assistant distribution engineer, Portland Division (1956); assistant to Vice Presient (1956-1959); superintendent of Electrical Maintenance and Construction (1959-1960); general division manager (1960-1968); elected Assistant Vice President (1968); Assistant Vice President, Engineering-Construction g (1969); Vice President, Engineering-Construction (1971-1977); O Executive Vice President (1977-1980); Vice Chairman of the Board (1980- ). O Amendment 13 O (January 1982)
. _ _ _ _ _ . . _ _ _ _ _ _ _ . _ . _ . _ _ m . _ . . _ _ _ _ .- _.
TABLE 13.1 Sheet'37 of 41 - NAME: -B. D. Withers
-TITLE AND DUTIES: -Vice President, Nuclear -~ responsible for corporate
. . activities related to PGE's nuclear. projects. t EDUCATION: Idaho State U - BS in Chemistry (1956); Naval Reactor Prototype Training, Idaho Falls, Idaho, AlW (1963), S5G (1972); i Westinghouse PWR Simulator Training (1974). i PROFESSIONAL AFFILIATIONS: PE - California; ANS. ) t i l ' EXPERIENCE: Phillips Petroleum - Chemist (1956-1958); Westinghouse' Electric Corporation - Chemist (1958-1962); Engineering Officer of the Watch at Naval Prototype Reactor Plant (1962-1964); Shift i Supervisor Navy Prototype (1964-1967); Operations Manager Navy Prototype (1967-1969); Bettis Atomic Power Laboratory - Manager, Operations Evaluation (1969-1971); Staff Management.in Naval Core Manufacturing-(1971-1972); Plant Manager, S5G Prototype - l (1972-1973); Portland General Electric Company - Assistant. t Superintendent, Trojan (1974-1977); Superintendent, Trojan (1977-1979); Manager, Environmental and Analytical Services - -(1979); Assistant Vice President, Environmental.and Analytical Services.(1979-1980); Vice President, Nuclear (1980- . ). 4 1 e i 1 I t i Amendment 13 , (January 1982) 1.
- - _ . . . . _ , _ _ . _ _ _ _ _ . _ . . . _ _ . _ . _ . . _ ~ _. , . , . . _ . _
TABLE 13.1-1 Sheet 38 or 41 e NAME: C. P. Yundt b TITLE AND DUTIES: General Manager, Trojan Nuclear Plant - responsible for operation of the plant in a safe, reliable, and efficient manner. Responsible for the safety of the plant staff and the general public. Responsible for operating the plant within the technical specifications and for complying with the provisions of the facility operating license. EDUCATION: OSU - BS in ME (1961); Westinghouse Fundamental Training and Reactor Operations (1971-1972); Westinghouse Design Lecture Series (1972); NUS Introduction to Nuclear Power (1969). PROFESSIONAL AFFILIATIONS: PE - Oregon, California. EXPERIENCE: San Diego Gas and Electric Company - junior engineer, Silver Gate Power Plant (1961-1962); junior engineer under efficiency engineer of Electric Production Department (1962-1963); assistant engineer under efficiency engineer of Electric Production Department (1963); plant engineer, Station B Power Plant (1963-1966); plant engineer, South Bay Plant (1966-1963); staff engineer, Resource Development Department (1968); Portland General Electric Company - mechanical engineer, Civil-Mechanical Branch (1968-1971); assistant plant superintendent, Trojan Nuclear Plant (1971-1974); Plant Superintendent, Pebble Springs Nuclear Plant (1974-1979); . O General Manager, Nuclear Station (1979); General Manager, Trojan Nuclear Plant (1979- ). 1 t 1 1 1 O Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 39 of 41 O V __ NAME: Gary A. Zimmerman TITLE AND DUTIES: Manager, Licensing Branch - responsible for managing the acquisition, maintenance and revision of state and federal licenses for thermal power plants. EDUCATION: USNA - BS (1964); US Navy Nuclear Power Training Programs; various Westinghouse PWR Engineering and Operations training programs. m C PROFESSIONAL AFFILIATIONS: PE - Oregon, Pennsylvania.
- EXPERIENCE: U. S. Navy - Engineering Officer of the Watch, responsible for operation and maintenance of naval nuclear powered submarines (1966-68); assistant to administrative officer at Puget Sound Naval Shipyard (1968-69); Portland General Electric Company -
engineering design review (1969-1970); Engineering Supervisor, Trojan Nuclear Plant (1970-1977); Licensing Section Supervisor, Generation Licensing & Analysis (1977-1980); Manager, Licensing Branch (1980- ). O (h \v/ Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 40 of 41 O Special Abbreviations Used in This Table hl AA Aeronautics and Astronautics AAAS American Association for the Advancement of Science 2 AE Architect-Engineer C AIChE American Institute of Chemical Engineers Cl AIF Atomic Industrial Forum M, ANS American Nuclear Society O' ASCE American Society of Civil Engineers 2 ASHRAE American Society of Heating, Refrigerating and Air Condition-C ing Engineers ASME American Society of Mechanical Engineers 2 ASQC American Society of Quality Control O Cl BA Bachelor of Arts BS Bachelor of Science 2 BWR Boiling Water Reactor C BYU Brigham Young University m CE Civil Engineering Cl ChE Chemical Engineering
- Certified Health Physicist CHP bl ECCS Emergency Core Cooling System
" Electric Club of Oregon ECO m EE Electrical Engineering EEI Edison Electric Institute h 2 FFTF Fast Flux Test Facility
=l General Electric Company GE GGA Gulf General Atomics Company HNC Higher National Certificate (Great Britain) - HPS Health Physics Society C HTSF High Temperature Sodium Facility % HVAC Heating, Ventilating and Air Conditioning M; ICOLD International Commission on Large Dams O IEEE Institute of Electric and Electronic Engineers Cl INMM Institute for Nuclear Materials Management INP-NUS Introduction to Nuclear Power Course by NUS Corporation m IRPA International Radiation Protection Association O IRFBR Liquid Metal Fast Breeder Reactor e
Cl MA Master of Arts
% ME Mechanical Engineering Master of Science Cl MS O
Amendment 13 (January 1982)
TABLE 13.1-1 Sheet 41 of 41 (\_ j'i Special Abbreviations Used in This Table NDE Non-Destructive Examination NE Nuclear Engineering l}} NELPA Northwest Electric Light and Power Association g;; NF-OSU Nuclear Fundamentals Course by Oregon State University 13 NMA National Management Association ; NSPE National Society of Professional Engineers l;d j NSS Nuclear Steam System (Babcock & Wilcox) ,, NSSS Nuclear Steam Supply System (W) [ v ONC Ordinary National Certificate (Great Britain) OSU Oregon State University PE Registered Professional Engineer 10 PE0 Professional Engineers of Oregon ; PhD Doctor of Philosophy h;3 PITT University of Pittsburgh PSU Portland State University l,g, PT, MT, Penetrant, Magnetic Particle, Ultrasonic, and Radiographic UT, RT Test-Course and Qualification PWR Pressurized Water Reactor l{f QA Quality Assurance lh, J RPG Reactor Power Generation RS Reactor Safety SAR Safety Analysis Report ,, SE Registered Structural Engineer SFPE Society of Fire Protection Engineers [ USN United States Navy USNA United States Naval Academy ,s UW University of Washington W Westinghouse Electric Corporation WSU Washington State University I f ~3 I
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BECHTEL QUALITY ASSURANCE PROGRAM
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i Bechtel Power Corporation (Bechtel) has established a quality assurance ^ program for regulation of all_ activities associated with their'responsi- O bilities in providing quality-related services for engineering design, construction management, and procurement assistance to Portland General Electric Company (PGE) on the Pebble Springs Nuclear Plant. Compliance with requirements of the quality assurance-program is' required by all i Bechtel entities. Bechtel's Scope of Responsibility differs from that indicated in BQ-TOP-1 in that Bechtel does not provide full procurement j- services. and -does not perform construction. Procurement assistance includes preparation of material requisition packages containing Technical Specifications and identification of quality requirements to be provided by contractor / suppliers; supplier source surveillance; and evaluation of bids, including contractor / supplier QA programs. Purchase orders and contracts are placed directly by PGE.
^
Construction management provisions for quality-related services -include receiving, receipt inspection and storage of PGE purchased materials. Bechtel QA is responsible for quality surveillance and audit over onsite contractor activities to assure implementation of their quality program,- including inspection responsibilities. nechtel Construction Management advises PGE on acceptability of completed work. The quality assurance program manuals and procedures applicable to the Pebble Springs project are identified in Bechtel's Pebble Springs Project i ~ Nuclear Quality Assurance Manual (NQAM) and the Project Quality Program l
- Document List.
1 i The quality assurance program used by the Bechtel Power Corporation dur-ing design, procurement and construction'of the Pebble Springs Nuclear Plant is described in the NRC-approved Bechtel Topical Report BQ-TOP-1, y I Rev. 2A, "Bechtel Quality Assurance Program for Nuclear Power Plants", subject to the following ruodifications and additions: b v i 1 Amendment 13 (January 1982)
Page 1, et al Replace all references to "Bechtel Thermal Power Organization (TPO)" with "Bechtel Power Corporation (BPC)". Page 3 Change " ANSI Standard N45.2.12-1974" to " Regulatory Guide 1.144, 1 l Rev. 1 (September 1980), ' Auditing of Quality Assurance Programs l l for Nuclear Power Plants', and ANSI /ASME N45.2.12-1977, ' Require-1 ments for Auditing of Quality Assurance Programs for Nuclear Power Plants'." l Add the following paragraph: l
" Regulatory Guide 1.146, ' Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants' (August 1980)."
l ' R O Change " Regulatory Guide 1.58" to " Regulatory Guide 1.58, Rev. 1 (September 1980)", and replace "(August 1973)" with "(1980)". I l Page 7, et al l Replace all references to " Materials and Quality Services (M&QS)" with " Material end Quality Services and Codes and Standards (M&QS/C&S)". Page 24 Add the following paragraph under Section 2.5, Personnel:
"QA staffing is based on the long-range projected work schedule and is periodically reevaluated and adjusted as necessary."
Amendment 13 2 (January 1982)
l
-Replace " Regulatory Guide'1.58 (August 1973)" with " Regulatory Guide 1.58, Rev.1:(September 1980)" in Subparagraph (2) of the' Q last' paragraph. .I Page 25 Delee Subparagraph (3) (first paragraph) and replace with:
" Auditor. Qualifications - Personnel performing audits will be qualified in accordance with the appropriate requirements of ANSI N45.2.23-1978 (Regulatory Guide 1.146)."
Page 32 Add the following paragraph after the first paragraph: O "For commercial.'cff-the-shelf' items where specific. quality assurance controls appropriate for nuclear applications cannot be imposed in a- practical manner, 4 special quality verification requirements shall be established and described to provide for an 'accep-table' item." Page 34 Add .the following paragraph af ter the second paragraph:
" Procedures are established for the review of procure-ment documents to determine that quality requirements are correctly stated, inspectable and controllable, and that there are adequate acceptance and rejection criteria."
O 3 Amendment 13 (January 1982)
Page 36 In the second sentence of the fourth paragraph, change " drawings and specifications" to " drawings, specifications and Q-List". Page 46_ Add the following after the words " planning documents" in the first sentence of the third paragraph:
", these documents provide for inclusion of mandatory hold points or witness points when applicable."
Page 48 Add only the words "which include designation of applicable witness and hold points" after the words " Test plans and procedures" in the third sentence of the second paragraph. Add the following paragraph after the third paragraph:
" Procedures are established and described to control altering of the sequence of required tests, inspec-tions and other operations important to safety.
Such actions should be subject to the same controls as the original review and approval. The QA organi-zation reviews and documents concurrence with these procedures." Page 49 Add the following to the fourth paragraph: l
" Procedures provide for the selection of measuring equipment compatible with the type and accuracy requirements of the operations to be performed."
O Amendment 13 4 (January 1982)
..P_ age 55 d Add the following'after:the first sentence-of the fifth-paragraph:
"An installation shall be considered to be in an 'as constructed' condition if it is installed within tolerances established by Project Engineering as indicated in the design output documents. . Completed quality verification records which correctly identify the 'as. built' condition of the plant, including material certification and test data for trace-ability, quality verification records, such as.
inspection and test reports, evidencing conformance to design documents, and nonconformance reports for repair and 'use-as-is' dispositions are placed in quality record files." O Page 57 8 Add to Subparagraph (3) of the second paragraph:
" Audits of Bechtel suppliers performing continuing work for one or more Bechtel projects are conducted as a minimum on an annual basis; audits of suppliers performing limited duration assignments are con-ducted at least once during the life of the contract.
The requirement may be waived when evidence exists of continuing satisfactory performance, including surveillance by the Procurement Supplier Quality Department. This waiver is based on an annual review by Procurement Supplier Quality with the concurrence of the Project Quality. Assurance Engineer. Results of these reviews are placed in supplier quality history files. O 5 Amendment 13 (January'1982)
The annual audit requirement shall not apply to standard off-the-shelf items and bulk commodities where required quality can adequately be determined by receipt inspec-tion or post-installation checkout of test." Replace BQ-TOP-1 Appendix A Fages A-16 and A-17 with the following:
" Regulatory Guide 1.144 AUDITING OF QUALITY ASSURANCE PROGRAMS FOR NUCLEAR POWER PLANTS (Revision 1, September 1980)
The requirements of ANSI N45.2.12-1977 as modified and interpreted by the regulatory position will be applied to the Bechtel quality program for safety-related items except as modified or interpreted below:
- 1.
Reference:
Standard Sections 4.3.2.4 and 4.5.1 (Investigation). As an equivalent alternative v to the requirement for the audited organization to investigate any adverse audit finding to determine and schedule appropriate corrective action, Bechtel's auditing organization may determine the investigatory action and corrective action, including action to prevent recurrence pertinent to adverse audit finding. These action:., are agreed to by the audited organization. Further, in Section 4.5.1, as an equivalent alternative to the 30-day response time, a response time appropriate to the finding is agreed to by the audited and auditing organizations.
- 2.
Reference:
Regulatory Section C.7, Standard Section 5.2 (Audit Records). Audit records O Amendment 13 6 (January 1982)
-___._._...._..-_._.m_. - -
_ _ _ . . ~ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ s i, o I.' shall include documents as' defined-in the ^ i' i standard and other. documents if necessary to. y: j support audit findings." i -. I i~ . i . i i i i~ - l 1 i l I , r
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1 [ t 7 . Amendment 13 f ._ (January 1982) I
'I APPENDIX A.9 This appendix contains responses to all applicable items discussed-in
.new paragraph'(f) to 10 CFR 50.34, entitled " Additional TMI-related Requirements" (47 FR 2286, dated January 15, 1982). Each requirement in ,
10 CFR 50.34(f) is restated in its entirety, followed by the PGE response. Applicable sections to the PSAR have also been revised as part of'this Amendment, where appropriate. ' The extent of all changes is summarized in the Amendment to Application that is a part of Amendment 13. All changes have been highlighted by a
" change indicator" mark as specified in the " Standard Format and Content
! .of Safety Analysis Reports for Nuclear Power Plants (Revision 1)", issued October 1972. The revision numbers and date are noted on the lower corner below tha sideline indicator. 6 1 ( l O i . f l i i l ) 1 O b A.9-1 Amendment 13 (January 1982) l-
~
NRC'10 CFR 50.34(f) REQUIREMENT Probabilistic Risk Assessment.
.(1) To s'atisfy the following requirements, the> application shall provide sufficient information to. describe.che nature of the studies, how they are to be conducted, estimated' submittal dates, and a program to. ensure that the results of such studies are factored-into the-F final design of the. facility. -All studies shall be completed no l later-than two years following issuance of the construction permit or manufacturing license. , (i) Perform a plant / site specific probabilistic risk assessment, the aim of which is to seek such improvements in the relia-bility of core and containment heat removal systems as are significant and practical and do not impact excessively on the plant. (II.B.8)[a) ,
i RESPONSE TO 10 CFR 50.34(f)(1)(i) O A plant / site specific probabilistic risk assessment (PRA) will be per-i formed, the principal objective being to seek such improvements in the reliability of core and Containment heat removal systems as are significant and practical and do not impact excessively on the plant. The PRA program will be structured to meet this objective, recognizing l the advanced state of design (engineering 40 percent complete) and fabrication (major plant components fabricated and in storage). k i-i i i l [a] Alphanumeric designations correspond to the related action
- plan items in NUREG-0660, "NRC Action Plan Developed as a
( Result of the TMI-2 Accident" and NUREG-0718, ' Licensing Requirements for Pending Applications for Construction
- Permits and Manufacturing License". They are provided
!- herein for information only. !O A.9-2 Amendment 13 (January 1982)
Risk Study The PRA study will identify the sequences and system / component failures which are the dominant contributors to core damage risk. The methodology used will be similar to that employed in WASH-1400, updated to be con-sistent with the IEEE/ANS efforts to establish a standard methodology. The initiating evcats to be considered will include those indicated in Table A.9-1, together with the applicable accidents and transients iden-tified in the Pebble Springs PSAR and in WASH-1400. These events will be screened to identify the basic set of initiating events requiring opera-tion of the key safety systems for core protect in and release mitigation. The FPA program will focus on core and Containment cooling systems in performing event tree / fault tree analyses, and will include consideration of environme.stal effects, human errors, common cause failures, interde-pendence of support systems, and system unavailabilities in the event tree / fault tree analysis. A component failure data base for use in system fault tree analysis will be utilized from recognized reference sources including WASH-1400 and IEEE-500. In addition, prototype specific failure data will be requested from vendors of selected components being supplied to Pebble Springs. Human errors will be considered in the development of the data base. Furthermore, uncertainty analyses will be performed to det cmine propa- ! gation of component failure data, including error ranges, through the l l fault trees. Sensitivity analyses will be performed by varying the i failure rates of key banis events which contribute to dominant event sequences in order to determine the effect on system failure rates and overall recults. The final report will follow the outline presented in Table A.9-2. O Amendment 13 A.9-3 (January 1982)
Performance of Study O Portland General Electric Company (PGE) has overall responsibility for the PRA study and will ensure that the study is performed by highly qualified personnel experienced in risk assessment methodology. PGE personnel will be actively involved in this study and will provide direction in the' development of the program. PGE retains ultimate authority and responsibility for implementing design improvements as a 4 result of this study. Application of Results to Final Design Acceptance criteria for system probabilistic risk analyses will be established during the initial phase of the program. These acceptance criteria will include both quantitative and qualitative considera-tions of potential design changes on plant cost, schedule and system availability. i The results of the probabilistic risk analyses will be evaluated using O*- the acceptance criteria to determine design or other changes. The results of the study will be used to improve reliability of component selection, specification and testing. Where practical, the results of the study will be used to identify improvements to_be considered for maintenance, procedures, operator training, and operating feedback, and to identify those areas where additional quality assurance activities ' would improve reliability of core and Containment cooling systems. Schedule i The program will commence following issuance of a construction per-mit (CP). The initial phase of the program is expected to take approximately 15 months and will consist of a preliminary PRA of the present Pebble Springs design. The final study will be completed and submitted to the NRC within 2 yr of CP issuance. O , A.9-4 Amendment 13 (January 1982)
Acceptance Criteria There are currently no established regulatory requirements or acceptance criteria for judging the acceptability of the PRA. Thus, the need for implementing changes in design, or operating, testing or maintenance procedures to achieve improvements will be based on judgmental acceptance criteria which are not directly related to licensing requirements. O O Amendment 13 A.9-5 (January 1982)
'NRC 10 CFR'50.34(f) REQUIREMENT.
A Q ' Auxiliary Feedwater System Evaluation
- 1
- . (1) To satisfy the following requirements, the application'shall provide 1-sufficient information to describe. the nature of ' the studies, how they are. to be conducted, estimated submittal' dates, and a program to- ensure tha* the results'.of 'such studies are factored ~into the final design of the facility. All . s tudies shall . be . completed - no later than two years following issuance of the construction permit or manufacturing license.
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~(ii) Perform an evaluation of the proposed auxiliary .feedwater.
system (AFWS),'to include.(applicable to PWR's only) (II.E.1.1): (A) A simplified AFWS reliability analysis using event-i- j tree and fault-tree logic techniques. (B) A design review ~of AFWS. (C) An evaluation of AFWS flow design bases and criteria. I' d RESPONSE TO 10 CFR 50.34(f)(1)(ii) A simplified reliability analyses of the Auxiliary Feedwater System (AFS) using event-tree and fault-tree logic techniques will be performed as j part of the PRA program (see response to 10 CFR 50.34(f)(1)(1)]. The l results of this evaluation will be compared with the results of the NRC staff's generic AFS evaluation published in Appendix III to NUREG-0611 ? and NUREG-0635. t The Pebble Springs Plant design includes: (a) four auxiliary feedwater j 4 pumps (two electric, two diesel-driven) arranged in a four-train con-figuration,'(b) auxiliary feedwater injection capability for up to 2 hr O.v I A.9-6 Amendment-13 (January 1982) 1
- , - . , - - . _ . . _ ._,,___c.~_,. _ . , _ _ . _ _ , _ . _ . - _ _ _ , . . . , , - . . , . . _ , , . . , . . _ , _ . . _ , _ . _ . _ _ . _ _ _ _ . _ , u .-__.-
following loss of all ac electrical power, and (c) high head High Pressure Injection (HPI) pumps which could be used in the feed and bleed mode of decay heat removal in the unlikely event of total AFS failure. These design features assure a very high AFS reliability relative to those systems evaluated by the NRC in NUREG-0611 and NUREG-0635. The AFS and HPI System are described in PSAR Sections 6.7 and 6.3, respectively. The design review of the AFS will be performed based on the acceptance criteria in Standard Review Plan (SRP) Section 10.4.9 (NUREG-0800). The AFS flow design bases and criteria will be evaluated to verify the ade-quacy of the calculated system requirements in meeting the Nuclear Steam System (NSS) design interface requirements. The above evaluations will be completed within 2 yr af ter issuance of the CP. Any modifications to AFS design identified by these evaluations will be implemented at that time. O O Amendment 13 A.9-7 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT
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I \ Impact of Reactor Coolant Pump Seal Damage Foll.owing_Small-Break LOCA with Loss of Offsite Power (1) To satisfy the following requirements, the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the results of such studies are factored into the final design of the facility. All studies shall be completed no 4 later than two years following issuance of the construction permit or manufacturing license. (iii) Perform an evaluation of the potential for and impact of reactor coolant pump seal damage following small-break LOCA with loss of offsite power. If damage cannot be precluded, provide an analysis of the limiting small-break loss-of-coolant accident with subsequent reactor coolant pump seal
,-~ damage. (II.K.2.16 and II.K.3.25)
(-) RESPONSE TO 10 CFR 50.34(f)(1)(iii) Reactor coolant pump (RCP) seal cooling is provided by either high-pressure seal injection from a makeup /HPI pump or component cooling water during RCP operation. Coincident loss of both sources of cooling must occur before all RCP seal cooling is lost. Loss of offsite power (LOOP) will cause the RCPs to stop, thus removing much of the heat input to the seal assemblies, but the effectiveness of component cool-ing water is also reduced. Seal injection is effective for running or non-running RCPs.
