ML20038A717

Amend 62 to PSAR

ML20038A717
Person / Time
Site:	Clinch River
Issue date:	11/30/1981
From:	ENERGY, DEPT. OF, CLINCH RIVER BREEDER REACTOR PLANT
To:
Shared Package
ML20038A715	List: ML20038A714 ML20038A717
References
NUDOCS 8111160211
Download: ML20038A717 (700)

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O PAGE REPLACEMENT GUIDE FOR AMENDMENT 62 CLINCH RIVER BREEDER REACTOR PLANT PRELIMINARY SAFETY ANALYIS REPORT

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Transmitted herein is Amendment 62 to Clinch River Breeder Reactor Plant Preliminary Safety Analysis Report, Docket 50-537.

Amendment 62 consists of new and replacement pages for the PSAR text O_ and Question / Response supplement pages. ,

Vertical lines on the right hand side of the page are used to identify question response information and lines on the left hand side are used to identify new or changed design infonnation.

The following attached sheets list Amendment 62 pages and instructions for their incorporation into the Preliminary Safety Analysis Report.

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8111160211 911113 PDRADOCK05000g B

4 AMENDMENT 62 PAGE REPLACEMENT CUIDE REMOVE THESE PAGES INSERT THESE PAGES Chapter 1 1.4-11 thru 14 1.4-11 thru 14 1.4-15. 16, 17, 17a 1.4-15, 16, 17, 17a 1.4-20, 21, 21a, 21b 1.4~20, 21, 21a, 21b 1.4-34, 35 1.4-34, 33 1.4-37, 38 1.4. '7, 38 1.4-40 1.4-40 Chapter 2 2.5-47, 47a 2. 5-47, 47a Chapter 3 3.1-17, 18 3.1-17, 18 3.1 73, 74 3.1-73, 74 1 3.2-9a 3.2-9a 3.2-11d, 11e 3.2-11d, 11e O 3.2-15a, 15b 3.8-3, 3a 3.2-15a, 15b 3.8-3, 3a 3.8-5, 6 3.8-5, 6 3.8-7b, 7c 3.8-7b, 7c 3.8-11, lla 3.8-11, lla 3.8-18a, 19 3.8-18a, 19 3.11-1, 2 3.11-1, 2 3Ai, ii 3Ai, ii 3A.2-1 thru 4 3A.2-1 thru 4 Chapter 4 4.2-112, 113 4.2-112, 113

- 4.2-439a thru f 4.3-9, 10 4.3-9, 10 Chapter 5 5xxia, xxib, xxii, xxiia 5xxia, xxib, xxii, xxiia 5.3-17, 18, 18a, 18b 5.3-17, 18, 18a, 18b 5.3-21, 21a 5.3-21, 21a 5.3-39b, 39c 5.3-39b, 39c 5.3-70b 5.3-70b 5.3-75b, 75c 5.3-75b, 75c 5.3-86, 87 5.3-86, 87 A

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l REMOVE THESE PAGES INSERT THESE PAGES l Chapter 5 j (Cont'd.)

5.3-100 thru 105 5.3-100 thru 105 5.3-111a 5.3-111a 5.5-9, 94, 10, 10a, 5. 5-9, 9a , 10, 10a ,

11, lla 11, lla 5.5-18b, 18c, 18d, 18e, 5.5-18b, 18c, 18d, 18da, 18e, 18f, 18g 18f, 18fa, 18g 5.5-24b, 24c, 24d, 25 5.5-24b, 24c, 24d, 25 5.5-27, 27a, 27b, 28, 5.5-27, 27a, 27b, 28,

, 28a, 28b 28a, 28b i 5.5-35, 35a 5.5-35, 35a, 35b f

5.5-46, 47 5.5-46, 47 5.5-52, 53 5.5-52, 53 5.5-58, 59 5.5-58, 58a, 58b, 59 3

5.7-2, 2a 5.7-2, 2a 5.7-9, 10 5.7-9, 10 l

i l Chapter 6 l 6.2-10a, 11 6.2-10a, 11 6.2-13, 13a

) 6.2-27, 27a 6.2-13, 13a 6.2-27, 27a 6.2-28, 28a 6.2-28, 28a I

Chapter 7

7.1-3, 4 7.1-3, 4 l 7.2-19 thru 23 7.2-19 thru 23 l 7.3-3, 4 7.3-3, 4, 4a 7.6-3b, 3c 7.6-3b, 3c Chapter 9 9.3-21 9.3-21 9.4-1, 2 9.4-1, 2 9.5-20 9.5-20 9.5-28 9.5-28 9.5-30 9.5-30 9.10-1, 2 9.10-1, 2, 2a 9.10-6 9.10-6 9.11-1 thru 7 9.11-1 thru 7 9.15-1, 2 9.15-1, 2 9.16-12 9.16-12 Chapter 15 15.1-98 thru 101 15.1-98 thru 101 15.7-9, 9a 15.7-9, 9a B

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'S Chapter 17 17A-23, 24 17A-23, 24 ,

17A-33, 34 17A-33, 34  !

17A-43, 44 17A-43, 43a, 44 17C-15, 16 17C-15, 16 i 17C-23, 24 17C-23, 24 i 17C-31, 32 17C-31, 32 17Fi 17Fi 17F1 thru 6 17F1, 2, 2a, .

i 3 thru 6  !

l Entire Chapter 17 Appendix I Entire Chapter 17 Appendix I i (including Title Page) (including Title Page)

(72 Pages) 17J-38 thru 53 17J-38 thru 53 l 17J-56, 57 17J-56, 57 17J-60 thru 73 17J-60 thru 67, 67a, 68 thru 72,  ;

72a, 73 Appendix B f I

i B-17, 18 B-17, 18 9 Appendix C C.4-15, 16 C.4-15, 16 i C.6-13, 13a C.6-13, 13a i

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AMENDMENT 62 QUESTION / RESPONSE SUPPLEMENT This Question / Response Supplement contains ar. Amendment 62 tab sheet to be inserted following Qi page, Amendment 61, September 1981.

Page Qi, Amendment 62, is to follow the Amendment 62 tab.

Replacement pages for the Question / Response Supplement are listed below.

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Construction Liaison Manager 45l The Construction Liaison Manager reports directly to the CRBRP Project Manager. As Construction Liaison Manager, his responsibilities 25 are to provide effective design guidance to enhance optimum construction sequencing, continue to develop construction advanced planning and se-45lquencingphilosophywhileworkingwiththeA-Eandconstructoraidin l25 ensuring that the CRBRP construction will be completed at the earliest possible time and at the least possible cost by prov"ing uniform plan-ning and logic to the construction sequencing.

25 45 LRM Safety and Reliability Manager The LRM Safety and Reliability Manager reports to the CRBRP l25 Project Manager and receives direction from the Division's fluclear Safety Manager in matters of corporate and divisional safei.y reliability and licensing policy. He is responsible for coordinating safety, relia- l25 bility and licensing activities among the RMs and for assuring that NSSS docurlentation reflects the Project's safety posture.

ARD Program Manager p The ARD Program Manager reports to the CRBRP Project Manager and is responsible for the coordination of ARD RM activities.

GE Program Manager The GE Program Manager reports to the CRBRP Project Manager and is reponsible for the coordination of GE mi activities.

AI Program Manager The AI Program Manager reports to the CRBRP Project Manager 34land is responsible for the coordination of AI RM activities.

CRBRP Systems Integration Manager The CRBRP Systems Integration Manager reports to the CRBRP Project Manager and is responsible for control and integration of the 45lNSSS design development and system interface including the LRM-AE interface 15 activities. and Amend. 45 July W 8 1.4-11 O

1.4.2.5.1.2 ARD RM Organization (Figure 1.4-6)

CPSRP Project Manacer The CRBRP Project Manager reports to the General Manager, Advanced Reactors Division and is responsible for discharging the tasks associated with the division's role as a Reactor Manufacturer (RM). In this capacity the Project Manager has the responsibility for all the technical and financial planning associated with the Westinghouse RM activities.

CPERP Reactor Plant Project Manager The CRBRP Reactor Plant Project Manager is responsible to the CRBRP Project Manager and through him to the ARD General Manager for the overall management of CRBRP RM activity at ARD. This includes the administration, design, documentation, procurement, shipment, and installation support of the NSSS systems, components, and licensing, safety and reliability related activities as well as all required software as assigned by the CRBRP Project Manager.

Reactor Engineering Manager The Reactor Engineering Manager is responsible to the CRBR') Reactor Plant Project Manager for establishing system requirements for the reactor enclosure, internals, and control rod systems; and the design, documentation, shipment, and Installation support of the reactor vessel, reactor internals, reactor primary control rod system, reactor guard vessel, reactor closure ,

head, and the components for the head accesc area and the reactor cavity and for the stress and thermal / hydraulic analysis of the permanent reactor components.

Pgactor Analysis and Core Design Manager The Reactor Analysis and Core Design Manager is responsible to the CRBRP Reactor Plant Project Manager for structures analyses, nuclear design, core thermal and hydraulic analyses, shielding analyses, and the design, documentation, and Installation support of the fuel and removable assemblies.

Procram Control and Deslan Intearation Manacer l The Program Control and Design Integration Manager is responsible to the CRBRP i

Reactor Plant Project Manager for establishing reactor system requirements, i integration of ARD systems (particularly interface control), maintaining cost and schedule visibility and control, planning, configuration management, and i preparation of RM procedures.

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Reactor Plant Procurement Manager i O The Reactor Plant Procurement Manager is responsible to the CRBRP Reactor Plant Project Manager for all CRBRP procurement. This includes establishing l Internal purchasing procedures to assure procurement is conducted in accordance with contractual and corporate procurement requirements.

Plant Engineering Manager The Plant Engineering Manager is responsible to the CRBRP Reactor Plant Manager for establishing system requirements for the reactor heat transport system, plant control, data handling, reactor and vessel instrumentation  ;

systems, plant protection systems, as well as the design, fabrication documentation, shipment, and installation support of the components in those systems. In addition, he is responsible for providing overall plant performance and reliability analyses, and the manufacturing engineering support for all ARD RM NSSS components.

CRBRP Licensing and Safety Manager The CRBRP Licensing and Saf ety Manager is responsible to the CRBRP Reactor Plant Project Manager for all activities necessary for licensing and the required safety analysis. He is responsible for assuring that nuclear safety and licensing requirements have been satisfied, for the preparation and coordination of licensing documentation generated within the ARD-RM, for l assuring that the required safety analyses are performed, and for directing safety analyses conducted by GE as the agent of the LRM.

RM Oualltv Assu ance Activities All ARD RM Quality Assurance activities are performed by the Division's Product Assurance Department which is totally independent from the RM Engineering and Procurement Organization. For the description of the Divisional RM Quality Assurance Organization and its duties and ,

responsibilities see Chapter 17 Appendix 17H. '

1.4.2.5.2 Burns and Roe. Inc. - Breeder Reactor Division (Figure 1.4-7)

Breeder Reactor Division Senior Corocrate Vice President The Senior Corporate Vice President and Director of the Brooder Reactor Division is the senior corporate officer assigned to the project and reports to the President. He draws upon the total resources of the corporation to assure that all necessary actions and support are forthcoming. He provides senior technical guidance as necessary. He assures that any problems requiring attention and resolution are being acted on in a timely manner.

O 1.4-13 Amend. 62 i Nov. 1981

Breeder Reactor Division Vice President The Vice President and Deputy Director of the Breeder Reactor Division reports to the senior corporate officer assigned to the project.

He provides guidance and direction to the Project Manager and the Project Quality Atsurance Manager in the conduct of the project. He performs special reviews of thE engineering and design work being conducted on the project and of progress being made. He assures that any problems requirins attention and resolution are being acted on in a timely manner. He contacts 451 senior representatives of DOE and the LRM as necessary to assure satisfactory completion of overall project efforts. 25 CRBRP Project - Project Manager The Project Manager reports to the division Vice President and is assioned overall reponsibility and authority for carrying out Burns and Roe's 45 l contractual commitments to DOE. He directs and coordinates all project.

activities in a manner to assure that all Burns and Roe efforts are proceeding in an integrated fashion which will support procurement and construction efforts and will produce a satisfactory technical product,

45) on time, and at minimum cost to DOE. He assures that the engineering and desian work by Burns and Roe provides a safe and reliable plant with minimum environmental impact, and a plant which has good operability, availability, maintainability, flexibility, inspectability, and prospect for future economy. He is the official point of contact for the project within Burr.s and Roe and assures that Burns and Roe's efforts are carried out in a satisfactory manner. He issues management reports and informa-tion concerning the project.

Assistant Project Manager The Assistant Project Manager reports to the Project Manager. He is responsible for monitoring the coordination of in-house engineering activities to ensure the design product, daily production and continuing output are all i

in conformance with contractual requirements. He performs such other duties 55 or tasks as he may be assigned by the Project Manager.

Contract Administration Manager l The Contract Administration Manager directs the contract l

administration functions for the project. He reports to the CRD Vice President and supports the CRBRP Project Manager as the central point of contact for the project on contract administration matters. Included in contract administration matters are preparation of documentation, compliance with notification provisions, cost segregation, and 53 negotiation.

1.4-14 Amend. 55 June 1980

Project Operations MaraSer The Project Operations Manager reports to the Project Manager and is responsible for the administrative, business, planning, scheduling, cost engineering and contractural systems of the Project. For the administrative and business systems he is responsible for Project cost control and reporting, manpower control, commitment control, and the formulation and monitoring of the Project data bank. He is also responsible for the Management Information Center and for development, custody and control of Project procedures together with the required indoctrination of Project personnel. For the planning, schedaling and cost engineering systems he provides the necessary controls and monitors overall Project progress and plant capital costs.

Under these systems he is also responsible for the preparation and maintenance 55l of all Project schedules. He is responsible for all project personnel training related to the above systems as required.

Procurement Manager The responsibilities of the Procurement Manager who reports to the Proj ect Manager are governed by the scope of work included in Burns and Roe's contract with the CRBRP Project Office. Where Burns and Roe has procurement support responsibility, the Procurement Manager is responsible for the preparation of the potential bidder's lists; review of technical specifications for procurement suitability; providing assistance to the procurement agency in the bidding process; administration of Burns and Roe support responsibilities for each subcontract and provides Burns and Roe contact with vendor subcontract administration personnel. Where Burns and Roe has complete procurement responsibility, the Procurement Manager is also responsible for the conduct of the bidding process including negotiations and award of subcontracts, and administration of subcontracts.

Quality Assurance Manager The responsibility and authority of the Quality Assurance Manager is discussed in Section 17E-1.3.

Project Office - Resident Manager The Project Office Resident Manager reports to the Project Manager and coordinates all Burns and Roe operations in Tennessee. He interfaces as ,

45l necessary and as directed with DOE, PMC, the LRM and the General Constructor.

He is supported by a Systems Integration Manager, Planning and Construction Liaison Manager (future Site Manager), Program Manager, and a Licensing and Environmental Representative. He is responsible for the activities conducted i

at the Project Office and at the construction site, other than Quality 45lAssurance. He shall keep the DOE Project Director advised on as frequent a basis as necessary of status and problems. He is empowered to speak and act for the Burns and Roe Project Manager where necessary. 25 1.4-15 Amend. 55 l June 1980 l

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1.4.2.5.3 General Electric Comoany (Figur_e 1.4-8)

The Advanced Reactor Systems Department (ARSD) is a part of the Energy Systems and Technology Division (ES&TD) of General Electric Co. (GE). The General Manager of the CE-ARSD reports to the Vice President and General Manager of the ES&TD and is responsible for organizing the resources to carry out such programs and for developing corporate programs that will lead to the eventual lcommercializationofAdvancedNuclearPowerPrograms, including LMFBR technology.

The GE-ARSD General Manager conducts review of progress being made on projects within the department and provides direction and guidance to the Section Managers reporting to him. He has the responsibility and authority to issue Department policy and to establish quality goals and objectives. (See Chapter 17, Appendix I for details of t5e General Managers' QA responsibility).

lTheGE-ARSDcorsistsofninesectionsandtheLegalOperation. Each section is headed by a Section Manager who reports to the General Manager anc' is responsible for an assigned area of responsibility as defined in the following paragraphs.

Clinch River Profect Section The manager of the Clinch River Project Section is responsible to the Department General Manager, GE-ARSD, for performance of work related to contracts on the Clinch River Breeder Recctor Plant Project. Major functional responsibilities (engineering, design, and supply) include engineering of the intermediate neat transport system, steam generator system, and decay heat emoval system, as well as primary and Intermediate sodium pump development, i+eam generator development, trace heating, IHTS control and instrumentation, RS projects, safety evaluations and Iicensing support cctivities. The vlinch River Project Section serves as the GE-ARSD interface with all other CRBRP Project participants. Functionally, the Clinch River Project Section is divided into five subsections (Figure 1.4-8), each with specific assigned responsibilities. Project management responsibilities include functional work performed in other GE-ARSD Sections, subcontractors, and vendors.

Design Engineering Section Tne Manager of the Design Engineering Section is responsible to the Department General Manager. The responsibilities of the Design Engineering Section in support of the Clinch River Breeder Reactor Plant Project include providing analytical and design engineering services in the areas of structural and thermal hydraulic analyses, safety analyses, rollabilIty engineering and SCRS Design. The Design Engineering Section also provides nuclear engineering support prImariiy reIated 1o the evaluation of critical experiments for the lClinchRiverCore,andsystemsEngineeringSupport.

O 1.4-16 Amend. 62 Nov. 1981

Develocment Engineering Section O The Manager of the Development Engineering Section is responsible to the Department General Manager for planning, organization and management of major programs of research :nd development, engineering test support, and experimental f acility design and construction for LMFBR pregrams. Such programs include work in support of assigned projects as well as the development of new systems and components for future LMFBR product lines.

Procurement Section The Manager of the Procurement Section is responsible to the Department General Manager for procurement of material, hardware and services.

Procurement locates and maintains adequate vendor sources of supply, executes all vendor relations in a f air and equitable manner, conducts vendor negotiations, awards and manages purchase orders and insures proper execution of all contractual mattars consistent with applicable General Electric policy, DOE / Government procurement regulations and other customer requirements as they relate to purchasing activity.

Product Assurance and Services Section The Manager of the Product Assurance and Services Section is responsible for ensuring an acceptable level of quality in all GE-ARSD products and services.

It is the responsibility of Quality Assurance to assure that all technical activities of the Clinch River Project, including those performed by i subcontractors, are consistent with the customer quality requirements and company quality policy (set Chapter 17, Appendix i for further detail). He also provides leadership and coordinates development of management systems and procedures, to guide and control all Department activities; and provides centralized engineering, technical and administrative support services for the Department.

Acolications Engineering and Planning Section The Manager of Applications Engineering and Planning Section is responsible to the Department General Manager for recommending goals and objectives and formulating and implementing strategies and action plans relating to the marketing of current Department services and products and related contract negotiation and administration and the market development for the Department's new products and services. Applications Engineering and Planning is also responsible for the negotiation and administration of all contractual matters related to the Clinch River Project.

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Technology and Soecial Project Section The Manager of the Technology and Special Projects Section is responsible for O coordinating and directing the overall manageme.it and execution of all projects undertaken by the GE-ARSD with the e>ception of those specif ically assigned to other sections in the department ay the Department General Manager. Similarly, he is responsible for coordinating the funding, reporting and measurement er progress of the department Cevelopment Authorizations (DA's). He provides the primary technical and programmatic interface between the Department and the Department of Energy (DOE) and other customer organizations on projects and related matters. He also provides technical and programmatic leadership and assistance to the Applications Engineering and Planning Section on project proposal and contract activities, product planning, product applications, and market development.

GE-ARSD Lega! Ooeration The GE-ARSD Legal Operation is staffed by the Department .ounsel who is l responsible to the Department General Manager for advice and counseling of department management regarding legal implications of contracts and otner arrangements which legally bind the Company. In addition, Counsel participates w ith other members of the staf f in the general operation of the business, advises on antitrust, Iabor, government regulatory, equal employment and other matters of legal significance, Counsel is assisted by patent counsel on matters involving patents and data.

GE-ARSD Financial Secticn The Manager of the Financial Section is responsible to the Department General Manager for reporting financial results of the Department, establishing the financial polIcles of the Department and providing f,Inancial service and counsel to the other GE-ARSD sections. In addition, the Financial Section is responsible for interpretation of financial contract language, establishment and negotiation of overhead rates, and development of operating budgets and long range forecasts of GE-ARSD.

GE-ARSD Eroloyee Relations Section The Manager of the GE-ARSD Employee Relations Section is responsible to the Department General Manager for identifying, developing and implementing relations programs responsive to the Department needs; for establishing goals, l objectives and assuring timely employment of qualified personnel. He also provides coordination, counseling and direction for all Department components in relations areas including Manpower Development and Equal Employment Opportunity and Minority Relations and maintains procedures and records and to assure compilance with federal and state laws in the areas of fair employment practices.

O 1.4-17a Amend. 62 Nov. 1981

Manager. Water Resources The Manager of Water Resources is responsible for supervising the systems analysis, cooling systems and alternative sections of the CRBRP-ER.

Manager. Advanced Develcoment The Manager of the Advanced Development Department supervises and coordinates the socioeconomic studies, data analyses and preparation of the socioeconomic and cost / benefit sections of the report.

1.4.2.5.6 Stone and Webster Engineering Corcoration (Figure 1.4-11)

The construction of the CRBRP is being undertaken by Stone and Webster Engineering Corporation (S&W) a wholly owned subsidiary of Stone & Webster, Inc. As general contractor, S&W will prepare the site, construct permanent plant structures and install both NSSS and B.O.P. components, systems and equipment.

CRBRP Senior Project Manager The Senior Project Manager for the CRBRP construction effort is a S&W Vice President and is the senior corporate of ficial responsible for S&W activities on the CRBRP Project. As Senior Project Manager, he will be responsible for coordinating all S&W headquarters and field operations required to perform the l construction of the Project in accordance with contract requirements. He

. reports to the President of S&W and is thus able to draw upon the required

\ corporate resources to assure the necessary support for the Project.

CPERP Deoutv Director of Construction The Deputy Director of Construction is a S&W Vice President and the Construction Manager of the CRBRP Project. As Construction Manager, he is responsible for the construction organization and assignment of construction personnel. He participates in establishing company-wide S&W construction policies and procedures.

CPERP Project Managers I'

Management of the S&W CRBRP construction activities is divided into two areas; control and production. Managers of these areas are accountable to the Senior

, Project Manager and work directly with the Project participants to support the l

Project schedule and budget. The Project Manager - Control is responsible for establishing Project construction criteria and determining the timing of and

directing of all Project construction criteria and determining schedules, estimates and expenditure forecasts. The Project Manager - Production is responsible for providing the necessary manpower and resources to meet the construction goals, coordinating with other groups and for the quality of the work.