- For Pebble Springs, the Makeup and Purification System (MPS) is designed to maintain seal injection to all four RCPs following LOOP, assuming failure of the operating makeup /HPI pump or failure of either diesel generator to start. Loss of all seal cooling following LOOP is therefore A not expected. A small-break LOCA would cause Engineered Safety Features A.9-8 Amendment 13 (January 1982)
(ESF) actuation of the MPS to the HPI operating mode. The design of the MPS/HPI also will maintain seal injection to all four RCPs in an HPI operating mode, assuming failure of the operating makeup /HPI pump or failure of either diesel generator to start. Loss of all seal cooling following small-break LOCA with LOOP is therefore not expected either. However, a postulated loss of all seal cooling, which would require a double failure for the MPS/HPI System following LOOP or small-break LOCA with LOOP, is not expected to result in unacceptable consequences. Prolonged loss of ooling to an idle pump will allow a temperature increase in the seal area due to leakage of the hot reactor coolant up through the seals. The rate at which heatup occurs is dependent mainly upon whether or not the seal return is isolated, the condition of the seals themselves, and the pressure in the Reactor Coolant System (RCS). The Pebble Springs design has double automatic isolation of seal return valves in series to ensure seal return isolation when needed. Instrumen-tation and alarms are provided in the control room which indicate seal injection flow and alarm on low flow. Should seal temperatures exceed approximately 350*F, it is expected that the elastomers which make up part of the seal assembly would start to soften and could begin to extrude before reaching 500*F. The amount of extrusion depends upon time, temperature and annular clearances. Dis-tortion and possible cracking of seal parts could also occur, depend-ing upon the temperature ramp rates endured. It is estimated that under worst conditions on a static pump, leakage could reach 5 gpm after 30 min and 10 gpm after approximately 1 hr. Beyond this time seal leakage would tend to stabilize, provided no subsequent attempt was made to operate the pumps. The combined leakage of approximately 40 gpm is well within the normal makeup capability. Operating procedures for Pebble Springs will prohibit RCP restart if seals have been subjected to concurrent loss of seal injection and component cooling water, resulting in seal tempera-tures exceeding 185'F. As described above, severe damage to the RCP seals is precluded by: O Amendment 13 A.9-9 (January 1982)
(1) Providing seal cooling by both high pressure seal injec-tion and through external heat exchangers using component O cooling water. (2) Design of the MPS to continue to provide i.igh pressure seal injection in the event of LOOP and single failure,
-or LOCA and single failure.
(3) Operating procedures to minimize temperature excursions on loss of cooling, reduce thermal shock on reestablishment of cooling, and prevent operation of a pump with poten-tially distorted or cracked seals. Based on the above, additional small-break LOCA analysis for a seal failure event is not required. O O A.9-10 Amendment 13 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT O
. Report on Overall Safety Effect of PORV Isolation System (1) To satisfy the following requirements, the application shall provide suf ficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the results of such studies are factored into the final design of the facility. All studies shall be completed no later than two years following issuance of the construction permit or manufacturing license.
(iv) Perform an analysis of the probability of a small-break loss-of-coolant accident (LOCA) caused by a stuck-open power-operated relief valve (PORV). If this probability is a significant contributor to the probability of small-break LOCA's from all causes, provide a description and evaluation of the effect or, small break LOCA probability of an automatic PORV isolation system that would operate when the reactor _ ,,' coolant system. pressure falls af ter the PORV has opened. (Applicable to PWR's only). (II.K.3.2) RESPONSE TO 10 CFR 50.34(f)(1)(iv) An analysis will be performed to demonstrate that planned upgrades to the 10RV block valve power supplies and associated controls will decrease the probability of a small-break LOCA caused by a stuck-open PORV. The report of this analysis will include a safety examination of the auto-matic PORV isolation to be applied to Pebble Springs and will be sub-mitted within 2 yr after the CP is issued. The analysis to be performed will demonstrate that the probability of a stuck-open PORV is not a significant contributor to the probability of a small-break LOCA due to all causes. It will seek to demonstrate that this can be achieved while retaining the present PORV and high
-, RCS pressure reactor trip setpoints. This will reduce the number of (v)
A.9-11 Amendment 13 (January 1982)
unnecessary reactor trips and retain the current runback capability of the NSS. The applicability of data analysis to any unique Pebble Springs features will be addressed. Historical safety valve failure rates for operating Babcock & Wilcox (B&W) plants will be used in an additional reliability analysis to show that the safety valves are not adversely affected by the automatic PORV block valve closure system and that they are not a significant contributor to the probability of a small-break LOCA due to all causes. Prior to the above analyses, several actions will be taken to decrease the probability of a small-break LOCA caused by a stuck-open PORV. These include upgrading the motive and control power supplies for the PORV block valve to safety grade, addition of an automatic PORV isola-tion signal and addition of anticipatory reactor trips to reduce PORV challenges. Power supplies to the PORV are currently capable of being supplied by a standby diesel generator. The automatic isolation signal will be developed from RCS pressure and PORV position and is intended to permit retention of the PORV setpoint to open at a pressure below the high pressure reactor trip. The power supplies for the PORV/ block valves and the additional anticipatory reactor trips are described in the the responses to 10 CFR 50.34(f)(2) (xx) and (xxiii), respectively. O Amendment 13 A.9-12 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT Hydrogen Control Systems Evaluation (1) To satisfy the following requirements, 'the application 'shall pruvide sufficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the results of such studies are factored into the final design of the facility. All studies shall be completed no later than two years following issuance of the construction permit or manufacturing license. t (xii) Perform an evaluation of alternative hydrogen control
- systems that would satisfy the requirements of paragraph (2)(ix) of this section. -As a minimum include consideration j of a hydrogen ignition and post-accident inerting system.
The evaluation shall include: (A) A comparison of costs and benefits of the alternative
.O systems considered.
1 (B) For the selected system, analyses and test data to ! verify compliance with the requirements of (2)(ix) of this section. (C) For the selected system, preliminary design descrip-tions of equipment, function, and layout. RESPONSE TO 10 CFR 50.34(f)(1)(xii) An evalustion of alternative hydrogen control systems will be performed for Pebble Springs. This evaluation will include consideration of a hydrogen ignition and post accident inerting system and a comparison of their costs and benefits. This evaluation will also include, for the selected system, analyses and test data to verify compliance with the requirements of 10 CFR 50.34(f)(2)(ix) and preliminary design descriptions l A.9-13 Amendment 13
! (January-1982)
of equipment, function and layout. The preliminary concept selected (distributed ignition system) is described in the response to 10 CFR 50.34(f)(2)(ix). ?GE will actively monitor the various industry-wide efforts on degraded core hydrogen control. The industry-wide efforts to be monitored will include submittals on individual plant dockets, research by national laboratories (eg, Sandia, Livermore), Electric Power Research Institute (EPRI), and the Industry Degraded Core Rulemaking (IDCOR) program. Also included in this effort will be the review of already completed testing programs (at TVA's Singleton Laboratory, hydrogen burn testing at Fenwal, and Lawrence Livermore Laboratory's igniter tests) and the monitoring of ongoing and future testing (Acurex, Factory Mutual, Hanford Engineering Development Laboratory, etc). PGE will incorporate the results of industry and NRC research programs that are applicable to the selected hydrogen control system. The hydrogen control systems evaluation will be completed within 2 yr following issuance of the CP. O O Amendment 13 A.9-14 (January 1982)
t NRC 10 CFR 50.34(f) REQUIREMENT Long-Term Training Simulator Upgrade (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (i) Provide simulator capability that correctly models the control room and includes the capability to simulate small-break LOCA's. (Applicable to construction permit applicants only.) (I.A.4.2) RESPONSE TO 10 CFR 50.34(f)(2)(1) The training program for the Pebble Springs licensed operators will include training on a simulator that correctly models the control room, including the capability to simulate small-break LOCAs. The simulator used will meet the functional requirements of ANSI /ANS 3.5-1981, " Nuclear
'- Power Plant Simulators for Use in Operator Training".
In addition, the licensed operator training program will meet the following: (1) ANSI /ANS 3.1-1981, " Selection, Qualification and Training of Personnel for Nuclear Power Plants". The Pebble Springs license candidate training program will be typical of that defined in Appendix A of ANSI /ANS 3.1-1981. (2) 10 CFR 55, " Operators' Licenses". (3) Regulatory Guide 1.149 (April 1981), " Nuclear Power Plant Simu-lators for Use in Operator Training", which meets the require-ments of NUREG-0660, Item II.K.3.54. These requirements will be accomplished in a timely manner to support startup and operation. a A.9-15 Amendment 13 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT O t j Long-Term Program Plan for Upgrading of Procedures (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate: that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (ii) Establish a program, to begin during construction and follow into operation, for integrating and expanding current efforts to improve plant procedures. The scope of the program shall include emergency procedures, reliability analyses, human factors engineering, crisis management, operator training, and coordination with INPO and other industry efforts. (Applicable to construction permit applicants only.) (I.C.9) RESPONSE TO 10 CFR 50.34(f)(2)(ii) o PGE will establish a program for the development of plant operating procedures during the construction period. The plan for this program will be developed within 2 yr following receipt of a CP. This program plan will be developed and implemented by the Plant General Manager, who will also be responsible for the training of operators. 1 i All plant procedures described in Section 13.5 of the PSAR will be reviewed and approved by the Plant Review Board (PRB) or subcommittees of the PRB. These subcommittees will be structured to ensure that the Plant General Manager will review all procedures with additional reviewers, l depending on the type of procedure. For example, the Operations Super-visor would be required to review and approve procedures involving operations, surveillance testing, etc. Specific requirements governing the review and approval of plant procedures will be included in the Pebble Springs Final Safety Analysis Reporc (FSAR) and Technical Specifications.
.h d A.9-16 Amendment 13 (January 1982)
I i ! ___ _ , , . - _ _ ,, - - - - - - - -- . - ~ - - - - - - - - - ~ --- -
Development of the program for preparation of operating procedures will include, but not be limited to, consideration of the following: O (1) Incorporation of the pertinent results from the human factors review of the control room, described in the response to 10 CFR 50.34(f)(2)(iii). (2) The results of the probabilistic risk studies, described in the response to 10 CFR 50.34(f)(1)(i). (3) Results from applicable portions of generic efforts on procedures, such as those being sponsored by the B&W Owners Group and currently under way, efforts by the Institute for Nuclear Power Operations (INPO), or other applicable industry activities that may become available. Emergency procedure improvemenmts will follow closely the efforts of the B&W Owners Group. (4) Evolving NRC criteria, as in NUREG-0737, Item I.C.1, cur-rently being applied to operating procedures for opera-ting plants and applicants for operating licenses. (5) Scheduling procedures development to support operator trainiag, including the training of operators during preoperational testing of completed systems, with plant-specific procedures. (The use of procedures during the l preoperational testing program is discussed further in Pebble Springs PSAR Section 14.1.1.1.1). (6) Development of suitable analytical bases for procedures. The emergency procedures for training will be documented l with references that identify the analytical or technical bases that demonstrate conformance to the plant safety requirements. Amendment 13 A.9-17 O (January 1982) 1
(7) Results_of the ongoing operating experience evaluation p program described in the response to 10 CFR 50.34(f)(3)(1). b (8) Crisis management techniques will be applied to emergency procedures governing the interaction of various activities between the control room, Operational Support Center (OSC), Technical Support Center (TSC), Emergency Opera-tions Facility (EOF) and the NRC Operations Center. Control room occupancy and shift turnover procedures will include consideration of emergency and crisis conditions. The procedures development program will also meet the applicable guidelines of ANSI /ANS 3.1-1981, " Selection, Qualification and Training of Personnel for Nuclear Power Plants", and ANS-3.2,
" Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants", Draft 8 (April 1981).
l l l-l V I- A.9-18 Amendment 13 (January 1982) l
_ _ _ . . _ .... . _ _. _ ._ .._m , _ _
'NRC 10 CFR 50.34(f) REQUIREMENT 4
Control Room Design Review (2) To satisfy the following requirements, the application shall provide ,i . sufficient information to demonstrate.that the required actions will be satisfactorily completed by the operating license stage.
-This information is of the type customarily required to sasisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues.
A (iii) Provide, for Commission review, a control room design that i reflects state-of-the-art human factor principles prior to
- committing to~ fabrication or revision of fabricated control room panels and layouts. (I.D.1)
RESPONSE TO 10 CFR 50.34(f)(2)(iii) 1 The Pebble Springs control-room design will be developed to reflect state-of-the art human factors principles. The primary objective of the control room design review will be to assure that the control room I ' provides the system status information, control capabilities, feedback l and analytic aids necessary for operations personnel to accomplish their functions effectively. - The control room design will be provided for IEC review prior to fabrication of the main control boards (MCB). The
- following provides preliminary design information and the approach that will be used for control room design and design review.
i Preliminary Design Information j Pebble Springs PSAR Section. 7.5 describes the control room system. Figure 7.5-1 shows the control room - control-board arrangement. This design includes several important features that improve the man-machine j interface in comparison to a THI-2 type control room. These features j include: i 1 (1) Use of equipment control miniature backlighted pushbutton i switch modules. The switches operate with a low voltage w
; A.9-19 Amendment 13 (January 1982) 4
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solid state logic system, and their use has resulted in an improved control board design due to: (a) Reduced control board length. (b) Reduced control board congestion (use of plug-in type modules). (c) Significant reduction of trouble alarms. Equip-ment pushbutton flashing 1.i.ghts will be used as equipment trouble alarms. (2) All controls and displays necessary to the operator are functionally arranged in the U-shaped front contro.1 boards. Safety-related systems (eg, AFS) are distinguishable on the control board by color-coded mimics. (3) The status of " critical safety-related systems" by auto-matically activated annunciation is available to the control room operator. The system will indicate inoperability at the safety-related " system train level". The train level of annunciation is independent of normal control room annunciation. Basic Design Approach and Methodology The basic goal of control room design is to provide sufficient devices (ie, controls, indicators and annunciators) functionally arranged to enable the control room operations personnel to efficiently and effec-tively direct or control the performance of the plant through all phases of normal or transient operation. This will be accomplished by performing a human factors review of the control room design using the guidelines of NUREG-0700, " Guidelines for Control Room Design Reviews" (September 1981). The scope of this review will include the control room design, including the control room arrangement and environment, and the MCB layouts. A O Amendment 13 A.9-20 (January 1982) {
systems / operations analysis of the control room design will be performed [) utilizing the guidelines of Appendix B of NUREG-0700. The design will be evaluated for conformance with the human engineering guidelines of Section 6 of NUREG-0700, which address control room workspace, communica-tions, ant.unciator warning systems, controls, visual displays, labels and location aids, process computers, panel layout and control-display inte-gration. This human factors review will be performed by individuals experienced in operations, systems analysis, human factors engineering, architectural engineering, nuclear engineering and instrumentation and control engineering. Figure A.9-1 is a preliminary plan for the design review of the Pebble Springs control room. The elements of this process are as follows: (1) Develop a detailed design review plan specifying the goals, e analytical tools, resources and schedule for the design effort. (2) Perform a systems / operations analysis to systematically define the equipment, personnel, and procedural data needed to meet all functional objectives of the control room, including safe operation of the plant. This analysis will be comprised of the following: (a) Function analysis. The purpose of this analysis is to define and evaluate all the functions and actions to be performed by operators in the control room. (b) Function allocation. The purpose of this analysis is to assign functions to operators or equipment based on a comparison of their capabilities and limitations to per-form the function. This will be accomplished by defining the performance capabilities of humans and machines based on the functions defined. The tasks required to perform each function are then defined and analyzed to establish design details, ie, elements of the control room design g specification for plant systems and training and operating
) -
procedure requirements for operators. 1 A.9-21 Amendment 13 (January 1982)
(c) Verification of function allocation. The purpose of this analysis is to verify the integration of humans and machines at the work station level. This will be accomplished by analyzing work station design, operations sequence, and work station links. (d) Validation of function allocation. The purpose of this process is to validate the control room configuration design against the functional requirements established from opera-tions analysis. This will be accomplished by simulating operations with a control room mock up or a control room simulator, or alternatively, by computer program synthesis of control room design and operations. (3) Document control room design specification. This document will contain the control room configuration design as well as design details for each individual work station. The results of the validation of the total control room design specifi-cation will be reported in the FSAR. A.9-22 O Amendment 13 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT N--) Plant Safety Parameter Display Console (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. 2 (iv) Provide a plant safety parameter display console that will display to operators a minimum set of parameters defining the safety status of the plant, capable of dis-playing a full range of important plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded. (I.D.2) RESPONSE TO 10 CFR 50.34(f)(2)(iv) N- / The Pebble Springs design will include a Safety Parameter Display System (SPDS) that will display.to operating personnel a minimum set of parame- + ters or derived variables representative of the safety status of the plant. The system will have the capability of displaying the full range
, of important plant parameters and data trends on demand. The system will also indicate when plant parameters are approaching or exceeding
< process limits. The SPDS will be designed in accordance with NUREG-0696 (February 1981) and will be located within the control room area based on an analysis I of the operator's needs and a functional analysis of the use of the SPDS. 4 This analysis will be an integral part of the final selection of the SPDS design and will be provided for NRC review as part of the control I room design review [see response to 10 CFR 50.34(f)(2)(iii)]. SPDS displays will also be provided in the TSC and the EOF. 10 1 A.9-23 Amendment 13 (January 1982) l
The SPDS will contain the minimum parameter set from which plant safety status can be determined quickly and accurately at a single display location. The plant functions presented will include, but not be limited to: reactivity control, reactor core cooling and heat removal from the primary system, RCS integrity, radioactive effluent control and Contain-ment integrity. The SPDS is expected to be a computer-based system of high quality and reliability and will be capable of displaying the full range of important plant parameters. It will be capable of functioning properly in the environments that are present during transient and accident conditions. liuman factors engineering will be incorporated into the SPDS design to enhance the ability of control room personnel in evaluating the safety status of the plant. Displays will be as simple as possible and will incorporate. human factor considerations into grcuping of parameters, patterns, and coding techniques to assist the operator in detection of unsafe operating conditions. Time-rate-of-change for selected parameters will be provided through trending or derivation in a manner designed to both optimize operator process communication and allow flexibility in the selection of variables for display. O Amendment 13 A.9-24 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT C\ V Safety System Status Monitoring (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (v) Provide for automatic indication of the bypassed and operable status of safety systems. (I.D.3) RESPONSE TO 10 CFR 50.34(f)(2)(v) The Pebble Springs design includes automatic indication of the bypassed and inoperable status of safety systems in conformance with Regulatory Guide 1.47 as described in PSAR Section 7.5. To the extent practical, inputs to the status monitoring system will be direct measurements of the desired variable. O() A.9-25 Amendment 13 (January 1982)
t i NRC 10 CFR 50.34(f) REQUIREMENT , ' Reactor Coolant System Vents (2) To satisfy the following requirements, the application shall provide l sufficient information to demonstrate that the required' actions ! will be satisfactorily completed by the operating license stage.
; This information is of the type customarily required to satisfy-~
10 CFR'50.35(a)(2) or to address unresolved generic safety issues. 4 ! (vi)~ Provide the capability of high point venting of noncondensible gases from the reactor. coolant system, and other systems that may be required to maintain adequate core cooling. Systems to-
, achieve this capability shall be capable of being operated-from the control room and their operation shall not lead to an
~ unacceptable increase in the probability of loss-of-coolant , 1 1-accident or an unacceptable challenge to containment integrity. (II.B.1) O ] . RESPONSE TO 10 CFR 50.34(f)(2)(vi) i 4
- High. point RCS vents, remotely operated from the control room, will be provided for Pebble Springs in accordance with SRP Section 5.4.12 (NUREG-0800). Vents will be provided atop each RCS hot leg, the reactor l i
vessel head, and the pressurizer. Redundant, independently powered i parallel valves will-be provided for each vent path, with power supplied ; from the Class lE ESF buses. i f' Flow rate will be limited by fixed orifices. These orifices are sized
- such that this flow rate is well below the flow capability of the normal
- makeup system and-thus well below the 10 CFR 50, Appendix A definition of l a LOCA. LOCA analysis to demonstrate conformance to 10 CFR 50.46 is
, therefore not required. I j Discharge paths-for all vents will be routed to a location which promotes mixing with the Containment environment. A.9-26 Amendment 13 (January 1982) i i
, - - ,m,....- - - - , - ,,,m, ,_-____-,,_-m,__, ,,,.-,_-m-mr.-,.--,-..__,,_,- -,m,.m.,,,-,,,,,,.,,,,e,,,.--,,--,_,_,.--
Assurance that a vent path, once opened, can be isolated is provided by series solenoid valves in each parallel path which will close upon loss of power. Direct position indication of all valves will be pro-vided in the control room. The potential for inadvertent operation will be reduced procedurally by removing power from the vent valves during normal operation. O l l O L:nendment 13 A.9-27 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT Plant Shielding ~to Provide Access to Vital Areas and Protect Safety Equipment ' for Post-Accident Operation (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2)'or to address unresolved generic safety issues. (vii) Perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain TID 14844 source term radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipment from the radiation environment. (II.B.2) RESPONSE TO 10 CFR 50.34(f)(2)(vii) A preliminary shielding design review has been performed for the Pebble Springs Nuclear Plant considering the current shielding design as pre-sented in Section 12.1 of the PSAR. This preliminary review was intended to clarify the assumptions to be utilized in future quantitative post-accident shielding reviews and to point out areas of potential shielding impact. A detailed shielding study will be undertaken in accordance with NUREG-0737, Item II.B.2, as part of the detailed design review process and will finalize any shielding design modifications based on post-accident dose rates in areas where access is required. Documentation will be pro-vided in the FSAR. In addition, the environmental qualifications of equipment required for accident mitigation will be studied further to ensure their proper functioning in a post-accident environment. O A.9-28 Amendment 13 (January 1982)
Systems Those systems important to accident mitigation or post-accident recovery are identified in Part B of Table A.9-8 [see response to 10 CFR 50.34(f) (2)(xxvi)]. These systems are postulated to contain highly radioactive materials. The use of temporary systems for post-accident cleanup will obviate the need for the normal liquid and solid radwaste systenc. The maieup and purification system (except the HPI portion) and the boron recovery system are automatically isolated in the event of a LOCA and are not required for accident mitigation or post-accident recovery. Further evaluation is needed to confirm that these latter systems are also isolated and unnecessary for post-accident functioning for non-LOCA events that result in postulated (nonmechanistic) core damage. Gaseous activity will be relieved to the Containment atmosphere through the use of the RCS vents such that the gaseous radwaste system will not contain high-activity noble gases post accident. These systems and others that are not expected to contain post-accident high-activity sources are identified in Part A of Table A.9-8 (see response to 10 CFR 50.34(f) h (2)(xxvi)]. Source Terms Radioactive source terms equivalent to those recommended in Regulatory Guides 1.4 and 1.7 and those specified in TID 14844 will be utilized in future detailed shielding reviews. The fractions of the nuclear core activity assumed to be released during the accident are presented in Table A.9-3. In applying these inventory releases to activity concentra-tions, dilution volumes and decay times will be considered. Airborne activity in the Auxiliary Building is not considered in the post-accident shielding review. Vital Areas The shielding reviews are intended to assure that the accident mitigation function is not degraded due to high radiation dose rates existing in Amendment 13 A.9-29 (January 1982) L
plant areas requiring access for ESF system operation or for activities
/^'S in non-ESF areas that are necessary for the recovery process. All areas \Ul (and pathways to areas) requiring operator access to aid in the mitiga-tion of or recovery from an accident are classified as vital areas.
Table A.9-4 lists plant areas important to post-accident recovery as well as the function (s) to be performed. Not all areas listed are clas-sified as vital areas since post-accident access may not be required. High activity source terms are also indicated in Table A.9-4 as applicable. The inherent redundancy of ESF systems allows the accomplishment of required safety functions in spite of many types of component malfunctions. Therefore, direct access into rooms containing highly radioactive compo-nents for maintenance purposes is not considered a necessary function in the short-term post-accident condition. Personnel Exposure Guidelines The shielding review is performed to verify that radiation exposures received by plant operators do not exceed the 10 CFR 50, Appendix A, V General Design Criterion 19 limits of 5 rem whole body or its equivalent to any part of the body during a post-accident situation. A guideline dose rate for vital areas requiring continuous occupancy (eg, control room, onsite TSC) is 15 mR/hr or less to ensure that General Design Criterion 19 limits will not be exceeded. In general, the dose rate for vital areas requiring infrequent access (Table A.9-4) should not exceed 5 rem /hr at the time of access, as a guideline dose rate, such that operator access would be possible but the frequency and duration of access would be limited. Cumulative operator doses will include the contribution received during ingress to and egress from vital areas and areas requiring continuous occupancy. Personnel Access Conclusions Table A.9-4 summarizes the results of the preliminary design review. Certain plant areas are expected to require additional shielding as noted. In the corridor areas of the Auxiliary Building adjacent to the
%)
A.9-30 Amendment 13 (January 1982)
HPI portion of the Makeup and Purification, Containment Spray and Decay Heat Removal Systems, dose rates well in excess of 5 rem /hr can be expected in the worst case short-term post-accident condition. However, the review of these systems has established that local access is not essential by virtue of system remote operation and instrumentation capabilities. At present, no added permanent shielding is anticipated in this area. Areas which are expected to require some additional permanent shielding are the primary sample lab and counting room area, and areas near the Containment isolation valves for external hydrogen recombiners [if installed; see response to 10 CFR 50.34(f)(3)(vi)] and for the hydrogen analyzers. For the sampling area, the extent of required additional shielding will depend on the method of obtaining the high activity samples and on whether or not the current area is utilized or an emergency sample station is provided. The existing counting room must be adequately shielded and ventilated to maintain an acceptable background level for analysis work (currently estimated at 1 mR/hr); alternatively, a separate facility may be more appropriate [see response to 10 CFR 50.34(f)(2)(viii)]. For the Containment isolation valves, sufficient shielding is necessary to allow manual valve positioning in the event that the power-operated valves of either train A or B fail to function. No additional permanent shielding is expected to be added to other plant areas since doses received by operators accessing vital areas can be controlled by the use of: (1) Post-accident operation procedures (2) Equipment remote control and monitoring capability (3) Decay time prior to access (4) Temporary shielding. O Amendment 13 A.9-31 (January 1982)
As design evolves and post-accident operatlng procedures are developed, (O) certain additional system changes may be found necessary to enhance system remote control and monitoring, such as additional remote valve operators. Equipment Qualification Equipment will be studied to determine the extent of necessary equipment /' system changes, if any, that may be required to ensure that proper environmental criteria are met during the post-accident period. Source terms, as recommended in Table A.9-3, will be utilized for determining the maximum anticipated integrated doses for equipment, except the initial Containment atmosphere source term will be the same as Source "B" in Table A.9-3. N n V A.9-32 Amendment 13 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT i(D
)
. Post-Accident Sampling !
(2) -To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily cornpleted by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (viii) Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID 14844 source term radioactive materials without radiation exposures to any individual exceeding 5 rem to the whole body or 75 rem to the extremities. Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree'of core damage (eg, noble U.ses, iodines and cesiums, and non-volatile isotopes),
/N hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations. (II.B.3)
RESPONSE TO 10 CFR 50.34(f)(2)(viii) Sampling System The current Pebble Springs process sampling system design provides remote sampling facilities and the capability for sacpling and analyzing various fluid systems during normal plant power and shutdown operation. The pro-cess sampling and analyses systems are described in PSAR Sections 9.3.2 and 11.4. A design and operational review of the Reactor Coolant and Containment atmosphere sampling systems, against the requirements of NUREG-0737, Item II.B.3, will be performed to determine the capability of personnel to promptly obtain a sample under accident conditions. Design features or shielding will be provided to assure that samples can be obtained and V j A.9-33 Amendment 13 (January 1982)
analyzed without incurring a radiation exposure to any individual in excess of 5 and 75 rems to the whole body or extremities, respectively. This review will be carried out in conjunction with the detailed plant shielding review [see response to 10 CFR 50.34(f)(2)(vii)] with due consideration given to the personnel occupational exposure criteria denoted above. Post-accident sampling design will include the capability for cbtaining grab samples of Containment air for hydrogen analysis, which is in addi-tion to the existing Containment Hydrogen Monitoring System [see response to 10 CFR 50.34(f)(2)(xvii)]. Capability currently exista for obtaining either undegassed or degassed reactor coolant samples. Radiological and Chemical Analysis Facility A design and operational review of the radiological and chemical analy-sis facilities, against the requirements of NUREG-0737, Item II.B.3, will also be performed to determine the capability to promptly analyze and quantify certain radionuclides that are indicators of the degree of core damage, hydrogen in the Containment atmosphere, dissolved gases,. boron, and chloride concentrations. Current plant design provides for a gamma spectroscopy system using Ge-Li and Na1 detectors (PSAR Section 11.4.3.3). This system is located in the plant counting room and may be rendered unusable during an accident by high background radiation levels due to direct radiation from airborne contamination and adjacent compon.nts and + ! piping. Therefore, the following alternatives are being cinsidered: l l (1) Upgrading the existing counting room design (access, ventila-tion, shielding) to provide low background radiation levels during accidents l i l (2) Providing a fixed area onsite in addition to and separate l from the existing counting room I l Amendment 13 A.9-34 l (January 1982)
t (3) Providing a mobile facility onsite in addition to and separate , from the existing counting room. i The radiological analysis capabilit/ will include provision to identify and quantify noble gases, iodines, cesium and non-volatile isotopes to levels corresponding to TID 14844 source terms. The chemical analysis 4 capability will also consider the presence of this source term. ? The final design of the sampling systems and analysis facilities will be presented in the FSAR. i t 1 i i l e J 4 i l i i ;. I i I I i i 1 A.9-35 Amendment 13 i (January 1982)
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NRC 10 CFR 50.34(f) REQUIREMENT
- Hydrogen Control i .
i (2) To' satisfy the following requirements, the application shall' provide .; 1 sufficient information to demonstrate that'the required actions j will be satisfactorily completed by the operating license stage. [- This information is of the type customarily equired to satisfy j 10 CFR 50.35(a)(2) or to address unresolved generic safety issues.- 4 j (ix) Provide a. system for hydrogen control that can' safely accommo-date hydrogen generated by the e_quivalent of a 100 percent
- euel-clad metal water reaction. -Preliminary design informa-i 1 tion on the tettatively preferred system option of those ;
i being evaluated in paragraph -('1)(xii) of this section is 1 sufficient at 'the construction permit stat e. The hydrogen' i 1 j control system and associated systems shall provide, with ! reasonable assurance, that: (II.B.8) 1 (A) Uniformly distributed hydrogen concentrations in the containment do not exceed 10 percent during and follow- ! ing an accident that releases an equivalent amount of i hydrogen as would be generated from a 100 percent fuel clad metal-water reaction, or that the post accident I atmosphere will not support hydrogen combustion. 1 ! (B) Combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could
- cause loss of containment integrity or loss of appro-ll l priate mitigating features. ,
I
- ) (C) Equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment j- integrity will perform its' safety function during and after being exposed to the environmental conditions attendant with the release of hydrogen generated by the >
f A.9-36 Amendment 13 (January 1982) l i a
equivalent of a 100 percent fuel-clad metal water reaction including the environmental conditions created by activation of the hydrogen control system. (D) If the method chosen for hydrogen control is a post-accident inerting system, inadvertent actuation of the system can be safely accommodated during plant operation. RESPONSE TO 10 CFR 50.34(f)(2)(ix) Pebble Springs will include a hydrogen control system capable of handling hydrogen generatad from a 100 percent active fuel-clad metal water reac-tion. The preliminary concept being considered will consist of a distri-buted ignition system with igniters located throughout Containment and in areas of potential pocketing so as to reasonably assure that uniformly distributed hydrogen concentrations will not exceed 10 percent during and following an accident. PGE will incorporate the results of industry and NRC research programs that are applicable to the selected hydrogen control system. Within 2 yr after receipt of the CP, design details describing the hydrogen control system will be provided to the NRC for review as part of the hydrogen con-
~
trol systems evaluation performed in response to 10 CFR 50.34(f)(1)(xii). These design details, including test data and analyses, will provide reasonable assurance that the hydrogen control system will perform in the manner as required by the above NRC rule. (A) Assuming that igniters are selected as the final design, the igniter system will be designed and the igniters will be strate-gically located such that the hydrogen generated from a 100 percent active fuel clad metal-water reaction will be ignited in a manner that provides reasonable assurance that uniformly distributed hydrogen concentration will not exceed 10 percent during and following an accident. Amendment 13 A.9-37 O (January 1982)
.(B)2 The relatively open configuration of the Containment incorporated
) N into the Pebble Springs design generally serves to. preclude pocket-l [d ing of hydrogen. Assuming igniters are selected as the final j design, igniters and/or vent paths will be strategically placed i throughout the Containment and local areas which have the potential j for pocketing hydrogen. These igniters and/or vent paths will l provide reasonable assurance that uniformly distributed hydrogen
- concentration in any area which has the potential for pocketing does not exceed 10 percent during and following an accident.
(C) The. equipment required to maintain Containment integrity and remove the-heat generated by a degraded core will be designed and 1 qualified to perform during and after exposure to cl.; environmental j conditions created by activation of the distributed ignition i system (if this method is selected). The design of these systems will specifically include use of ] l Containment temperature profiles that result from the postulated hydrogen burn events as input to the analyses or tests which will ; demonstrate equipment survivability and/or qualification. .
; 'The location of components associated'with these systems and method of protection (if required) will be described in the FSAR.
] (D) Inerting as a hydrogen control measure is not proposed for the Pebble Springs Containment design. Therefore, this item is not applicable. f i The final design for hydrogen control will be described in the FSAR. 2 i i i i A.9-38 Amendment.13 ;
.(January 1982) i i
_ . , _ . , , . . , . - - . _ . . _ . , . _ . _ _ _ , . . - _ . , ~ , , . . . , . _ . , _ ~ . _ . , _ . . . . , , . . . , _ . . _ _ , , . _ . , , , _ . - . . _ , , - . . . , . , . _ _ . . .
NRC 10 CFR 50.34(f) REQUIREMENT p\~s Testing Requirements (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (x) Provide a test program and associated model development and conduct tests to qualify reactor coolant system relief and safety valves and, for PWR's, PORV block valves, for all fluid conditions expected under operating conditions, transients and accidents. Consideration of anticipated transients without scram (ATWS) conditions shall be included in the test program. Actual testing under ATWS conditions need not be carried out' until subsequent phases of the test program are developed. (II.D.1) (V} RESPONSE TO 10 CFR 50.34(f)(2)(x) Pebble Springs will implement the results of the industry testing neces-sary to qualify the RCS power-operated relief, safety and PORV block valves for all fluid conditions expected under operating conditions, transients and accidents. A generic program has been developed by EPRI to verify the operational characteristics of PWR safety and relief valves and to provide assurance that these systems can perform as required to prevent overpressurization of the primary coolant boundary. The " Program Plan for the Performance Testing of PWR Safety and Relief Valves", Revision 1, July 1980 has been submitted to the NRC staff. The experimental data, together with foreign relief valve test results, will be used to validate a computational methodology for assessing the hydraulic / structural performance of PWR safety / relief valve discharge piping systems on a plant unique basis. O b A.9-39 Amendment 13 (January 1982)
A test program will be performed in which safety and relief valves of the same model to be utilized for Pebble Springs are tested under fluid conditions which are representative of those calculated to occur during anticipated operational transients and postulated accident sequences specific to the plant. A combined test and analysis program will be performed to evaluate the adequacy of analytical methods utilized for the Pebble Springs safety and relief valve discharge piping response. The effect of as-built relief and safety valve discharge piping on valve operability will be accounted for, and the discharge piping and supports will be designed for the loads resulting from expected operating condi-tions for design basis transients and accidents. O Amendment 13 A.9-40 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT
-.)
Relief and Safety Valve Position Indication (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xi) Provide direct indication of relief and safety valve position (open or closed) in the control room. (II.D.3) RESPONSE TO 10 CFR 50.34(f)(2)(xi) Reactor Coolant System PORV and safety valves will be provided with positive position indication in the control room. This will be derived from position switches or acoustic acaitoring devices located on each valve or discharge pipe. Position indication for PORVs will be provided by Class IE transducers, signal conditioning hardware and indicator lights on the plant control console. Safety valve position will be either Class 1E or non-Class lE provided with a backup indication, such - as discharge pipe temperature or pressurizer relief tank level. Indica-tion will also be on the plant control console. Alarms for both PORY and safety valve position will be provided. The operator will be provided with appropriate procedures to be followed in the event of a stuck-open valve. I O v A.9-41 Amendment 13 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT ps ( Auxiliary Feedwater System Automatic Initiation and Flow Indication (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that'the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xii) Provide automatic and manual auxiliary feedwater (AFW) system initiation and provide auxiliary feedwater system flow indication in the control room. (Applicable to PWR's only.) (II.E.1.2) RESPONSE TO 10 CFR 50.34(f)(2)(xii) The current Pebble Springs AFS design complies fully with the require-ments of NUREG-0737, Item II.E.1.2. PSAR Section 6.7 discusses the AFS b ( ,/ and provides cross-references to the appropriate parts of Chapter 7 which describe the initiating circuitry. AFS Initiation The AFS automatically provides feedwater to the steam generators in response to any of the following events: (1) Loss of normal onsite power and both preferred offsite power sources; ie, loss of power to the 4.16-kV ESF buses. (2) Low water level in either steam generator [via the Essential Control and instrumentation (ECI)]. (3) Engineered Safety Feature Actuation System (ESFAS) actua-tion. The ESFAS automatically initiates the AFS, in A.9-42 Amendment 13 (January 1982)
i conjunction with the ECI, in the event of low RCS pres-sure, or high Containment pressure, or low steam gener-ator pressure. A logic subsystem called Feed Only Good Generator (FOGG) is incorporated in the ESFAS to ensure proper auxiliary feedwater distribution to the steam generators in the event of a main steae line break. (4) Tripping of both main feedwater pumps. This is a non-Class lE anticipatory (of low steam generator water level) initiation of auxiliary feedwater. No credit is taken for this trip in the PSAR Chapter 15 accident analysis. Non-Class lE main feedwater pump trip signals are buffered from the auxiliary feedwater control cir-cuitry by Class lE isolation devices. (5) Loss of reactor coolant pumps. (6) Reactor trip on neutron flux-to-main feedwater flow. The AFS can also be manually actuated and controlled either from the control room or from the auxiliary shutdown panel. The automatic initiation signals and circuits are designed so that a single failure will not result in the loss of AFS function. The automatic initiating signals and circuits to the four AFS flow trains (Train A and Train B for each steam generator) meet the criteria for independence and redundancy for a single failure. These automatic signals and circuits are derived from two fully independent safety-related channels, power sources, logic cabinets and manual control stations. The automatic and manual initiating signals and circuits are powered from independent and redundant ESF buses and are designed in accordance with Class lE requirements. O Amendment 13 A.9-43 (January 1982)
The design provides for_ testing capability of the initiating signals
} and circuits. Initiating signals such as ESFAS and loss of power can be manually tested and initiated from the main control room panels.
. The ECI circuits _ for auxiliary feedwater valves are testable at the ECI panels, which are located outside the-control room. .The main feedwater pump trip signal can be locally initiated by tripping the local non-Class IE pressure switches. These signals can be simulated to test individual. components of the AFS. Concurrent aatomatic initiation of auxiliary feedwater to the steam generator or concurrent automatic isolation of the steam generator will not be degraded, as these actua-tions will override any testing feature.