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CRPRP Project OualIty Assurance Manager The Project Quality Assurance Manager is responsible for assuring that an O

adequate quality assurance program is established, implemented and documented to meet the requirements of Appendix B, 10CFR50 and RDT F2-2, Aug.,1973 with Addenda I dated 12/73, Addenda 11, dated 3/74 and Addenda ill, dated 7/11/75, within the scope of the S&W construction effort. He receives quality assurance guidance from the S&W Manager of Qual!ty Assurance in S&W I Headquarters.

Senior Site Construction Reoresentative The Senior Site Construction Representative is in charge of the construction organization at the site and directs the day-tc-day activities. He responds to the goals set by the S&W Project Managers and acts under the guidance of the Deputy Director of Construction.

Suoerintendent of Field Quality Control The Superintendent of Field Qual f ty Centrol is in charge of the quality control organization at the site and directs the day-to-day activities. He is responsible for the implementation of the quality assurance program at the construction site and acts under the direction of the S&W Project 0A Manager.

Corporate administration, corporate policy, and corporate resource support are received frce the Manager, Field Quality Control Division in S&W Headquarters.

Contract Administrater The Contract Acministrator provides IIalson activl1les related to the S&W contract with DOE reviews contract related material, monitors perfornance and provides the interface with DOE on contractual matters related to construction site activity. The Contract Administrator acts under the direction of the Senicr Ccnstruction Site Representative.

Engineering Liaison Engineering Liaison is responsible for providing tne S&W interface in the offices of the Architect-Engineer. Acting under the guidance of the Senior Site Construction Representative, he is responsible for providing S&W Input to the design and engineering process and for providing S&W with timely information on engineering and design matters which impact construction.

Suoerintendent of Cost and Scheduling The Superintendent of Cost and Scheduling acts under the direction of the Senior Construction Site Representative and supervises the project site cost and scheduling program to provide coordinated and Integrated cost and planning control necessary for the canpletion of the construction ef fort in accordance with master schedules and projected costs.

Construction Administrator Reviews design for constructibilih and furnishes technical assistance to the Senior Site Construction Representative in planning and execution of the construction program with special attention to areas unique to sodium systems.

He is responsible for the daily contact with the Reactor Manufacturers.

Amend. 62 1.4-21 flov . 1981

Suoerintendent of Construction Acting under the direction of the Senior Construction Site Representative, the Superintendent of Construction is responsible for the construction of a complete and operating plant in accordance with engineering plans and specifications and planned schedules for the least cost consistent ::!th good quality.

Assistant Suoerintendent of Construction Engineering Under the direction of the Senior Construction Site Representative, the Assistant Superintendent of Construction Engineering directs all S&W Construction engineering activities for the Project. He directs and controls the distribution of engineering documentation, requisitions permenent plant materials and coordinates with Field Quality Control and the Architect-Engineer in the resolution of problems encountered during the construction phase.

Assistant SuoerIntendent of Construction Services The Assistant Superintendent acts under the direction of the Senior Construction Site Representative and is responsible for providing the personnel, purchasing, accounting and office service functions necessary to support the construction effort so that it may proceed in accordance with plant and specifications and according to schedules and budgets.

Safety Suoervisor Under the direction of the Senior Construction Site Representative, the Safety Superviser is responsible for the administration of the construction site safety, accident and fire prevention programs, ensuring adherence to Federal, State, and Local safety regulations and fire ordinances and the S&W safety program.

Chief Construction insoector Under the direction of the Senior Construction Site Representative, the Chief Construction Inspector administers the quality inspection program in areas assigned to the construction department.

1.4.3 INTERRELATIONSHIPS WITH CONTRACTORS AND SUPPLIERS PMC has contracted with Westinghouse Electric Corporation, acting through its Advanced Reactors Division (ARD), to perforin the function of Lead Reactor Manufacturer (LRM) for design, manufacture, and provision of test support for the Nuclear Steam Supply System (NSSS) for the Clinch River Breeder Reactor Plant. Westinghouse also has RM responsibilities and has subconfracted with General Electric Company Energy Systems and Technology Division and Rockwell International ( Atomics internat f or al Division, Al) to provide the design and manufacture of certain systems for the NSSS. PMC has assigned the administration of its contract with ARD to DOE.

1.4-21a Amend. 62 Nov. 1981

O PMC has contracted with Burns and Roe, Inc., to orovide the architect-engineer services required for the Project. Burns and Roe has subcontracted with Law Engineering and Testing Company to carry ott investigations to determine +he suitability of the site geology in support of foundation designs for permanent structures.

Burns and Roe also has a subcontract with Holmes and Narver, Inc.

to provide services in liquid metal engineering technology. PMC has assigned the administration of its contract with Burns and Roe, IL Inc. to DOE.

25 PMC has contracted with Westinahouse Electric Corporation to 45; provide servi:es needed in the preparation of the Environmental Report

, for the Project and to perform certain other associated tasks. PMC has assicned the adminis ration of its contract with Westinghouse to 45 DOE. 25 PMC has contracted with Stone & Webster for the construction of the plant. Stone and Webster may subcontract portions of the work to others. PMC has assigned the administration of its contract with Stone and !ebster Engineering Corporation to DOE. 25 t.5l The DOE oruvides R&D information in support of the CRBRP Project

througn its LMFBR base technology programs being carried out by its national laboratories and contractors. A description of related base technology programs is provided in Section 1.5.

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Anend. 54 1.4-21b May 1980

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l OFFICE OF j THE CHAIRMAN AND i PRESIDENT I

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] GENION VICE PRESIDENT rit0 JECT MAN AGER DEPHTV DIAECTOA rA0Jter ouAuty

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jl g OAKRIDGE i j SENION SITE "*=""***==~h*===========..== CASAP FA0JtcT j CONS T AUCTION - .J MEFAESENTATIVE GA MANASLA i

CONTRACT F ADMINISTRATOR p ENGINEERING j ^ LsAlsoN CONSTRUCTION ADMINISTRATOR l I I I 1 1 1 1

ASSISTANT ABIIIIANT

) SUPEAINitkOf M T 3AfgTy SUPE AlN TEN 0tNI SuPERINTEN0f NI EU'EI i or CONSTAucil0N gur(AVISOA of of COST 8. OF CONSTRUCTION j stAv CEs CONSrpuCil0N '" I I8 SCHEDUUNG ENGINEERING t I  !  !  !

A - - A A A l trerNO PROJECT ORGANIZATION l AtaroNslaiutt CLINCil RIVER DREEDER REACTOR PLANT

-~~~ COMMUNICATION f. LI AISON STONE L WEBSTER EHolNEERING CORPORATION i

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g- rosos.4 l 5!0 Figure 1.4-11. Stone and Webster Engineering Corporation Organization i

d. Mat foundation has no surcharge.

O With these very conservative assumptions, the bearing capacity for the foundation has been calculated, using the relationship outlined in refer-ence (136) to be about 800 KSF. This calculated bearing capacity exceeds all anticipated static loads by a factor greater than 130, 38 l 2.5.4.12. Techniques to Improve Subsurface Conditions l Based on inspection of the core from 115 drill holes, including l borings 41, 54 and 55, which were specifically evaluated by NRC, and 41 borings in the immediate vicinity of the Nuclear Island and Emergency Cooling Tower, it is concluded that the Unit A Upper Siltstone is the optimum bearing stratum for the Nuclear Island structures located on a common mat, the Emergency Cooling Tower and the Fuel Oil Storage Tanks. It is also concluded that the engineering characteristics of the foundation stratum are consistent below the proposed base of the common mat (Elevation 715). The results of a detailed geophysical investigation including refraction, cross-hole and up-hole methods, in-situ and laboratory test data and examination of the walls of exploratory holes by geophysical logging devices have supported these conclusions. The significant factors considered in the evaluation of the approximately 400 foot wide band of siltstone and :.he bordering Unit A and B limestone included depth of weathering, susceptibility to development of solution activity and extent of recorded variations in the engineering characteristics. The low pemeability, infrequnet O joint spacing, inhetent resistance to solutioning and adequate available strength ensure the satisfactory bearing capability of the silstone stratum at the foundation grade elevation of 715 for the Nuclear Island and 765 for the Emergency Cooling Tower.

The interface of the underlying Unit A Limestone and upper silt-stone stratum was also checked in some detail with two lines of core bor-ings spaced 50' apart with a 50' to 70' spacing between borings. It 38l has been concluded that the limestone stratum within the immediate static influence of the Nuclear Island structures is competent below Elevation 715. Inclided boring 56 had indicated soil layers extending to Elevation 704, however, the Nuclear Island structures are located 59 approximately 100 feet to the east of this zone.

To confim the homogeneity and satisfactory bearing quality of the foundation strata, a test grouting program centered around boring 55 has been conducted on the west side of the Plant Island extending through the silstone and Unit A limestone. Water pressure test results, negligible grout take and geophysical logging indicated that foundation conditions applicable to that zone below the Top of Continuous Rock were reasonably consistent for all borings completed in this area. It is concluded that foundation treatment of the siltstone and underlying Unit A limestone will not be required and the results of the test grouting program are essentially 8 confirmatory in scope. Photographs were taken of tN core from many of the borings shortly af ter removal from core barrel and these are available O for inspection if additional information is required. Resul ts of this investigation are provided in Appendix 2-C.

Amend. 59 Dec. 1980 2.5-47

X Subsequent to the completion of the site investigation for the Nuclear Island, the width of the common mat supporting the Category I structures increased slightly due to a series of design changes. Consequently, a minor portion (10' - 20') of the mat may be supported on Unit A Limestone. It is apparent f rom the results of the test grouting program that this would not result in a change in the foundation conditions for the Nuclear Island.

Some dental work is anticipated, which will consist primarily of removing loose materials including any pinnacles of rock which may be evident af ter excavation, and filling local depressions with mass concrete. The rock surface will be air cleaned and relief points or vertical cracks will be treated with a slush grout application to prevent entry of rain and run-of f.

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0 2.5-47a Amend. 62 Nov. 1981 L

I l Criterion 8 Reactor Design j

O The reactor and associated coolant, control, aad pro'ection systems i shall be designed with appropriate margin to assure that specitied acceptable fuel design limits are not exceeded during any condition of normal operation including the effects of anticipated operational occurrences.

RESPONSE

' This criterion is satisfied by the following two design bases.

a. Fuel Residence Time In the first core loading, the fuel rods are limited to a peak 51 l pellet burnup of 80,000 megawatt days per metric ton of heavy metal (mwd /T). For later cores the peak burnup increases to 51 l 115,000 mwd /T with an average burnup of 80,000 mwd /T. These a peak burnup limits are achieved by limiting the in-core residence time and optimizing the fuel management scheme. The duration of the first cycle is 128 full power days (FPD) and the second

! cycle is 200 FPD. These cycle lengths are consistent with the initial core peak pellet burnup limit of 80,000 mwd /T. For all operating cycles after the first two, the cycle length is increased

to 274 FPD and the maximum fuel assembly residence time is two cycles. All fuel and inner blanket assemblies are discharged as

51 a batch after two cycles under equilibriurr core conditions. Mainten-O ance of fuel rod structural integrity is a design basis should an Unlikely Fault occur during the fuel residence time.

i

b. Power Distribution Limits The power distribution limits are derived from the maximum allowable

, peak heat generation rates for nominal and anticipated operational conditions which, when combined with the rod mechanical and thermal 4 design parameters, assure that incipient fuel melting does not

! l occur in the fuel pellet with peak power. The superimposed effects 51 of fuel depletion and control rod insertion patterns on the radial power peaking factors is included in this assessment. The peak fuel

, pellet linear power in the core at any time-in-life, which includes i

the highest radial and axial power factors,15% overpower conditions and 3a nuclear and engineering uncertainties, is less than that which results in fuel melting.

3.1-17 Amend. 51 J Sept. 1979

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Criterion 9 Reactor Inherent Protection The reactor and associated coolant systems shall be designed so that in the pcwer operating range the net ef f ect of the prompt inherent nuclear feedback characteristics tend to compensate f or a rapid increase in reactivity.

FESPONSE The f olicwing design basis satisfies this criterion:

The Doppler effect provides the prompt negative reactivity feedback which is required to mitigate the effects of reactivity transients (rapid power increases). Theref ore, the f uel temperature (Doppler) coef f icient shall be strongly negative when the reactor is critical. The negative Doppler coefficient is obtained through the inherent use of f uel with a large proportion of U-238. The Doppl er coef f icients f or each major f ueled reactor region have been calculated at the beginning and end of cycle for both the f irst and equil ibrium cores w ith FFTF-grade (low Pu-240) plutonium fuel (See Teb l e 4.3-16 ) . In al l cases, the Doppl er coef f icients are strongly negative.

The analysis of accident conditions, presented in Cnapter 15, uses conservative values of the Doppler reactivity feedback coef ficient (nominal value less 3 uncertainty).

At I cw pow er/ f l ow ratio operating conditions as during the reactor startup, positive bowing reactivity ef f ects are predicted. The net reactivity feedback during th i s pcw er-to-flow ratio range is evaluated to ccnservatively envelope all possible combinations of bowing and compensating negative reactivity effects. For certain assumptions on assembly bowing behavior a net positive lreactivityfeedbackispredictedoveraportionofthelow pow er-to- f l ow ratio range. The PFS can safely accommodate all design basis events initiated in the startup range when the above worst case of f ects are considered. Studies have been perf ormed f or a range of startup overpower transients. These have canonstrated that, even negl ecting the ef f ects of the plant protection system, the integrated reactivity feedback f rom the point of the initiation of the l transient up to f ull power temperatures is always negative. Consequently, i

reactor pcwer and temperatures are bounded even when worst case reactor l feedback characteristics are utilized. Max!u n temperature values f all well belcw values which are expected for normal power operation demonstrating satisfactory reactor inherent protection, j As th e pcw er-to- f l ow ratio approaches 1.0 and at higher power-to-flow ratios

(>1.0), reactor assembly bow ing react!v ity is r.egative and enhances the effect c' negative Doppler which is discussed above.

O 3.1-18 Amend. 62 Nov. 1981

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l i I i

G Each line will be provided with one automatic ' solation valve in side containment and one automatic isolation valve outside contain-ment. (See Table 6.2-5.) Simple check valves are not used as  !

containment isolation valves outside containment.

The isolation valves outside containment will be located as close l to the centainment as practical and the automatic isolation valves i

are designed to take the position that provides greater safety upon loss of actuating power. Appropriate measures will be taken to minimize the probability or consequences of an accidental

rupture of these lines or lines connected to them to assure adequate i safety. (More details are provided in section 6.2.4.2.)

l 32

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i Amend. 32 Dec. 1976 7 3.1-73  ;

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Criterion 47 - Primary Containment Isolation l

Each line that connects directly to the containment atmosphere and penetrates primary reactor containment shall be provided with containment isolation valves as follows, unless it can be demonstrated that the containment isolation provisions f or a specif ic cl ass of lines, such as instrument lines, are acceptabl e on some other def ined basis:

(1) One locked closed isolation valve inside and one locked close isolation valve outside containment, or (2) One automatic isolation valve inside and one locked closed Isolation val ve outside contai nment, or (3) One locked closed isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic Isolation valve outside containment, or (4) One automatic Isolation valve inside and one automatic isolation valve outside containment. A simple check valve may not be used as the automatic Isolation valve outside containment.

Isolaticn valves outside containment shall be located as close to the containment as practical and upon loss of actuating power, automatic isolation valves shall be designed to take the position that provides greater safety.

Resoonse:

The f olicwing lines penetrate the reactor containment and are directly connected to the containment atmosphere:

Containment Ventilation Air Supply Line Containment Ventilation Air Exhaust Line Containment Vacuum Breakers Each of these lines, except the containment vacuum creakers wil l be provided l wIth three confinetent/ containment isolation val .es, two immediately outside the conf inement and one i nside the containment, with independent actuating trains.

The valves and associated actuators will close on loss of air or electrical pcwer. Because the system operating pressures are low and the closure times required f or the containment isolation valves are f our seconds, tne dynaa.ic f orces resul ting f rom the inadvertent closure under operating conditions wil l not challenge the Integrity of the valves or connecting piping. However, a quick acting automatic relief damper will be provided in a branch duct between the Air Supply Line containment isolation valves and the supply f ans in order to relieve any excess pressure on the ductwork cr!ginated by the activation of the contai nment isol ation valves. A relief damper is provided in the exhaust air line between the isolation valves and the exhaust f ans, to relieve the vacuum in the exhaust duct after the isolation valves close, in addition, upon containment isolation, the contai nment ventil ation supply and exhaust f ans are automatical ly stopped.

3.1-74 Amend. 62 Nov. 1981

TABLE 3.2-2 (Continued)

PRELIMINARY LIST OF SEISMIC CATEGORY I MECHANICAL SYSTEM COM.%NENTS AND ASSIGNED SAFETY CLASSES (3)

Safety Qual ity Components Ciass(1) Group (11) Location (2)

Impurity Monitoring and Analysis System Primary Pl ugging Temperature Indi-cation Package 3 C RG Primary Sodium Sampling Package 3 C RG Ex-Vessel Plugging Temperature Indication Pac;: age 3 C RSB Ex-Vessel Sodium Sanpling Package 3 C RSB IHTS Sodium Characterization Package (3) 3 C SGB Fuel Fail ure Monitoring System Cover Gas Monitoring Subsystem 3 C RSB Failed Fuel Location Subsystem l Continuing Reactor Cover Gas 3 C RSB l

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3.2-9a Amend. 62 Nov. 1981

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l TABLE 3.2-3 (Continued) t O l Emergency Plant Service Water System i Pump Motors [

ECT Fans  !

l Makeup Pump Motors f i

Temp. Control Valve O prators [

l t

i Temp. Transmitters  !

Temp. Indicator ControlIers 1

Pressure Differential Switches f Level Switches Cables ,.

Auxillary Liquid Metal System  !

Valve Operators O OverfIow Thermocoupies i Local Panels

, i

Pump Motors l

Control Cabinets i

Capacitor Cabinets Transf ormer Cabinets l l i

! EVST Thermocouples '

I l

Control Room Panels l  !

j Cables l I

l EVST Cell Termperature Thermocouples .

i  !

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l 3.2-11d Amend. 62 Nov. 1981 ,

I

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TABLE 3.2-3 (Continued) l)

Inert Gas Receiving and Processing Systems Containment Isolation Valve Operators  ;

i Expansion Tank Equalizatien Valve Operator Cables Plant Control System Nain Centrol Panel i SCRAM Breaker Cubicle Cables Reactor and Vessel Instrumentation System Reactor Cool ant Operati ng Level Instrumentation Cables Flun Penitoring System Flux Monitoring Instrumentation Cab i nets instrument Crawers Cables Plant Protection System Contair. ment Isolation System Cabir.e+5 Reactor Shutdown System Cabinets Plant Protection System Cabinets 61 Cables O

3.2-11e Amend. 61 Sept. 1981

l i

TABLE 3.2-5 (Continued)

PRELIMINARY LIST OF ASME CODE CLASSIFICATIONS FOR SEISMIC CATEGORY l MECHANICAL SYSTEM COWONENTS Component Code / Code Class (I) Location (2)

Emergency Plant Service Water System ASME-Ill/3 SGB,DGB Emergency ChIIIed Water System ASME-1Il/3 SGB,CB,DGB, RSB,RCB Normal ChiiIed Water System ASME-1II/3 RCB Auxiliary Mechanical Systems for Diesel ASME-Ill/3 DGB Generators Fuel Oil Storage and Transfer System including:

Diesei Fuel 01i Storage Tanks ASME-1Il/3 YARD Fuel 011 Transfer Pumps ASME-1II/3 DGB Fue1 01l Day Tanks ASME-1II/3 DGB Cooling Water System including:

Water Expansion Tank ASME-1II/3 O

v Jacket Cooling Heat Exchanger Water Temperature Regulating Valve ASME-lil/3 ASME-Ill/3 DGB DGB DGB Starting Air System including:

Air Storage Tanks ASME-Ill/3 DGB Lubrication System Including:

Lubricating Oil Heat Exchanger ASME-Ill/3 DGB Lube Oil Filters and Strainers ASME-Ill/3 DGB Control Room Heating, Ventilating, and ASME-lil/3 CB Air Condition System isolation Valves Non-Sodium Fire Protection System SGB,CB,DGB Seismically Qualified Water Supply ASME-lil/3 DGB Piping, Valves, and Valves l&C j

RCB Penetration, Valves, and Valves l&C ASME-Ill/2 SGB,RCB Standpipe System (Nuclear Island) Note (9) RSB,RCB l

i Piping and Valves lO 3.2-15a Amend. 62 Nov. 1981

TABLE 3.2-5 (Continued)

PREL IMINARY L IST OF ASME CODE CL ASSIF ICATIONS O FOR SEISMIC CATEGORY I MECHANICAL SYSTEM COMPONENTS Component Code / Code Ciass(1) Location (2)

Standpipe System Seismic Category l Note (9) DGB Pumps Fuel Failure Monitoring System Cover Gas Monitoring Subsystem ASME-Ill/3 RSB Failed Fuel Location Subsystem Contal nI ng Reactor Cover Gas ASME-IIl/3 RSB Notes:

(1) ! ncl udi nr, appl icabl e code cases.

(2) RG - Re actor Contai nment Buil ding IB - 1:.termediate Bay of the SGB l SGB - iteam Generatcr Bullding CA6 - Reactor Service Area of the RSB CB - Control Building DGB - Diesel Generator Bullding (3) Only piping f rom containment isolation valves to the +Ilter intake; f il ters and discharge ductwcrk per Reg. Guide 1.52.

(4) System wil l meet the requirements of Reg. Guide 1.52 (5) Cut to First isolation Valve (6) Within Dual isolation Valves (7) Downstream of Isolation Valve (8) Downstream cf First Isolation Valve (9) Non-Saf ety Rel ated, Seismic Category I O

3.2-15b Amend. 62 Nov. 1981

Tolerances The Containment Vessel as constructed shall not exceed the tolerance requirements of NE-4000 of ASME-Ill for fabrication or erection. The dimensional control procedures shall meet the requirements of RDT STD F3-15T.

l The out-of-plumb tolerances shall not exceed 1/500. The out-of-roundness

tolerance shall not exceed 1/2 of one percent of the nominal inside diameter.

3.8.2.2.3 AcoIIcable NRC Regulations and Regulatorv Guides NRC Reculatorv Guides The applicable regulatory guides are listed below.

1.10: Mechanical (Caldwell) Splices in Reinforcing Bars of Category l l

Concrete Structures (Revision 1, January 2, 1973).

1.11: Instrument Lines Penetrating Primary Reactor Containment (March 10,1971) 1.12: Instrumentation for Earthquakes (Revision 1, April,1974) a l 1.13: Spent Fuel Storage Facility Design Basis (December,1975)

! Attachment I 1.15: Testing of Reinforcing Bars for Category 1, Concrete Structures (Revision 1, December 28, 1972) 1.19: Nondestructive Examination of Primary Containment Liner Welds (Revision 1, August 11, 1972) 1.29: Seismic Design Classification (Revision 2, August 1976)

! 11.55: Concrete Placement in Category 1, Structures (June 1973) j 1.57: Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components (June,1973) 1.60: Design Response Spectra for Seismic Design of Nuclear Power Plants (Revision 1, December,1973)

I

1.61

Damping Values for Seismic Design of Nuclear Power Plants (Oct.

i 1973) i 1.63: Electric Penetration Assemblies in Containment Structures for Light-Water-Cooled Nuclear Power Plants (Revision 2, July,1978)

1.85

Materials Code Case Acceptability - ASME Section lit, Division 1, 1976 l

() 1.92: Combining Modal Responses and Spatial Components in Seismic Response Analysis (Revision 1, Feb.,1976)

- 3.8-3 i Amend. 62 l Nov. 1981 l

1.102: Flood Protection for Nuclear Power Plants Rev. 1 (September 1976) 1.117: Tornado Design Classification (September 1976) 1.122: Development of Floor Design responso Spectra for Seismic Design of Floor-Supported Equipment or Components (September 1976) 1.124: Design Limits and Loading Combinations for Class i Linear-Type Component Supports 1.143: Design Guidance for Radioactive Waste Management Systems, Structures and Components Installed in LWRs.

Of the above, Regulatory Guide 1.63 is applicable af ter the following changes:

1. Deleting "Iight-water-cooled" wherever it appears.

2. Replacing " Appendix B to 10 CFR Part 50" wherever it appears with "RDT Standard F2-2".

3. Replacing " General Design Criterion 50 of Appendix A to 10 CFR Part 50" wherever it appears with "CRBRP GDC 41".

4 Replacing " loss of coolant accident" with " containment design basis accident".

Construction No special construction techniques are anticipated for this containment vessel.

O 3.8-3a Amend. 62 Nov. 1981

3.8.2.3 Loads and Loading combinations O 3.8.2.3.1 Design Loads The following loads shall be used in the design of the Containment Vessel and Appurtenances.

D -

Dead Load, including the weight of the steel containment vessel, penetration sleeves, equipment and personnel access hatches, and other attachments supported by the vessel, plus loads due to concrete shrinkage.

L -

Live Loads, as applicabie, including:

1. Penetration Loads (including seismic), as applicable

2. FIoor Loads - 100 PSF

3. Walkways -200 lbs per I inear f oot

4. Equipment and Personnel Airlock Floor Load - 300 PSF or 40,000 lbs moving concentrated load

5. Emergency Airlock Floor Load -200 PSF or 10,000 lbs.

6. Polar Crane Loads (Ref.1)

7. Construction Loads *

8. Mezzanine -200 PSF

9. Painters Line Anchor - 2,000 lb. In any horizontal direction

10. Interior Scaffold -2,000 lb. each on any 2 adjacent clips Support Clips - combined with a Dead Load on all clips of 200 lbs.

O V

each.

P i

- Internal Design Pressure (or Transient Pressure Loads)

P e

External Design Pressure P

t Testing Pressure T

o - Thermal loads due to temperature gradient through walls under normal operating conditions.

T' -

Thermal loads due to temperature gradient through walls f rom accidents, such as major sodium fires.

T -

Thermal load under testing temperature conditions.

t E -

Loads resulting f rom an Operating Basis Earthquake (OBE)

E' -

Loads resulting from a Safe Shutdown Earthquake (SSE)

• A concrete placement load, resulting f rom using the vessel shell below operating floor elevation as the f ormwork for placing the reinforced concrete walls, and loads that are imposed by concrete f orms when constructing the conf inement shel l. A snow load will be considered also during the construction period.

3.8-5 Amend. 62 Nov. 1981

1 45 R - Accident loads due to Extremely Unlikely Faults (such as interfacing loads from inner cel l concrete structures) .

Note (11 The crare live load shall include, as appropriate, the vertical Impact load and the lateral thrust load as determined in accordance with Ref erence 3.

Design Pressures and Temoeratures The design pressures and the associated design temperatures shall be as' specifled below:

Internal Design Pressure 10 psig External Design Pressure 0.5 psig Design Temperature 2500F 30 The design of the containment vessel may also consider a transient design pressure end attendant temperature loading due to extremely unlikely faults.

Details of this information will be provided in the FSAR.

The operating condition contair. ment a1Tnosphere temperature and pressure are as folicws:

Operating Condition Tanperature = 700F Operating Condition Pressure = 0.0 psig Lowest Service Metal Temperature = 15 F 3.8.2.3.2 Loading Combinations The loading combinations for which the vessel and its appurtenances are to be designed shall be, but nor limited to, those specified in Table 3.8-1. The contairment design requirements and limits shall be in accordance with ASE- 1 I i, ArtIc! e NE-3000.

l For conditions where seismic Iceds are involved, the design analysis l requirements and criteria as contained in Section 3.7 shall also be met.

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l Amend. 61 3.8-6 Sept. 1981 l

r 's U Earthquake Loads and Analysis The containment vessel will be designed for seismic effects based on seismic loadings to be provided by seismic analysis to be performed on the overall structural arrangement of the Nuclear Island. The vessel will be analyzed with these loadings in accordance with Appendix 3.7-A CRBRP Seismic Design Criteria.

In the seismic analysis, it will be assumed that any direct connection between the containment vessel and other structures, such as piping and locks.has sufficient flexibility to preclude any coupling of the containment vessel except as specified in the following paragraphs.  ;

O LJ Amend. 33

3. 8 -7b Jan. 1977 l

The seismic analysis will include the local effects of the air locks vibrating as independent systems. The seismic effects of this independent vibration wIII be added directiy to alI other seismic effects.

The Equipment / Personnel airlock will be supported entirely by the containment vessel shell.

3.8.2.5 Structural Acceotance criteria A certified stress report shall be prepared for the Containment Vessel which will meet the requirements of Section ill of the ASME Code. In addition, the Buckling Stress Criteria contained in Appendix 3.8A will be imposed on the vessel cesign.

3.8.2.5.1 Pressure Tests A pneumatic pressure test shall be made on the Containment Vessel, airlocks, and equipment hatch at a pressure of 11.5 psig. Both inner and outar doors of the airlocks w ill be tested at this pressure. All pneumatic tests shall meet the requirements of Appendix J to 10CFR50, NE 6000 of ASME lit, and division 2 of ASME lil, as applicable. Exceptions to Appendix J to 10CFR50 are Identified in Section 6.2.1.4 3.8.2.5.2 Leakage Rate Test Folicwing successful completion of the pressure test, a leakage rate test at 10 psig will be performed with the airlock inner doors closed. The allowable leakage rate of the steel containment shall be 0.1% by volume of the containment in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and shall meet the requirements of 10CFR50 Appendix J.

The vessel plate seam welds will be Inspected and tested before placement of adjacent concrete in accordance with the requirements of Section lil, Articles NE-5000 and NE-6000 of the ASME BPV Code. Around all penetrations through the outer wall, sufficient space will be left between concrete and the penetration assembly such that permanent inspection of penetration leaktightness can be performed.

3.8.2.5.3 Ooerational Testing After ccepletion of the airlocks, including all latching mechanisms, Interlocks, etc., each airlock will be given an operational test consisting of repeated operation of each door and mechanism to determine whether all parts are operating smoothly without binding or other defects.

O 3.8-7c Amend. 62 Nov. 1981

i

c. Code for Concrete Reactor Vessels and Containments of ACI ASE Committee (ASE B&PV Code, Section lil, Division 21975.

45 d. Applicable State Codes

e. Specificatloa for the design and construction of Reinforced Concrete 61 chimneys (ACI-307-69). (See below)

ACI-318 wilI be extensively used for the design of the internal structures.

ACI-318 Is generally based upon ultimate load design. Since loading j combinations for the Internal structures require that ultimate capacity of a section be always greater or equal to the imposed _ load combination, this code is best appropriate for the design of the above structures. Chapter 18 of ACl-318 will not be invoked since this is not relevant to the structures under 4

discussion. Applicable portions of Appendix A of ACl-318 will be applied to i

the design. Since ACl-318 does not f ully cover design requirements for thermal stresses due to temperature gradient, the recommendations of ACl-307 y

j 61 will be used for guidance in addition to ASE Section ill, Division 2 for specific areas. '

AlSC specification wIlI be appiled to the structural steel members such as steel embedments, beams, equipment and pipe supports and restraint siructures.

j 15 l 3.8.3.2.2 Structural Soecifications 1 i See subsection 3.8.4.3 for the appropriate structural specifications.

4

• 3.8.3.2.3 NRC Reculatorv Guides

, The design will meet requirements or basic intent of the following NRC

15 Regulatory Guldes

s 61l a. 1.10 Mechanical (Cadweld) Splices in Roinforcing Bars of Category l Concrete Structures (Revision 1, 1/73) 44

b. 1.15 Testing of Reinforcing Bars for Category i Concrete Structures 1 (Revision 1, 12-28-72) i

, c. 1.28 Quality Assurance Program Requirements (Design and Construction)

d. 1.29 Seismic Design Classification (Revision 1, 8/73)

! e. 1.55 Concrete Placement in Category 1 Structures (6/73)

! f. 1.60 Design Response Spectra for Seismic Design of Nuclear Power

, 45 Plants (Revision 1, December,1973) i O

3*8-11 Amend. 61 Sept. 1981 r7-ve- f'gmwW wg ymesty .

nrwiyg==egr,q-yeum e %g9,99 m 9%ep,,.,y--,yp.9.p-+ >ge77 h g= +r4 *

g. 1.61 Damping Values for Seismic Design of Nuclear Power Plants (Oct.

1973)

h. 1.69 Concrete Radiation Shield for Nuclear Power Plants (12/73)

1. 1.92 Combine Modal Responses and Spatial Components in Seismic Response Analysis (Revision 1, Feb, 1976) l J. NUREG - 0554, SIngie-FaIiure-Proof Cranes-for NucIear Power Plants 3.8.3.2.4 ASTM Standards Ali ASTM Standards to the extent they are referenced in the codes and standards noted in Section 3.8.3.2 and further specifically identified in other parts of Section 3.8.8, will be applied to the design of the facility.

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3.8-11 a Amend. 62 l

Nov. 1981

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

i j

Since the walls, ceilings and floors of each cell are considered as two-way

! slabs, the applied loads, used in the analysis, will be prcportioned to the

one-foot wide strips, in orthogona' directions, according to the ratio of their relative stif f nesses.

. The cell design will be verified )y using a three dimensional finite-element analysis with the computer program NASTRAN. The cell and adjacent structures will be represented in the mathematical model which will include the Interaction with the containment shell and the exterior concrete wall. The appropriate loads and load combinations will be used in the analysis. 33 Further detailed analysis will be perf ormed in areas of load concentration and penetrations as noted in 3.8.3.4.3. The reactor cavity is treated as a hollow cylinder for structural analysis. When in-house computer programs are used, their correctness will be verified againsi acceptable published programs. All vertical loads will be transferred to the foundation mat by three principal structural elements, v!z (a) walls of PHTS cells (b) perimeter wall around j containment and (c) reactor cavity.

3.8.3.4.2 Analvsis for Seismic Loads J

Equivalent static seismic loads as developed from the dynamic analysis of the structure wilI be transferred through the horizontal slab diaphragms and vertical shear walls to the foundation mat. The details of seismic analysis are described in Section 3.7.

3.8.3.4.3 Analysis for Ocenings Structural analysis will be performed around openings in walls and slabs particularly where concentrated loads from thermal effects are Induced. The 31 design will account for all the stresses in those areas a7d proper 28 reinforcement wilI be provided for the rellef of such stress concentration.

l 3.8.3.4.4 Liner Analysis i

i The liner-anchor system will be designed and analyzed in accordance with the requirements and criteria specifled in paragraph 3.0 of PSAR Appendix 3.8-B.

37 Liner analysis is discussed in PSAR Section 3A.8.3.5.

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l l 3.8-18a Amend. 61 Sept. 1981 i._.,__,_ _ _ _ _ _ _ . _ _ _ _ _ _ . , _ _ . _ _ _ _ _ _ - _ . . . _ . - , . . _ . _

3.8.3.4.5 Radiation Generated Heat Effeet Adequate heat removal capacity will be provided for the reactor cavity so that the radiation generated heat does not cause temperatures of the structural materials in excess of the ASME Code, Section lil, Division 2 requirements.

Where radioactive piping penetrates the concrete walls, adequate insulation or heat removal measures will be provided to control the temperature of structural materials wIthin Code limits.

Radiation generated heat will be produced as a function of the position of the Reactor core with respect to the structures and the temperature diriributions will be calculated. In PHTS cells, significant heat will be generated from other sources such as piping. In all such instances, adequate cooling capacity wIii be provided to Iimit the temperature of the structural materials.

3.8.3.4.6 Reinforcement Design The reinforcing steel will be proportioned to meet the requirements of ACl-318. The bond and anchorage requirement of ACI-318 will be observed, since the Interior structures primarily provide a confinement function. The reinforcement for each wall or slab will principally consist of a set of orthogonal bars on each face with additional reinforcement provided in areas of penetrations or load concentration.

3.8.3.4.7 Structural Steel Design The structural steel components other than the cell liner systems and reactor O vessel support ledge will be designed to the requirements of AISC specifications as identified in Section 3.8.2.2.1, When steel parts are stressed into the plastic range, an energy absorption check will be performed to assure the functional integrity.

3.8.3.4.8 Reactor Vessel Succort Ledge Design The reactor vessel support ledge will be designed in accordance with Subsection NF of the ASME Code, Section lil, Division 1. The ilnear analysis method will be used. A finite element computer analysis will be performed for the support ledge.

O eni M 3.8-19 Nov. 1981

3.11 ENVIRONMENTAL DESIGN OF MECHANICAL AND ELECTRICAL E00lPMENT 3.11.1 Egt: foment Identification The safety related systems which are required to function during and following an accident are identified in Section 3.2. Worst case environmental conditions including temperature, pressure, humidity, chemical and radiatlan exposure which result from a postulated design basis accident have been defined for each location. Reference 13, PSAR Section 1.6, describes the environmental qualification basis for such equipment and the program that will be followed to assure the basis is satisfied. The objective of the qualification basis and the qualification program is to conform to IEEE Standard -323-1974 "lEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations." One aspect of anticipated CRBRP Environmental Conditions different from those for a typical light water reactor plant is that some accident environments may include a sodium oxide aerosol. The CRBRP Equipment Environmental Qualification program qualifies prototypic equipment to high concentrations of sodium oxide aerosols. Equipment will bo qualified for environments of low concntrations of sodium oxides as required through a generic test program.

Where possible, the equipment comprising the safety related I&C systems is located in controlled atmospheres (e.g., control room). Safety related equipment located outside controlled atmospheres will be designed to operate through, or be protected from, the worst environmental conditions for which the equipment must perform.

O

() 3.11.2 Oualification Tests and Analvses The program of environraental qualification tests and analysis for Class 1E Equipment is described in Reference (13) of PSAR Sec tion 1.6, "CRBRP Requirements for Environmental Qualification of Class 1E Equipment." This document establishes the qualification program which will be conducted to qualify Class IE equipment located in different areas of the CRBRP and sets forth the documentation to be completed for qualification. The entire program is designed to conform to the IEEE Standard 323-1974.

3.11.3 Oualification Test Results The results of any qualification tests will be documented as specified in i Reference 13 of PSAR Section 1.6' and summarized, as appropriate, in the FSAR.

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3.11-1 Amend. 62 Nov. 1981 i /

3.11.4 Loss of Ventilation i Al plant locations containing safety related control and ,

electrical equipner:t, that need a controlled envircraent to maintain the required operabilit.y, are to be provided with redundant air coriditioning 31d/or ventilatirn facilities for the needed environmental coatrol. Analytical

, information on the various local environmental conditions in the plant is given in the corresponding sections in Chapters 2, 3, 6, 9 and 15.

As described in Section 3.11.1 above, the safety-related equipment is designed to operate successfully at environmental extremes.

3.11.5 Special Considerations l Enclosures containing safety related equipment will be designed l tu withstand the effectr, of normal, emergency and faulted conditions

! expected on the system. The effects of sodium spillage or fire will be protected against by placing redundant safety related equipment in l separate corpartments or rooms or by enclosing the safety related equipment l in nonabsorptive, noncombustable, explosion proof casings. Applicable code requirements including those of the National Electric Code will be satisfied, as appropriate.

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O 3.11-2 N

CHAPTER 3A SUPPLEMENTARY INFORMATION ON SEISMIC CATEGORY I STRUCTURES TABLE OF CONTENTS Page No.

3A SUPPLEMENTARY lNPORMAT10N ON SEISMIC CATEGDRY I STRUCTURES 3A.1 inner Cell System 3 A.1 -1 3A.1.1 Functional Design 3 A.1 -1 3A.I.2 Design Bases 3 A.1 -2 3A.1.3 Design Description 3 A.1 -3 3A.1.4 Design Evaluation 3 A.1 -3b 3A.1.4.1 Structures 3A.1-3b 3A.1.4.2 Inerted Celi Atmosphere Control System 3 A.1 -4 3A.1.5 Test 1.1g and Inspection 3 A.1 -5 3 A.1.6 Instrumentation Requirement 3 A.1 -5 3A.1.7 Materials 3A.1-5 3A.I.8 inner Barrier 3 A.1 -5a 3A.2 Head Access Area 3A.2-1 3A.2.1 Head Access Area Functional Design 3 A.2-1 3A.2.1.1 Design Bases 3A.2-1 3A.2.1.2 Design Description 3A.2-1 3A.2.1.3 Design Evaluation 3 A.2-3 3A.2.1.4 Testing and Inspection 3 A.2-3 3A.2.2 Head Access Area Heat Removal Systei 3 A.2-3 3A.3 Control Building 3A.3-1 3A.3.1 Design Bases 3A.3-1 3 A.3.2 Design Description 3 A.3-1 3 A.3.3 Design Evaluation 3 A.3-2 3 A.3.4 Ter'Ing and Inspection 3 A.3-2 3 A- 1 Amend. 62 Nov. 1981

_ TABLE OF CONTENTS (Continued)

Page No.

3A.3.5 Instrumentation Requirements 3A.3-2 3A.4 Reactor Service Building (RSB) 3A.4-1 3A.4.1 Design Bases 3A.4-1 3A.:.2 Design Description 3A.4-1 3A.4.2.1 Reactor Services Building 3A.4-1 3A.4.2.2 Auxiliary Coolant Fluid 3A.4-3 39l 3A.4.2.3 Deleted 3A.4-3 3A.4.2.4 Heating and Ventilation 3A.4-3 3A.4.2.5 Sodium Fire Protection 3A.4-3 3A.4.2.6 Recirculating Gas Cooling 3A.4-3 3A.4.2.7 Reactor Refueling 3A.4-4 3A.4.2.8 Nuclear Island General Purpose Maintenance 3A.4-4 Equipment 3A.4.2.9 Auxiliary Liquid Metal 3A.4-5 3A.4.2.10 Inert Gas Receiving and Processing 3A.4-5 3A.4.2.11 Impurity Monitoring and Analysi' 3A.4-5 3A.4.2.12 Fuel Failure Monitoring 3A.4-5 l

3A.4.3 Design Evaluation 3A.4-5 l 3A.4.4 Tests and Inspection 3A.4-6 l

3A.4.5 Inctrumentation Requirements 3A.4-6 l

i l

3A.5 Steam Generator Building 3A.5-1 3A.5.1 Design Bases 3A.5-1 3A.5.2 Design Description 3A.5-1 3A.5.2.1 Heat Transport and Steam Generator System 3A.5-3 3A.S.2.2 Steam Generator Auxiliary Heat Removal System 3A.5-3 3A-ii Araend 39 May 1977

3A.2 HEAD ACCESS AREA 3A.2.1 Head Access Area Functional Design The Head Access Area (HAA) provides electrical and gas services, access platf orms and the environment necessary f or reactor operation, ref ueling, inspection, surveillance and maintenance.

3 A.2.1.1 Design Bases The HAA air temperature must be compatible wi th instrumentation, equipment, seals and personnel access. The temperature of all routinely accessible surf aces of HAA equipment will be limited to 1250F maximum f or metallic surf aces and 1400F f or insulated surf aces. Stairways, ladders, walkways, and platf orms will be designed to OSHA standards. The radiation dose rate will be Ilmited to values as def ined in Section 12.1.

3 A. 2.1.2 Design Descriotion The HAA is open to reactor containment at the operating floor elevation (816').

The general arrangement is shown on Figure 3A.2-1. The HAA is 44 feet square and 14 feet deep (below operating floor level). The top surf ace of the reactor head and the HAA floor are at the same elevation (802').

Electrical services enter the HAA by way of wall penetrations leading into terminal boxes and/or condult. Electrical cables are routed vie wireway and conduit (separated according to signal type) to the equipment on the rector O5 closure head. During refueling the service bridges are swung aside and the cable transfer machine provides the services needed on the rotating plugs.

Gas services enter the HAA via pipes through the south and east walls. Gas services to the rotating plugs are also provided through the service bridges and cable transf er machines.

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3 A.2-1 Amend. 62 Nov. 1981

Major items within the HAA include:

Small Rotating Plug / Riser Assembly in-Vessel Transf er Machine (IVTM) Port Nozzle IVTM Port Adapter (Preparation f or Ref ueling)

IVB4 (Refueling)

Intermediate Rotating Plug / Riser Assembly Primary Control Rod Drive Mechanisms (PCRDMs)

Secondary Control Rod Drive Mechanisms (SCRDMs)

Shield & Selsmic Support (S&SS)

Liquid Level Monitors (2)

Upper Internals Structure (UIS) Columns (4)

UlS Jack!ng Mechanisms (UISJMs) (4)

IRP Pl atf orms and Ladders IRP Cable Transfer Machine & Cooling Duct SRP Drive Unit & Lock Large Rotatino Plug / Riser Assembly Fuel Transf er Port and Lower Adapter FTP Upper Adapter & Floor Valve (Ref uellrg)

Liquid Level Monitors (2)

LRP Pl atf orm & Ladders Cable Transfer Machine and Cooling Ducts IRP Drive Unit & Lock Off Head items Reactor Vassel Support Ring & Cavity Seal Head Heating System Cabinets LRP Drive Uni is ( 2) and Lock HAA Electricci Equipment & Wiring Radiation Monitors (4)

FL ux Monitor Preampl t is s Remote Data Acquisitic r Terminal HAA Service Bridges and Ref erence Units HAA PCRDM Cooling Pipes HAA Cooling Ducts HAA Gas Service Equipment Seismic Recorders North & South Service Platf orms & Stairs HAA Plug Drive Control Box EVTM Cavity Beams & Columns O

3A.2-2 Amend. 62 Nov. 1981

3A.2.1.3 Design Evaluation A Head Access Area which provides ready access to operating personnel enhances both the safety and availability of the plant by permitting surveillance and Inspection of the many systems and component in this area during operation.

This will provide a good basis for decisions that must be made during operation.