Normally, the AFS responds to an automatic actuation signal, but the AFS can be manually started and controlled at any time from the control room or from the auxiliary shutdown panel. Steam generator level is normally controlled by the ECI by modulating the AFS pump discharge control valve. Single failure of a manual control circuit will not cause loss of function of the AFS. The AFS consists of two 4.16-kV motor-driven and two diesel-driven auxiliary feedwater pumps and associated valves. All AFS pumps and power-operated valves are controlled by Class 1E power. The motor-driven pumps are designed to be automatically sequenced onto their' respective ESF buses and to inject water into the steam generator within 40 sec of the initiating signal (s). -The standby diesel generators are adequately sized to accommodate these loads. The diesel-driven pumps and all AFS discharge valves are not required to be sequenced. The AFS design does not use the instrument air system for operation. The AFS control circuits are designed with main control room control board-mounted switches as backup features to the automatic initiating signals. This design complies with the requirements of IEEE Standard 279-1971, Section 4.17. Therefore, failure of automatic initiating signals and circuits will not cause the loss of manual system initia-f tion capability from the control room. 1 A.9-44 Amendment 13 { (January 1982)
AFS Fluw Indication The current AFS instrumentation design complies fully with the criteria of NUREG-0737 by providing safety grade auxiliary feedwater flow indi-cation. PSAR Section 7.5 describes this paramater, which is included in the Post-Accident Monitoring System (PAMS). The present desigr. provides safety grade Class lE auxiliary feedwater flow indication for each of the four auxiliary feedwater flow trains to the steam generators. Flow indicators are located at the control room boards and at the auxiliary shutdown panels. The auxiliary feedwater flow instrument channels are powered from the Class 10 battery-backed 120-V vital instrument ac buses. The A and B auxiliary feedwater flow instrument channels for each steam generator are powered by the Channel A and B vital instrument ac buses, respec-tively. Tne design meets the guidance for emergency power diversity set forth in Branch Technical Position ASB 10-1 of SRP Section 10.4.9. O O Amendment 13 A.9-45 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT p-, i I Reliability of Power Supplies for Natural Circulation (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.25(a)(2) or to address un esolved generic safety issues. (xiii) Provide pressurizer heater pover supply and associated motive and control povar interfaces sufficient to establish and maintain natursi circulation in hot standby conditions with only onsite power available. (Applicable to PWR's only.) (IT.E.3.1) RESPONSE TO 10 CFR 50.34(f)(2)(xiii) /'~'s i Two of the 10 pressurizer heater groups will receive power, from either
\/ /
the offsite power source or the onsite standby diesel generators, via redundant ESF (ie, Class lE) 480-V motor control centers. The power supplies and controls of one heater group are redundant and independent from those of the other heater group. Adequate separation exists between the cables to the two heater groups as the heaters are on dif-ferent sides of the pressurizer vessel and at different elevations. Each heater group provides enough capacity to establish and maintain natural circulation at hot standby conditions. The two heater groups can be manually connected to their respective buses from the control room as soon as sequencing of ESF loads onto the 4.16-kV buses has been completed (approximately 1 min). The standby diesel generators have been sized to accommodate these heaters without shedding any ESF loads. Therefore, no instrumentation or operational criteria is needed to prevent overloading a diesel generator. Although not necessary, the standby diesel generators have adequate design margin (~'n capacity to provide power to the heaters concurrent with the ESF loads \ ) v A.9-46 Amendment 13 (January 1982)
required for a LOCA. Procedures and training shall be established, prior to operating license issuance, to make the operator aware of when and how the required pressurizer heaters should be connected to the ESF buses. Connection of the pressurizer heaters to the ESF buses can be completed within the minimum time period for which heat input from the pressurizer heater is required for maintenance of natural circulation conditions. Motive and control power devices which interface with the ESF buses are qualified in accordance with Class lE requirements. O l I l 1 O Amendment 13 A.9-47 (January 1982)
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NRC 10 CFR 50.34(f) REQUIREMENT Isolation Dependability (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xiv) Provide containment isolation systems' that: (II.E.4.2) (A) ensure all non essential systems are isolated auto-matically by the containment isolation system, L .1 (B) for each non essential penetration (except instru-ment lines) have two isolation barriers in series, (C) do not result in reopening of the containment isolation valves on resetting of the isolation signal, , (D) utilize a containment set point pressure for initiating containment isolation as low as is compatible with f normal operation, 4 (E) include automatic closing on a high radiation signal for all systems that provide a path to.the environs. RESPONSE TO 10 CFR 50.34(f)(2)(xiv) 4 Prior to the TMI-2 accident, PGE decided to incorporate the improved B&W ESFAS-II into plant design to replace the ESFAS-I design currently described in PSAR Section 7.3. A generic description of the ESFAS-II design is provided in Section 7.3 of B-SAR-205. The Containment Isola-I tion System (CIS) is described in PSAR Section.6.2.3 and complies with the recommendations of SRP Section 6.2.4. j A.9-48 Amendment 13 (January 1982)
The BEN safety grade ESFAS-II design includes Containment isolation initiation on diverse parameters in conformance with SRP Section 6.2.4. Parameters for the Containment isolation function are: (a) low reactor coolant pressure, and (b) high Containment pressure. PGE currently defines essential systems as those systems or portions of systems which penetrate Containment and which must be operating or capable of operation without operator action in order to: (1) Mitigate the consequences of an accident, or (2) Maintain the plant in hot standby, or (3) Prevent degradation of the reactor coolant pressure boundary. Essential system Containment isolation valves (except where special circumstances dictate otherwise) are normally open or automatically opened on actuation of the ESFAS and may be selectively closed by an operator after it is established that the use of the essential system will not be required. Table A.9-5 lists the essential systems for Pebble SpringsI *l. Non-essential systems are those systems or portions of systems which penetrate Containment and which do not meet the conditions for essential systems noted above. Containment isolation valves for non-essential systems are either normally closed or automatically closed on actuation of the ESFAS. Table A.9-6 lists the non-essential systems for Pebble SpringsI *l. j [a] The next revision to Regulatory Guide 1.141 will contain further I guidance on the classification of essential versus non-essential systems which will be used by PGE to reevaluate the categorization in Tables A.9-5 and A.9-6. l Amendment 13 A.9-49 (January 1982) l
(A) All non essential systems use either administratively controlled manually closed valves (or flanges) or are automatically isolated b by the ESFAS. (B) Each non-essential penetration (except instrument lines) has two isolation barriers in series (see PSAR Table 6.2-23). (C) Resetting the isolation signal to a group of Containment isola-tion valves will not cause the valves to automatically reopen. A distinct second deliberate step to change the valve position at the normal control switch is necessary to cause the valves to reopen. (D) Tha Containment setpoint pressure that initiates Containment isola-tion for non-essential penetrations will be set at the minimum compatible with normal operating conditions. The value of the revised setpoint and its basis will be discussed in the FSAR. In determining the revised setpoint for initiating Containment isola- {'} V tion, the Containment pressure history during normal operation for similar operating plants will be taken into consideration. The pres-sure setpoint selected will be far enough above the maximum observed (or expected) pressure inside Containment so that inadvertent Contain-ment isolation does not occur during normal operation from instrument drift or fluctuations due to the accuracy of the pressure sensor. If a margin in excess of 1 psi above the maximum expected Containment pressure is utilized, justification will be provided. (E) Safety grade Containment high airborne radiation (separate from ESFAS) is another parameter used in addition to an ESFAS signal for automatic closing of the Containment purge supply aad exhaust systems and the Containment drain sump pump discharge line. I A.9-50 Amendment 13 (January.1982)
NRC 10 CFR 50.34(f) REQUIREMENT l v Purging (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xv) Provide a capability for containment purging / venting designed to minimize the purging time consistent with ALARA principles for occupational exposure. Provide and demonstrate high assurance that the purge system will reliably isolate under accident conditions. (II.E.4.1) RESPONSE TO 10 CFR 50.34(f)(2)(xv) The Pebble Springs Containment purge systems are described in PSAR Sections 6.2.2.2 and 6.2.2.3. The purge systems are desigt.ed to reduce the concentration of radioactive noble gases, halogens and particulates in the Containment atmosphere as required during normal plant operations in order to allow access while limiting personnel exposures to less than the dose limits of 10 CFR 20 'or occupational exposure and Appendix I to 10 CFR 50 for gaseous radioactivity releases. The Containment purge isolation valves are designed to meet the criteria of Branch Technica! Position CSB 6-4, Section B, and the NRC Staff Interim Position of October 23, 1979 for valve operability. The purge system design was found acceptable by the NRC as discussed in Section 15.6 of Supplement No. 5 of the Pebble Springs Safety Evaluation Report (NUREG-0013). 9 m (vI A.9-51 Amendment 13 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT n Design Evaluation (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily. required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xvi) Establish a design criteria for the allowable number of actua-tion cycles of the emergency core cooling system and reactor protection system consistent with the expected occurrence rates of severe overcooling events (considering both antici-pated transients and accidents). (Applicable to B&W designs only). (II.E.5.1) RESPONSE to 10 CFR 50.34(f)(2)(xvi) - Analyses have been performed for other B&W 205 fuel assembly (FA) plants to address the above requirement. The methods used for these analyses are applicable to the Pebble Springs. Nuclear Plant and the results are generally applicable. These analyses will be reviewed for changes due to Pebble Springs specific requirements and specifications and will be modi-fied as appropriate prior to the operating license stage. A description of the analyses to be performed to identify the most severe overcooling events is presented below. This is followed by a description of the plan to justify the actuation cycles of the Emergency Core Cooling System (ECCS) and Reactor Protection System (RPS). Finally, Pebble Springs ; changes either being considered or planned to reduce primary system sen-sivity are addressed. Identify the Most Severe Overcooling Events These analyses will include more than one transient type to address dif-ferent frequency of occurrence classifications and to assure the most A.9-52 Amendment 13 (January 1982)
severe cases are indeed included in the evaluation. A qualitative assess-ment of possible nonmitigative operacor actions in the 0- to 10-min time frame will be provided. This assessnent will provide an indication of what operator action is anticipated daring the initial phases of an over-cooling transient. The ana?.yses will identify the frequency of the RPS, ESFAS, and operator action for mitigation of the transient. Details of the initial condi-tions, computer codes and basic assumptions used in the analyses will be provided. Maximum RCS coolant inventory shrinkage results from a decrease in the pressure and temperature of the coolant at a maximum rate, without a com-pensating coolant makeup addition. Several sensitivities and differing conditions will be analyzed to provide greater insight into the steam void formation and collapse which could occur and its subsequent effect on core cooling. These additional studies will be performed on the steam line break since this accident is expected to result in RCS voiding; whereas it will be shown the limiting moderate frequency event analyzed does not produce voiding as a result of RCS cooling inventory shrinkage. In selecting the limiting anticipated transient, safety analysis report and operating plant overcooling events will be reviewed. The overfeed by main feedwater after reactor trip is believed to be the most overcooling transient at this time. A B&W certified computer code, such as TRAP 2, will be used in the above analyses. This computer code is a nodal type, digital simulation capable of handling rapid overcooling transients that may result in two phase fluid conditions in the RCS. The types of overcooling events to be considered include: (a) those which constitute the inititating event, (b) those which result from single fail-ure following any initiating event, and (c) those which are made more severe from single failures following the initiating overcooling event. O Amendment 13 A.9-53 (January 1982)
Justify the Design Criterion for the Actuation Cycles of the ECCS and RPS
-p Required ECCS and RPS actions necessary to protect the core will first be summarized. No operator action will be assumed within 10 min for mitigation in the analysis.
The types of transient cycles to be used in evaluating the acceptable number of design cycles is provided in PSAR Table 5.1-8. The number of design cycles for each transient type will be factored into the analysis. This data is the basis on which the stress evrluation is performed for the plant and will be contained in the plant Technical Specifications. The adequacy of the number of design cycles can also be inferred from operating plant data. The actual arrival rate for RPS and ESFAS actuation
, to date on plants of B&W design to the rates allowed by the design basis will be compared in the analyses.
Recommended Changes to Reduce Primary System Sensitivity Much of the concern expressed about the " sensitivity" of the B&W Once Through Steam Generator (OTSG) PWR design is based on the operational experiences with the currently operating 177 FA plants and particularly that experience accumulated since the accident at TMI-2. It is important to recognize that the normal evolution of design that has occurred on the generation of B&W plants now under construction as a result of new regulatory requirements, improvements in the state-of-the-art in hardware and the feedback of operating experience has resulted in the incorporation of several new features. These features serve to improve the reliability of the systems and equipment and thereby reduce the probability of the challenges to the safety systems, to improve the response of the NSS to those events that do occur, and to provide better capability to mitigate < the events which occur. These changes include: O A.9-54 Amendment 13 (January 1982)
(1) Addition of a two channel, Class lE ECI system to control auxiliary uedwater and provide post-accident monitoring instrttment ation for the operator (2) Initiation uxiliary feedwater by the IEEE-279 ESFAS (3) Addition of FOGG logic to the ESFAS to help ensure that auxiliary feedwater is delivered to the intact steam generator following secondary system breaks (4) Moving the pressurizer level sensing taps to the top and bottom heads of the pressurizer to expand the range of level indication (5) Raising the level of the OTSG with respect to the level of the reactor. Notwithstanding the above, design studies of various aspects of the B&W NSS will be conducted with the objective of minimizing, to the extent practicable, the actual number of design basis transients. These studies will be divided up into categories with the following objectives: (1) Retain the basic design operating characteristics of the OTSG PWR (2) Improve the reliability of the systems whose failure can lead to overcooling transients (thereby enhancing plant availability and reducing the frequency of challenges to the safety systems) (3) Further improve the renponse of the NSS to the transients which do occur (4) Improve the capability to mitigate these transients. O Amendment 13 A.9-55 (January 1982)
NRC 10 CFR 50.34(f) REQUIREME?Tr
,~
t I
\-# Additional Accident Monitoring Instrumentation (2) Ta satisfy the following requirements, the application shall provide suf ficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.
This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to addrcss unresolved generic safety issues. (xvii) Provide instrumentation to measure, record and readout in the control room: (A) containment pressure, (B) contain-ment water level, (C) containment hydrogen concentration, (D) containment radiation intensity (high level), and (E) noble gas effluents at all potential, accident release points. Provide for continuous sampling of radioactive iodines and particulates in gaseous effluents from all potential accident release points, and for onsite capa- ,/S bility to analyze and measure these samples. (II.F.1) 1 4
%J RESPONSE TO 10 CFR 50.34(f)(2)(xvii)
(A) A continuous indication of Containment pressure will be provided in the control room. Measurement and indication capability will include the range of three times the Containment design pressure to -5 psig. The Containment pressure monitor will meet the cri-teria of NUREG-0737. (B) A continuous indication of Containment water level will be provided in the controt room. A narrow range instrument will be provided to cover the range from the bottom to the top of the Containment sump. A wide range instrument will be provided to cover the range from the bottom of the Containment to the elevation equivalent to a 600,000 gal capacity. The Containment water level monitor will meet the criteria of NUREG-0737.
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( ) \_/ A.9-56 Amendment 13 (January 1982)
(C) A continuous indication of hydrogen concentration in the Containment atmosphere will be provided in the control room. Measurement capa-bility will be provided over the range of 0-to-10 percent hydrogen concentrar. ion under both positive and negative ambient pressure. The Containment hydrogen monitor will meet the criteria of NUREG-0737. (D) In-Containment high-range radiation monitors will be provided. Such monitors will be redundant, physically separated and qualified to function in an accident environment. The Containment high-range radiation monitors will meet the criteria of NUREG-0737, including the specifications of Table II.F.1-3. (E) Noble gas effluent monitors will be installed with an extended range designed to function during accident conditions as well as during normal operating conditions. Multiple monitors will be provided to cover the ranges of interest. These monitors will meet the criteria of NUREG-0737, including the specifications of Table II.F.1-1. This instrumentation is known to be commercially available and space is available for transmitter locations in the plant. The display location in the-main control room may be in dedicated post accident panels or adjacent to or integrated with the existing normal range instrumentation display. l Continuous sampling of radioactive iodines and particulates in gaseous l effluents will be provided in the manner specified in NUREG-0737, 1 1 Table II.F.1-2, as described below: l (1) Sample collection: The release points with high-range noble gas effluent monitors will also have particulate and iodine sampling capability. Iodine samples will be taken with a charcoal or silver-zeolite cartridge and particulate samples with a filter. The post-accident iodine and particulate samples are extracted from the Amendment 13 A.9-57 O (January 1982)
release point' via the same sample line as the monitoring line. (2) Sample transport: The sample cartridges will be placed in a portable shielded cask and taken to the counting room. (3) Sample analysis: Capability for the analysis of sample cartridges will be provided. Design of the counting facility will consider the design basis sample. The precise location of the sample collection station will be selected upon completion of the post-accident shielding study [see response to 10 CFR 50.34(f)(2)(vii)]. The location selected will-assure that a worker involved in the sample collection and transport operation will not receive an exposure greater than 5 rem to the whole body and 75 rem to the extremities. O O A.9-58 Amendment 13 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT Identification of and Recovery from Conditions Leading to Inadequate Core Cooling (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xviii) Provide instruments that provide in the control room an unambiguous indication of inadequate core cooling, such as primary coolant saturation meters in PWR's and a suitable combination of signals from indicators of coolant level in the reactor vessel and in-core thermocouples in PWR's and BWR's. (II.F.2) 7~ RESPONSE TO 10 CFR 50.34(f)(2)(xviii) b Inadequate core cooling (ICC) is defined as insufficient reactor core heat removal to preclude continuing fuel clad temperature increase. Therefore, ICC can only occur subsequent to sufficient loss of RCS inventory to partially or totally uncover the reactor core. Approach to ICC is defined as a continuing decline in RCS inventory leading to core uncovering. The real measure of the degree of ICC, however, is the temperature to which core components actually rise. Instrumenta-tion to assist the operator in detecting the approach and onset of ICC, including RCS level indication and core exit thermocouples, will be provided for Pebble Springs. Anticipatory warning of conditions which could lead to the approach of ICC is provided by two saturation margin meters monitoring hot leg temperatures and RCS pressure. These saturation margin indicators, part of the ECI described in PSAR Section 7.4, will display the temperature n v margin to saturation in either loop. The inputs to each indicator will A.9-59 Amendment 13 (January 1982)
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consist of RCS pressure (pressurizer) and two hot leg temperatures (one loop A and one loop B). Alarms uill be provided to warn the operator of low margin to saturation (approach to saturation) These indicators will be qualified to Class lE requirements. Because ICC can only occur with substantially reduced RCS inventory, loss of RCS inventory will be used to indicate the approach of ICC. This indication of decreasing inventory may be decreasing pressurizer level or loss or subcooling margin. Continued loss of RCS inventory will be identified by loss of level in the hot legs (high points of the RCS). This is monitored by narrow range level instrumentation applied to the tops of the hot leg piping in each loop and wide range level instrumentation applied from the top to the bottom of the hot leg piping. This instrumentation will be part of the ECI and qualified to Class lE requirements. The narrow range level instrumentation assists the operator in determining when insufficient RCS inventory is available for natural circulation to continue, ie, when the level in the upper hot leg bend is below that required for flow to the steam generator. Further loss of inventory will result in decreasing level in the hot leg piping and steam generator primary side. Additional wide range instruaentation monitors level from the top to the bottom of each hot leg. Use of this instrumentation also assists the operator in moni-toring progress to restore RCS inventory. It should be noted that substantial voiding will not occur in the core unless the hot leg has l l drained. Level indication is also provided in the upper reactor vessel for detection of void formation there. Its range is from the bottom of l l the hot leg piping to the top of the reactor vessel head. The wide range and upper vessel level instrumentation will be part of the ECI and qual-ified to Class 1E requirements. Continued loss of RCS inventory below the bottom of the hot leg connections to the reactor vessel results in the core becoming covered with a steam-water mixture (froth) or eventu-ally becoming partially uncovered. It is under these conditions that the core can become inadequately cooled. The onset of ICC and its severity are monitored by incore thermocouples located in the incore detector assemblies (described in PSAR Section 7.8) near the tops of the A.9-60 O Amendment 13 (January 1982)
i e fuel pins. These temperatures will be displayed in the control room. Should the thermocouple temperatures exceed predetermined values, the operators will further depressurize the primary and secondary systems to maximize HPI and low pressure injection (LPI) flow, and start one or j more RCPs to promote all possible core cooling. The incore thermocouples will be qualified as Class lE. Class IE backup indication will be pro-l vided for at least 16 incore thermocouples. Indication for the remain-- i
- ing thermocouples will be non-Class lE.
, Analysis will be performed to support design and use of this instrumen- ! -tation, including accuracy and error analyses for the level' monitors f and correlation of the incore monitor thermocouple readings to fuel i temperatures under a range of ICC conditions. ? I 1 1 I 1 l J i i i i i 1 j A.9-61 Amendment 13 l l (January 1982)
. ._ - ~ _ . - . . - = . . - - . - - , - .
} NRC 10 CFR 50.34(f) REQUIREMENT
- Instrumentation for Monitoring Accident Conditions (2) To satisfy'the following requirements, the application shall provide
- j. sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.
- This information is of the type customarily required'to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues.
[ (xix) Provide instrumentation adequate for monitoring plant j conditions following an accident that includes core
- damage. (II.F.3)
RESPONSE TO 10 CFR 50.34(f)(2)(xix)- The Pebble Springs design will include instrumentation to monitor plant variables and systems during and following an accident-in accordance (y with defined design bases.
- '\
l 4 The present plant design will be reviewed against the guidance of 1 Revision 2 of Regulatory Guide 1.97 (December 1980) and ANSI /ANS-4.5-1980,
" Criteria for Accident. Monitoring Functions-in Light-Water-Cooled Reac-
.i tors". Those recommendations of Regulatory Guide 1.97 not-in the current design (as described in PSAR Section 7.5) will be incorporated into the Pebble Springs design, or a suitable alternate will be.provided for those items that challenge the state-of-the-art. For instrumentation which constitutes suitable alternates to the guidance of Regulatory Guide 1.97,
[ conceptual design information and justification for their adequacy will be submitted for NRC staff review prior to equipment procurement. The instrumentation needed to meet the intent of Regulatory Guide 1.97 will be addressed in the Pebble Springs FSAR.
- A.9-62 Amendment 13 ;
(January 1982) i
NRC 10 CFR'50.34(f) REQUIREMENT ~ O Power Supplies for Pressurizer Relief Valves, Block Valves, and Level Indication (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xx) Provide power supplies for pressurizer relief valves, block valves, and level indicators such that: (A) level indicators are powered from vital buses; (B) mo'.ive and control power connections to the emergency power sources are through devices qualified in accordance with requirements applicable to systems important to safety; and (C) electric power is pro-vided from emergency power sources. (Applicable to PWR's only.) (II.G.1) O RESPONSE TO 10 CFR 50.34(f)(2)(xx) f (A) Two channels of pressurizer 'le t i l indication instrumentation are provided as part of the Class lE ECI. Level indication readout is provided in the control room and at the auxiliary shutdown panel. Each channel is powered from a separate Class lE 120-V vital instrument ac bus. These buses are capable of being supplied from either the offsite power source or the standby diesel generators when offsite power is not available. Each bus is also backed by the Class IE 125-V de battery system.
-(B) The motive and control power circuits for the PORV and block valve will be connected to the Class 1E ESF buses through devices quali-fled in accordance with safety grade requirements.
O A 9-63 Amendment 13 (January-1982)
(C) Motive and control components of the POP.? are powered from a non-Class 1E, battery backed 125-V de bus and a 120-V ac uninterruptible bus, respectively. These sources are capable of being supplied by a standby diesel generator via the non-Class lE isolated 4.16-kV bus. The current design has the provision for manually opening the valve from the control room. The motive power for the PORV block valve will be supplied from an ESF 480-V motor control center and control components will be powered from a battery-backed Class lE 120-V vital instrument ac bus. The current design has the provision for manually opening or closing the block valve from the control room. The motive and control power for the block valve will be supplied from a different emergency power source (diesel generator) than those which supply the PORV. O O Amendment 13 A.9-64 (January 1982) l l
l NRC 10 CFR 50.34(f) REQUIREMENT Analysis and Upgrading of Integrated Control System (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy
.10 CFR 50.35(a)(2) or to address unresolved generic safety issues.
(xxii) Perform a failure modes and effects analysis of the inte- 1 grated control system (ICS) to include consideration of failures and effects of input and output si.gnals to the ICS. (Applicable to B&W-designed plants only). (II.K.2.9) RESPONSE TO 10 CFR 50.34(f)(2)(xxii) A generic failure modes and effects analysis (FMEA) for the ICS, Topical Report BAW-1564, has been submitted for the B&W operating plants. - The ( NRC staff has not completed its review of this report, and PGE has not performed a detailed review of its applicability to Pebble Springs. When the staff's review is completed and prior to the submittal of the FSAR, PGE will perform such a review and either establish that BAW-1564 is appl? cable to Pebble Springs or perform an FMEA for the ICS specifi- - cally for Pebble Springs. BAW-1564 does not address the response of the ICS to power supply failures. Therefore, an FMEA will also be performed for the ICS and non-nuclear instrumentation (NNI) for power supply failures.