Radioactivity in the Head Access Area will be monitored to detect leakage of gas f rom any of the seals in the head. Detectors will be located in the head access area. The HAA air will be recirculated once every two minutes. This will reduce the chance of any local buildup of leaking gas. The relatively low cover gas pressure (10 Inches of water) and the use of double seals with buf f er gas between will minimize leakage and dilute any cover gas which may leak into the HAA.

3A.2.1.4 Testing and Insoection There are no testing and in-service inspection requirements for the Head Access Area.

3A.2.2 Head Access Area Heat Removal System HAA heat loads include heat from the reactor vessel support ledge, HAA floor, closure head and heaters, and miscellaneous electrical equipment. Cooling is provided by an independent air conditioning system located in an adjacent Q

(/

cell. The cool pool / focused flow ducting system directs cooling air to critical regions on the closure head. Warmed air returns to the air conditioning system via a grille located in the south wall of the HAA.

Loss of the heat removal system would result in a natural convective air flow that would transfer heat to the reactor containment building atmosphere, in the worse case (loss of of f-site power and loss of power to the HAA cooling system) the R G atmosphere is expected to rise to a maximum of 1200F from its normal temperature of 80 to 850F resulting in a temperature rise within the head access area which will not exceed 400F. Equipment in the HAA will remain functional at an R G atmospheric temperature of 1200F.

O 3A.2-3 Amend. 62 Nov. 1981

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$$ Figure 3A.2-1. CRBRP Head Access Area

4.2.1.5.2 Fuel Assembiv Fabrication and Site Examination OualItv Assurance Provisions A. Fabrication Examination A comprehensive quality assurance program will be employed during all phases of the fuel assembly fabrication to ensure compliance with design parameters to a high degree of certainty. During final design, the details of the program were documented as quality conformance inspection plans, in the formalization of these plans, all available Regulatory Guides and the DOE Reactor Development and Technology Standards used for FFTF fuel assembly fabrication were considered. For the fuel rods, fact assembly duct tubes and fuel assembly hardware, dimensional inspection will typically be 100%. In Tables 4.2-59A and 4.2-59B are lists of typical inspection characteristics in the plan for the fuel rod and fuel assembly and their various components.

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4.2-112 Amend. 62 l Nov. 1981

B. On-Site Acceotance Tests To assure that damage during shipping, loading and unloading or from sabotage O

has not cccurred, a standard inspection procedure w il l be perf ormed on every fuel assembly entering the plant site Similar to the inspection f or incoming light water reactor and FFTF assemblies, this procedure consists of:

a. Examination of the shipping container for broken seals, dents, penetrations, sheared bolts or any sign of shipment damage and verif Ication that the shipping container shock Indicators were not tripped.

b. A standard visual inspection of the assembly for dents, nicks, and gouges, especially in the area of hexagonal load pad corners, shield block corners, the inlet nozzle, piston ring and discrimination post, and correct assembly identification by verifying handling socket identification, discrimination post gecrnetry and assembly serial number.

Defects determined during visual examination shall be photographed and selective dimensional inspection of external features shall be performed using manual general purpose tools that are commercially available, This inspection would be perfcrmed to determine if defects are of such an insignificant nature to be acceptable. A photographic record will be maintained for all accepted defects.

4.2.1.5.3 Blanket Assembiv Planned Desien Verification Tests The inner and radial blanket assemblies have no equivalent in the FFTF O reactor. The assembly hardware (Inlet nozzle, outlet nozzle, and duct) are simlier to the CRBRP core f uel assembly hardware, however, signif icant differences between the fuel and radial and inner blanket assemblies require The design verification tests presented here. The blanket rods have outside diameters and f uel pellets which are similar to LWR f uel recs. Thus, this LWR fuel red fabrication technology can be used to fabricate and inspect the blanket reds.

l l

l O

4.2-113 Amend. 62 Nov. 1981

I TABLE 4.2-59A QUALITY CONFORMANCE INSPECTION CHARACTERISTICS - FUEL ROD AND COMPONENTS Fuel Rods End Cap-t'o-Clad Welding Qualifications General Weld Throat Thickness 1

Fuel Rod Component Placement Destructive Test Components and Materials Nondestructive Test Gas Tagging Weld Surface Tag Gas Composition Internal Weld Number of Rods per Tag Destructive Test Fuel Rod Dimensions Nondestructive Test Length Weld Diameter Fuel Rod Bow Wire Wrapping Concentricity End and Weld Location End Cap Angular Orientation Overall Length Fuel Pellet Stack Orientation i Pellet Type Number of Turns Other Conformance Tension Weight Wire-Cladding Gap Length and Pellet Spacing End Cap-to-Wire Wrap Welding Before Rod Assembly Qualification After Rod Assembly Wire Weld Strength Volatiles Content Surface Characteristics Gas Cleanliness Moisture Surface Passivation

0xygen-to-Metal Ratio Residual Chloride and Flunride

, Axial Blanket Pellet Stack Components Confonnance Fuel Rods Weight Radioactive Surface Contamination Length Smearable Volatiles Content Fixed Gas Surface Defects Moisture Surface Roughness 0xygen-to-Metal Ratio Identification Fuel Rod Fissile Content Workmanship Bonding Gas Assembly Sequence Composition Tag Gas Weight Welding Chamber Helium Purity Capsule Loading Pressure Gas Release Fuel Rod Integrity Lotting l

Handling and Storage During Processing j Archive Samples 1

4.2-439a l

Amend. 62 i Nov. 1981 ,

TABLE 4.2-59A (Continued)

Fuel Pellets Axial Blanket Pellets Plutonium Fom and Properties Uranium Fom and Properties Isotopic Concentration Purity - Minimum Uranium Other Characteristics Individual Impurities Uranium Fom and Properties Oxygen-to-Metal Ratio Fuel Pellet Fissile Concentration Gas Content Uranium Concentration Moisture Content Fuel Pellet Plutonium Concentration Weight per Unit length Individual and Total Impurities Diameter and Length Lot Maximum Radius or Chamfer Core Zone Perpendicularity Americium Content Surface Condition Oxygen-to-Metal Ratio Process Sequence Pellet Pellet Manufacture Lot Average Lotting and Identification Gas Content Storage During Processing Moisture Content Archive Samples Pellet Lot Average Tao Gas Capsule Homogeneity - Figure of Merit Plutonium Concentration Regions Dimensions Grain Size, Porosity, Secondary Capsule Materials Phases Tag Gas Composition Fuel Pellet, Weight Per Unit length Blending Gas Fuel Pellet Dimensions Isotopic Ratios Length Xe to Kr Ratio Diameter Impurities Dish Diareter Tag Gas Capsule Dish Depth Capsule Integrity Perpendicularity Identification Pellet Weignt Cleanliness Surface Condition Residual C1 and F1-Components Cefects Surface Roughness Appearance Weld Procedure Cleanliness Laser Drilling Procedure Process Sequence EB Weld Thickness Pellet Manufacture Weld Integrity Lotting and Identification Maximum Diameter Storage Curing Processing Helium Leak Test Archive Samples Tag Gas Capsule Filling Lotting Archive Samples 4.2-439b Arend. 62 Nov. 1981

k TABLE 4.2-59A(Continued)

Uranium Dioxide Tubing (Cladding) l Maximum Particle Size Chemical Composition j Particle Distribution Ingots

, Surface Area Finished Tubing l Isotopic Content Cold Work I Impurity Levels Tensile Properties 1 0xygen-to-Metal Ratio Hardness Uranium Content Ductility and Soundness i Sinterability Flaring l Analytical Samples Burst Pressure

! Sampling Grain Size i

Lotting Inclusion Content Intergrannular Attack Plutonium Dioxide Carbide Precipitation '

Corrosion Resistance Sieve Analysis Surface Condition Particle Distribution Surface Roughness Surface Area Tube Hollows 1 Plutonium Isotopic Concentration Finished Tubing Americium Content Surface Marring Impurity Levels Residual C1 and F1 0xygen-to-Plutonium Ratio Passivation Plutonium Content Cleanliness Packed Materials Ovality After Calcining Wall Thickness and Eccentricity Loss on Ignition Straightness

, Sinterability Penetrant Examination Screening Halogens and Sulfur i Blending Tube i Analytical Samples Finished Tubing

, Sampling Ultrasonic Examination Lotting Melting Tube Making '

Heat Treatment Lotting Identification 4

4.2-439c Amend. 62 Nov. 1981

4 i

TABLE 4.2-59A (Continued)

Wrap Wire l Chemical Composition Ingots Finished Wire Cold Work Tensile Properties Hardness I

Ductility Grain Size l Inclusions Intergranular Attack Carbide Precipitation Corrosion Resistance Surface Condition l Surface Roughness l Surface Marring l Residual C1 and F1 Cleanliness l Dimensions Heat Treatment Lotting Identification O

4.2-439d Amend. 62 Nov. 1981 i

TABLE 4.2-598 QUALITY CONFORMANCE INSPECTION CHARACTERISTICS FOR FUEL ASSEMBLY AND COMPONENTS Fuel Assembly Duct Tubes Fuel .'ssembly Components CMmical Composition Component Materials Cold Work Components Mecitanical Properties Fuel Assembly Hardness Gas Tag Grain Size Enrichment / Component Match Inclusions Fuel Rod Position Intergrannular Attack Fissile Content Carbide Precipitation Fuel Rods Corrosion Resistance Assembly Surface Condition Americium Content Surface Roughness Fuel Rods Eurface Marring Assembly Pesidual C1 and F1 Dimensions Cleanliness Lengths Dimensions Weld Surfaces Wall Thickness Straightness Penetrant Examination Alignment Ultrasonic Twist Melting Tight Bundle and Duct Tube Gap Tube Making ps Fuel Assembly Weight Heat Treatment Cleanliness Lotting

')

Residual C1 and F1 Repair and Rework Surface Texture Archive Sample Surface Marring Identification Identification Fuel Assembly End Components Workmanship Processes Machining Welds Process Archive Samples Air Flow Test Handling and Storage During Processing Fabrication Records Packaging 4.2-439e Amend. 62 Nov. 1981

TABLE 4.2-598 (Can'inued)

Piston Rings Hard Facing Chemical Compsi tion C :e:Ical Composition Grain Size Perticle Size Tensile Pr perties Strength Hardness Hardness Surface Requirements Metallograpnic Surfuce Roughness Surface Condition Surface Marring Surface Roughness Surface Indications Surface Marring Dimensions Cleanliness Outside Diameter Residual C1 and F1 Roundness Dimensions Flatness Hard Surface Application Hook Free Gap Lotting Width Identification Thickness Archive Samples Coating Thickness Perpendicularity Cleanliness Heat Treatment Coating Lotting Identification O

4.2-439f Amend. 62 Nov. 1981

d. The gamma background at the detector location. The gamma background will be a primary f actor in determining Instrument sensitivity at a O given location.

e. The abIIIty of the SRFM to accurateIy determine the reactivity worth of in-core control rods by means of the inverse kinetics rod drop technique. The application of this IKRD technique at the ex-vessel detector locations is required to properly calibrate the SRFM system f or sub-critical Ity determination.

Extensive analyses at ARD and both analyses and experiments at Oak Ridge National Laboratory (Reference 1 and 2) have been performed in support of the ex-vessel SRFM system. Particular emphasis was placed on investigating the five nuclear characteristics listed above. The significant results of the analyses and experiments performed to date are summarized below.

Calculations were performed to assess the magnitude and spectrum of the neutron flux at the ex-vessel SRFM location during shutdown conditions. These calculations were performed for beginning-of-life conditions (all fresh fuel, FFTF-grade plutonium in the core). The minimum shutdown flux at the SRFM locations (beginning-of-life conditions) was calculated to be approximately 0.1 ny, which corresponds to about 4 counts per second at each BF 3 proportional counter. The magnitude of this count rate which is smaller than during any subsequent refueling sequence, assures good counting statistics for monitoring subcriticality and refueling operations. Additional calculations have shown that the neutron flux is almost fully thermalized ( 85% below 0.1 ev) at the SRFM location, eight Inches behind the front face of the graphite O block. This enhancement of the thermal flux inside the graphite block has been confirmed by experiments performed by ORNL at the Tower Shield Facility near Oak Ridge, Tennessee (Reference 1).

To investigate the effect of core configuration on count rate, the homogeneous core configuration was modified by employing different banks of control rods to maintain a fixed reactor power level and K . For these reactor configurations, the flux level at the SRFM va$d by less than 10%. This result shows that the ex-vessel detectors are not sensitive to changes in the homogeneous core configuration during constant power operation. The detector response is proportional to the power level of the reactor. Similar calculations will be repeated for the present heterogeneous core layout and the results will be reported.

Regarding the possibility of the detector monitoring neutron flux from sources other than the core, analyses have shown that the flux monitoring requirements for the SRFM can be satisfied with the background associated with a maximum discharge fuel assembly withdrawn to a point 64 inches above its fully inserted position for any core location.

'O 4.3-9 Amend. 62 Nov. 1981 l

The count rate due to background is minimized by shielding in the form of boron carbide slabs which surrounds all sides of the graphite block except the front face. The shielding is used to reduce the count rate from neutrons which are scattered into the graphite block f rom the reactor cavity walls.

Normal ref ueling procedures wIll require that no assembly of any type be in the FT&SA whlie any assembly is being Inserted the last 64 inches into the core or withdrawn the first 64 inches from the core so thaT all three SRFM detectors will have an unhindered view of the core. For this case, the requirement of (c) above are imposed.

The gamma dose at the SRFM location immediately af ter shutdown has been analyzed in detail to assure that the sensitivity of the BF3 neutron counters is not adversely affected. The type of BF3 neutron detectors to be used in CRBR have a minimum sensitivity of 40 counts per second/ thermal equivalent nv for gamma dose rates less than 100 R/hr. When the gamma dose exceeds 100 R/hr the detector sensitivity f alls of f rapidly. Calculations have shown that the local gamma dose rate at the SRFM location is less than 100 R/hr with appropriate shielding in front of the graphite block and in the other locations as required.

A principal function of the SRFM is to determine the subcritical reactivity of the CRBRP based on proper calibration of the Instrumentation near critical.

The recommended method for calibrating the SRFM detectors is a two step procedure. First, a known value of negative reactivity must be established.

This is accomplished by using the SRFM count rate trace that results from scramming one or more control rods to determine the reactivity worth of the scrammed rods. This is known as the Inverse kinetics rod drop (lKRD) technique. Secono, the calibration constant, which relates the subcritical reactivity to the count rate, must be determined. This is accomplished by inserting the previously measured reactivity worth (the same control rods described above) and noting the corresponding count rate. This same calibratica constant is then used to imply subcritical reactivity when all the control rods are inserted and the reactor is fully shutdown.

This procedure depends strongly on the accurate determination of the negative reactivity worth of the scrammed control rods by means of the IKRD technique.

ORNL has perf ormed numerous rod-drop experiments (Ref erence 2) in the Tower Shiel d Facil ity in addition to analytical calculations and both have supported the conclusion that reactivity interpretations, based on the change in count rate at the ex-vessel detectors, are consistent with in-core detectors. The experiments and analyses performed to date have not included the ef f ect of neutron streaming in the reactor cavity. Future analyses will investigate these reactor cavity ef f ects and the results will be included in the FSAR.

The neutron source multiplication technique is employed to monitor the subcritical reactivity state of the reactor during the loading to critical and all subsequent f uel reloadings. The relationship between the steady-state SRFM de'ector count rate and the subcritical reactivity is derived f rom the point klaetics equations:

O 4.3-10 Amend. 62 Nov. 1981

5.3-28 Deleted 5.3-29 Deleted 5.3-30 Deleted i 5.3-31 Deleted 5.3-32 Deleted 5.3-33 Deleted 5.3-34 Deleted 5.3-35 Deleted 56 5.3-36 Pipe Hanger Clmp Assembly 5.3-134 5.3-37A Shear Pin Assembly 5.3-135 5.3-37 B Pipe Hanger / Snubber Arrangement 5.3-136 Vertical Pipe Clmp Assembly 5.3-137 )

5656l 5.3-38 5.3-39 S/N Curves of 18/12 and 18/12/0.05C Alloys at 12920F 5.3-138 5.3-40 Deleted 5.3-139 5 .3 -41 Deleted 5.3-140 56 5.3-42 Precipitation Reactions in Type 316 Stainless Steel Solution Treated at 23000F f or 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and Water Quenched 5.3-141 25 Intermediate Sodium Pump Isometric 5.4-36 41l 5.4-1 Intermediate Sodium Pump 5.4-36a l 5.4-1 A 44 i

t lO 5-xxia Amend. 59 Dec. 1980

i 5.4-2 IHTS Components and Piping Ster-m Generator Celi 5.4-37 5.4-3 Intermediate Pump Characteristics 5.4-38 5.4-4 Deleted 5.4-5 Deleted 5.4-6 Loadings Considered I. Crack Growth Analysis 5.4-41 5.5-1 Steam Generation System Hydraulic Profile 5.5-54 5.5-2 Steam Generator Module 5.5-55 ,

1 5.5-2A Main Steamline Isolation Valve 5.5-56 5.5-3 Transwrap Model of the IHTS and Rellef System 5.5-57 l

l l

l l

i l

l l

O 5-xxib Amend. 62 Nov. 1981

1%GE 5.5-4 Steam Drum OutiIne 5.5-58 5.5-4A Predicted Pressure History in lHX for Sodium-Water Reaction Design Basis Leak in Evaporator 5.5-58a 5.5-4B Leaksite Pressure History for Sodium-Water Reaction Design Basis Leak in Evaporator 5.5-58b 5.5-5 Pipe Whip Test Configuration, Test 3 and 4 of Reference 9 5.5-59 5.5-6 Preliminary Seismic Model of the Stean, Generator Module 5.5-60 5.5-7 Seismic Model Sodium Dump Tank 5.5-61 5.5-8 Seismic Model Water Dump Tank 5.5-62 5.5-9 Seismic Mcdel Reaction Products Separation Tank 5.5-63 5.6-1 Heat Load On SGAHRS (U-1B Event) 5.6-36 5.6-2 SGAHRS Response (U-16 Event) 5.6-37 5.6-3 Na Temperature Response (U-18 Event) 5.6-38 5.6-4 Temperature Estimates for Early Stages of Transient Primary Loop Temperatures 5.6-39 5.6-5 Temperature Estimates for Early Stages of F-2 Transient Reactor Temperatures 5.6-40 5.6-6 Decay Power After Shutdown EquiiIbrium Cycle Conditions Maximum Values 5.6-41 5.6-7 PACC Air Side Isolation Schematic (Shown During Normal Plant Operation - PACC Hot Standby) 5.6-42 5.7-A Typical Start-up Characteristics of Reactor Heat Transport System 5.7-8aa 5.7-B Typical Start-up Characteristics of Steam Generator System 5.7-8b 5.7-C Typical Shutdown Characteristics of Reactor Heat Transport System 5.7-8c 5.7-D Typical Shutdown Characteristics of Steam Generator System 5.7-8d 5-xx!I Amend 62 Nov. 1981

l 9

44 f 5.7-1 Expected System Parameters vs. Power Level 5.7-9 5.7-2 Expected Loop Flow Parameters vs. Power Level 5.7-10 5.7-3 Reactor Vessel Outlet Temperature vs. Time for Reactor Trip from Full Power (fiinimum Decay Heat) 5.7-11 1

5.7-4 Primary Pump Sodium Temperature vs. Time for Uncontrolled Rod Withdrawal from 100 Power 5.7-12 i L 5.7-5 Primary Pump Sodium Terperature vs. Time for Uncontrolled Rod Withdrawal from Startup with Delayed flanual Trip 5.7-13 5.7-6 Intermediate Pump Sodium Temperature vs. '.

Time for Loss of Steam Generator Load

(Dumping of Water / Steam Side of Bath Evaporators and the Superheater) 5.7-14 5.7-7 Intermediate Pump Sodium Temperature vs.

Tine for Inadvertent Opening of Superheater ou tl e t 5.7-15 5.7-n Superheater Inlet Sodium Temperature vs.

Time for Primary Pump flechanical Failure 5.7-16 5.7-9 Check Valve Pressure vs. Time for Primary Pump flechanical Failure 5.7-17

5.7-10 Intermediate Pump Sodium Temperature vs.

Time for Saturated Steamline Rupture 5.7-18 l

5.7-11 Primary Pump Sodium Temperature vs. Time for Loss of Primary Pump Pony flotor with Failure of Check Valve to Shcr 5.7-19 l

i Amend. 44 5-xxiia April 1978 O

i l

, [ Valid for 297 51 5 922 K, 75-1200 F]

u = -1.806 - 227.391 (C+N) + 1218.392 T l/2

-4

+ 1.127 T (C+N) - 8.446 x.10 T2 (C+N) (2) standard deviation = 2.645

[ Valid for 366 5 T 5 866 K, 200-1100 F]

c = 29.595 - 89.898 (C+N) + 468.414 T-II (3) standard deviation = 2.955 l

j [ Valid for 477 5 T 5 922 K, 400-1200 F]

4 The above three equations are valid for (C+N) 50.13 weight percent and apply to plate, bar, pipe, and forged material, o 4

c is expressed in percent, and T is in degrees Kelvin.y and ou are in ksi, lf In the c m of solution annealed Type 316 stainless steel o

y = 88.474 + 211.640 (C+N) - 0.206 T (C+N) i

{ - 5.762 Tl/2 + 0.111 T (4) standard deviation = 2.806

[ Valid for 297 5 T 5 977 K, 75-1300 F]

t u = 8.563 - 461.665 (C+N) + 1064.154 T -1/2 (5)

! 17

-3 l + 2.722 T (C+N) - 2.330 x 10 T2 (C+N)

, standard deviation = 6.458 .

[ Valid for 297 5 T s 977 K, 75-1300 F]

-1/2 c = 28.000 - 99.139 (C+N) + 653.926 T (6) standard deviation = 6.703

[ Valid for 297 5 T 5 977 K, 75-1300 F]

' The above three equations are valid for (C+N) ( 0.10 weight percent, <

but at room temperature,1100 F, and 1300 F the range of validity may be increased to 0.167, 0.141, and 0.141, respectively. The equations may be used for plate, bar, pipe, rod sheet, strip, and forged material.

5.3-17 Amend. 17 Apr. 1976

Based on the above equations a general comparison may be made between Type 304 and Type 316 stainless steel. From Figures 5.3-1 and 5.3-2 it may be seen that for alloys containing equal quantitles of (C+N) the ultimate tensile strength and ducti!Ity of Type 316 stainless steel is far superior. However, the yield strength of Type 316 is only slightly higher, 5.3.2.2.2 Stress Ruoture Procerties The rupture strength of austenitic stainless steel is also strongly dependent on C+N content as shown in Figures 5.3-3 and 5.3-4. To a good approximation the rupture strength is a linear function of C+N.