\
A.9-65 Amendment 13 (January 1982) i .. .. .. . ..
NRC 10 CFR 50.34(f) REQULJ!fENT in.
.( I
\ hard-Wirad Safety-Grade Anticipatory Reactor Trips (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.
This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xxiii) Provide, as part of the reactor protection system, an anticipatory reactor trip that would be actuated on loss of main feedwater and on turbine trip. (Applicable to B&W-designed plants only). (II.K.2.10) RESPONSE TO 10 CFR 50.34(f)(2)(xxiii) An anticipatory reactor trip actuated on loss of main feedwater and tur-fg bine trip will be provided on Pebble Springs. b The loss-of-feedwater reactor trip utilizes a neutron flux-to-main feedwater flow (MFW) function. Neutron flux and main feedwater flow are continuously monitored. A reactor trip occurs whenever flux exceeds a variable limit set by the magnitude of MFW flow signal [K(MFW) + B]. Additionally, a loss of both main feedwater pumps will also initiate a , reactor trip. This reactor trip permits channel functional test at power. Testing of the circuitry will be included in the RPS monthly surveillance tests. The RPS checkout procedure will include provisions to demonstrate the operability of all circuitry. A reactor trip on significant decrease in steam generator level is not used for Pebble Springs. An analysis will be performed to demon-strate that for the steam generator level trip to be anticipatory to A.9-66 Amendment 13 (January 1982)
-s - -
the neutron flux-to-main feedwater flow trip, the steam generator level trip will impact operability (spurious trips). The turbine trip signals used to initiate reactor trip will originate with pressure switches actuated by decreasing oil pressure in each turbine trip oil system. Thase non-Class IE signals will be buffered signals isolated from the Class IE portions of the RPS by isolation devices (located within the main control room area) which meet the guidance of Regulatory Guide 1.75. O l O Amendment 13 A.9-67 (January 1982)
.l NRC 10 CFR 50.34(f) REQUIREMENT O-1
' Upgrade Licensee Emergency Support Facilities (2) To satisfy the following requirements, the applicatior. shall provide sufficient information to demonstrate that the required actions
~
will be satisfactorily completed by the operating license stage.
.-This information is of the type customarily-required to satisfy 10 CFR 50.35(a)(2) or to address unresolved ~ generic safety issues. l (xxv) Provide an onsite Technical Support Center, an onsite Opera-tional Support Center, and,. for construction permit appli-cations only, 'a nearsite Emergency Operations . Facility.
(III.A.l.2) RESPONSE TO 10 CFR 50.34(f)(2)(xxv) Emergency response facilities will be provided in accordance with NUREC-0696, " Functional Criteria for Emergency Response Facilities" , (Final Report, February 1981). Preliminary information for the Pebble Springs emergency facilities is provided below. Technical Support Center Two concepts are under consideration for design of the TSC. Option A involves separate TSCs for each generating unit. The TSC would { be located in the Control Building of each unit in an area adjacent to the control room (Figures A.9-2 and A.9-3). This TSC concept would have the same structural and habitability design as the control room. i -; i Option B involves a single TSC, common to both generating units, situated . i in an-area between Units 1 and 2 (Figures A.9-2 and A.9-4). For this !- TSC location, adequate plant ingress will be provided to ensure the walking time from the TSC to the control room of either unit does not l exceed 2 min.
)
A.9-68 Amendment 13 (January 1982) l 4
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The TSC for this option will be a well engineered non-Seismic Category I structure able to withstand adverse conditions reasonably expected during the design life of the plant, including earthquakes, high winds and floods. The ventilation system will be manually actuated, non-Seismic Category 1 with high-efficiency particulate air (HEPA) and charcoal filters. The TSC selected will be large enough to meet the minimum space criteria of Section 2.4 of NUREG-0696 and all functional criteria related to radiological monitoring, communications, instrumentation, the technical data system, power supplies and records management. Operational Support Center The preliminary location for the OSC is shown in Figure A.9-2. A pre-liminary floor plan of the OSC is shown in Figure A.9-5. The OSC will have direct communications with the control room and with the TSC so that the personnel reporting to the OSC can be assigned to duties in support of emergency operations. The preliminary location selected for the OSC is in an area (Turbine O Building) where habitability would not be comparable to the control room. The emergency plan will include procedures for evacuation of OSC personnel in the event of a large radioactive release and performance of OSC functions by essential support personnel from other onsite locations. Emergency Operations Facility The EOF is tentatively planned to be located on property owned by PGE at its Boardman Coal Plant site, about 17 miles east of the Pebble Springs TSC. This location will permit designated Pebble Springs plant personnel to report to the EOF within 1 hr. A preliminary floor plan of the "0F is shown in Figure A.9-6. The EOF vi'.1 be a well engineered non-Seismic Category I structure designed in accordance with the Uniform Building Code and able to Amendment 13 A.9-69 O (January 1982) .
withstand adverse conditions of high winds and floods. The EOF will be fs I large enough to meet the minimum space criteria of Section 4.4 of
}
NUREG-0696. The EOF will also be designed to meet all other functional criteria of NUREG-0696 related to radiological monitoring, communications, instrumentation, the technical data system, power supplies and records management. l As an engineering alternative to an EOF located at Boardman, secondary consideration is being given to an EOF located within 10 miles of the TSC. A preliminary concept would involve an EOF located onsite but ] outside the plant's protected area, either in the basement of the Administration Building (see Figure A.9-2) or as a separate facility nearby. Either concept would have a floor plan similar to that shown in Figure A.9-6. The onsite EOF selected would be designed to meet the additional functional criteria of Section 4.2 of NUREG-0696. i e s
. \~,)
1 A.9-70 Amendment 13 (January 1982)
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I i NRC 10 CFR 50.34(f) REQUIREMENT
. Primary Coolant Sources Outside the Containment Structu're (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions
^ will be satisfactorily completed by the operating' license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xxvi) Provide for leakage control and detection in the' design of systems outside containment that contain (or might l contain) TID 14844 source term radioactive materials follow-ing an accident. Applicants shall submit a. leakage control program, including an initial test program, a schedule for re-testing these systems, and the actions to be taken for. i minimizing leakage from such systems. The goal is to minimize potential exposures to workers and public, and to i provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency. (III.D.1.1) RESPONSE TO 10 CFR 50.34(f)(2)(xxvi) i The Pebble Springs Plant design includes many engineering features to reduce the potential for leakage from radioactively contaminated systems and to mitigate the effects of such leakage (see Table A.9-7 for details). However, in the event of an accident resulting in severe core damage, additional leak reduction and control measures are warranted to ensure that radioactive leakage outside Containment will not unduly restrict access or cause excessive offsite exposures. These additional measures are described below. Following an accident resulting in severe core damage, the primary method of reducing leakage is to isolate all non-essential systems outside Containment from potential sources of highly radioactive fluids, thereby O A.9-71 Amendment 13 (January 1982)
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mc--.me.,.se a w -- va v - ' e-- --*--*e--v-+-w-* *wwn------',--- ,w-- - - - ------T ,-'-- - -- '~ +-=mw--*- - --
eliminating unnecessary leak paths. Table A.9-8, Part A, provides a list of systems or portions of systems which can be effectively isolated and, therefore, are not expected to contain highly radioactive fluids. Although these systems are designed to minimize leakage in order to keep radiation exposures to personnel as low as reasonably achievable during normal operation, they art not designed or shielded to process highly radioactive fluids. A review will be conducted prior to FSAR submittal to ensure that the interface between piping of systems is adequately shown on the piping and instrumentation diagrams (P& ids) to preclude incidents such as reported in IE Circular 79-21 (and NRC letter dated October 17, 1979) and that all potential release paths for radioactive fluids (such as relief valve discharges) have been considered. One of the systems effectively isolated is the Caseous Radwaste and vent collection system (GRS). Highly radioactive gases are not expected to accumulate in this system since the letdown portion of the makeup system will be isolated by the ESFAS and the RCS can be vented inside Contain-ment, if necessary [see response to 10 CFR 50.34(f)(2)(vi)]. A design change which will provide a backup to isolation of the GRS and additional flexibility in post-accident operations is the addition of a line which will allow gases to be discharged from the vaste gas surge or decay tanks to the Containment via an existing penetration (see Figure A.9-7). This change will provide a method of returning any radioactive gases accumulating in the Auxiliary Building systems to Containment if required. Since this is only a backup to isolation, helium or an alternate method of leak checking the GRS is not considered to be necessary. The systems which could become highly contaminated are those required for mitigation or recovery from the accident and which thereby form an extension of Containment (see Table A.9-8, Part B for a complete list). In addition to the leak-tight features listed in Table A.9-7, two additional provisions will be provided to collect or contain any leakage from these systems: (1) Leak-off, drain, vent and overflow connections in the Auxiliary Building or penetration arcas will be hardpiped to and collected O j Amendment 13 A.9-72 (January 1982) i l
in the Auxiliary Building sumps. The primary method of detect-h o ing leakage from safety train systems will be the rate at which
- leakage is collected in the respective Auxiliary Building sump.
Specific sources of leakage may be detectable by flow, tempera-ture or sonic detectors. Two design changes will be required
~
to control the spread of radioactive fluids leaking to'the Auxiliary Building sumps from highly cont'aminated systems (see Figure A.9-8). (a) The Auxiliary Building sump pump discharge piping will be modified so that the sump effluent can be pumped to Con-tainment through an existing penetration. (b) The Auxiliary Building sump vent piping will be modified so that the sump gases will be diverted from the vent collection header to the fuel handling area exhaust system (FHAES) for particulate and charcoal filtration prior to release. O (2) A slight negative pressure will be maintained by the FHAES.in compartment s outside Containment, which have piping and compo-nents of highly contaminated systems,-so that halogens and particulates will be filtered prior to release. The FHAES currently takes suction from the Containment spray pump rooms, l decay heat removal pump rooms and decay heat exchanger rooms !. for this purpose and will be expanded to take suction from the penetration areas, pipeways and makeup pump rooms (see Figure A.9-9).
' Pebble Springs is currently designed with sufficient connections in systems which could become highly contaminated to allow periodic leak rate testing. Supplemental programs will be developed for initial and periodic maintenance and leak rate testing of the systems outside
. Containment which may contain highly radioactive fluids or gases [ following an accident. The systema which will be included are listed i
- y in Part B of Table A.9-8. Where possible, the testing will be performed i -
A.9-73 Amendment 13 I (January 1982)
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. - , - , ---+-v,.,r -
e
by inspecting potential leakage areas such as flanges, pump seals and valve packing while the system is maintained at operating pressure. An alternate means of testing will be developed for systems that do not contain water except during actual operation following an accident. The acceptance criteria for leakage will be determined for individual components within the system. Excessive leakage will be repaired in a timely manner consistent with system operation and the magnitude of the leak. Testing will be done at refueling cycle intervals. The results of the initial leak rate tests will be reported to the NRC. s G
~
s ii O Amendment 13 A.9-74 (January 1982)
I NRC 10 CFR 50.34(f) REQUIREMENT Inplant Radiation Monitoring (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. {' This'informatio'n is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. i (xxvii) Provide for monitoring of inplant radiation and airborne radioactivity as appropriate for a broad range of routine and accident conditions. (III.D.3.3) 1 RESPONSE TO 10 CFR 50.34(f)(2)(xxvii) Normal monitoring of inplant radiation and airborne radioactivity is described in PSAR Sections 12.1.4 and 12.2.4. Portable airborne iodine samplers and. sample analysis equipment will be available onsite prior to the issuance of the operating license', in i accordance with NUREG-0737, Item III.D.3.3. This equipment will not be purchased for several years, but it is expected that it will be cart mounted and backup battery powered. Plant personnel will be trained in the use of this equipment under both routine and emergency conditions. Details will be provided in the FSAR. i 4 4 1 i I l
\
A.9-75 Amendment 13 (January 1982) L
- . _ . , - - _ - , _ . . - _ _ , _ . _ - _ - _ _ - _ . - . ____--.....__,,,_.__..,__._,_r . , . , _ - - , , - .
NRC 10 CFR 50.34(f) REQUIREMENT i %) Control Room Habitability (2) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10 CFR 50.35(a)(2) or to address unresolved generic safety issues. (xxviii) Evaluate potential pathways for radioactivity and radiation that may lead to control room habitability problems under accident conditions resulting in a TID 14844 source term release, and make necessary design provisions to preclude such problems. (III.D.3.4) RESPONSE TO 10 CFR 50.34(f)(2)(xxviii) The control room habitability systems include radiation shielding, air ventilation and filtering systems, fire protection, personnel protective equipment and first aid, utility and sanitary facilities. These systems are incorporated in the Pebble Springs design to permit access to and occupancy of the contral room during all modes of operation including an accidental release of toxic or radioactive gases. The design of these systems and facilities as described in PSAR Section 6.4 meets General Design Criterion 19 and is consistent with applicable guidelines of SRP Section 6.4 and Regulatory Guides 1.78 and 1.95 (Revision 1, January 1977). The Pebble Springs design includes redundant non-Seismic Category I chlorine and ammonia detectors which isolate and monitor the control room outside air supply as described in PSAR Sections 2.2.3.2 and 6.4. During the TMI-2 accident, the control room was contaminated via internal pathways. The causes of contamination at TMI-2 were: (a) lack of ade-quate control room access control, (b) access by contaminated personnel, v A.9-76 Amendment 13 (January 1982)
(c) doors that were left open, and (d) the inability to accurately monitor the control room atmosphere in the recirculation mode. These difficulties will be precluded for the Pebble Springs Plant by provision of a dedicated TSC and an onsite OSC to be used as staging areas for emergency support personnel as discussed in the response to 10 CFR 50.34(f)(2)(xxv). The control room atmosphere will be continuously monitored for radioactivity as described in PSAR Section 11.4.2.2.8. The control room is also served by the Area Radiation Monitoring System described in PSAR Section 12.1.4. Portable iodine monitors (see response to 10 CFR 50.34(f)(2)(xxvii)] will be available for use ir the control room for checking on airborne iodine contamination. O i l l O Amendment 13 A.9-77 (January 1982) l
NRC 10 CFR 50.34(f) REQUIREMENT ,I I \ Procedures for Feedback of Operating, Design and Construction Experience j (3) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met. This information is of the type customarily required to satisfy 10 CFR 50.34(a)(1) or to' address the applicant's technical qualifica-tions and management structure and competence. (i) Provide administrative procedures for evaluating operating, design and construction experience and for ensuring that applicable important industry experiences will be provided j in a timely manner to those designing and constructing the plant. (I.C.5) RESPONSE TO 10 CFR 50.34(f)(3)(1) As the lead Applicant and project sponsor, PCE has the primary responsi-bility for assuring that applicable operating, design and construction experience is factored into the Pebble Springs Nuclear Plant. i PCE, Bechtel and B&W will each have administrative procedures for the evaluation of operating, design and construction experience. The proce-dures of each company will complement and overlap each other to assure j that applicable industry experience is incorporated into the design and construction of the Pebble Springs Nuclear Plant. The following is a f description of those procedures: i , Organizational Responsibilities Within PGE, the Technical Functions departments will be responsible for reviewing and categorizing the information received from outside the Project and identifying those experiences which may be of interest to Pebble Springs. Technical Functions will also be responsib?.e for cate-gorizing these experiences such that operating, design and construction i ! A.9-78 Amendment 13 (January 1982)
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o
~
e experience are directed to the respective organizations within PGE for their refiew and use. Design and construction experiences from within the Project will be directed to the Pebble Springs Project Manager, who will be responsible for reviewing, screening and directing the information to the appropriate organization in PGE. l I l Bechtel and B&W are responsible to PGE for implementing feedback programs l for design, operating and construction experience within their respective organizations as described in the following sections. Administrative and Technical Review Steps; and Recipients of Information General: PGE has contracted for design and construction of Pebble Springs with Bechtel and B&W. As part of its responsibilities B&W has established and maintained, within its Nuclear Power Generation Division, a formal service adviscry communication system that is designed to provide B&W-Owner-Operator with a broad coverage of B&W operating and maintenance information and recommendations. In addition, B&W routinely reviews other available industry experience for applicability to the equipment and services it supplies for the Pebble Springs Nuclear Plant. Similarly, Bechtel reviews available industry experience for applicability to the design, construction and other activities it provides for Pebble Springs. In addition, PGE is responsible for advising Bechtel and B&W of operating design and construction experience data uniquely available to PGE such as from utility owners groups. PGE: PGE functions within the program for review of operating, design and construction experience to: (a) review and approve Bechtel's and B&W's programs, (b) audit and monitor Bechtel and B&W implementation of their programs, (c) furnish data uniquely avellable to PGE or unlikely Amendment 13 A.9-79 (January 1982)
to be available to Bechtel and B&W and (d) provide direction to Bechtel () for incorporating and implementing design and construction experience into the Pebble Springs design. Operating, design and construction experience information from external , sources enters the PGE program from two general categories: regulatory agencies and industry sources. Examples of documents reviewed are as follows: (1) Regulatory Agency Information: (a) Licensee Event Reports (b) Regulatory Guides (c) Regulations (10 CFR and 49 CFR) 4 (d) IE Bulletins, Circulars, Orders and Notices fe's . (e) NUREGs (f) Standard Review Plan (including Branch Technical Positions) (2) Industry Sources: (a) Topical Reports from B&W and the Nuclear Safety Analysis Center (b) IEEE, ANS, ANSI and ASME Codes and Standards (c) Licensee Event Reports (d) NSAC/INPO Significant Events Evaluation Information Network l (e) Owners Group Activities
, v 1
I A.9-80 Amendment 13 (January 1982)
As external information enters the PGE system, it will be directed to the Technical Functions area. There it will be categorized, screened for applicability and documented. The review at this stage is two-fold in purpose: (a) to reduce the quantity of information received to manageable amounts by culling out information clearly not relevant to Pebble Springs, and (b) to broadly categorize the information into operations, design or construction categories. Technical Functions will transmit the information to either the PGE Pebble Springs opera-tions organization, in the case of operational information, or the PGE Nuclear Projects Engineering Department (NPED), in the case of design information, or the Nuclear Projects Construction Manager, in the case of construction information, along with a specified time for disposition of the items. PGE will provide continuous assessment of the efficacy of the experience feedback programs at PCE, Bechtel and B&W by using a commitment tracking system which will provide feedback as to the ulti'nate resolution of the information sent out to the Project. The same commitment tracking will be utilized for PGE internally generated experience information. For design experience, NPED has the primary responsibility for resolving concerns once the information is available within Technical Functions. NPED will review the information and direct it to the appropriate disci-pline leader for a determination of the necessary action. The engineer may consult with either or both Bechtel and B&W to evaluate the concern. From this point, normal design control processes are used. For construction experience, the Manager of Nuclear Projects Construction will have the primary responsibility for resolving concerns once the information is available within Technical Specifications. He may use assistance from NPED and either or both Bechtel and B&W, as appropriate. Construction concerns that affect plant design will be resolved in accor-dance with the Project's normal design process. Information on operating experience will be received from Technical Func-tions by the Plant General Manager, who will perform a more detailed O Amendment 13 A.9-81 (January 1982)
review of the information. This review will determine if the information should be factored into operations planning activities or if it is of sufficient concern to pursue with Bechtel and B&W. If warranted by the nature of the item, the Plant General Manager will cons' ult with NPED, obtaining assistance as necessary, and recommend a course of action. In some cases, the appropriate action will be decided without involving Bechtel or B&W, particularly if it is in plant maintenance or operations. The operational or maintenance concern may then be resolved as part of the normal process of operator training or procedures development. Bechtel: Both on and off Project personnel have the responsibility for identifying _ and resolving design and operations feedback concerns. Sources utilized for feedback include: (1) NRC Inspection and Enforcement Bulletins, Circulars and Notices (2) Licensee Event Reports (3) INP0/NSAC Significant Operating Experience Reports and Signifi-cant Event Reports (4) Various internal Bechtel sources. Bechtel receives these documents through direct distribution, the Bechtel Licensing Information System, the Atomic Industrial Forum, etc. The discipline design groups are responsible for determining the applica-bility of the concern to Pebble Springs and for written disposition. Significant experience feedback, if applicable, is also incorporated into generic engineering documents such as design standards, guides and specifications. These generic engineering documents are utilized in developing Project-specific documents. O A.9-82 Amendment 13 (January 1982)
The Bechtel organization also reviews information transmitted to them by B&W (experience feedback from operating B&W plants), PGE, and other sources of unique experience. Items applicable to Pebble Springs will be resolved in the design or, if significant enough to warrant a PGE decision on the resolution, submitted to PGE for review and approval. Such submittals may be in the form of design documents submitted for PGE review, studies or I correspondence. PCE then reviews the Bechtel recommendation for dis-position as described earlier. The Bechtel Construction Engineering Staff obtains construction experi-ence data through reports from the field, review of IE Bulletins, Circulars and Information Notices and review of construction practices at various Bechtel nuclear project sites. The significant experience data obtained from these sources is communicated to the Site to alert construction personnel to potential problems that may be encountered during the construction phase. In addition, Project-level construction reviews are held to discuss and O avoid problems that may have arisen during construction or as a result of feedback. Problem resolutions are incorporated in the various construction-related manuals and instructions. Babcock & Wilcox: Pebble Springs will have either a Site Operation Manager (construc-tion and startup phase) or a Service Manager (operating phase) assigned from B&W. If at any time a problem is encountered in the field which affects B&W supplied equipment, a Site Problem Report (SPR) is written and transmitted to both PGE and B&W offices in Lynchburg, Virginia. This SPR will describe the problem encountered and also may suggest a solution. The SPR is assigned to a Task Engineer at B&W's offices. The Task Engineer will study the SPR and if it is generic, a Multiple Contract Problem (MCP) Report will be written and the MCP will be issued to all affected projects for resolution. If the problem is a potential safety Amendment 13 A.9-83 (January 1982)
concern (PSC), then a PSC report will be initiated elso. Pebble Springs
/S will also receive the above reports which are generated from other sites or other sources if they are applicable.