Although the figures show that Type 316 SS has a higher rupture strength than Type 304 SS for a given temperature, the ef f ect of a change in the C+N on ihe rupture strength is basically the same for each alloy. As shown in Figure 5.3-5 a given change in C+N changes the rupture strength by the same amount for each alloy at a given temperature. There is a temperature effect however with a change in C+N influencing the rupture strength more at the lower temperatures. The percentage change in rupture strength for Type 304 and Type 316 SS for a given change in C+N is shown in Figures 5.3-6 and 5.3-7.

A comparison of allowable design stresses for the unstabilized austenitic stainless steels, as specified in the ASME Boller and Pressure Vessel Code, Section Ill, and ASME Code Case 1592 is given in Table 5.3-11. Note that the use of L grade materials is restricted to a temperature of less than, or equal to, 800 F.

5.3.2.2.3 Fatigue Procerties A comparison between the fatigue behavior of Type 304 and Type 316 stainless steel has been made for temperatures between 806 and 1202 F. The Type 304 contained 0.053 and 0.052 weight percent carbon and nil ogen, respectively, and the Type 316, 0.06 and 0.048 weight percent carbon and nitrogen, respectively.

For equivalent heat treatments and test temperatures Type 316 has a shorter fatigue life than Type 304 at the same total strain ;ange. If the comparison is based on the stress amplitude, however, Type 316 is superior. This is because for a given stress the corresponding strain range for Type 316 is somewhat smaller than that for Type 304 owing to the higher yield strength of Type 316 stainless steel. Hence, a selection of Type 304 or Type 316 for fatigue resistance will depend on whether the farigue condition will be strain or stress controlled.

The ef fect of carbon content on the f atigue behavior of 18 CR-12 Ni iron based alloys has been studied. It was shown that at elevated temperatures in the range of 1112 to 1472 F, for a wide range of stress amplitudes, decreasing the carbon level from 0.05 to 0.004 weight percent causes a much shorter fatigue l if e and lower endurance l imit. This was attributed to the beneficial ef f ect of carbides in ef f ectively blocking grain boundary sliding and migration, thereby inhibiting the nucleation of grain boundary cracks. At room tempera-ture, however, it has been shown that for Type 316 stainless steel fatigued at strain amplitudes in the range i 1 to i 4 percent, the presence of 2 to 3 volume percent of chromium carbide can decrease the fatigue life to about on one-third of that in solution-treated material. This was attributed to 5.3-18 Amend. 62 Nov. 1981

I the brittle fracture of the grain-boundary carbides. A comprehensive research program is underway on investigating the effect of interstitial i O, transfer on the fatigue of LMFBR materials. This includes work at Westinghouse and Argonne flational Laboratory on interstitial loss effects in Types 304 and 316 stainless steel, and a General Electric Company effort on the effects of carburization in both austenitic and ferritic alloys. The Westinghouse and Argonne work will be mainly at elevated test temperatures whereas the latter program will be focused on lower temperatures (<1000*F) where carburization of austenitic stainless steel prevails. 1 No reduction of fatigue properties due to carbide precipitation is expected in either Type 304 or Type 316 austenitic stainless steel at the operating temperatures of ghe CRBR. Various investigators have shown that at temperatures above 750 F there is either no effect or a beneficial effect due to the precipitation of carbides on the fatigue properties of either Type 304 austenitic stainless steel. Driver (Ref. 5.3-42) has showg that a 0.05C austenitic stainless steel, heat treated 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 1292 F to produce M23C6 carbides, had a higher fatigue strength than ag 0.004C austenitic stainless steel. Driver's results for tests at 1292 F are shown in Figure 5.3-39. Kanazawa and Yoshida (Ref. 5.3-43) have also observed increased fatigue strength of both Tyge 304 and fype 316 which they attribute to aging at temperatures of 750 F and 1110 F during fatigue testing. Cheng et al (Ref. 5.3-44) have shown that aging both Types 304 U

and 316 stainless gteel at 1050 F (for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />) had a beneficial effect on the 1050 F fatigue life. For material aged at 1200 F (for 160 p hours) there was no significant difference in fatigue life over that of Q annegledmaterial. Brinkman et al (Ref. 5.3 45) 1100 F fatigue properties after aging at 750 3F or 1110 F.

have ghown improved In terms of high temperature fatigue crack propagation rate, the effect of carbides in Type 304 and Type 316 appears to be negligible or slightly beneficial . James and 0 Knecht (Ref. 5.3-46) found no effect of aging for 1150 hours0.0133 days <br />0.319 hours <br />0.0019 weeks <br />4.37575e-4 months <br /> at 891 F on the crack paopagation rate in Type 304 or Type 316. Mahoney and Patton (Ref. 5.3-47) have found that crack g

propagation rates at 1200 F v;ere not dependent on efther carbon content or carbide morphology af ter aging for 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> at 1300 F. James (Ref. 5.3-6) has shown a decrease in the crack growtg rate of goth Type 304 and Type 316 stainless steel. Material aged at 1000 F or 1200 F had crack growth rates lower than those of the annealed material as shown in Figures 40 and 41.

The improvement in the elevated temperature fatigue properties of aged Type 304 and 316 stainless steel has been attributed to carbides effec-tively blocking grain boundary sliding and migration and to a blunting effect when a crack encounters a carbide particle. (Ref.5.3-42).

Although the above results are based on short time aging data, relative to CRBRP operating life, the results 6re considered represen-tativc of those that will be observed in CRBRP. Weiss and Stickler (Ref.

5.3-48) have shown the precipitation of carbides in Type 316 to follow the time-temperature curve given in Figure 5.3-42. Precipitation of the , 25 0 5. 3-18a Amend. 25 Aug. 1976

carbicos in Type 304 would f ollow a similar curve). It is apparent f ran this curve that the carbido precipitates will be present for long aging times. This has been conf Irmed by data of Mochal et al (Ref. 5.3-49). They observed that carbides considered to precipitate in 316 SS throughout the duration of a 5 year test program which included approximately 30,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 1225 F.

Although these daia indicate that carbide precipitation will occur over a very long period in CRBRP, the effect on the fatigue properties is expected to be more immediate. James (Ref. 5.3-6) has shown that increasing the aging time f rom 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> to 6000 hours0.0694 days <br />1.667 hours <br />0.00992 weeks <br />0.00228 months <br /> did not signif icantly change the aging ef f ect.

As notea above it has been shown that there may be a reduction in the room l temperature f atigue properties of aging austenitic stainless stenis. Barnby end Pearce (Ref. 5.3-50) observed a 1/3 reduction in the f atigue lif e of Type 316 stainless steel which they attributed to the presence of 2-3 volume percent carbices.* However, Mahoney ano Patton (Ref. 5.3-47) did not observe any signif icant dif ference in f atiguo crack propagation rates in Type tanperature crack propagation rates in Type 304 stainless steel af ter aging f or 3000 hours0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> at 1200 F.

O

• Test temperature not stated buT assumed to be roan temperature.

O 5.3-18b Amend. 62 Nov. 1981

Both PHTS and IHTS pumps require a shaf t seal to ef f ect a zero leak seal from O- cover gas to atmosphere. This shaf t seal, shown schematically in Figure 5.3-14A, is an oil lubricated, double rul;bing face seal. The seal has e shaf t driven internal oil circulator and an Integral air to oil heat exchanger, with of I supply to make up for of I leakage past the rubbing faces. 01l leakage f rom the seal assembly into the sodium coolant is prevented by +wo barriers.

The fIrst barrier is an oil dam approximately 1.2 inches above the lower f ace seal. The normal leakage f rcrn the lower f ace seal is diverted by this oil dam into the oil leakage drain passage into the lower seal leakage collection l reservoir. A second barrier is the collar above the drop down seal located just above the purge labyrinth. This collar extends beyond and over the labyrinth, thereby shunting any oil to a drain plenum. For oil to penetrate into the sodium, three things must happen:

o Fail ure of the oil dam o Failure of the collar to divert oil l

o Overflow of the plenum drainage over the drop down seal lip A positive pressure is maintained in the shaf t seal oil at all times by means l of an oil supply tank which will be pressurized abcve the loop operating pressure. The oil feed line to the seal will be oriented to preclude seal drainage in the event of a Iine break. The seal is capable of many hours of operation on the sel f contained fluid.

The oil system supporting the shaf t seal contains three tanks, each of which will have a level probe, thereby permitting monitoring of total oil inventory, its location, and permitting calculation of seal leak rate. The lower seal leakage collection tank is sized to hold the entire system's oil inventory of approximately 41 gallons.

O 5.3-21 Amend. 62 Nov. 1981

Oil vapors which may potentially be drawn f rcm the lower seal leakage collection tank into the tank ullage during draw down (pump speed up) are retarded f rcrn such passage by treans of a split flow purge gas f eed of recycled argon into the purge labyrinth. This gas f eed sp!Its and flows up and down the shaf t f rcrn the f eedpoint. This gas input is ficw controlled at the inlet, and l f low controlled at the discharge f rcrn the lower seal leakage collection tank.

If reed pressure into the tank is detected to be Icw (by the gas f eed systern) the discharge of gas f rcm the tank will be closed. in event of gas line l rupture at the oil tank discharge, the orificing by the line will retard loss of cover gas pressure. The gas f eed to the purge labyrinth w ill be made f rcan two supply lines, with an automatic switchover to auxiliary gas, so as to provice dependable gas supply during and af ter SSE events.

Radioactive vapors f rcm the tank ullage are prevented f rcrn escape to the atmosphere by the two barriers consisting of the gas downf icw at the purge labyrinth and the oil lubricated double shaft seal. Radioactive purging !s continuous by means of the bubbling in the standpipe, which is connected to RAPS.

O O

5.3-21a knend. 62 Nov. 1981

classified as faulted for the affected steam generator module. Differential pressure between the primary and Intermediate sides during this event is conservatively evaluaied bv assuming that the primary side pressure is that resulting from pony motor speed (approximately 6 psig). For the rest of the loop, the occurrence is classified as an emergency event.

For the unaffected loops, the event is similar to the reactor trip from full power. Decay heat removal is maintained throughout the two remaining loops.

The transient responses of temperature, flow and pressure on both the primary and intermediate side of the IHX in the affected loop are presented in Figures 5.3-18A through 5.3-18G. Particular attention is directed to Figures 5.3-18F and G which show intermediate side short term and long term pressure effects.

In evaluating the structural adequacy of the IHX, with respect to the check valve slam, the dynamic nature of the primary sodium pressure history is being accounted for by using dynamic load factors. The factor will be applied to the maximum primary pressure, which in turn, is used to determine the pressure-Induced primary stresses. These primars stresses are Iimited by the emergency condition allowables of Code Case 1592, Paragraph 3224, as modified by RDT F9-4T. The fatigue damage associated with the cyclic nature of the pressure l history is accounted for per Paragraph T-1400 of Code Case 1592. The description of the pressure pulses for the sodium-water reaction and check valve closure is included in the equipment specification. The curves define the amplitudes, duration and number of cycles.

Rapid check valve closure can only occur as a result of primary pump mechanical j failure. The event involves a postulated instantaneous stoppage of the Impeller of one primary pump, while the system is operating at 100% power. The failure may be a seizure or breakage of the shaft or impeller. Primary system sodium flow in the affected loop decreases rapidly to zero as the pumps in the unaffected loops seat the check valve (thereby causing a rapid check valvo closure or stam). A reactor trip will be initiated by the primary intermediate flow ratio subsystem. Sodium flow in the intermediate circuit of the affected loop decays as in a reactor trip from full power, modified by changes in natural circulation head. The event is characterized by a down transient in the hot leg of the Intermediate circuit of the affected loop. The transient responses of temperature, flow and pressure on both the primary and intermediate side of the IHX in the affected loop are presented in Figures 5.3-18H through 5.3-18M. Particular attention is directed to Figure 5.3-18J which shows primary pressure ef f ects.

Both the sodium water reaction and check valve closure events are classif ied as emergency events for the IHX. As such, the IHX designer is required to determine which of the six emergency events is most severe to the IHX. The selected event is then appiled with a periodicity of two consecutive occurrences during the first three years of operation, and thereafter five times over the remaining 27 years (or once every six year period). If vendor analysis indicate either as the most severe event, the occurrence of the two consecutive events wIII be moved to the most stringent time in the Iif e for the event to occur. Preliminary analysis indicates that damage from either of these events wIll be insignificant.

5.3-39b Amend. 62 Nov. 1981

, .w.s,p. . . ,y _ . - _

1 l

l Pumo Inelastic analyses of the pumps wcs required to demonstrate coafarmance with O

59 27 the ASME Code. Paragraph 4 of RDT Standard F9-5T, Sept. 1974 gives a description of acceptabIe methods for time-Independent eIastIc-p!astIc analysis and time-dependent creep analysis. Scme of the computer programs 11sted above have !nalastic capabilities, and will be used wherc applicable.

For the purposes of loads and cnalysis the pump R-Spec divides the pump into four areas. These are: Subcomponent I which consists of the pump tank, Subcomponent 2 which is the upper inner structure including the pressure bulkhead, Subccmponent 3 which is ihe rotating machinery and Subcomponent 4 which is the static hydraulics.

Suucomponent 1 is designed to the ASME Boi1er and Pressurc Vesse; Code Section Ill, Subsection NB Class 1 and Code Case 1592 where applicable. The cone and cylinder are designed mainly by dynamic stiffness requirements. These include seismic loads and the necessity of keeping the natural frequency of the structure above the operating speed of the impeller. SAP IV and the "NASTRAN" 59 ccmputer codes were used for this analysis. The analysis has been qualified by comparing ihe results of one analysis against the other. The sphere sealing ring and cone-sphere support ring are designed by sealing ring leakage which requires elastic response during rormal and upset conditions. A failure will reduce pump efficiency below plant criteria. These areas have been analyzed by 3D global analysis using NASTRAN. 1,e nozzles are designed by pressure, pipe 59 nozzle Iceds, ard thermal Transients. The failure modes associated are creep and creep fatigue. 2D elestic analysis s required. The design is being made with sufficient space for thermal baf fles and liners to keep it elastic as much as is possible. But it may be necessary to qualify it using simple inelastic analysis. Hydraulic leakage test data has been ob+ained which determined the 51 l relation of sealing gap to leakage rate.

59 l Subcomponent 2 conf orms to the same Code requirements as Subcomponent 1. The upper closure piate and radiation shield are designed by the design pressure and temperature requirements. Elastic f ailure is the predominant rrode. The heat shield has sieady state thermal gradients which are determined by a 2D 59 axisymmetric model and stresses are calculated with a 2D stress model. The motor stand has been designed by the stiffness requirements of the motor and seismic loads. The principle falIure mode is excess vibration ieading to fatigue falIures.

59 Subcomponent 3 can be removed and inspected arter an emergency or f aultea event and repaired before the plant is placed in service again. Therefore, this section was designed and analyzed to the ASPE Boller and Pressure-Vessel Code, Section ill, Subsection NB for Clasc 1 Components and Code Case 1592 where applicable. However f or mergency m p's Code Case 1592 is used and the design 59 rules for lead controlled stressr . ac" : on 3227 ) applies. Strain deformation and fatigue analysis need only c -f , w ed up to the emergency event and the limits will epply only to t' u o ollit" to operate at pony motor speed after the event. This area o M fgned by critical frequency gl req u i rement s, inertial loads, ton ea o'd thermal transients, it was analyzed with a 2D axisymmetric modcl. The loads. ~3used by bearing misalignment were accounted f or. A general 1/2 scale model hydraulic performance test was run 26 using water as the pumped flu!d. This test provided Information on the pump 59 5.3-39c Amend. 59 Dec. 1980

cladding. For the Intermediate heat exchanger, a recent stuuy indicates that O in the hottest regions of the Intermediate heat exchanger the carbon level will fall belt, the minimum permissible value of 0.04 weight percent during service.

A wall thickness allowance was used to counter resulting strength loss (see Section 5,3.3.5).

Of the heat transport system components only the reactor vessel, vessel head, and guard vessel are likely to experience significant irradiation. Current work will evaluate the effect of neutron dose and spectra on mechanical degradation of these components and sufficient shielding will be used to ensure that the ilfetime neutron fluence is maintained with acceptable limits.

1 0

l

! 5.3-70b l Amend. 62 Nov. 1981 f

l

l

References:

$ t

42. J. H. Driver, The Effect of Boundary Precipitates on the High-

Temperature Fatigue Strength of Austenitic Stainless Steels,

' Metal Science Journal, Vol. 5, March 1971, p. 47.

43. K. Kanazawa and S. Yoshida, High-lemperature Rotating Beveling 4

Fatigue Behavior of the Austenitic Stainless Steels, SUS 304-B and j 316-B, Trans. Nat. Research Inst. Metals (Japan), June 1974, i 16 (3) 90-98.

t 44 C. F. Cheng, C. Y. Cheng, D. R. Diercks, and R. W. Weeks, Low-Cycle Fatigue Behavior of Types 304 and 316 Stainless Steel at LMFBR Operating Temperature, Fatigue at Elevated Temperatures, STP 520, i August 1973, p. 355-364.

1 i 45. C. R. Brinkman, G. E. Korth and R. R. Hobbins, Estinates of l Creep-Fatigue Interaction in Irradiated Austenitic Stainless Steels, i Nuclear Technology, Vol .16, October 1972, p. 297.

46. L. A. James and R. L. Knecht, FatiSue-Crack Propagation in Fast- l Neutron-Irradiated Types 304 and 316 Stainless Steels, Nuclear i Technology, Vol. 19, Sept. 1973, p. 148.

47. M. W. Mahoney and N. E. Patton, The Effect of Carbide Precipitation on Fatigue Crack Propagation in Type 316 Stainless Steel, Nuclear Technology, Vol. 23, July 1974, p. 53.

l

48. B. Weiss and R. Stickler, " Phase Instabilities During High

' Temperature Exposure of 116 Austenitic Stainless Steel," Met.

Trans., 3 April 1972, p. 651-866. ,

49. N. L. Mochel, C. W. Ab tr.an, G. C. Wiedersum and R. H. Zong, 4 Performance of Type 316 Stairdess Steel Piping at 5000 psi and  !

1200 F, Vol . XXVIII, Prcwiirigs of the American Power Conference, j

. 1966.

i

! 50. J. T. Barnby a,J F. M. Peace, The Effect of Carbides on the High ,

l Strain Fatigue Resistance of an Austenitic Steel; Acta Metallurgica, l

Vol . 19, Dec.1971, p. 1351. ,

51. T. 1. Daly and G. G. Elder, " Final Report of Clinch River Breeder Reactor i Pinr.t Piping and Valve Technical Supoort Team", Westinghouse Advanced l Reactors Division, Madison, Pa. , WAh0-D-0051. September 27, 1974.

i

52. Deleted

53. Deleted 56 25 iO l tmend. 56 i

i 5.3-75b Aug. 1980

+

References:

54. P. Soo, " Selection of Coolant-Boundary Materials for the Clinch River Breeder Reactor P1 ant", WARD-D-0010, August 1974.

55. S.A. Shiels, S.L. Schrock, and L.L. France, " interstitial Transfer Program impact Assessment Report, Part iI - Topical Report", WARD-NA-3045-3, Septe-N' 1973.

56. G.V. : , "An Evaluation of Yield, Tensile, Creep and dupture Strength of Wrough: 304, 316, 321, and 347 Stainless Steels at Elevered Temperatures",

ASTM Data Series Publication DS 5S2, American Society for Testing and Materials, February 1969.

57. P. Soo (Compiler), " Analysis of Structural Materials for LMFBR Cooient-Boundary Components - Materials Property Evaluations", WARD-3045T3-5, November 1972.

j 58. Deleted

59. R. A. Leasure, "Effect of Corbon and Nitrogen and Sodium Environment cn the Mcchanical Properties of AusTonetic Stainless Steels", WARD-D-0277, November 1979.

O l

l l

l O

5.3-75c Amend. 62 Nov. 1981

. O O O i

TABLE 5.3-11 S (3 x 105h) AND SmVALUES FOR AUSTENITIC STAINLESS STEELS

• S t (ksi) Sm (ksi)

TEMPERATURE 304 and 316 and 304 L 304 and 316 L 316 and (oF) 304 L 304 H 316 L 316 H (SA-182) 304 H (SA-182) 316 H 100 N/A N/A N/A N/A 16.6 20.0 16.6 20.0 200 N/A N/A N/A N/A 16.6 20.0 16.6 20.0 300 N/A N/A N/A N/A 16.6 70.0 16.6 20.0 400 N/A N/A N/A N/A 15.7 18.7 15.5 19.2 P 500 N/A N/A N/A N/A 14.7 17.4 14.4 17.9 w

g 600 N/A N/A N/A h/A 13.9 16.4 13.5 17.0 cn 700 N/A N/A N/A N/A 13.4 15.9 12.8 16.3 800 N/A 20.4 N/A 20.8 13.0 15.1 12.315.8 900 N/A 16.0 h/A 19.3 N/A 14.6" N/A 15.68*

1000 N/A 9.3 N/A 14.0 N/A 14.08 N/A 15.4" 1100 N/A 5.7 N/A 7.8 N/A 13.2e' N/A 14.8H 1200 N/A 3.4 N/A 4.5 N/A 12.7 N/A 14.6 "

' Data are f rom ASME Boller and Pressure Vessel code, Section lil,1974 Edition with Addenda through Summer 1975 and ASE Code Case 1592-7, and apply to Class 1 components. For other aditf or's of the Code, these values may vary somewhat.

• • For Type 304 and Type 316 grades only.

N/A = Not applicable.

5$

<g W

TABLE 5.3-12 PRIMARY HEAT TRANSPORT SYSTEM CODE AND SEISMIC CATEGORY MATRIX F9-4T Req'ts Primary RDT For NucIear System Component ASME Code Seismic Components At Comoolent Standardsl Section/ Class Categorv Elevated Temo.

Primary Pump E3-2T June 1974 111/1 i Note 2 IHX E4-6T April 1975 lil/1 i Note 2 Circk Valve El-18T May 1975 Ii1/1 i Note 2 F:ow Meter C4-5T April 1974 111/1 I 27 . G ard Vessels E10-2T July 1973 Ill/2 i Note 3 Piping & N/A 111/1 I Note 2 Fittings Pipe Hangers Supports a E7-0T ay 1972 III/ I Snubbers Subsection NF Thermal Insul. N/A N/A lil* -

Trace Heating N/A N/A I Notes:

1. Component standards used as gu! dance in equipment specification preparation, not invoked.

2. This standard, mod! fled only as Indicated in the equipment specification, is to be applied in its entirety to all structures in the component.

3. Constructed to rules of Class 1 but not hydrostatically tested or code stamped. An elevated temperature supplcrnent to the equipment specification, equivalent to RDT Standard F9-4 with modifications and code cases, will be used.

• Thermal Insul ation Is f unctlonal ly sei smic Category i I i, however, it is designed to the requirements of Seismic Category I, utilizing static analysis 59 rather than dynamic analysis.

O f, mend. 5^)

5.3-87 rec. 1980

--.-. - __1__.-_.-. .-. .. . . . . .. -. _ . _ . - _ -.

I 1

1 400 O 4 2'c 360 -

MI (C'M og' *gt.01 ,.

320 -

E E 280 -

z e-CD E

g 240 -

t; w

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,15HQ + .0313 (C+N) i 3

@ 160 -

s-i 649'c 120 -

og ,gg,g3 + .0331tt+NI 80 -

M

, ,g 36.20 + .0324 (C+N1 40 -

= 14.07 + .0166 (C+N) - 816'C oR I I i 0

400 600 800 100 12 0 14 0 1600 i

C+N CONTENT ppm

.t

Figure 5.3-3. Effect of C+N Content on the Rupture Strcngth

! of Type 304 SS (Taken from Ref. 59)

' ~

O 5.3-100 Amend. 62 Nov. 1981

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

O 320 28 0 -

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ga (C*"I 3 . \ob S' * ~g a- O B

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, pp . 0255 (C.N) e 40 -

,,g = 18 48 + .0120 (C+ NI - 816 C

' I  ! l 0

0 200 400 600 800 1000 1200 1400 1600 C+N CONTENT ppm Figure 5.3-4. Effect of C+N on the Rupture Strength of Type 316 SS (Taken from Ref. 59) 4425-2 Arrend. 62 5.3-101 Nov. 1981

O

.08 f

.07 -

i s i e d

.06 O TYPE 304 SS

7 -

e TYPE 316 SS 2

z O &

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! 500 f 700 800 900 i.'

, TEMPERATilRE C 1,

i, Effect of C+N Content on the 10 3Hour Rupture 4

, Figure 5.3-5.

! Strength of Types 304 and 316 SS as a Function 447. ;

of Temperature (Taken from Ref. 59) 5.3-102 Amend. 62 1 Nov. 1981 i

l i

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1400 ppm .0375 - 9.65 x 10 5 (y) 1000pp -

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1200 ppm 0372 - 9.5 x to 5 (y) -

4

• 05 10c0 ppm .0437 - 10.9 x 10 5 (T) 000 ppm --

i. $

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600 ppm .0520 - 14.35 x 10 5 (7) See ope ~

{ .06 - -

i j i i i i i i i I e I i i i 1 500 600 700 8 00 g al TEMPE R ATU RE (DC)

I f@1 m e, q

t

- ro Figure 5.3-7. Change in Rupture Strength of Type 316 SS Due to Chanae in C+N Content (Taken from Ref. 591 l

i i

b

.--, , G

O FIGURE 5.3-8 HAS BEEN DELETED O

i l

l 5.3-105 Amend. 62 Nov. 1981 i

l ARGON f S ')PPL Y 9  !

OIL SUPPLY

\ TANK L~ - ~ =

' UPPER SE AL EM" Oll"'

UPPER LEAKAGE W SEAL {' <

',,,7 << g- AIR q COLLECTICN l DR AIN RUBBING / ,,

TANK FACE l [ y

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.l i

', f ,Q'

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L/ f OIL AND G AS' y _l'il'l'. /6, f f DRCP DOWN SEAL ,

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PUMP COVE A ,- 'l GAS ,

81 31341 Figure 5.3-14A. Schematic of Typical Seal

" Oil System f:1end. 62 Nov. 1981 5.3-111a

Design loading used for flexit lity and seismic analysis for the determination of adequate piping supports wl.i include all expected transient loading p)s

(_, conditions. Spring-type supports will be provided for the initial dead weight loading during hydrostatic testing of steam systems to prevent damage to piping supports.

Test and insoection in-service inspection is considered in the design of the main steamwater and feedwater supply piping. This consideration assures adequate work!ng spece and access for the inspection of selected pipe segments.

After completion of the installation of a support system, all hanger elements will be visually examined to assure that 1..ey are correctly adjusted to tnear cold setting position. Upon hot start-up operations, thermal growth will be observed to confirm that spring-type hangers are functioning properly. Final adjustment capability will be provided for all hanger or support types.

5.5.2.3.4 Steam Generator Module The steam generator module shown in Figure 5.5-2 is a shell and tube heat exchanger with fixed tubesheets. Flow is counter-current, with sodium on tne shell side and water / steam on the tube side. The evaporator modules transfer heat from the sodium and generate 50 percent quality steam from the subcooled recirculation water. The steam-water mixture exiting from the evaporator is separated into saturated water and caturate* steam in a steam drum. The e- superheater modules transfer heat from the socium to superheat the saturated

(,s) steam to the temperature required for admission to the turbine.

The Atomics internatlonal - Modular Steam Generator (MSG) was a 32.1 Mwt maximum power, hockey stick designed unit used as the basis for the CRBRP Steam Generator design. The salient features of the MSG unit are as follows:

. Maximum Power 32.1 Mwt

. Temperature 930 F

. Pressure 2530 psig

. Startup/ Shutdown 37 Cycles

. Tube Design 158 Tu';es 5/8 in. 0.D. x 139 mil. wall

. Length 66 ft

. Material 100% Ferritic Steel - 21/4 Cr-1 Mo For further details see Reference 4.

Evaporator and superheater modules are identical in all respects except for the l

Inlet orifices that may be added to the cvaporator tubes at the lower tubesheet to increase the evaporator water flow stability margin. Each module consists of a 53 1/2 inch 0.D. shell containing a tube bundle with 757 5/8 Ir. 0.D. x 0.109-inchwall tubes. The design employs O 5.5-9 Amend. 62 Nov. 1981

autogeneous buttwelded, tube to tubesheet joints. The shell and tube material is 2-1/4 Cr-1 Mo steel. There are two evaporator modules 41} and one superheater module per loop. The evaporator modules operate with a recirculating ratio of 2:1.

The steam generator design requires that each loop (two evaporators and one suoerheater) develop 325 MWT at rated full load. The design life 41l of the steam generator module is 30 years with items that cannot be reason-ably expected to last 30 years being replaceabic during in-service inspec-tion periods. The steam generator is designed to withstand the normal, j upset, emergency and faulted operating conditions in accordance with I CRBRP Criterion 26. The steam generator module is also designed to with-stand the loading combinations indicated in Section 3.9.2.2.

41l The materials used in design of the steam generator module major l components are as follow.  !

Pressure Boundary (2-( CR-1 Mo-Ref. 3, Vol .1, Section 2.2)-

Shell Forgings - SA 336. Class F22A  !

Shell Plate - SA 387, Grade L2, Class 1 t 41 Tubesheet - ADT M2-19 with Stional provision 2 and Code Case  :

Forgings 1557-2 Tubing - RDT M3-33 as modified to limit silicon and carbon content for weldability and carburization consid- i erations (Reference 5). I Steam i ead Studs and b..tc (Irao 718-Ref. 3, Vol .1, Section 2.5):

i O

Steam Head Studs and Nuts - RDT M2-15, Grade 718 l

The steam generator nodule supplier will provide procedures for welding and heat treating in accordance with the requirements specified i 41 I in the Code as modified by RDT E15-2NB (see Section 5.5.1.2). Since the tube to tubesheet weld is the most critical joint in the steam gen u ator ,

module, this weld is being supported by a separate development program.  ;

Welding qualification is cortrolled by the Code as modified by RDT F6-5 l 41 I and RDT E15-2NB (see Section d.5.1.2).

Material integrity prior to placing the steam generator, in service will be assured by complying with the ASMZ Code Section III which requires weld radiography, tubing ultrasonic testing, plate ultrasonic testing, tubing hydraulic testing, component pressure testing and helium leak testing.

Material considerations are indicated in Sections 5.5.1.4 and 5.5.3.11. Section 5.5.3.1.5 indicates the tests being conducted to i support thc steam generator design. It is not anticipated that back-up '30 matecials will be required.

5. 5- 9a O

Amend. 41 Oct. 1977

The following is a description of the steam generator modules as shown in

/^% Figure 5.5-1.

() a. Tube-Bundle Arrangement The tube size was selected as 5/8 inch OD x 0.109 inch wall tubing based on optimization studies performed as part of the prototype design program. The tubes are positioned on a 1.22 inch triangular pitch. The effective tube hear transfer length is 46 feet. The tubes are supported along their length by Inconel 718 tube spacer plates in the active heat transfer regioi and by inconel 718 tube support bars in the elbow region. Axial span between the tube spacers and the tube support bars were determined by anal f sis and tests to keep the tube vibration levels significantly below acceptable levels.

The tube bundle is enclosed in a 36.5-inch I.D. shroud along the active hear transfer length to insure proper flow distribution. An elbow shroud is provided in the hockey stick region to protect the outer shell from the direct effects of wastage in the event of a water / steam to sodium leak. Sodium flow enters the module through the uprer sod!um nozzle, and turns up into the flow distributing annulus. The sodium toen enters the tube bundle over the top of the shroud. The sodium then flows into and down through the tube bundle. At the lower end of the module, the sodium exits the tube bundle below the shroud, turns up and flows through the annuius, and flows out the exit nozzle (s).

b. Other Internals p Six 1/2-inch thick thermal baffle plates cover the face of the lower tubesheet, t 'j These baffles serve several purposes: 1) they protect the lower tubesheet from sodium outlet temperature transients, 2) they partially fill the volume below the level of the 6-inch drain nozzle, and 3) they act as wastage baffles in the, area of the tube to tube-sheet welds. The plates are attached to the sodium l side of the tubesheet by six studs. The baftle plates are gang drilled to reduce relative tube hole misalignments.

The stud hole clearance is cmaller than the tube hole clearance by a factor of two. This condition allows lateral motion of the baftle to occur without placing loads upon the tubes.

The studs also place a bearinp !oad on the baftie plates. The bearing Ioad resists lateral motion of tho Loftle plates.

Clearance between the tube and baffle tube hole is .062 inch. The area is large enough to vent the proacts of a sodium-water reaction.

O a

5.5-10 Amend. 62 Nov. 1981

The effects of sodium-water reaction to the wastage baftles have been tested (Reference 17). One of the primary objectives of the test series was to verity that the reaction products escape from the narrow annulus between the plates anc tubes, 0.062 inch around. The report concludes, "these is a clear indication of flow away from the leak site and no severe building of solid reaction products occurred in the gap or the outside of the plates".

At the upper tubesheet, wastage baffles are also provided. The thermal wastage baffle consists of four (4) Inch long hexagonal sleeves over each tube which are retained by bars spanning the tube bundle. The bars are attached to a retaining ring. The upper wastage baftle serves the same purpoce as the lower wastage baffle.

l The Alloy 718 Inconel plate and seals in the annular region serve to protect the hockey stick region from thermal eddies induced by the sodium inlet flow.

This arrangement also provides an adequate rollef path for sodium-water reaction pressure that would occur in the hockey stick region in the unlikely event of a large tube rupture. An Alloy 718 inconel plates seal the exit annulus at the lower end of the active region.

The support ring-shroud joint serves to: 1) support the shroud within the shell, and 2) prevent flow from bypassing the tube bundle by blocking oft the annular stagnant gap between the shroud and the she!! at the upper end of the outlet flow annulus.

O 1

0 5.5-10a Amend. 62 Nov. 1981

An upper header thermal liner and an inlet nozzle thermal liner are proviced to

/'" mitigate the effects of system sodium transients.

(_h/

c. Shell Arrangement (1) Major Comoonents of Shell The shell connects to an upper and lower tubesheet, and consists of two reducers, an elbow, an inlet header " tee" section, an outlet header " cross" l section, a main support section and a main shell section. These components have been sized structurally to contain postulated maximum large leak SWR conditions as well as meet design operating conditions.

(2) Shell Penetrations Each superheater and evaporator module is fitted with one inlet sodium nozzle and two outlet sodium nozzles. Present intermediate sodium loop arrangement drawings show both superheater outlet nozzles being used, while only one of the two outlet nozzles is used on each of the two evaporator units. The spare evaporator exit nozzles are capped. The inlet sodium nozzle is a 30-Inch nozzle that attaches to the 4 1/4-inch thick Inlet sodium header in the direction of the hockey stick. The 30-inch nozzle is reduced to a 26-inch, 1-inch thick wall pipe, which will be mated to the loop piping. The two outlet sodium nozzles are 22-inch nozzles that attach at 900 to the direction of the hockey stick to the 4 1/4-inch thick outlet sodium header. The 22-inch nozzles reduce to 18-inch, schedule-60 pipes, which will be mated to the loop piping.

f-~~ The purpose of the oversized nozzles in regcrd to the piping size is to provide

\ space in the nozzles for thermal liners and to reduce ficw velocities in the inlet / outlet regions.

Two 8-inch sweepolets are attached to the reducers located at both tubesheets.

These serve as ports to inspect the final closure welds. Also, one of the ports on the lower reducer is attached to a 6-inch schedule-80 pipe by a transition section to provide for rapid drainage of the lower stagnant end of the modules, should it be required. Again, the purpose of the transition section is to provide for possible lining of the nozzles. A one-inch drain is also provided through the lower tubesheet to drain the lower thermal baftle region. A three-inch sodium bleed vent is provided in the hockey stick end of the module to provide for: 1) venting during initial filling of the shell side, and 2) a small sampling flow to a hydrogen detector to allow detection of any small leak in that region during operation.

(3) External Attachments Bolt-on spool pieces provide steam heads at both ends of the module. Bolted connections are employed to facilitate access to the faces of tubesheets O

5.5-11 Amend. 62 Nov. 1981

for easy access for maintenance and in-service inspection. The spool piece consists of a steam head, straight piece of pipe, and a flange &

to mate with the steam piping. A flexitallic gasket is employed to W effect a seal between the spool piece and the tubesheet.

Steam Generator Inspection Access to the heat transfer tubes of the steam generator is readily obtained by removal of a piping spool-piece and the steam head which is bolted to each end of the steam generator module. The steam head is 72 inches long and is attached to the module by 24 studs which are threaded into the tubesheet. The steam head is also bolted to a flanged pipe spool piece en the onposite end. When the steam head is removed, the outer face of the tubesheet is exposed for inspection. Furthermore , the i

i I piping spool-piece car, be removed to give additional clearance beyond the 41 72 inch opening provided by rcmoval of the steam head.

l The inner diameter of the heat transfer tube is readily available for inspection by ultrasonics, eddy current and/or other suitable means which will be determined acceptable at the conclusion of a development program (now in progress). The outer surface of the heat transfer tubes cannot be readily inspected since the'shell of the steam generator is a fully welded assembly. However, it is expected that the above tube in-spection techniques will give sufficient information on the condition of the tubes to provide assurance of integrity of the sodium / water boundary.

1 5.5.2.3.5 Steam Drum The steam drum, shown in Figure 5.5-4, is a horizontally mounted 82 inch 0.D. , 35 f t. long cylinder with hemispherical heads (42 ft. overall g

length). Most of the major nozzles are located in a vertical plane through the steam drum centerline. These consists of one 12 inch steam 41 outlet nozzle located at vessel midpoint and directed vertically upward, two 16 inch riser nozzles (evaporator return) located at approximately cylinder cuarter points and directed downward, four 10 inch dnwncomer nozzles (recirculation pump suction) spaced evenly along the cylinder and directed downward, one 6 inch continuous drain ngzzle located in one head and directed downward normal to the head at a 45 angle to the vertical, and one 10 inch feedwater inlet nozzle Incated in the opposite head and ,

directed downward normal to the head at a 450 angle to the vertical. l The only nozzle that is not copienar with the vessel centerline is the auxiliary feedwater nazzle. This is a 4 inch nozzle located on the same 0 head as the main feedwater inlet nozzle in a vertical plane rotated 45 28 Pam the vessel centerline; the nozzle is directed downward normal to the head at a 45 0 angle to the vertical.

O Amend. 41 Oct. 1977 5.5-ila

O The natural frequency of the steam tubes in air and water (d e e The damping rate of the tubes in air and water e Tube, support plate, shroud, liner, baffle and shell response to prototypic flow-induced forces e Vortex shedding frequencies in the cross flow region Flow distribution tests were conducted for sodium flow rates from 10%

to 100% of rated flow for the superheater and at 40 and 100% for the evaporator mode. Pressure measurements were made using two and three dimensional pitot probes; approximately 30 penetrations were made in the model. Flow measurements were made with a magnetic flowmeter having a calibrated accuracy of + 260 gpm.

Schedule of Tests Testing with the HTM was conducted between July 1975 and June 1976.

Summary of Results Vibration tests indicated that the vibration levels are small and that excessive stress levels should not occur. Preliminary analysis determined the maximum vibration induced tube stress to be about 2000 psi (13.8 FT MD). This would occur in the crossflow region. The maximum tube U peak-to-peak displacement amplitude was measured to be 6 mils and The corresponding tube stress was occurred in the static region.

determined to be 400 psi by preliminary analysis. Vortex shedding was found to be not a dominant source of tube excitation.

The flow tests showed that the flow became uniform at an L/D of 21 (109 cm below the bottom of the inlet window) and remained uniform until about 15 cm above the outlet window. Relatively strong mixing occurred between the main body flow and the lower stagnant region.

However, this condition did not appear to persist in the region of the tubesheet where near-stagnation conditions are desirable.

b. Large Leak Tests Objective The objective of the large leak tests is to support establishment of adequate design and operational methods to accommodate large sodium-l water reacticns within the steam generator system of a LMFBR.

41 The large leak tests will provide data in support of efforts to validate interim and advanced computer codes for large leak SWR 36' analysis programs.

O k._)

Amend. 41 5.5-18 b Oct. 1977

Data will be obtained from the large leak tests for assessing the potential of secondary tube failures.

Data will also be obtained on relief system performance and on cleanup and recovery techniques employed to return the system to operation following the majo tube leaks.

Program Descriotion The tests wlit utilize, as a test article, the Atomics International modular steam generator (MSG) thermal hydraulic model previously tested in the sodium components test installation ($CTI) at Santa Susana, California.

The model has been converted to a large leak test article by replacing selected tubes with tubes designed foi controlled rupture upon signal command. The large leak tests will be conducted in the Large Leak Test Rig l (LLTR), located at the Energy Technology Engineering Center (ETEC) at Santa Susana, California. The water injection system of the LLTR provides water under the desired conditions for the large leak injectico device (LLID).

The LLID for these tests consists of a cylinder with a pneumatic piston which applles an axial load to the circumferentially weakened tube. The gas pressure is applied to the piston, the tube is pulled apart at the weakened spot, creating a guillotine type failure.

The MSG test article is an approximately 70' Lg. x 16" 1.D. hockey stick steam generator containing 158 steam tubes each of which is 0.625 inches 0.D. with a wall thickness of 0.109 inches. The tube material is 2-1/2 Cr-1 Mo steel and the tubing is spaced on centers 1.042 inches apart. The intentional rupture tube for a given test is pressurized by water / steam frcm the LLTR water injection system tanks. The remaining tubes are suppl led by conrnon headers at each end of the steam generator.

The Al-MSG was not originally designed to accommodate large leak testing and thus there are no provisions for tube bundle removal and insertion, instrument accessibility, or internal disassembly. The steam generator has l'een assembled with four (4) inplace pre-weakened rupture tubes, and a capability for an additional fcur (4), using replaceable rupture tubes in the horizontal short leg of the Al-MSG.

The Al-MSG instrumentation consists of approximately 130 thermocouples, 13 high-frequency pressure transducers, and 18 strain gauges. These sensors are provided to monitor bubble growth and pressure wave propagation, mechanical effects deformation, and to define the potential for secondary tuce failures.

The test data capability provided for the modified Al-MSG includes:

1) Pressure wave magnitude and propagation mapping.

2) Bubble growth mapping.

O 5.5-18c Amend. 62 Nov. 1981

3) Visual damage inspection.

4) Results of pneumatic and helium leak tests.

5) Ultrasonic signature comparison of tube wall thickness with original (full treatment depends on technique development in the bend regions).

6) Measurements of water / steam injection flow conditions.

7) Piping and vessel transient pressure, temperatures, and strain.

8) Relief system transients including pressure, temperature, flow, and strain responses at representative positions in the system out to and including the stack.

9) Transient pressures and temperatures in reaction products tank resulting from the entry of unreacted steam into the tank.

10) Vent system flow vs. time, and particulate size, distribution, and chemical composition in the vent stack.

Six large leak tests have been conducted in the MSG during the Series I LLTR tests. The first three investigated double-endcd guillotino (DEG) breaks in the evaporator. The first test was of a break near the sodium outlet window.

The second test, a break about twenty f eet above the lower tube sheet. The third test, was of a break in the hockey stick region. The remaining three

) tests investigated a DEG in a superheater, DEG with inert gas and a large s_,/ break (equivalent to 3 DEGS) in the superheater.

The results and data obtained from the LLTR Series I tests demonstrate that the TRANSWRAP Code predictions of pressures and velocities resulting from large SWR events are conservative (Reference 25). The methodology which validates the TRANSWRAP code for use in the design of the CRBRP SWRPRS utilized the following procedures:

1) Theinjecgionflowtransientsonthewatersidewerecomputedwiththe RELAP/ MOD code (Reference 9).

2) These flow injection rates were input to the TRANSWRAP code to compute pressures throughout the sodium side of the system. The claculations were based on a sodium water reaction with an assumed hydrogen yield of % of the injected water to hydrogen gas and a resulting bubble temperature of 1700 F.

3) A dynamic rupture disc model was incorporated into the TRANSWRAP code to conservatively predict pressures within the LLTR.

4) A static rupture disc model was used in TRANSWRAP to conservatively predict velocities throughout the system and pressures within the LLTR SWRPRS n

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5.5-18d Amend. 62 Nov. 1981

The Series ! Test Article was disassembled and examined f ollowing the sodium-water reaction tests. This examiniation (reported in Reference 28) showed no evidence of secondary tube f ailures. Tube deformation and localized wastage was found in the regions of the tube rupture sites. The maximum wastage found was 0.019 inch adjacent to the test no. 2 site. Wastage at other tube rupture sites was in the order of 0.004-0.005 inch.

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5.5-18d a Amend. L2 Nov. 1981 l

c. Few Tube Test (FTT)

(/)

s, Objective The objectives of the Few Tube Test Program were to (1) conduct endurance tests of the Few Tube Test Models (FTTM) to evaluate tube / tube support wear and the reliability of tube-tubesheet welds under long-term operating conditions, and (2) obtain performance data for operating conditions ranging from natural circulation to full power.

Test Facilltv Breeder Test Facility (BTF), Steam Generator Test Rig (SGTR), General Electric Company (GE), Advanced Reactor Systems Department, San Jose, California.

Comoonent Characteristics The physical configuration of the FTTM's were similar to the reference hockey stick design of the CRBRP steam generator evaporator /superheater configuration. The FTTM's were operated in a side-by-side position in the General Electric SGTR. The assemblies were f abricated f rom 2-1/4 Cr - 1 Mo low alloy ferritic steel. The evaporator employed 7 active tubes and the superheater employed 3 active tubes (4 tubes were plugged during fabrication). The tube-to-tube support interface details were designed to be representative of the CRBRP units. The length and radil of the FTTM tubes were selected to be the same as the CRBRP steam generator shortest p)

(,, row of tubes on the basis that this condition represented the worst combination of tube-to-tube support movement and side forces in the,CRBRP steam generators.