The B&W Customer Service Department provides formal Site Instructions to provide Owners with a broad coverage of operating and maintenance information and recommendations. The resolution of the above problems (SPR, MCP) and/or concerns (PSC) and other information and/or instruc-tions may be transmitted to PGE as a Site Instruction if applicable. The Site Instruction procedure is designed to collect, process and disseminate information and recommendations pertinent to: (1) Unique operating conditions and experiences (2) Improved methods, techniques and procedures for operating and maintaining B&W supplied equipment (3) Plant performance improvement and equipment upgrading O)
'\- (4) Safety, licensing and other regulatory matters.
In addition to the Site Instruction, B&W provides various Site Support Documents. These include Test Guides, Test Specifications and Operating Specifications, Emergency Operating Specifications, and Plant Limits, Precautions and Setpoints. i The major sources of information used to prepare Site Instructions and Site Support Documents include SPRs, MCP Reports, PSCs, safety and licensing reports, reports and instructions prepared by engineering organizations, B&W and Vendor Equipment Instruction Manuals, Startup and Preoperational Test Reports, and NRC Documents. Information received j from the NRC includes IE Bulletins, Circulars, and Information Notices, NURECs, Regulatory Guides and SRPs. These NRC documents are received by the B&W Licensing Sectica and are disseminated within B&W after review to determine which organizations should receive them. These documents are O A.9-84 Amendment 13 (January 1982) g -,r., w. - , , _ . . - - _ _ . . , . . - . . . - . , , , . , . - _ , .,,,__y . . . _ . - . _ . ,
an important source of input for the Site Instructions and Site Support Documents which B&W supplies to PGE. Avoidance of Extraneous and Unimportant Information; and Avoidance of Conflicting or Contradictory Inforetation Within PGF, the Technical Functions area will assure the avoidance of extraneous and unimportant information through its normal screening process. Technical Functions will also assure that potentially con-flicting or contradictory information is identified and transmitted to the appropriate organization for resolution. Bechtel, through its normal screening procesk, will provide for the avoidance of extraneous and unimportant info rmation. Practical Interim Audits PGE will assure compliance with these requirements by monitoring and periodic audits of PGE, Bechtel and B&W implementation of their programs. PCE will audit the implementation of experience fcedback as part of their auditing of quality-related design and const co : tion activities at PGE and at B&W and Bechtel. O Amendment 13 A.9-85 (January 1982)
NRC 10 CFR~50.34(f) REQUIREMENT Expand QA List (3) To satisfy ~the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. i This information is of the type customarily required to satisfy 10 CFR 50.35(a)(1) or-to address unresolved ~ generic. safety issues. (ii) Ensure that the. quality assurance (QA) list required by Criterion II, App. B, 10 CFR Part 50 includes all structures, systems, and components important to safety. (I.F.1) RESPONSE TO 10 CFR 50.34(f)(3)(ii) i The quality assurance list (Q-List) includes all items and activities affecting safety as defined by Regulatory Guide 1.29 and Appendix A to ! 10 CFR 50 or described in PSAR Section 3.2. Bechtel Project Engineering is responsible for preparation and mainte-nance of the Q-List. Each revision of the Q-List contains the issue date, approval date and authorized signature. The Q-List and revisions thereto require the approval of the Bechtel Project Engineer.and Chief Nuclear Engineer. For items that fall within the B&W scope of work, input to the.Q-List will be implemented based on B&W recommendations.
'All changes to the Q-List are reviewed and approved by PGE.
The Q-List may be expanded as a result of ongoing activities related to i the TMI-2 event. Any such items (eg, hydrogen control or additional post-accident monitoring systems) will be added to the list at the appropriate time using existing procedures. f
- To add further verification to the Q-List, a systematic analysis of j plant equipment will be performed using the criteria and analysis i
approach of ANS-51.1, " Nuclear Safety Criteria for the Design of I A.9-86 Amendment 13 (January 1982) 1 L _ _ -. _ _ _. _ _ _ __.. _ _ .._ _ . _ ._ _ . _ _ . _ _ _ _
Stationary Pressurized Water Reactor Plants", Draft 2, October 1981. The structures, systems and components identified by this analysis as important to safety will be checked against the existing Q-LLst, and modifications to the Q-List will be made as appropriate. For any addi-tions to the Q-List, the QA Program will be applied to all subsequent system design, procurement, construction and operation activities. O O Amendment 13 A.9-87 (January 1982)
-NRC 10 CFR 50.34(f) REQUIREMENT O
Develop More Detailed QA Criteria (3) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met. This information is of the type customarily required to satisfy 10 CFR 50.34(a)(1) or to address the applicant's technical qualifica-tions and management structure and competence. (iii) Establish a quality assurance (QA) program based on considera-tions of: (A) ensuring independence of the organization per-forming checking functions from the organization responsible for performing the functions; (B) performing quality assurance / quality control functions at construction sites to the maximum feasible extent; (C) including QA personnel in the documented review of and concurrence in quality related procedures asso-ciated with design, construction and installation; (D) estab-O lishing criteria for determining QA programmatic requirements; (E) establishing qualification requirements for QA and QC per-sonnel; (F) sizing the QA staff commensurate with its duties and responsibilities; (C) establishing procedures for mainte-nance of "as-built" documentation; and (H) providing a QA role in design and analysis activities. (I.F.2) RESPONSE TO 10 CFR 50.34(f)(3)(iii) 3 As described in Section 17.1 of the PSAR, the Quality Assurance Program ! for design and construction of the Pebble Springs Nuclear Plant consists l of five major elements: (a) the Portland General Electric Company pro-gram, (b) the Bechtel Power Corporation program, (c) the Babcock & Wilcox Company program, (d) the fuel fabricator program and (e) the audit sur-veillance plan - fuel. O A.9-88 Amendment 13 (January 1982)
The PGE quality assurance program that is applied to the design, procure-ment, construction and preoperational testing of the Pebble Springs Nuclear Plant is described in the NRC-approved Topical Report PGE-8003, Revision 2A, " Nuclear Projects Quality Assurance Program for Design, Construction and Preoperations". PGE previously submitted a proposed Revision 3 to PCE-8003 to the NRC in June 1978. PCE intends to update this topical report to fully address the requirement above and the revised acceptance criteria of SRP Section 17.1 (NUREG-0800). This update (PGE-8004, Revision 4) is planned to be submitted for NRC review and acceptaace in March 1982. In the interim, the following is provided in response to the additional acceptance guidance developed by the NRC's QA Branch. This guidance was informally provided to pending CP appli-cants in April 1981. The format and content of this response is consistent with and responsive to the subject NRC staff guidance and is sufficient to demonstrate that the above requirement has been met. Summaries of the major contractor's QA Programs for construction are included as Attachments 1, 2, 3 and 4 in PSAR Section 17.1. Attachment I describes the Bechtel QA Program; Attachment 2, the B&W QA Program; , the nuclear fuel supplier's QA Program (to be furnished later); and Attachment 4, audit surveillance plan - fuel (to be furnished later). These summaries have been updated as appropriate to support the following response. Independence of Functions For the Pebble Springs Nuclear Plant, Bechtel has been designated respon-sible for supplier control in accordance with Section 7, Pages 38 through 43 of the Bechtel Topical Report BQ-TOP-1, Revision 2A. The Bechtel QA program requires that inspection personnel be independent of the individual or group performing the activity being inspected. Bechtel Procurement Supplier Quality, Site Construction Quality Control and Quality Assurance comprise the quality group, which is delegated the authority and responsibility for inspection and verification functions. O Amendment 13 A.9-89 (January 1982)
Where reference is made to the appropriate QA organization in the f'~s following responses, depending on the organization and the function to k-
% be performed, this may be either the QA or the quality control group of the involved organization. In either case, this group is independent of the group performing the work.
2Al Verification of conformance to established requirements at (IB2)[a] all levels (except for designs) is accomplished by individ-uals or groups within the appropriate QA organization who do not have direct responsibility for performing the work being verified. 9 2A2 Bechtel Site Construction Quality Control will be responsible (10B1) for surveillance inspection of the site contractors' activi-ties and audits of the contractors' QA program. 2A3 Verification of PGE suppliers' activities during fabrication, (7A2) inspection and testing, and shipment of materials, equipment f-s and components will be planned and performed by Bechtel Pro-(,,) curement Supplier Quality in accordance with written procedures to assure conformance to the purchase order requirements. These procedures, as applicable to the method of procurement, provide for: (1) Specifying the characteristics or processes to be witnessed, inspected, or verified, and accepted; the method of surveillance and the extent of documentation required and those responsible for implementing these procedures. (2) Aclits, surveillance or inspections which assure that the supplier complies with the quality requirements. [a] Numbers in parentheses correlate with the numbers in Section 17.1 of the NRC Standard Review Plan (NUREG-0800). They are provided for reference purposes only. A.9-90 Amendment 13 (January 1982)
2A4 Receiving inspection will be performed by Bechtel Site Con-(7B1) struct. ion Quality Control to assure: (1) The material, components or equipment is preperly identified and corresponds to the identification on the purchase document and the receiving documentation. (2) Materials, components, equipment and acceptance records satisfy the inspection instructions prior to installation or use. (3) Specified inspection, test and other records (such as certificates of conformance attesting that the materials, components and equipment conform to specified require-ments) are available at the Pebble Springs Nuclear Plant prior to installation or use. 2A5 Correct identification of materials, parts and components (8B3) will be verified and documented by Bechtel Site Construction Quality Control prior to release for fabrication, assembling, shipping and installation. 2A6 Procedures are established for recording evidence of accept-(9B2) able accomplishment of special processes using qualified procedures, equipment and personnel. Bechtel Site Construc-tion Quality Control will verify the recorded evidence and document the result. 2A7 Inspection and test results will be documented and evaluated, (10C3) and their acceptability determined by a responsible indi-(11C1) vidual or group. Bechtel Site Construction Quality Control as a minimum evaluates, verifies and documents completeness of this activity. O Amendment 13 A.9-91 (January 1982)
2A8 Follow-up action will be taken by the QA organization to (16.3) verify proper implementation of corrective action and to
\ close out the corrective action in a timely manner.
QA/QC Functions at Construction __ Site 2B1 The person at the construction site responsible for directing (1C3) and managing the site QA program will be Bechtel's Field QA Engineer who has appropriate organizational position, respon-sibilities and authority to exercise proper control over the QA program. This individual is free from non-QA duties and can thus give full attention to assuring that the QA program at the plant site is being effectively implemented. In addi-tion, the PGE Nuclear Projects QA engineering staff personnel will maintain weekly audits / surveillance of onsite activities to assure PGE management that all QA program requirements are being properly implemented. p 2B2 PGE and Bechtel site QA personnel are involved in plant (IB6) activities important to safety and are kept abreast of work schedule and construction activities by periodically attending 1 construction status meetings. In addition, Bechtel site QA 7 personnel actively participate in day-to-day planning, sched-uling, and construction status meetings to review problem areas and evaluate these areas to determine if they are chronic and/or are developing a trend. Problem areas are also evalu-ated by Bechtel to determine to what extent corrective action is taken and its effectiveness. Bechtel site QA personnel ensure that there is adequate QA coverage relative to procedural and inspection controls, accep-tance criteria, and QA staffing and qualification of personnel to carry out QA assignments. i C A.9-92 Amendment 13 (January 1982)
Procedures Review and Concurrence 2Cl The PGE QA Program for the Pebble Springs Nuclear Plant is 0 (2Bla) described in the QA manual for design, construction and 2C2 preoperations (PGE-8003). An executive Policy Staterent (2Blb) discusses PGE's dedication to a strong QA Program in the design, construction and preoperational testing of nuclear power plants. This Policy Statement and the Introduction require compliance with the Quality Assurance Manual for all activities that affect quality of those structures, systems and components that are safety-related. Fire protection and radioactive waste management systems will be included in this program. Procedures are reviewed by appropriate QA personnel during preparation for inspections, surveillance, implementa-tion reviews and audits to ensure consistency with QA program commitment s . Additionally, these procedures are reviewed and concurred with by the appropriate QA organization prior to issuance. 2C3 Procedures are established for the review and documented con-(4A1) currence of procurement documents by QA personnel tc catermine that: (a) quality requirements are correctly stated, inspect-able, and controllable, (b) there is adequate acceptance and rejection criteria and, (c) that procurement documents have been prepared, reviewed, and approved in accordance with QA Program requirements. To the extent necessary, procurement documents will require contractors and subcontractors to pro-vide an acceptable QA Program. 2C4 Procedures for the review, approval and issuance of documents (6A2) (including procedures, instructions, specifications and con-struction drawings) and changes thereto are established and described to assure technical adequacy and inclusion of appro-priate quality requirements prior to implementation. These documents are reviewed and are concurred with in writing by the appropriate QA organizations for quality-related aspects. Amendment 13 A.9-93 (January 1982)
l 2C5 . Inspection' procedures, instructions or checklists provide-(10Cl) for the following as reviewed and concurred with by the D appropriate Bechtel QA organization for_ quality-aspects and other technical organizations as appropriate: (1) -Identification'of characteristics and activities to be inspected. (2) A description of the method of inspection. (3) Identification of the individuals or groups respon-sible for performing the inspection operation. (4) Acceptance and rejection criteria.
.(5) Identification of required procedures, drawings, and specifications and revisions.
(6) Recording inspector or data recorder and the results of the inspection operation. (7) Specifying necessary measuring and test equip-ment including accuracy requirements. 2C6 Test procedures or instructions are reviewed and concurred .
-~ (1181) with by the appropriate Bechtel QA' organization for quality aspects and by other technical organizations'for technical aspects and provide as required for the following:
'(1) The requirements and acceptance limits contained in applicable design and procurement documents.
(2) Instructions for performing the test. . (3) Test prerequisites such as calibrated instru- { mentation, adequate test equipment, and instru-mentation including their accuracy requirements, L-A.9-94 Amendment 13 _(January 1982) l
completeness of item to be tested, suitable and controlled environmental conditions, and provi-sions for data collection and storage. (4) Handatory inspection hold points for witness by owners, contractor, or inspector (as required). (5) Acceptance and rejection criteria. (6) Methods of documenting or recording test data and results. , (7) Provisions for assuring test prerequisites have been met. 2C7 Procedures are established for calibration (technique and (12.3) frequency), maintenance and control of the measuring and test equipment (instruments, tools, gauges, fixtures, reference and transfer standards, and nondestructive examination equip-ment) that is used in the measurement, inspection and moni- h toring of structures, systems and components. The review and documented concurrence of these procedures are described and the Bechtel organization responsible for these functions is identified. Requirements for such description and identifi-cation are included in procurement documents, as appropriate for contractors and suppliers. 2C8 Procedures are established to control the cleaning, handling, (13.2) storage, packaging and shipping of materials, components and systems in accordance with design and procurement requirements to preclude damage, loss or deterioration by environmental conditions such as temperature or humidity. The appropriate ' QA organization reviews and documents concurrence of these procedures. Amendment 13 A.9-95 O (January 1982)
2C9 Procedures are established to_ indicate the inspection, test
]# (14.1) and operating status of structures, systems and components throughout fabrication, installation and test. The appro- ; priate QA organization reviews and documents concurrence with these procedures.
2C10 Procedures are established to control the application and
; (14.2) removal of inspection and welding stamps and status indicators such as tags, markings, labels and stamps. The appropriate QA organization reviews and documents concurrence with_these procedures.
2C11 Procedures are established to control altering the sequence of (14.3) required tests, inspections and other operations important to safety. Such actions should be subject to the same controls - as the original review and approval. The appropriate QA organi-zation reviews and documents concurrence with these procedures. 2C11a The status of nonconforming, inoperative or malfunctioning i. ~ (14.4) structures, systems or components is documented and identified to prevent inadvertent use. Bechtel is responsible for the identification and the tagging or marking of nonconforming items during construction activities. The marking or tagging of nonconforming items during the preoperation phase will be accomplished and controlled by the plant staff. 2C12 Procedures are established for identification, documentation, (15.1) segregation, review, disposition and notification to affected organizations of nonconforming materials, parts, components, and as applicable to services (including computer codes) if disposition is other than to scrap. The procedures provide identification of authorized individuals for independent review of nonconformance, including disposition and closeout. O j A.9-96 Amendment 13 (January 1982)
~. ._, . . _ _ ___ __ . _ _ __ _ _ _ _ - - _ .
2Cl3 PGE requires that its contractors have procedures and a system (15.2) for controlling nonconforming items. The contractor's QA organization is involved in documenting concurrence to the disposition, satisfactory completion of the disposition and corrective action. PGE receives nonconformance reports that are dispositioned
'* repair or use as is". These nonconformance reports will be reviewed and analyzed by PGE's Nuclear Project Quality Assurance Department and either the Nuclear Projects Engineer-ing Department or the Resident Engineer.
PGE's Nuclear Projects Quality Assurance Department reviews nonconformance reports to assure: (1) Disposition has been satisfactorily completed and closed out. (2) Cause of the nonconformance has been determined and, when applicable, proper effective action has been taken to preclude repetition. (3) Chronic problems and trends have been reviewed and analyzed and forwarded to management. (4) Nonconformance reports pertaining to conditions out-lined in the requirements of 10 CFR 50.55(e) have been properly processed. The following examples include situations which may warrant a corrective action request: (1) Repeated failure to follow approved procedures after previous violations have been reported. O Amendment 13 A.9-97 (January 1982)
(2) Nonconformance which, due to their repetition or (~'} impact (potential or actual) upon quality, should be brought to management's attention for special action. (3) Repeated failure to implement action to correct deficiencies discovered in audits by the commitment date if the lack of such action may contribute to
- a failure of the quality program.
i (4) Repeated disregard for documentation requirements. PGE's QA engineers review and analyze nonconformance reports and corrective action requests and follow the action taken to assure that it is effective. The analysis will consist of evaluating chronic problems and trends to aid in determining adequate QA/QC staffing, and in determining where QA/QC efforts should be concentrated in regard to witness and hold points, scheduling of surveillance and monitoring activities, ON training, etc. For PGE issued nonconformance reports, QA and Nuclear Projects Engineering approval / disapproval is required for dispositioning. . QA documents satisfactory completion of the dispositioning and . corrective action. i 2C14 Procedures are established indicating an effective corrective (16.1) action program has been established for contractors and sup-pliers performing quality-related activities. The appropriate QA organization reviews and documents concurrence with the i l procedures. Criteria for Determining QA Requirements 2D1 PCE's QA organization and the necessary technical organizations (2B3) will participate early in the procurement process (via A.9-98 Amendment 13 (January 1982)
specification review / approval) to identify the extent QA controls are to be applied to specific structures, systems and components. Bechtel Engineering considers the importance of design features and characteristics when defining technical, inspection and test requirements in the Technical Specifications. Bechtel Engineering, with QA participation, utilizes a unique ordering approach when specifying the QA criteria for procurements and j contracts. Bechtel's Quality Control and Procurement Supplier l Quality Representatives consider the specification requirements when preparing inspection instructions. The " graded approach" has been utilized by Bechtel for applying QA criteria to non-Q-related items when formalized QA programs are required. For items determined to be important to safety where specific QA controls cannot be imposed in a practical manner, an evalua-tion will be made by Bechtel to determine appropriate quality verification requirements to be applied during receipt, instal-lation or testing to provide the necessary assurance that the
'cem(s) meet project requirements.
2D2 For commercial catalog or "off-the-shelf" items where specific (7B4) QA controls appropriate for nuclear applications cannot be imposed in a practicable manner, quality verification require-ments shall be established and described to provide the necessary assurance of an acceptable item. 2D3 An effective inspection program will be established. Inspec-(10A) tion program procedures will provide criteria for determining the accuracy requirements of inspection equipment and criteria for determining when inspections are required or define how and when inspections are performed. The appropriate QA organiza-tion participates in the above functions. O Amendment 13 A.9-99 (January 1982)
2D4 Procedures will be established and described to identify, in (10D2) pertinent documents, mandatory inspection hold points beyond which work may not proceed until inspected by a designated inspector. 2D5 A test control progras will be established to include certain (llA1) proof tests prior to, installation and preoperational tests.- Procedures will provide criteria for determining accuracy requirements of test equipment and criteria for determining when a proof test is required'and how and-when testing activi-ties are to be performed'. - 2D6 Audits are conducted and the results analyzed by the appro-(18B1) priate QA organization. Audit reports indicate any quality problems and the effectiveness of the audited area. Re-audits of deficient areas are conducted as necessary to assure implementation of corrective action and recurrence control. Audit results are reported to management for. review and assessment. Qualification of QA/QC Personnel ', 2El Training, qualification.aud certification programs will be (2D) established so that: _ (1) Personnel responsible for performing activities affecting quality will be inotructed as to the purpose, scope and implementation of the quality-related manuals, instructions and procedures. (2) Personnel verifying activities affecting quality will be trained and qualiff ed (commensurate with their formal education and experience) in the principles, technfques and : requirements of the activities being performed. O G { A.9-100' Amendment 13 (January 1982)
(3) For formal training and qualification programs, documentation includes the objective, content of the program, attendees and date of attendance. (4) Proficiency tests will be given to those personnel performing and verifying activities affecting quality, and acceptance criteria vil be developed to determined if individuals are properly trained and qualified.
.(5) Certificate of qualifications will clearly delineate
' r the specific functions personnel are qualified to perform, and the criteria used to qualify personnel
- in each function.
(6) Proficiency of personnel performing and verifying activities af fecting quality will be maintained by coatinuing application of skills, retraining,
' reexamining, and/or recertifying as determined by management or program commitment. lh
. 2E2 A qualification program for inspectors and exaulners (including (10B2) NDE personnel) will be established and documented, and the qualifications and certifications of inspectors and examiners will be kept current.