The tube and tubesheets at each of the tube bundles provided the sodium-water barrier. The main shell assembly provided containment for the sodium, support for the internals, mcunting points for the sodium inlets and outlets, and instrumentation feed-throughs. Rupture discs for sodium-l water reaction pressure relief were prcvided in the sodium inlet and outlet lines adjacent to the test article.

Removchte, flanged steam / water headers wer e bolted to each tube sheet to

l provide a transition between the steam tubes of the models and the steam /

water lines of the SGTR.

Test Descriotion Summary l The overall scope of the Few Tube Model testing program include steady-state thermal performance testing, long term endurance testing and post-test examination. Thermal / hydraulic and stability performance O

V 5.5-18e Amend. 62 Nov. 1981

l tests were conducted to confirm thermal performance predictions for multi-tube evaporator and superheater operation based on single tube test l results over a range of conditions. The performance verification testing was to include oper ation with a plugged tube and with and without evaporator tube orifices for operating modes from natural circulation decay heat removal to fuii tube power.

Long term endurance testing (10,000 - 12,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of system operation) l would have tested the sodium-water boundary design and included performing simulated CRBRP steam generator load changes to impose full relative thermal expansion movement, tube support side forces, and associated bending stresses in the tube /tubesheet joints. The long term endurance tests would have included startups and shutdowns from ambient to full power cyclic load changes, and periods of full power operation.

l Ancillary tests were conducted in conjunction with the Few Tube Test Program which further increased the quantity of technical data derived l from this effort. These additional test areas were leak detection and prototype rupture dise Integrity. Operation of prototypical CRBRP leak l

detectors provided proof testing of the leak detector module designs.

Signal response times as a func+ ion of simulated leak rates in the SGTR l were determined.

Test Results The Few Tube Test Program was concluded prior to fully achieving the test objectives. Testing of the Few Tube Test Models (FFTM) was conducted over O. the period of July - December 1978 at which time testing was terminated to investigate anamolles in FTTM performance. Subsequently, the test models were disassembled and examined. This extensive post test examination provided dat3 that initiated design improvements to the steam generator modules. The design changes reduce potential restraints to thermal expansion of internals by increasing clearences between parts where relative ation is expected and modifying material specifications to avoid material couple problems such as gc!!Ing and adhesive wear,

d. DNB Tests ObiectIve Tne overall objective of the test program was to verify that excessive damage to the CRBRP evaporator tubes will not be produced by operation with departure from nucieate boliIng (DNB) or iIquid fiIm dryout in conjunction with maximum specified CRBRP water chemistry conditions.

The minimum objective of the test program was to demonstrate that the CRBRP evaporator tubes wiII not suf f er earIy f alIure; 1 hat is, have a 1if e of 5 years or greater.

i O 5.5-18f Amend. 62 Nov. 1981

1 l

l Test Facility Breeder Test Facility (BTF), General Electric Company (GE), Fast Breeder Reactor Department (FBRD), Sunnyvale, California.

Comeonent Characteristics The DNB Effects Test section is a 16' long single tube sodium heated evaporater which employs a 5/8" OD x 0.109 inch wall 21/4 Cr - 1 Mo i

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5.5-18f a Amend. 62 Nov. 1981

tube. Under the Endurance Test Conditions, the tube contains the

/7 nucleate boiling, DNB and phrt of the post-DNB boiling regions of the CRBRP evaporator. Water chemistry is controlled to CRBRP parameters

(._)

43 l with the exception that NaOH is added to maintain Na ion at or above the maximum expected level, and thermal / hydraulic conditions are controlled to maintai n the DNB condition that represents the maxi-mum tube wall temperature oscillations.

Test Descriotion Summary Test conditions were selected which represented the worst case conditions to which the CRBRP evaporator tubes could be subjected:

(a) Maximum o(AT) or tube wall temperature oscillations as produced by part load conditions of 64% load and (b) Maximum sodium hydroxide conditions in the evaporator water. Based on being able to detect a tube wall thickness change of 1 mil, a test period of approximately 4 months was determined to be necessary, i.e..

15 mils _

3 mils _

1 mil 5 ';e r s 1 year 4 months Baseo on 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> / day and 30 days / month, the test period of 4 months is equivalent to 2880 hours0.0333 days <br />0.8 hours <br />0.00476 weeks <br />0.0011 months <br />.

43 l The test program was comprised of the following phases:

e Initial thermal conditioning run of s400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> for establishing

(]

V proper film condition inside the tube. During this period, thermal / hydraulic test data were taken to establish the endurance test condition.

43) e Endurance testing of about 3000 hrs wich in-situ nondestructive exan,ina tion s af ter about 1000 nes and after test completicn.

! e Post-Test Destructive Examination.

l Schedule of Tests l

DNB Effects testing was initiated in December 1975 and was completed in Novenber 1976.

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Summary of Results 43l  !

A total of 2820* hours were accumulated at endurance tests conditions consisting of a) the expected worst case water chemistry condition allowed by the CRBRP soecification (i.e. , 6 ppb of sodium),

j and b) the largest temperature cycling condition predicted for the plant evaporator A( AT) of 100 UF , -The sodium ion concentration 43{ was maintained at or above the maximum CRPRP limit by adding sodium hydroxide 36 y)

• See Footnote,following page.

Amend. 43 5.5-18g Jan. 1978

p V

size of a secondary f ailure caused by wastage is dif ficult to predict, it is expected to be smaller than the leak rate corresponding to a double-ended guiilotine faliure.

The third mechanism for falIing a tube Is overheating from the thermal effects of a SWR caused by a leak in an adjacent tube. This could cause heatup in the adjacent tube and a decrease in tube strength until the tube bursts from the internal pr essure. The time f or tube f ailure was analytically investigated using boundary conditions of a temperature of 27000F (the adiabatic SWR2 temperature) and heat transfer coef fIclents as high as 10,000 Btu /hr-f t 0F.

The minimum computed time for failure was 0.4 seconds for an evaporator tube and 0.3 seconds for a superheater tube. Measurements during large SWR tests show peak temperatures of 23000F and heat transfer coeffIctents of approximately 2000 Btu /hr-ft2 cF (Ref. 15). These measured conditions, if applied uniformly around the circumference of a tube and over a significant longitudinal area, would require longer times than computed above to produce an overheating failure. Further, establishment of the conditions that could cause such a secondary f ailure require a large initiating leak.

9 It is concluded that alI three leak growth mechanisms require time to develop; hours for small leak self-wastage, tens of seconds to minutes for wastage, and tenths of seconds to seconds for overheating. The secondary falIures that 39 could potentially result from these mechanisms are expected to yield water leaks considerably less than that of a single double-ended guillotine f ailure.

An estimate of the worst plausible leak development sequence would be as follows. A small leak (less than 10-3 lb/sec) is postulated to develop. This J Ieak will be assuied to not be detected by the operators through readings f rom the hydrogen Ieak detectors. This Ieak then grows by seIf-wastage to a Icak of 10-2 or 10-3 lb/sec (a leak size that gives maximum wastage rates on an adjacent tube). This leak continues, if no mitigating action is taken by the operators, until the adjacent tube wall is penetrated. The second leak is assumed to be smali enough to not activate the SWRPRS by bicwing out rupture disc, but large enough to yield a SWR which overheats an adjacent tube to the point of failure. This f Inal overheating f ailure is assumed to activate the SWRPRS. The result of this worst plausible sequence is leaks for which the total leak flew rate is not expected to exceed that of a single double-ended guilIctine falIure. 9 FThe Design Basis Leak is not intended to represent a realistic sequence, but, rather, to provide a conservative basis for computing design loads. The DBL is defined as a equivalent double ended guillotine (EDEG) failure of a steam

' generator tube which is followed by two additional single EDEG failures, spaced at 1.0 second intervals, to a total of 3 EDEG. This sequence is superimposed on a system which has been pressurized by an undetected moderate sized leak to

.; 59 Just below the rupture disk t'urst pressure.

The design basis leak for the CRBRP steam generators contains safety margins in Experimental data 59lthetimingandalsothemagnitudeoftheassumedfailures.on (Ref. 15, 16, 1,7, 24) the consequences of a sud event, justi fy the design basis leak postul ated f or the CRBRP steam generator.

59 l Although much of the data is not strictly prototypic of the CRBRP steam O~ generator modnt es, the results demonstrate the ef fects of SWR reactions in

' LMFBR steam generators using sim!Iar materials, tube wall thickness, pressures, 9 5.5-24b Amend. 59 30 Dec. 1980

water injection rates and sodium temperatures. Japanese, German and US large Icak SWR tests have produced no secondary failures.

The Japanese have conducted seven large roak SWR tests ranging from seven to n seconds. The Germans have conducted five large leak tests of durations 4 to 9 seconds. Six large leak tests (in neer-prototype configurations) have been conducted in the U.S. The U.S. test have ranged from 3 to 40 seconds in duration. Significant wastege was observed in only one U.S. test in which one tube in the leaksite region exhibited a 0.016 inch reduction in wall thickness.

This corresponded to a wastego rate of 0.016 inch /sec.

A full-scale leak progression test has been conduc ed in a steam generator of prototypic dimens!cns and materials. This test, Large Leak Test kig (LLTH)

Series ll Test A-3 (Ref. 27), was initiated by repidly pulling apart a pre-notched tube to expose an injection tube containing a pre-drilled 0.040 inch diameter hole. This hole, representing the self-wastage leak depicted in Step 5 of Figure 15.3.3.3-1, was aimed a a target tube two rows away. The aiming and spacing had been previously determined by bench scale experiments to yleid the maximum wastage rate on the target tube. Observed serendary failure For reference, an EDEG failure erec (two cross sequenceistabulatejbelow.

secticns) is 0.26 in .

TIME-SECGNDS SECONDARY FAILURE FAILUPE AREA-IN.2 0 Injection Tube Pre-drilled 0.0013 in.2 60 1 0.01 /

72 2 95 3 0.029 108 4 0.029 114 RUPTURE DISC CPENED 114-120 5 0.11 114-120 6 0.200 114-120 7 0.170 The first secondary required one ininute to develop. Further secondary f ailures occurred at 6 to 12 second spacing. The size of each secondary was less than of a EDEG. These results are a conservative representation of how an acutal leak would progress because (1) the initial leak was aimed and spaced to produce maximum wastage on the target tube (wastage is observed in bench scale tests to be sensitive to configuration), (2) the third secondary failure occurred in the injection tube itself which was of non-prototypic (thin) wall 15Ickness, (3) the tubes contained initial:y subcooled water which was static; as the test progressed the water flashed and was expelled into the supply system - the tubes were thus undercooled, and (4) the sodium was stetic. The CRBRP Design Basis Leak is conservative in both the magnitude of and timing of these secondary leaks.

5.5-24c Amend. 62 Nov. 1981

British and Russian tests have demonstrated that an Initial tube failure can O produce secondary tube failures, in the 15 tests of the NOAH series Gef. 20) there were three 1ests in which pressurized Cr-Mo secondary tubes f ailed. No mere than one dabe failed in any test. Times to failure varied from 2.5 to 4.5 seconds. ... 'hree Super-NOAH tests (Ref. 21) there was one secondary '

failure attributed to wastage / overheating. Time to f ailure was 14.5 seconas.

Dumm's 'Ref. 15) sodium-water reaction test number 5 had pressurized secondary tubes 1.one of which failed, r

The Russians (Ref. 17) report which a small leak (13.2x10~gsodiumheatedsteamgeneratorleaktotin lb water /sec injected through a 0.030" hole) caused two tubes to fall in a period of 100 seconds. The first failure was attributed to direct wasiege (from the jet Impinging on the tube). The second (large) teak caused a blowout type (overheating) failure of the second tube.

In every case, the rnechanism producing f ailures has been of the type which requires a substantial time lag between the occurrence of the primar / event and the initiation of secondary SWR, and the magnitude of the secondary failure was less than the equivalent of one double-ended guillotine failure.

This delay or lack of coherence in propagation and the limited extent of secondary tube failures has been a significant finding. These results demonstrate that water / steam injection from any secondary failures occurs at a time when potential pressure or shock ef f ects are mitigated by the sequence of conditions re 'Iting from the primary rupture event, i.e., by the production of a large volume of reacrion product gas within the steam generator shell, and by actuation of the SWRPRS rupture discs. Also, while high frequency O (1000 Hz) acoustic pressure pulses typically occur in the first fee, b milliseconds of a guillotine-type rupture event, it has been determined that they produce no shell loading problems due to their low energy content (Ref.

15). Therefore, on the basis of large SWR experience to date, no mechanism has been found effective in causing even a single secondary failure c7 a time scale rapid enough to contribute to substantial IHX Icadir.gs. In addition, the number of equivalent double-ended guillotine secondary tube failures thaT have occurred in any test (approximately one - Ref. 16) is significantly less than the two assumed for the DBL.

In ihe case of small primary (initiating) leaks, the leak growth mcchanisms identifled through tests do not cause instantaneous secondary failures and do not cause secon Gry water leaks equivalent to a double-ended gulliotine tube break. Any deley in time to fall the additional tubes would reduce the i pressures resulting from this event. Based on existing data and ar,aiyses, the destgr; basis leak of a 1-tube double-ended guitlotine failure folicwed at 1.0 second Intervals by single EDEG f ailures to a total of 3 EDEG f ailures will result in a conservative IHTS and SGS design. Analysis (Ref. 25) of data from LLTR test series (Ref. 24) has verified analysis methods used in assessing the conservatism of the DBL.

O 5.5-24d Amend. 62 Nov. 1981

P 4

59 The three lube event is not intended to represent a realistic, mechanistic scquence, but rather it providec e basis for calculating loads Scr the design of components and piping which are be!Iey?d to be conservatlye f or the large nember of rrechanistic sequences involving secondary f ailures which c.m be postulated.

Prellt inary ar,alyses have indicatcd that edjacent tubes sould not f all wt.an 4

abjected to the ceap calculated pressures f or the double ended gui! ktine falIure of one tube. Mechanical failure of Ijacent tubes due to whipping of the initially f ailed tube is also considerec anlikely. It a series o5 tests it was demonstrated I

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L O' d 5.5-25 Amend. 59 Dec. 1980

c. Pressure and volume in cover gas spaces are computed,

d. Rupture disc bursting and subsequent flow of sodiva into the relief system are computed.

e. Pressure pulse reflections end resultant reinforcements and raref ectr is are treated.

f. The mo4 fen of the bubble-sodium interface is compute:d.

g. Friction effects are included for all components end piping. Resultant acoustic wave attenuation is computed.

h. Pressure pulse attenuation due to flow into tanks and reservoirs is computed.

Important conditions and assumptions built into TRANSWRAP Include:

a. Instantaneous conversion of 65% of the injected water to hydrogen gas.

The 65% yield has been determined (Ref. 25) to be conservative through analysis of U.S. large leak test data (Ref. 24). The British (Ref. 20) and Germans (Ref. 15) have Inferred, respectively, 55% and 50% hyd'rogen yleids from their large leak tests.

b. The Na 2 0 remains in the H2 bubble but has no effect on pressure or volume of the bubble.

c. The reaction products are in thermal equilibrium.

l d. The effective hydrogen bubble tempereture is 17000F. This has been determired (Ref. 25) to be conservative through analysis of U.S. large leak test data (Ref. 24).

o. The pump cover gas experiences Isentropic compression (expansion) as a perfect gas,

f. The rupture discs are represented by dynamic models which have been conservatively calibrated against prototype rupture disc tests (Rets.

26, 27).

The TRANSWRAP code has been validated through analysis (Ref. 25) of the LLTR Series I test aara (Ref. 24).

O 5.5-27 Amend. 62 Nov. 1981

The RELAP code (Ref. 9) has been cailbrated (Ref. 25) against the LLTR Series I large leak test date (Ref. 24) and used to compute a maximum water injection ratio of 38./ lb/sec. This corresponds to 12.5 lb/sec from each of three double ended tube failures. The expanding hydrogen bubble is ireated within the TRANSWRAP code as a continuem in which there are no graclents. The msss of hydrogen contained in the bubble Ircreases as the reaction proceeds and the bubble interacts hydrodynamically with the flowing sodium at its bounderles.

The bubble is considered to be a perfect gas. The sodium-water reaction is assumed to be instantaneous with a 65% yield of hydrogen plus 1.2 lb/sec f rom the precursor leak.

Prediction of local hot spots is not within the scope of the TRANSWRAP Code.

l The tube over-heating analysis presented earlier in this section is a conservative bourding estimate for local hot spot effects. The probability of l

local hot spots ;s small considering the excessive amounts of high conductivity sodium and the high degree of turbulence which can be expected at the leak site.

Analvticat Model Each compo-ent in the IHTS is represented in the TRANSWRAP model as an equivalent c.i cular cross section pipe (or a comb! nation of equivalent pipes) of specified length, diame+er, elesticity, resistance, and associated initial conditions of flow rate ano pressure distribution. The model employed for the large Si'R analysis is represented in Figure 5.5-3.

The method of characteristics as developed in References 6 and 7 is incorporated in the TRANSWRAP Code. The equations solved between nodal points on the characteristic grid are (p. 23, Deference 6).

a dH + GV + fV!Vl = 0 a dt dt 2D 2"+a dt

_ g dH + dV + fV V = 0 a dt dt 2D L= -a l dt l

where b = pressure head g = gravitational constant V = fluid velocity, a = acoustic velocities l f = friction factor D = pipe diameter I t = t!me X = distance (axial) l l

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5.5-27a Amend. 62 Nov. 1981

Attenuation of head or acoustic pulse between nodal points, i.e., through straight runs of pipe, is thus implicitly recognized through the friction factor f, which is input for each calculational setpent within the model .

j O Three types of junction points or joints connecting straight runs of pipe are available within TRANSWRAP; butt joints, tee joints and end joints (dead ends). Attenuation of acoustic pulses in passing through each type of junction is incorporated through the boundary conditions imposed on the above equations. The boundary conditions, as derived in Reference 7, for each junction type are given below.

(1) Head due to reflections at a dead end:

H = 2 F( t - h) l F = travelling wave function L = len;;th from disturbance evaluated from boundary conditions to dead end The meaning of this expression is that at a dead end, the reflected wave is equal to the incident wave and is of the same sign, (2) At a butt joint with change of diameter, both reflection and transmission are recognized through the Actors:

Reflection Factor = (A j/a j ) - (A2/a2)

(Aj /aj ) + (A2/a2) 2Ajja)

Transmission Factor =

{A/ap+(A/a) j 2 2 where A is cross-sectional flow area and a is acoustic velocity.

(3) Transmission and reflection factors at a tee joint are:

2(Aj /aj )

Transmission Factor j= ( A /a))+( 2 2 A3 /a 3)+( A /a )

(Aj /a))-(A2/a2 )-A3/a3)

Reflection Factor = ( A)/a))+(X 2 2 /a3)+( 3 A /a )

This treatment of attenuation through friction factors and boundary conditions ,

is generally conservative in that the one-dimensional model cannot account for multi-dimensional attenuation-effects, e.g., in the transmission of an acoustic wave cround an elbow. Preliminary measurements in water (Reference 8) indicated a reduction in acoustic wave magnitude of from 10 to 30% in passing through an elbow. This is about an order-of-magnitude higher than what is computed in TRANSWRAP using one-dimensional fraction factors.

6 5.5-27b Amend. 6 October 1975

Results TRANSWRAP calculations have been performed for both evaporator and suoerheater O tube f611ures for the Design Basis Leak. Calculated peak pressures in the major components are shown in Table 5.5-11. The predicted pressure history in the IHX for an evaporator DBL is shown in Figure 5.5-4A. The pressure remains at the steady state value untii the fIrst pressure putse arrives. Thereafter, the pressure is affected by reinforcements end rarefactions. The initial disturbance occurs at 90 milliseconds after initiation of a moderate-sized leak which falls to burst the SWRPRS rupture discs. The delay corresponds to the transit time from the leaksite to the IHX via the cold leg. Between 90 and 480 ms the pressure builds up as the leak continues. The pressure decreased between 460 and 700 ms as the burst of the superheater rupture disc and the venting through both the evaporator and superheater Sodium-Water Peaction Pressure Reilef System (SWRPRS) reilef 1Ines depressurized the system. As seen from Figure 5.5-4b (Leaksite Pressure History for SWR DBL in Evaporator), the two additional EDEG (at 1.3 and 2.3 seconds) have little effect on the IHX pressure since by 1.3 seconds the leaksite is cushioned by about 120 f t 3 of recction product gas.

O O

5.5-28 Amend. 62 Nov. 1981

For the Design Basis Leak in the superheater, the peak pressures in all the Ol major components are lower than for the DBL in the evaporator leak sequcnce except in the IHX. The lower pressures are the result of the lower water mass flowrate throughout the transient.

The steady-state sodium flow rates and pressure drcps throughout the IHTS prior to the steam generator tube rupture are represented in the TRANSWRAP model.

Following the tube rupture, the expanding bubble of sodium / water reaction products is treated as a continuum of perfect gas which interacts hydrodynamically with the flowing sodium. For the Design Basis Leak in an evaporator, the bubble is predicted to expand away from the leak in both directions, i.e., the sod!am which at steady state flows into the evaporator is reversed while the sodium flowing out of the evaporator is accelerated. As the sodium originally within the evaporator below the leak site is displaced by the reaction products, it is driven through the evaporator outlet tee. The bulk of the sodium and reaction products are expelled through the Interconnected SWRPRS relief line. However, a portion is also predicted to flow towards the pump.

The SWRPRS roltef line is cleared of liquid sodium after about 4 seconds. Gas blowdown through the cleared relief line decreases the bubble pressure. Peak system pressures occur during the first second of the event. it is expected that the sodium ficw in the pump suction line will reverse before the gas bubble reaches the pump. The sodium will then drain back toward the relief line (Iow point in the system). Loop draining will be completed by manual opening cf the sodium dump valves.

in the unlikely event that the flow in the IHTS does not reverse and the gas bubble reaches the pump, no damage to the coolant boundary of the pump is expected. It is conservatively assumed that the sodium / gas interface reaches the pump inlet abcet 8 seconds after the CWR is initiated. However, all PHTS and IHTS pumps are tripped by the PPS by approximately 1 second after SWR initiaticn. Per the specified pump transient, by seven seconds the pump inlet pressure is reduced to the order of 50 psi, and the pump speed will be reduced to the order of 40% full speed.

Since the pump main motor is tripped long before the bubble could arrive at the pump Inlet, there is no possibility of pump overspeed and subsequent missile generation. Uneven hydraulic loads and loss of sodium would eventually result in bearing damage and seizure of the pump.

O O

5.5-28a Amend. 62 Nov. 1981

O For the Design Basis Leak in the superheater, the reaction products bub-ble is predicted to expand away from the leak site in both directions also.