QA Staffing 2F1 PGE-8003 will be updated to describe the Project QA organiza-(IAS) tions of PCE and Bechtel and to provide organizational charts which indicate onsite and offsite personnel. Criteria for determining the QA/QC staffing needs include work load and schedcle, and personnel ef ficiency. a The methods used to establish the work load and schedule include work packsging, development of Epecific work plans (including knendment 13 A.9-101 O (Jaiscary 19S2) { __ _ - - -
' s::- .
~ -
c.
~
+ '
3 3 Lingpections), identification of areas of inspection (QA/QC) '
"'s ' expertise needed, and participation in day-to-day staff, plan-
~
ning and. schedule meetings. .i lE
~ .
The factors considered in determining personnel efficiency 1 ' include experience and training,~ quality of work' coming to the , i "~ inspector or enaminer, quantity of work coming to the inspector
- or . examiner (work flow)', procedure ef fectiveness, and produc-tivity of personnel.
l The criteria for determining if QA/QC staff is adequate include the degree of~uninspected work, quality problems detected after
- icipection or examination, inspector or examiner preparation i time, close-cut of nonconformances and audits, follow-up cf
~
corrective actions, and complaints. 1 2F2 _ Long-range matching of QA/QC resources with work load will be (-) accomplished by QA review of projected work forces of the utility and.its contractors. This review will permit recruiting
~
and training activities to be carried out in such a manner as to
- provide trained QA' personnel necessary to assure the quality of
~
4 - work. QA/QC personnel will reevaluate staffing levels periodically (i~e, monthly) to assure they are adequate and then modify as i . necessary. Effectiveness of the staffing program will be assured by QA
, parti'cipation in work planning, surveillance and audit.
~
) -
,As-Built Documentation
- = ,,
' 2G1 PGE-8003 describes the scope of the document control program (6A1) and' includes "as-built" drawings in the document control l
2G2 system. Project procedures will'be established to provide i (6C1) for the preparation of "as-built" drawings and related A.9-102 Amendment 13 (January 1982) i !
!=
O ar-, 4 r , - - , .,-J.--, -t .w-. ,-.m , , , . . , . . w, . wr-. . ,.,,r- .ww_, e .re--i.~#+-- -- --
documentation in a timely manner to accurately reflect the actual plant construction. "As-built" drawings will be an item on the checklist prior to turnover to the preoperational test group. QA Role in Design and Analysis 2111 Procedures will be established to require documented che.ks to (3El) ensure the dimensional accuracy (including tolerances, ac.?ept/ reject criteria and inspectability) and the completenesa of the drawings and specifications. QC inspections of safety-rt: lated activities will be conducted using procedures or inspection checklists developed from the engineering specifications and drawings for the system, component or structure. 2112 Procedures will be established to require that design documents, (3E2) including drawings and specificctions, be reviewed by individ-uals knowledgeable and qualified in QA/QC techniques to assure that the documents are prepared, reviewed and approved in accor-dance with procedures and that documents contain or reference the necessary QA/QC requirements, such as inspection and test requirements, acceptance requirements, and the extent of docu-menting inspection and test results. l l Amendment 13 A.9-103 O (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT Dedicated Containment Penetration (3) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met. This information is of the type customarily required to satisfy 10 CFR 50.34(a)(1) or to address the applicant's technical qualifi-4 cations and management structure and competence. (iv) Provide one or more dedicated containment penetrations, equi-valent in size to a single 3-foot diameter opening, in order not to preclude future installation of systems to prevent containment failure, such as a filtered vented containment system. (II.B.8) RESPONSE YO 10 CFR 50.34(f)(3)(iv) (} The Pebble Springs Containment design will provide a single dedicated
\~ / 3-ft diameter penetration in order not to preclude the installation of systems to prevent Containment failure, such as filtered vented Contain-ment systems. This dedicated penetration will be capped and seal welded and will meet all requirements of existing penetrations. Space inside containment will be allocated for the Containment penetration assembly and a future inboard isolation valve, if required.
O
/
A.9-104 Amendment 13 (January 1982)
m.- . .-. .. _ _ . _ . , _ __m. . _ - . . . . . _ _ . . _ - - _ _ _ . . l NRC 10 CFR 50.34(f) REQUIREMENT ( Conta'inment Integrity and Hydrogen Control' l i (3) To satisfy the_.following requirements, the application shall provide 4 sufficient information to demonstrate that the required actions will be' satisfactorily completed by the operating license stage. J This information is of the type customarily required to satisfy i
'10 CFR 50.35(a)(1) or to address the applicant's technical quali-fications and management structure and competence.
i i (v) Provide preliminary design information at a level of detail consistent with that normally required at the construction permit ste.ge of review sufficient to demonstrate that: (II.B.8) [ (A) (1) Containment integrity will be maintained (i.e., for steel containments by meeting the requirements of the ASME S Boiler and Pressure Vessel Code, Section III, Division 1, ; t ' Subsubarticle NE-3220, Service Level C Limits, except that
- evaluation of instability is not required, considering
, pressure and dead. load alone. For concrete containments 2. i by meeting the requirements of the ASME Boiler and Pressure j 4 Vessel Code, Section III, Division 2, Subsubarticle CC-3720, Factored Load Category, considering pressure and , dead load alone) during'an accident that releases hydrogen
~
generated from 100 percent fue1 clad metal-water reaction accompanied by either hydrogen burning or the added pres-J. 4 sure from post-accident inarting assuming carbon dioxide is the inerting agent, depending upon which option is chosen for control of hydrogen. As a minimum, the specific code requirements set forth above appropriate' for each type of containment will be met for a combination of dead i
- t load and an internal pressure of 45 psig. Modest devia- ~
I tions from these criteria will be considered by the staff, if-good cause is shown by an applicant. Systems necessary to ensure containment integrity shall also be demonstrated - . 1 A.9-105 Amendment 13 ) [ (January 1982)
, .n-w,, - -~ .-,-,-,m,, , ,- ,~~.m. ,, ,,-m,y.,.-y-n, -,,. .,------r ,s , - , r,,m- ,n-r. w-
to perform their function under these conditions; (2) Subarticle NE-3220, Division 1, and Subarticle CC-3720, Division 2, of Section III of the July 1, 1980 ASME Boiler and Pressure Vessel Code, which are referenced in 50.34(f)(3)(v)(A)(1) and 50.34(f)(3)(v)(B)(1), were approved for incorporation by reference by the Director of the Office of the Federal Register. A notice of any changes made to the material incorporated by reference will be published in the Federal Register. Copies of the ASME Boiler and Pressure Vessel Code may be purchased from the American Society of Mechanical Engineers, United Engineering Center, 345 E. 47th Street, New York, NY 10017. It is also available for inspection at the Nuclear Regulatory Commission's Public Document Room, 1717 11 Street NW, Washington, D. C. (B) (1) Containment structure loadings produced by an inad-vertent full actuation of a por,t accident inerting hydrogen control system (assuming carbon dioxide), but not including seismic or design basis accident loadings will not produce stresses in steel containments in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsub-article NE-3220, Service Level A Limits, except that evaluation of instability is not required (for concrete containments the loadings specified above will not produce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code, Section III, Division 2, Subsubarticle CC-3720, Service Load Category). (2) The containment has the capability to safely withstand pressure tests at 1.10 and 1.15 times (for steel and concrete containments, respec-tively) the pressure calculated to result from carbon dioxide inerting. O Amendment 13 A.9-106 (January 1982)
RESPONSE TO 10 CFR 50.34(f)(3)(v) (A) The Pebble Springs Containment is designed so its integrity is maintained during an accident that releases hydrogen generated from 100 percent active fuel clad metal-water reaction accompanied by hydrogen burning. The Containment final design will be included in the FSAR. The preliminary Containment design is in accordance with the ASME Code, Section III, Division 2 for 60 psig. To assure that the pressures and temperatures resulting from the combustion of hydrogen released f rom a 100 percent active fuel clad metal-water reaction are within the Pebble Springs Containment structural capability, Pebble Springs has been compared to Pilgrim 2. Pilgrim 2 was selected because: (1) Both have large dry Containments designed by the same A-E (2) Both have post tensioned, reinforced concrete-steel lined Containments (3) Both Containments are of similar size having a relatively open arrangement l (4) Pilgrim 2 was the lead NTCP plant with a large dry Containment that has done preliminary analysis. l t While both plants have different nuclear steam system suppliers (B&W [ for Pebble Springs, CE for Pilgrim 2), the amounts of zirconium subject to the metal water reaction are similar, with Pebble Springs-having about 9 percent more zirconium. Other parameters, such as spray flows, heat sink data, and fan cooler capacity are shown in Table A.9-9 for comparison. Even though Pebble Springs has more zir-L conium, the peak pressure in Containment will be approximately the
,A same, since igniters will burn the hydrogen as it reaches a certain i
1 A.9-107 Amendment 13 (January 1982)
concentration, not the entire amount to be released. Based on the similarities between Pebble Springs and Pilgrim 2, there is reasonable assurance that the pressures and temperatures resulting from the combustion of hydrogen are within the Pebble Springs Containment structural capability. Final plant design will include consideration of the following aspects of a hydrogen burn in Containment: (1) A variety of accident sequences which are characterized by: (a) Eeing important contributors to risk. (b) Accident evolution times that are sufficientiv long to allow for practical cooling-recovery (eg, WASH-1400 accident sequences S2D and S2H). (2) Hydrogen and steam production and release rates into the Con-tainment will be analyzed with deterministic analysis based on the above-described sequences or with parametric analysis to envelop these release rates. (3) Igniter performance and endurance characteristics (if the option chosen for hydrogen control continues to be a deliberate ignition system). The design of systems necessary to ensure Containment integrity will specifically include use of Containment temperature profiles that result from the postulated hydrogen events as input to the analyses or tests which will demonstrate equipment survivability and/or qualification. The location of components associated with these systems and method j of protection (if required) will be described in the FSAR. l l O Amendment 13 A.9-108 (January 1982)
l ~.. - - -" '- - -'- 7 s t i I I !-. . (B) 'Inerting as a hydrogen control measure is not proposed for the Pebble
- Springs Containment design.= Therefore, this ites_is not applicable, i '
! i 6
+
1 i l i l
- I I
t P l f '-i l l t f l l l i 1 ! -i I ! l ! i 1
-i i-I I
j ! A.9-109 Amendment 13 ; ! (January 1982) !
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NRC 10 CFR 50.34(f) REQUIREMENT m Dedicated Penetrations (3) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met. This information is of the type customarily required to satisfy 10 CFR 50.34(a)(1) or to address the applicant's technical qualifications and management stracture and competence. (vi) For plant designs with external hydrogen recombiners, provide redundant dedicated containment penetrations so that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere. (II.E.4.1) RESPONSE TO 10 CFR 50.34(f)(3)(vi) The Pebble Springs Nuclear Plant design basis includes requirements for hydrogen recombiners and a backup hydrogen venting system as a means of combustible gas control inside Containment. The decision whether to 'use internal or external recombiners has not been made (see PSAR Section 6.2.4.2). If the external Containment Hydrogen Recombiner System is chosen, four dedicated Containment penetrations will be provided (two per recombiner) in order to meet the single-failure requirements for both Containment isolation and post accident recombiner operation. Figure A.9-10 shows one design which is capable of meeting the above requirements. If it is decided to use permanently installed internal hydrogen recom-biners, no penetrations would be necessary for the recombiners. The previously noted arrangement for the hydrogen vent system would still apply. 0' l G A.9-110 Amendment 13 (January 1982)
NRC 10 CFR 50.34(f) REQUIREMENT Organization and Staffing to Oversee Design and Construction (3) To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met. This information is of the type cus'tomarily required to satisfy 10 CFR 50.34(a)(1) or to address the applicant's technical qualifi-cations and management structure and competence. (vii) Provide a description of the management plan for design and construction activities, to include: (A) the organizational and management structure singularly responsible for direc-tion of design and construction of the proposed plant; (B) technical resources directed by the applicant; (C) details of the interaction of design and construction within the applicant's organization and the manner by which the appli-cant will ensure close integration of the architect engineer and the nuclear steam supply vendor; (D) proposed procedures for handling the transition to operation; (E) the degree of top level management oversight and technical control to be exercised by the applicant during design and construction, including the preparation and implementation of procedures necessary to guide the effort. (II.J.3.1) RESPONSE TO 10 CFR 50.34(f)(3)(vii) PSAR Sections 1.4 and 13.1 have been revised to describe the PGE corporate organizational structure for design, construction and operation of the Pebble Springs Nuclear Plant. The following supplements material in Sections 1.4 and 13.1 and describes PGE's program for management of design and construction activities and for transition to operation. O A.9-lli Amendment 13 (January 1982)
Organizational and Management Structure PGE has overall responsibility for the design, procurement, fabrication, construction, preoperational testing, operation and quality assurance (QA) activities for the Pebble Springs Nuclear Plant. PSAR Sections 1.4 and 13.1 describe PGE's organizational and management structures for the Pebble Springs Nuclear Plant. Figure 13.1-1 shows the overall Pebble Springs organization and indicates the interface relation-ship between PGE, Bechtel, BW and General Electric (GE). Figures 13.1-2, 13.1-3 and 13.1-4 show PGE's organization for management of design, con-struction and operation of the Pebble Springs Nuclear Plant. Technical Resources Directed by the Applicant (1) Staffing Levels: During Pebble Springs construction, PGE will maintain an in-house staff of engineers and managers to oversee the design, procurement, fabrication and construction management activities and verify con-formance with applicable regulations, codes and design criteria. In specific ccses where PCE's in-house staff is not sufficient to meet Pebble Springs responsibilities, temporary technical support is assigned from in-house line organizations or outside consultants contracted to work under the direction of PGE personnel. PGE will systematically develop manpower plans annually based on projected work requirements developed by cognizant managers. These manpower plans will be periodically reviewed and updated as required by actual workload. (2) Level of Education and Experience: PGE has and will continue to retain a highly trained and capable technical staff to meet the responsibilities of managing the design and construction of Pebble Springs. Table 13.1-1 lists PGE's current technical resources and experience. In addition to these resources, O Amendment 13 A.9-112 (January 1982)
p there-is a wide range of technical expertise within the PGE corporate organization covering all major engineering disciplines plus some of the more highly specialized fields (eg, instrumentation and control engineering, metallurgy and materials, chemistry, health physics). Also available are program supervisory expertise in geology / seismology, meteorology, fisheries and aquatics and general environmental and con-struction impact disciplines. If a technical issue arises that is outside the scope of PGE's technical staff's engineering capabilities, services of outside experts may be utilized to assist in resolving the issue. (3) Training and Experience Feedback: In addition to hiring experienced individuals, PGE has an active technical training program. All professionals have the opportunity and are expected to attend courses, seminars and workshops as a means of staff development. They also remain cognizant of current industry concerns and activities associated with the following: (a) Edison Electric Institute Nuclear (b) Electric Power Research Institute (c) llealth Physics Society (d) Owners groups (e) Atomic Industrial Forum (f) Nuclear Safety Analysis Center (g) Institute of Nuclear Power Operations (h) American Nuclear Society. In addition, PGE also uses temporary assignments of personnel to other organizations an a means of staff development.
%J A.9-113 Amendment 13 (January 1982)
It should be noted that the above participation significantly broadens the experience of PGE personnel. Additional information on the feed-back of design, construction and operating experience is provided in the response to 10 CFR 50.34(f)(3)(1). Details of the Interaction of Design and Construction Activities (1) General: The following supplements the material in PSAR Sections 1.4 and 13.1 on the interaction of design and construction activities by PGE and the principal contractors, Bechtel, B&W for the nuclear steam system, and GE for the turbine generator. Establishment of the division of responsibility and the means of assuring close integration of the work is established in contractual documents, project procedures, and the B&W Balance of Plant (B0P) Criteria document. PGE has the overall responsibility for the design, construction and operation of the Pebble Springs Nuclear Plant in accordance with NRC regulatory requirements, including the quality assurance requirements of 10 C"R 50, Appendix B. PGE also has the responsibility for pro-viding management oversight of principal contractor activities, approving basic design criteria and releasing design documents. Bechtel is responsible for project engineering management, planning, cost control, engineering, construction management, contract adminis-tration, quality control, quality assurance, and BOP preoperational testing. Bechtel is also responsible for design interface control and/or coordination among Bechtel, B&W and GE and between Bechtel and its contractors. Bechtel will perform its services in accor-dance with applicable Federal, State and local codes and regulations including the quality assurance requirements of 10 CFR 50, Appendix B. PGE monitors and evaluates Bechtel's performance of these responsi-bilities by requiring Bechtel to obtain PGE's approval of the basic design criteria, release of selected design documents prior to pur-chase or construction, and acceptance upon completion of construction. O Amendment 13 A.9-114 (January 1982)
. . .. - _- . - , - -- . =- - - - - - - - . .
~
B&W-is. responsible for design and fabrication of the NSS',. including preparation of design documents and procurement of related hardware.
\ Bechtel reviews these documents to provide interface coordination between the NSS and' BOP. PGE also reviews the B&W design and inter-face with BOP systems.- Otherwise B&W has authority to determine the NSS design subject to PGE QA surveillance. B&W prepares interface-criteria, safety analyses, other~ design information, test procedure l
guidelines and technical support for NSS installation.~- B&W is accountable to PGE to perform its services and provide NSS designs 4 and equipment in accordance with applicable Federal, State and local codes and regulations, including the quality assurance. requirements-of 10 CFR 50, Appendix B. I GE is responsible for design and fabrication of the turbine genera- , , . tor, which does not include activities subject to:the~ quality assur-l ance requirements of 10 CFR 50, Appendix B. (2) Oversight of Design: 4 PGE has the overall responsibility for the design of the Pebble Springs Nuclear Plant, including management of all design activities. l
- The Pebble Springs Project Manager and technical managers under the direction of the General Manager, Technical Functions (who' reports to the Vice President, Nuclear) will manage the design and procure-ment of Pebble Springs.~ The Pebble Springs Project Manager will be
! accountable for the cost, schedule and quality of Pebble Springs and { will assign responsibilities to the various support groups under the General Manager, Technical Functions. These support. groups manage l the contracts of Bechtel, B&W, GE and outside consultants and provide technical direction to the principal contractors. The Purchasing and Materials Management. Department, under the Vice President, Operating Services, is responsible for PGE's contracting
- and procurement activities. The Purchasing Department acts on the l
!O A.9-115 Amendment 13 (January 1982)
advice and recommendation of the Nuclear Projects Engineering Depart-ment under the General Manager, Technical Functions. The Nuclear Projects Engineering Department has primary responsibility for pro-curement technical matters, whereas the Purchasing Department is principally responsible for commercial aspects. Both departments participate in all meetings and decisions. In addition to its specific control aspects over design and procure-ment activities, PGE monitors the quality, cost and timeliness of other activities performed by the principal contractors. Management oversight of contractor design activities is facilitated by the issu-ance of several status and performance reports which are directed to various levels of management. Also, copies of correspondence among contractors are sent to PGE for information. (3) Oversight of Construction: The Nuclear Projects Construction Manager and his resident engineer-ing staff will be responsible for management oversight of contractor construction activities on Pebble Springs. The Nuclear Project's Construction Manager will report to the General Manager, Technical Functions, as shown in PSAR Figure 13.1-3. The Nuclear Projects Construction Manager and his staff are responsible for construction overview of contractor performance. The contractors and subcontrac-tors under Bechtel construction management are responsible for con-struction activities that conform to design quality requirements. The Nuclear Projects Constructica Manager and his staff monitor con-struction activities, approve schedules, field procurements, selected invoices and other financial controls, monitor coepliance with permit and license requirements, monitor procedure compliance with permit and license requirements, monitor procedure compliance, monitor coor-dination of Bechtel field engineering with Bechtel home engineerng staff and coordinate the Bechtel Field Construction Manager's turn-over of plant systems to PGE. O Amendment 13 A.9-116 l (January 1982)
Quality assurance responsibilities are described in Section 17 of
,-~
4 [j the PSAR. PGE provides construction oversight through PGE's Nuclear Projects Quality Assurance Department which is responsible for moni-toring the QA aspects of site construction, including review of con-tractor site procedures, audits and surveillance of construction,.
' identification of quality problems and monitoring of their resolu-tion; and acceptance reviews of components, constructed structures and completed systems. The PGE Site QA engineers interact with principal contractor site organizations through Bechtel and with the PGE home office QA organization.
PGE will assure that approved procedures exist for construction man-agement and control prior to the start of each safety-related con-struction activity. These procedures will reflect the organization and conform to applicable regulatory requirements, contractual arrangements, and PGE's QA Program (PGE-8003). Procedures will exist for each organizational element involved in safety-related construc-tion oversicht activities. (~~s
\-- Transition to Operation 4'
PGE's Vice President, Nuclear, has overall responsibility for the Pebble Springs design, construction, fuel, QA and operation. This Vice President functions under the oversight of PGE's President. PGE's staff function-ing under the direction of-the Vice President, Nuclear, will oversee the Pebble Springs design and construction. The Trojan Nuclear Plant opera-tions staff is also under the direction of the Vice President, Nuclear. They will be a basic source of experienced operations personnel for Pebble Springs and will greatly facilitate the transition from construction to operation. The PGE resident engineering staff, physically located at the site during conatruction and startup, will be another resource for actual transfer to the operations or engineering support groups. The PGE tech-nical staff responsible for review and approval of plant design will also be available, as a technically cognizant resource, during Pebble Springs i startup and operation. O (_-) A.9-ll7 Amendment 13 (January 1982)
. -__ _ . _ . . . -- .. .- - - _. , ~ .
Once Pebble Springs becomes operational, PGE will provide the required technical support necessary to assure safe and reliable plant operation. This support will be consistent with the guidelines suggested in NUREG-0731. Prior to start of Pebble Springs operation, PGE will consider various organizational alternatives to ensure that: (a) unambiguous management control and effective lines of authority and communication are maintained among the organizational units involved in the management, technical sup-port and operation of Pebble Springs; and (b) any potential conflict with the application of resources to non-nuclear functions of the utility are minimized. PGE intends to employ the operating staf f with ample lead time for them to learn the Pebble Springs design and operation as discussed in PSAR Section 13.2. Furthermore, it is PGE's personnel policy to open new tech-nical staff positions to internal staff and to encourage transfers. Thus, Trojan Nuclear Plant operations personnel, as well as engineering and man-agement personnel involved in the Pebble Springs design and construction phases, will be encouraged to transfer to Pebble Springs operating staff positions as they become available, facilitating the transfer of experi-ence to plant operation. B&W, the NSS supplier, will provide instruction manuals for various pieces of NSS equipment. These manuals will include operation and main-tenance instructions which will be used as references during formation of the Pebble Springs startup, maintenance and operating procedures. PGE may request additional procedure guidance from B&W during all phases of plant construction or operation. This will help ensure that plant opera-tions reflect the engineering expertise in plant design. Operating and maintenance procedures will be written by the plant staff with assistance, as necessary, from the startup organization, principal contractors and consultants with PWR experience. During this period, the operating staff will have the opportunity to interface directly with personnel in the design organizations. Bechtel and B&W will provide inputs to the procedures and will review completed draft procedures as appropriate so that design information is accurately reflected. These O Amendment 13 A.9-118 (January 1982)
procedures will be developed on a schedule which will permit their use / for operator training and the startup test program. D The operating staff will be directly involved in the preoperational and startup test programs. The startup organization will be under PGE's direc-tion and will be an integrated group including Bechtel, B&W and GE person-nel. The integrated nature of this group should facilitate communications between these organizations and thus enhance the transfer of design and equipment performance information to the plant staff. The trial use of plant procedures during the test program is described in PSAR Section 14.1.1. This process should provide further assurance that design information and base line data are incorporated into the plant operating procedures. Management Oversight PCE corporate functions, responsibilities and authorities are summarized ,f'" in PSAR Sections 1.4 and 13.1. PCE, under a joint ownership agreement \s_s' with other utilities, has sole responsibility and is fully authorized to act for the owner utilities with respect to design, construction and operation of the Pebble Springs Nuclear Plant. PCE exercises top-level management oversight by assigning the responsi-bility for design, construction and operation of Pebble Springs to the Vice President, Nuclear. The Vice President, Nuclear, regularly reviews status and progress information, and will be informed of significant pro-ject decisions, issues, problems and project plans for resolution of issues and problems. Regular meetings will be held by the Vice President, Nuclear, to discuss project design and construction status and problems. The Vice President, Nuclear, will also hold periodic project status meet-ings with the owner utilities. The Pebble Springs Project Manager will provide periodic reports to PGE's Vice President, Nuclear, the General Manager, Technical Functions and b V A.9-119 Amendment 13 (January 1982)
the Pebble Springs owner utilities. These reports will identify pro-gress, current difficulties and planned activities over the next report-ing period and will ensure that top-levei management is aware of Pebble Springs activities. PGE's Vice President, Nuclear, and/or the General Manager, Technical Functions, will be in frequent communication and hold regular meetings with Bechtel and B&W management so they will be informed of the project status, management and technical issues, and plans for the future. Additional management oversight is provided by a project engineering com-mittee, consisting of members from each owner utility, who will consider significant matters related to the engineering and construction of the plant and make recommendations for PGE's action. ) I O O Amendment 13 A.9-120 (January 1982)
TABLE A.9-1 INITIATING EVENTS FOR RISK STUDY
- 1. LOCA i
- 2. Transients
- 3. Steam /Feedwater line breaks
- 4. Steam generator tube rupture
- 5. Failures during cold shutdown operation
- 6. Station blackout, loss of AC/DC.
O O Amendment 13 (January 1982)
TABLE A.9-2 OUTLINE OF RISK STUDY REPORT I. INTRODUCTION II.
SUMMARY
III. METHODOLOGY OVERVIEW A. Event trees B. Fault trees C. Quantification of accident sequences D. Treatment of uncertainties IV. SYSTEM DESCRIPTIONS A. Performance requirements B. Actuation C. Environment considerations D. Dependency diagrams for support systems power, cooling and lubrication V. CORE MELT PROBABILITIES A. Dominant sequences B. Dominant cut-sets VI. PLANT MODIFICATIONS THAT ADDRESS DOMINANT SEQUENCES A. Improvement in reliability expected B. How factored ~into design, equipment purchase, fabrication, procedures, operations, etc C. Basis for not implementing more reliable alternatives VII. APPENDICES (DETAILS OF STUDY) Amendment 13 O (January 1982)
a. TABLE A.9 HICH-ACTIVITY SOURCE TERMS I 'l Percent of. Equilibrium Source-Tera- Component Core Inventory Containment Noble Gases 100 % Airborne Halogens' 25 %
" Source A" Reactor Coolant Noble Cases 100 %
" Source B" Halogens ~ 50 % 1 Solids 1%
Containment ' Halogens 50 % Sump Solids 1%
" Source C"
[a] Source A. dilution consists of the Containment free volume. (ignoring the water volume. occupied by ECCS operation). Source B is utilized for non-LOCA events where a nonsech-anistic type failure of.the reactor core occurs. The l activity is diluted by the water volume of the RCS. For j i systems which operate after primary system depressuriza- !- tion, credit is taken for noble gas removal. Source C L is utilized for post-LOCA conditions in those systems utilized te recirculate Containment sump water. Source C , is diluted by the RCS, borated water storage tank and l core flooding tank volumes. Radioactive source strength for each system will be a function of the decay time period available prior to system actuation. f
\
Amendment 13 (January 1982).
O Q 'J TABLE A.9-4 Sheet 1 of 4 POST-ACCIDENT ACCESS REQUIREMENTS Access Required 3 Equipment ' Source Area Term [a] Function 1 2 3 Comments [c] DHRS NA LPI from BWST X Automatic initiation - important parameters monitored from control room C Sump recirculation (LPI) X Automatic initiation - important j parameters monitored from control room B Shutdown cooling X Initiated from control room C HPI feed (" piggyback") X Initiated and monitored from control room MPS (HPI) NA HPI from BWST X Automatic initiation - important parameters monitored from control room C HPI from DHRS feed X Automatic initiation - important (" piggyback") parameters monitored from control room CSS NA Spray from BWST X Automatic initiation - important parameters monitored from control room NA Hydrazine injection X Automatic initiation - important parameters monitored from control room C Sump recirculation X Automatic initiation - importtnt parameters monitored from control room CCWS II NA DHR' heat exchange cooling X Normally operating system - monitored locally and remotely to ensure normal operation NA Containment cooling- X Normally operating system - monitored locally and remotely to ensure normal operation NA ESF pump seal and lube oil X Normally operating system - monitored
,sf g cooling locally and remotely to ensure normal operation g g, mB Normally operating system - monitored qg CCWS[d] NA Standby DC heat exchange X
" locally and remotely to' ensure normal g[ operation D
{ v
)
G G TABLE A.9-4 Sheet 2 of 4 Equipment Source Access Required II Area TermI *l Function 1 2 3 CommentsI #l SWS I 'l NA CCW heat exchange cooling X Automatic initiation of Category I portion - monitored locally and remotely. to ensure normal operation AFSI *I NA Cool RCS by providing X Automatic initiation - system monitored OTSG feed locally and remotely to ensure normal operation Standby NA Provide emergency power X Automatic initiation - system monitored j diesel locally and remotely to ensure normal generators operation and auxiliaries *l Fuel handling [f] ESF room ventilation X Automatic initiation - system monitored area exhaust locally and remotely to ensure normal equipmentI 'l operation Hydrogen A Reduce Containment hydrogen X Potential Containment isolation valve recombiners concentration operation - system design not complete at this time , Hydrogen A Monitor Containment hydro- X Potential Containment isolation valve analyzers gen concentration operation manually Hydrogen A Reduce Containment hydrogen X Access required for manual operation of vent system concentration (backup to system isolation valves. Further access equipment recombiners) may be required to ensure normal operation Containment A Detercina Containment X
,s pressure- atmosphere pressure g sensing lines uo Shielded to ensure post-accident exposure g y Control room NA Operations control X limited to 5 rem whole body (CDC 19)
Qg n
;; ~ Control room NA Maintain habitability of X Post-accident ESF filtration initiated oo u HVAC control room automatically - access may be required to 13 maintain normal operation
s
('~N )
TABLE A.9-4
.v h Sheet 3 of 4 Equipment Source Access Required II Area TermI *I Function 1 2 3 CommentsI *I Cable NA Routine inspection X Access may be required to ensure normal spreading operation rooms-Electrical / NA Routine inspection X Access may be required to ensure normal-
.switchgear operation room Auto gas NA Normal processing X Access may be required to ensure normal analyzer operation CRS panel NA Normal processing X Process return line to the Containment to be added DRS panel NA Normal processing X Auxiliary C Collect ESF pump and heat X Vent to be routed to ESF room ventilation Building exchanger room floor drains exhaust. Liquid effluent to be pumped drain sumps back to Containment Effluent routed to RC bleed holdup tanks CRS panel II NA Normal processing X SRSI *I NA Normal processing X (including baler)
Onsite TSC NA Engineering support to X Subject to post-accident exposure limit. operators of 5 rem whole body (GDC 19) Onsite OSC NA Staging area for operations X Subject to post-accident exposure limit personnel of 5 rse whole body (GDC 19)
,,I 4 Shielding changes will be made. The Aid in determining extent X g g Primary A, B extent of change is dependent upon the g g sampling and C of core damage Q@ system I*
final design of the lab area utilized for C I8} n taking high-activity post-accident samples 5- Normal processing X Access.may be required to ensure normal gu SFPCS[d] NA operation f
,/ '-
i i
I
?
/
T .c. A.9-4 ShNt'4cf4 Equipment Source Access kequired l Area ' Term [a] ' Function ' 1 2 3 Comments ICI ESP HVAC NA Cool ESF equipment Fan coil units compartments X ESF CHWS *l NA Supply chilled water to ESF X Access may be required to ensure normal equipment rooms, control operation room and switchgear room ' [a] From Table A.9-3. [b] 1 = none, 2 = infrequent, 3 = continuous. Areas requiring infrequent or continuous access are classified as vital areas. [c] Instrumentation necessary for assessing system operation will be upgraded as necessary to meet environmental qualification requirements consistent with the source terms of Table A.9-3. [d] Complete access may not be achievable until 1 day or more after the accident. [e] Immediate access should be possible. [f] Collects halogens on carbon filters from ESF room leakage. [g] Depends on type of sample taken. CI EB B B-G!a" e
'T TABLE A.9-5
() , ESSENTIAL SYSTEMS '
~
Makeup and Purification System
- High pressure injection portion
- Reactor coolant pump seal injection portion s.
Decay Heat Removal' System - Low pressure injection portion *l
- Recirculation sump suction portionII ,
Containment Spray System ^ Component Cooling Water System
- Containment air cooler portion r -
Auxiliary Feedwater System "I Main Steam System
- From S/G to MSIVs Containment Pressure Sensors s
~
. ~
_s s [a] Power to the low pressure injection line isolation valves is s . racked out to prevent inadvertent closure of the valve ~.~ , d [b] Recirculation sump outboard isolation valves are normally closed and opened automatically on two out of three low level signals '- from BWST. The inboard Containment isolation valves are normally-open and manually operated from the control room board. s [c] Auxiliary feedwater control valves- ara controlled by the ECI; ' auxiliary feedwater control and isolation:yalves are closed ' automatically by the FOGG system. .
~
O 1 Amendment 13 (January 1982)
s
?. _, ..
i - x-
, TABLE A.9-6 Sheet 1 of 2 NON-ESSENTIAL SYSTEMS o .
Makeup and Purification System s
- RCS normal makeup and letdown portions [a]
.:7 R'eactor coolant pump seal return portion [b]
- j. -
e r Core flood tank fill _ Fire Water System
', ' Miscellaneous'Cas Supply System Decay Heat Removal. System Drop line portion Auxilie.ry spray. port' ion I,- , Fuel' Transfer System Component Cooling Water System
)1
; CRDM cooling Iater service portion Reactor coolant pump service portionI #l f >
Letdown-cooler. service portion a] ^
~ .i .,
. Reactor coolant pump seal return cooler portion 7
4 Clean Radioactiv.e Waste System Dirty Radioactive Waste System i f Main Feedwater System [d] I ' Main Steam Systeta b, ~
~ *
. Portio downstream of the MSIVs II
.e
' Primary and Containment Sampling Systems l
l Containment Purge Systems I*l Hydrogen Recombine'r. System IIl I~ ' I Containment Chilled Water System i 4 +-
- _ Steam Generator' Vacuum Heatup Systems l,' ) '
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., 1 i - - . Amendment 13
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- (January 1982)
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TABLE A.9-6 Sheet 2 of 2
-[a] Normal letdown (including CCW supply to the letdown coolers) can
~ -
be reopened for RCS inventory control, if necessary, or to degas for cooldown if the RCPB is intact and significant fuel damage 4 has not occurred. [b] RCP seal return (including CCW supply to the seal return coolers) can be reestablished if RCPs are to remain operating or be restarted. _
, [c] Current design assumes that a test will be conducted to verify that the.RCPs can operate for 30 min without CCW or operator action. If the test is not conducted or is unsatisfactory, safety-grade instrumentation will be added to initiate auto-matic protection of the Plant or the CCW system to the RCPs will be reclassified as an essential system and upgraded to Seismic Category I, Quality Group C requirements.
[d] The main steam isolation valves and the main feedwater isolation va1ves will be closed by ESFAS signals. Main feedwater control s and bypass control valves are closed by buffered non-Class lE Channels A and B ESFAS signals. [e] All Containment purge valves are closed on high airborne radia-tion levels in Containment. Containment low purge isolation valves can be reopened if it becomes necessary to use the hydrogen vent system. [f] Hydrogen recombiner penetrations may not be necessary if internal hydrogen recombiners are used. O Amendment 13 (January 1982)
O O O TABLE A.9-7 Sheet 1 of 2 LEAK REDUCTION MEASURES OF POTENTIALLY RADIOACTIVE SYSTEMS System? (E = existing provistori, M = modifications) MPS DHRS CSS CABRS GRS CRS DRS SRS SCS SFPCS Physical Design Provisions to Prevent Leakage Welded construction to the maximum extent practical E E EI "l E E E E E E E Component overflows and drains hardpiped to i collection tanks E E E E E E - - - - Isolation valves with stem leak-off hardpiped to collection tanks or of diaphragm, plug or sealed design E E E[a] E[d] E E E E E - l Pumps with mechanical seals with leak-off hardpiped to collection tank E EI *l E E E E EI - - - Relief valves which can be removed for testing E E E E E E E E - E Component vents connected to gaseous radwaste and vent collection system or sumpe E EI E E E E I8I E E - Additional Measures to Control Leakage hfAutomaticisolationofpotentiallyhigh g g radioactive sources E E E - - - - - M - ma EkESFfiltrationofventilationexhaust M[h] E/M E/M - - - - - - - es
TABLE A.9-7 Sheet 2 of 2 C$ n E 8-na Systems M$ (E = existing provision, M = modifications) MPS DHRS CSS CABRS CRS CRS DRS SRS SCS SFPCS 3-O" Testing ani Prevention Measures to Detect Leakage _ E E E E E E E E EI '} E Hydrostatic (pneumatic) test per applicable code E E E II - - - - - E Inservice inspection of welds E l Periodic leak rate testing M Il M{ I M Il - - - - - - [a] Except for MPS chemical addition tank and CSS hydrazine storage tank. [b] Only portion of DHRS and CSS used during recirculation mode of operation. l l [c] Drains are directed to DRS through open drains. [d] RC bleed evaporators and associated valves are B&W supplied. [e] B&W has previously investigated providing two seal leak-off connections for the DHR pumps; however, to accommodate two connections, the DHR pump shafts would have to be lengthened. [f] Pumps have double mechanical seals; leak-off is directed to open DRW drains. l [g] Tanks are vented to VCH; standpipes are vented to atmosphere; pumps are vented to sumps. i [h] Only portion of MPS used during piggyback mode of operation. [i] Only pump cischarge lines will be hydrostatically tested. [j] Only the boric acid addition portion of the system. 9 - O e
TABLE A.9-8 Sheet 1 of 2 ISOLATED AND POTENTIALLY HIGHLY CONTAMINATED SYSTEMS [%./ (A) Radioactively contaminated systems which are not expected to contain highly radioactive fluids: (1) Makeup and Purification System (MPS)[a] a) Letdown portion b) Purification portion c) RCP seal injection portion d) Normal makeup portion (2) Decry Heat Removal System (DHRS) a) BWST suction portion (3) Containment Spray System (CSS) a) Hydrazine addition portions - (4) Chemical Addition and Boron Recovery System (CABRS)[a] (5) Spent Fuel Pool Cooling and Cleanup System (SFPCS) L/ (6) Clean Radwaste System (CRS) (7) Dirty Radwaste System (DRS) (8) Solid Radwaste System (SRS) (9) Gaseous Radwaste and Vent Collection System (GRS)I*I [a] These systems are not expected to be utilized in the post-accident (LOCA and non-LOCA) condition if significant core damage is present and are therefore not considered to contain high activity sources. Reactor coolant degasification is considered the only essential post accident function related to these systems and alternate means of degassing exist via the RCS vents inside Containment which do not result in the movement of high activity fluids outside Contain-ment. Further analyses will be conducted to confirm that these systems are not required for post-accident mitigation or recovery for non-LOCA accident scenarios (intact RCS) that result in significant core damage. (n) v Amendment 13 (January 1982)
TABLE A.9-8 Sheet 2 of 2 (B) Radioactively contaminated systems which are expected to contain highly radioac ive fluids: (1) Decay lleat Removal System (Low Pressure Injection and Recirculation Portion) (2) Containment Spray System (3) Makeup and Purification System (High Pressure Injection System During Piggyback Mode) (4) External liydrogen Recombiners (if required) (5) Hydrogen Vent System (Downstream of Containment Isolation Check Valve on Supply Side and Upstream of Filters on Discharge Side) (6) Post-Accident Sampling System (Pressurizer, Containment Atmosphere and Hydrogen Content, and Decay Heat Removal) (7) Containment Pressure Sensing System (8) Auxiliary Building Sump Collection System (SCS) O Amendment 13 O (January 1982)
. _ _ .. . - . . - _ .m . _ . _ . _ _ . -_ _ .- . _ _ . _ _ _ . _ . _ - _ .. _ -_ __. __
i TABLE A.9-9 COMPARISON OF CONTAINMENT PARAMETERS s-FOR PEBBLE SPRINGS AND PILCRIM 2 Item Parameter Pilgrim 2 Pebble Springs Containment Diameter (ID) (ft) 143' 130 6 Volume (ft )- 2.5 x 10 2.45 x 10 6 Initial Temp (*F) 120 120 P
- Spray System Design flow rate 3,600 3,000 per pump (gpm) '
Temp (*F) 120 90 i Fan Cooler Number of fans No credit taken 2 of 4
- System used post-accident for fan coolers in Pilgrim analysis i
Flow rate (cfm) - 32,000 Heat removal rate - 94.76 x 106 (with (Btu /hr) air entering at > 280*F)
- Passive' heat Area (ft ) 235,000 259,000 i sinks L
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, (January 1982) 4 1
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Nuclear Povver Plant V Control Roon. Design Perform Operations Analysis to Defene Functions and Interactions Re41 locate ~~ ir {~ Analyre and Allocate Each Function to: l
- Personnel
- Machines - l
- Both Personnet and Machines i
I 1f I Verafy and Validate the Allocation of Functeons l
, + l Verify Functions Verify Functions Allocated to Personnel l
Allocated to Machines I 1f l Analyse Tasks to: Analyze Tasks to Define:
- Define Performance Parameters - Human Performance Parameters
- Data Needs p - Specify Desip of Machines
- Identify Equipment . - Decision Points l
g N - Measure Performance l
' I Analyze and Verify: l
- Work Station Design
- Operations Sequence l
- Operator Workload
- Human Error Rete l
- Work Station Links
_ _ _ _ ___ __ _ _l It i Specify Control Room Configuration l l Valedate System l integration ( - Synthesire Operations l i _ . _ _ _ _ _ _ _ _ _ _J 1r " Document Control Room Design Specifications Figure A.9-1 Preliminary Plan for Control Room Design Review
Amendment 13 (January 1982)
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O O O FRONT ENTRANCE RECORDS DECON CONFERENCE WC STORAGE ROOM ROOM OFFICE WC l 0FFICE KITCHEN COMMAND OFFICE COMMUN TIONS RAC CENTER MAINTENANCE OFFICE RE A R SHOP
?iECHANICAL AND COMPUTER ELECTRICAL ROOM EQUIPMENT ROOM Figure A.9-4 Preliminary Floor Plan for 0 10 20 30 40 50 Technical Support Center (Option B) 1 I I I I i gg Amendment 13 I (January 1982)
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. MECHANICAL AND MAINTENANCE E12CTRICAL AND EQUIPIENT R
EQUIPMENT REPAIR STORACE ROO'1 SHOP Figure A.9-6 Preliminary Floor Plan for 0 10 20 30 40 50 Emergency Operations Facility i I I I 1 e gg Amendment 13 (January 1982)
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Figure A.9-7 Gaseous Radwaste and Vent Collection Modifications Amendment 13 (January 1982)
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@; * @ .j REMOVAL PUMP m u e' u R O O M '"A" llo ol EL.700'-Op ww 1AIN MENT SERAY PUMP Figure A.9-9 Fuel Handling Area Exhaust ROO M *A" System Modifications Amendment 13 (January 1982)
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s* Figure A.9-10 External llydrogen Recombiner Containment Isolation and Power Supply Scheme Amendment 13 (January 1982) O
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