Reaction products above the leak site and sodium which normally flows into the superheater at e expelled through the SWRPRS relief line connected to the super-heater inlet tee. Since flow reversal in the superheater sodium inlet line does not occur because of continued flow from the pump, reaction products cannot enter I

the IHTS hot ieg. Reaction products below the leak site are predicted to accel-erate downstream toward evaporator relief lines. The sodium drains from the loop through the relief lines and the sodium dump lines similar to that described l14 above for the evaporater event.

Conclusions Based on the information provided above, it is concluded that systems 43l and components Msigned to the ASME Section III categories given in Table 3.2-5 6 using the loadings given in Table 5.5-11 will maintain their f

1 Q'

l Amend. 43 Jan. 1978 5.5-28b 1

1

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

p References to Section 5.5

'

• 1. W.H. Yunker, " Standard FFTF Values f or the Physical and Thermophysical Properties of Sodi um" WHAM-D-3, July 6,1970.

• 2. J. A. Bray, "Some Notes on Sodium / Water Reaction Wcrk," CONF-710548, pp.

187-205, July 1912.

• 3. Nuclear System Materials Handbook, Hanford Engineering Development 41 l Laboratory, TID-26666, Volume 1, Section 2-2 1/4 Cr-Mo, pp.1.0-1.2, 6 Rev. O, August 14, 1974.

• 4. R.B. Harty, "Modul er Steam Generator Final Project Report," Atomics i nter nati ona l , TR-097-330-010, September 1974. )

5. Nuclear Systems Materials Handbook TID 26666, 1974. 2

• 6. V.L. Streeter and E.B. Wyl le, Hvdrau l ic Trans ients, McGraw-Hil l, New York, 1967, Ch. 2 and 3. ,

• 7. John Pickf ord, Analysis of Surge, MacMil le ., London, 1969, pp. 32-37.

e 8. D.J. Cagliostro, S.J. Wiersman, A.L. Florence, Stanford Research Institute Final Report P.O.190-C1H88GX, " Pressure Pulso Propagation in a Simple Model of the intermediate Heat Transport System of a Liquid 6 Metal Fast Breeder Reactor," June 1975.

c I )

L'

• 9. "REL AP4/ MOD 5 a Cceputer Progran f or Transient Thermal-Hydraul ic Analysis of Nuclear Reactors and Related Systems," prepared by Aerojet Nuclear Conpany for U.S. Nuclear Regulatory Ccmmission and Energy Research and Development Administra+ ion under Contract E (10-1) - 1375,

-N -

, pt 6.

26 59

• 10. J.N. Fox, R. SalvatorI, H.J. ThalIar (H NES), " Experimental Bending Tests on Pressurized Piping Under Static and Simulated Accident Cord!tions" TRANSACTIONS, ANS Power Division Conf erence on Power Reactor Systems and Components, September 1-3, 1970.

11. "Draf t Design Basis f or Protection Against Pipe Whip," ANSI N176, June, 59 l 3974,

• 12. "LLTR Serir.s i Test Request", General Electric Co., Fast Breeder Reactor Departn ent Specif Ication No. 22A3924, Rev. O, Prepared Under Contract No. AT (04-3)-893-10, LMFBR Steam Generatcr Dcvelopment Program, Task 189-13 882 ( B) .

• 13. Gudahl, J.A. and Magee, P.M., "Microteak Wastage Test Results,"

GEFR-00352, March 1978.

• References annotated with an asterisk support conclusions In the Section. 9 26 Other references are provided as background information.

\ ,6

+

1 i

1 5.5-35 Amend. 59 Dec. 1980 l

• 14. Greene, D.A., Gudahl, J.A., Hunsicker, J.C., "Exper v e<ial Investigation of the Wastage t' Steam Generator Mawr!a s by Sodium /

Water Reactions," GEAP-14094. anuary 1976.

• 15. Dumm, K. et.al., Experimental mu Theoretical In, ot c. 1 on Safety of the SNR Straight-Tube Design Steam Generster with Soalum-Water Reactions, INTAT72.12 (ERDA-TR-27), INTERATOM, Aprii 1972.

16. J. A. Bray, "Some Notes on Sodium / Water Reactir.n Work," Paper presented at the Specialists Meeting on Sodium Waier Reactions, CONF-710548, held May 16-21, 1971, Melekess, USSR.

• 17. B.V. Kuplin, et.al., " Study of Na and H O2 'ateractions in a One Megawatt Modular Steam Generator," Peper present a at the Specialists Meeting on Sodium-Water Reactions, CONF-710548- HW d May 18-?I ,1971, Melekess, USSR.

18. Liquid Metal Engineering Center - Failure Data Hanabcok, LMEC Memo -

69-7, Volume 1, Atomics International, Al-F T 996-13-00, August 15, 1969.

19. Reactor Primary Coolant System Rupture Study, Quarterly Report No. 23, October-December 1970, GEAP 10207-23, January 1971.

20. J.A. Bray, et al., " Sodium / Water Reaction Experiments on Model P.F.R.

Heat Exchangers- The NOAH Rig Tests," TRG-Reports-1519,1967.

21 . J.A. Bray, "A Review of Some Sodium / Water Reaction Experiments,"

British Nuclear Energy Society Journal, Vol. 10, No. 2, Aprl! 1971.

• 22. P.B. Stephens, D.N. Redgers, et.al., "DNB Effects Test Program Final Report," GEFR-00100(L), June 1977.

• 23. J.C. Whipple, et.al., "U.S. Program for Large Sodium / Water Reaction Tests," GEFR-SP-039, Noveraber 1977.

j 24. R.L. Eichelberger, " Sodium-Water Reaction Tests in LLTR Series I, Final

! Repert," ETEC-78-10, Ju l y 15, 1978.

l 25. J.0. Sane, et . al . , " Eval uat ion of Sodium-Water Rea: tion Tests No.1 1 Through 6 Data and Comparison with TRANSWRAP Analysis Series I Large Leak Test Program, Volumes I and lI," GEFR-00420, June 1980.

l 1

i 26. J.C. Whipple, et.al. " Evaluation of LLTR Series ll Test A-2 Results,"

l General Electric Advanced Reactor Systems Department, July 1980, Prepared for U.S. Department of Energy under Contract No.

De-ATc3-76SF70030, Work Package AF 15 10 05, WPT No. SG037.

l l

l 5.5-353 Amend. 62 Nov. 1981

i i

I 27. J. C. Amos, et.al, " Evaluation of LLTR Series ll Test A-3 Results," General l

Electric Advanced Reactor Systems Department, November 1980, Prepared for U.S. Department of Energy under Contract No. DE-ATc3-76SF70030,  ;

Work Package AF 15 10 05, WPT No. SG037.

l 28. J.O. Sterns, " Metallurgical Evaluation of the Modular Steam Generator l

(MSG) af ter LLTR Testing," ETEC-78-12, Sept.1978.

• Refer 9nces annotated with an asterisk support conclusions in the Section. t Other references are provided as background Information.

i i

l i

i 1  !

O

! l 2 I l

4 l

t I

l l

i

till l 5.5-35b ,

Amend. 62 i i

Nov. 1981 l

l

i i

1 t TABLE 5.5-6

! SGS LOADIfiG C0t1DITI0f15 35l ASME III Code Class 3 system components will be designed considering

the following load combinations

1 Pumps (3ecirculation Loop)

! Operating Condition Component Load Stress Limit i See flote 1 Pump Case Design Pressure Section III

! Design Temperature Allowable Stress l Cover Design Pressure Section III l Bol ti ng Design Temperature Allowable Stress j Safe Shutdown Earthquake 1

J Pump Thrust j Weight '

Gasket Loads  !

flote 1: Design pressures and temperaturas of the recirculation system com-  !

) ponents are established using pressures and temperatures occurring i

during emergency and faulted transients. The design temperature is not exceeded during these transients. The design pressure may be exceeded by not more than 10% during these transients. flormal and I upset conditions are not controllina.

2 Valves (Recirculation Loop and fiain Water / Steam) 35l The valve pressure retaining parts designed to ASf1E - III Class 3 will withstand seismic forces and pipe loads of the SSE as well as design '

I pressure and temperatures. On other parts, if earthquake needs are to be l considered, the following applies:

Operating Condition Loads Upset 1. flormal Operating

2. OBE Faulted 1. fiormal Operating

2. SSE Amend. 35 5.5-46 Feb. 1977

TABLE 5. 5-7 SGS Piping and Their besign Characteristics NO. NO. ASME CODE COMONENT PER PER SEC. til DES IGN PlPlNG AND HEADERS SIZE LOOP FtANT (1 ASS REQUIREMENTS

1. Steam Generator Subsystem & Feedwater Subsys+ ,

SGB Wall to Drum Feedwater Isolation VM ve 10", sch. 160 1 3 3 3000 psig. 500 F Feedwater Drum isolation Val w to Steam Drum 10", sch. 140 1 3 3 2200 psig, 650 F Drum to Pump Inlet Header 10", sch. 140 4 12 3 2200 psig, 650 F Pump Headers (inlet) 18", sch. 140 cn 1 3 3 2200 psig, 650 F Pump Inlet Header to Pump 18", sch, 140 1 3 3 2200 pstg, 650 F g Pump to Pump Discharge Tee 12", sch. ;60 1 3 3 2450 pstg 650 F i Pump Discharge Tee 12", *ch. 160 1 3 3 2450 psig, 650 F

$ Pump Dischargo Teo to Eveporator Isolation Valve 10", sch. 160 2 6 3 2450 psig, 650 F Evaporator isolation Valve to Evaporator 10", sch, 160 2 6 3 2400 psig, 650 F Evaporator to Drum 16", sch. 1 h' 2 6 3 2200 psig, 650 F Drum to S.H. 12", sch. 140 1 3 3 2200 psig, 650 F S.H. to isolation Valve 16", sch. 160 1 3 3 2200 psig, 650 F lsolation Valve to SG8 Wall 16", sch. 160 1 3 3 1900 psig, 935 F Startup Feedwater Control Valve Piping sch. 160 4", 1 3 3 3000 psig, 500 F

2. SWRfRS Sodium Rupture Disc Discharge Lines to Separator Tanks 18" nom., var.

Well, 24" nom.,

Tar Wall, 26.46" nom., 2.2" wall 3 7 3 300 psig, 800 F Separation Tank to SGB Roof 16", sch. 40 1 3 3 125 psig, 200/800 F SGB Roof to Flare Tip 16", sch. 40 1 3 ANSI 125 psig, 1000F B31.1 2CL C'

e S 9

O O O Table 5.5-10 SWR DESIGN BASIS ASVE CODE CATECORY OTHER STEAM GENERATORS AND IHTS EQUlFNEliT l

LEAK DESCRIPTION (1) FAILED STEAM MNERATOR AFFECTED RPST IN TFE AFFECTEC LOOP 44 SrnalI Leak in One Tube Upset Normal Upset

, One ECEG5 Folicwed by Two Faulted Faulted Emergency i Additional Single EDEG Fallures at One Second 61 Intervals (Total 3 EDEC's)

, (1) Seo Section 5.5.3.6 for detalled descriptions and basis 61l

• Equivalent double ended gulllatine F'

i'

. W X l' RBQ.

=

i.

e

TABLE 5.5-11 Calculated Results for Large SWR Design Basis Leak

• lHX Failed Pump Peak Peak Pressure in Adjacent Time to Failure Peak Unit Peak Pressure Steam Generators, PSIA Clear First Relief Location Pressure Pressure PSIA 8 Sec. # Sec. Line, Seconds PSIA R Sec. PSIA A Sec. Eva3 orator Suoerheater Evaporator $331 3 95 373 320 337 8 0.412 8 0.420 0 0.436 8 0.391 8 0.364 4.24 Superheater 304 333 311 254 -

8 0.311 8 0.548 8 0.619 0 0.438 3.65

• Water Iajection rate = 1.2 lb/sec f or 01 t = 0.3 see Precursor Leak At t = 0.3 sec, one EDEG occurs. At t = 1.3 sec., one additional EDEG occurs. At t = 2.3 sec.,

one more DEG occurs. (total 3 EDEG)

W I

W

(,s)

< (D

= 3 w

e CD C wN O O O

O O O I r

= I42 FI.-2.25 12" STEM OUTLET j

4' AUX ICIrMAlfR _\

r 'N f

~

f; 82.0 0.p. r r o /

~i n - n .

'd

~i

~

W  :

'Q ~d

{

? 17 W -

6" CONTifiOUS DPAIN 10" WATER 16" INLET

• OUTLET N0ZZLES

.I Figure 5.5-4 Steam Drum Outline EE 5a 2

$2

400- - - -

300 -

Y l

t '

200 e h

j- h

U

~n i

l 1

\

?

b

.\

)

u 100 m ~ ~ ~ - - ~ - -  ! - --

N u

a.

g y 0 500 1000 1500 2000 2500 3000 3500 4000

? TIME (SEC0t1DS) -

THOUSAtlDTHS Figure 5.5-4A Predicted Pressure History in IHX for SWR DBL in Evaporator O O O

1 O O O 400

- - - - - - -~ ~ - - ~ - -~

---

I------------

350 - - - - - - - - - - - - - - - - -

t

! 300 -

-L-- -- -- - - - -~

--- L -- - -- -

1 1

P I 250

? ,\

C P2 I

" 200

~---- ~ -- ~~ ~ ~'

- - ~ ~ + -~

N 8 /

/

w E 150 Ng a_

i 100

'-^ ~-- - - - - - - -- - - - - - - - '- - --^ ^-- ---

-I ~ --- ' -~~

0 500 1000 1500 2000 2500 3000 3500 4000 4500 EF TIME (SECONDS) -

THOUSANDTHS

?3 g..

$ili Fioure 5.5-4b. Leaksite Pressure History for l Sodium Water P e a.y _' i o n DBL in Eyaporator

[ P0lNT OF IMPACT APPROXIW4TELY AT

.g DISCHARGE END OF V/ HIPPING PIPE q ,

MIDPOINT OF THE 50" SECTION OF THE WHIPPirjG PIPE AND r.11DFPAN OF THE g AT;Cl!01E 0 l'? PACT PIPE TO 525 GALLON 50" PRE 550RE VESSEL TYP WITH GG0"F,2000 PSI A WATER a

f I m

PIPE

~

'{'

m 61" >[ SUPPORT l IMPACT 24"

• PIPE  ;, -

r1 A f~l F~l

'I f

anvil '

k N d l

/ / i..__ 4 8 " ----.-

TYP

>cCTION A-A ee Pe" on Figure 5.5-5. Pipe Whip Test Configuration, Tests 3 And 4 Of Reference 9 e 9 -

9

5.7.1.2 Plant Shutdown Plant shutdown encompasses the operations that take 1he reactor plant from 40%

reactor thermal power to either hot standby or ref ueling conditions. Figures 5.7-C and 5.7-D present the typical shutdown characteristics of the HTS and SGS.

At the 40% thermal power position, the plant control is shifted from automatic load follow to manual control. At this point, steam flow to the condenser via the bypass system is established in order to maintain proper steam presst. e.

Reactor power is then decreased at a rate which lImits the primary hot leg cooldown to 150 F/hr, while the primary and intermediate sodium flows are maintained at flows corresponding to 40% power. The turbine is tripped at l app m imately 10% power cnd the generator output breaker opened. The reactor power is then further decreased to the decay heat level at a rate which limits the primary hot leg cocid e rate to 150 F/hr. During this phase of the shutdown, the superbarcr Meam flow is also reduced and the Protected Air Cooled Concenser (FACC) of the Steam Generator Auxiliary Heat Removal System is placed into operation. When the PACC .3 removing all the decay heat, the lsteambypassflow is secured.

ThereactorandsodiumpumpsarenottrippeduntiltheHTShotandcoldleg) temperatures are approximately equal (at the tot standby condition of 600 F and the control rods are properly positicoed; this minimizes the system temperature transients as the pumps shif t f rom main to pony motors. If the plant is to be maintained at hot standby corditions, the system temperatures p will be controlled by operation of the PACC and electrical resistance heaters.

Q For those cases where the reactor plant temperatures are to be reduced to refueling conditions, the cooldown is accomplished by heat removal from the SGSsteamdrumbgthePACC. This cooldown operation is controlled so as not to exceed the 50 F/hr limit. As in the case of hot standby conditions, the reactor plant is maintained at refueling conditions by operation of the PACC and the electrical resistance heaters.

5.7.2 Load Followino Characteristics The plant is designed to follow changes in load at a maximum rate of 3% of rated thermal power per minute over the range of 40% to 100% of rated thermal power. In addition, the plant is designed to respond to step load changes of i 10% of rated thermal power. The plant follows these load changes in a smooth method and avoids tripping of the reactor and dumping of steam Load fofIow changes are implemented using the control configuration described in Cnapter 7. The capacity to follow load changes is accomplished by automatic adjustment of rod positions in concert with changes to both primary and intermediate sodium flow. Nominal expected steady state temperatures over the range of 40% to 100% rated thermal power (975 MWt) for the primary sodium, Intermediate sodium, steam, feed water, and evaporator water side inlet are presented in Figure 5.7-1. Nominal expected steady state flows over the range of 40% to 100% of rated thermal power for the primary sodium, intermediate sodium, feed water, and cvaporator are presented in Figure 5.7-2. Note in Figure 5.7-2 that evaporator flow is essec.tlally constant over the range of m 40% to 100% of rated thermal power and that primary sodium flow varies linearly with power over the same range. These temperature and flow profiles,

[V )

while not unique (due to possible variation in actual heat transport system 5.7-2 Amend. 62 Nov. 1981 1

B component perf ormance), are typical of those expected f or the 40% to 100%

power operation of CRBRP.

5.7.3 Transient Effects To provi;1e the necessary high degree of integrliy for the equipment in the Heat Transport System, the transient conditions selected f or equipment structural evaluation are based upon a conservative estimate of the magnitude and frequency of the temperature and pressure transients resulting from verious operating conditions in the pl .?. The transients selectea are representative of operating conditions lich presently should be considered to I

O i

l l

O 5.7-2a Amend. 62 flov. 1981 l

\

I i l l 960 - -

g PRIMARY NOT LEG 940 - -

920 - -

INTERMEDIATE NOT LEG 900 - -

880 -

STE AM (TUR8INE THROTTLE) -

860 120 700 - -

PRIMARY COLD LEG 680 - -

C C~

660 - -

640 - -

cc E

E 620 -

Z INTERMEDIATE COLO LEG 600 -

~

~

EVAPOR ATOR INLET (W ATE R) 560 -

540 480

~

460 -

440 -

FEEDWATER 420 -

400 -

I I I I 380 20 40 60 80 100 120 THERMAL POWER (%)

Figure 5.7-1. Expected Loop Thermal Parameters Vs. Power Level Constant Primary Flow to Power Level Ratio Constant Steam Pressure at 1450 psig Linear Primary llot Leg Temperature 5185-2 3

x 5.7-9 Amend. 62 ilov. 1981

15 i j g ,

26 14 -

, = -

2.4 13 -

E v APOR# T O R 22

~

x 12 20 -x I I d 11 -

p ai,3 q ,.

U "a 9

2

'O 16 9 2

5 5 h

~ ~

I INT E RVE DI ATE o

o 8 -

f -

12

{o a o 7 - -

10 "

6

- F E E DW ATE R _ 0g 5

06 4 0.4 1.1 1.7 10 -- / -

16 PRIM AR Y -

09 -

INON BAS (LINtp -

15 M

Bl S"!

to c:g 08 -

1.4

.o a, 07 -  ! ' -

1.3

=t2  %

e smx 06 -

his -

1.2 i

INTE RME DI AT E {g 04 -

10 03 I ' ' '

09 20 40 60 80 100 120 THERMAL POWER PO Figure 5.7-2. Espected Loop Flow Parameten Vs. Power Lesel Constant Primary Flow to Power Lesel Ratio Constant Steam Prewure at 1450 psig Linear Primary llot Leg Temperature

3. ,e - 10 knend. 62 Nov. 1981 e

I

Argon and nitrogen supply lines shall have two automatically initiated containment i sol ation val ves. One valve is located inside the containment, O while the other is located outside of the containment as close as practical to the containment. The valves are back pressure regulated valves which close automatically if the supply side pressure drops below a preset limit. This assures that breaching of the gas system boundary outside of containment results in isolation of the line penetrating containment. Remote manual actuation i s al so provided. Closure of the valves f or loss of supply side pressures assures that f ailures in the system outside of containment combined with postulated events within the containment cannot result in release of radioactive gases.

Nitrogen exhaust line to CAPS shall have two automatically Initiated containment isolation valves. One valve shall be located Inside while the other is outside of contair. ment as close as practical to the containment.

Ranote manual actuation is also provided.

For closed systems penetrating containment, the requirements of GDC 54 and 57 shall be met. Specifically, each line penetrating containment which is part of a closed system other than the lHTS shall have at least one Isolation valve (either automatic, locked closed, or capable of remote manual operation). The isolation valve shall be located outside of containment as close to the containment as practical. Provisions f or testing the operation of the isolation valves and to determine that valve leakage is within acceptable limits shall be included. The valves shall close on loss of electrical signal or loss of air. The valves includej in this category are delineated in Table 6.2-5 and identif ied in Table 6.2-5A.

}

The IHTS boundary is maintained intact as a containment boundary by providing in-depth protection f or the piping and components. The seismic category 1 Steam Generator Building (SGB) provides a barrier capable of withstanding the extremes expected in the environmental conditions. It assures that tornado wind loadings and missiles will not result in damage to the IHTS. The SGB I design also incorporates internal structures to prov!de miss!Ie protection f or the IHTS (from Internally generated missiles), separation between heat removal paths, piping restraints and impingement shields. These design features provide assurance that events initiated externally to the IHTS will not result in loss of the IHTS boundary integrity. The IHTS piping and components are also designed to maintain their integrity following the Saf e Shutdown Earthquake (SSE).

l The integrity of the lHTS boundary will be monitored through a combination of j inservice Inspections, sodium to air leak detection, and maintenance of a 10 ps! (minimum) IHTS to PHTS p in the IHX and leak detection f or IHX leaks (IHTS level probes and IHTS sodium radiation monitoring). These mechanisms provide assurance that leakage between the primary and Intermediate systems, as well as leakage f rom the IHTS to the air will be detected early and corrective action initiated bef ore signif icant amounts of IHTS sodium are leaked.

D 6.2-10a Amend. 62 Nov. 1981

The in-depth protection and Integrity monitoring provisions assure that the containment f unction of the IHTS will be maintained through all expected environmental conditions and external accidents. However, even I f a l oss of l Integrity were to occur it will not result in unacceptable of f-site radiological consequences. Section 15.3.3.3.2 presents the results of an evaluaticn of the potential radiological consequences of a sodium-water reaction occurring with an undetected leak in the IHX. Thi s eval uation shows that the site boundary doses are considerably less than the dose Iimits given in 10CFR20.

Based on the f oregoing, the addition of Isolation valves would add nothing in the way of public safety. In fact, the addition of isolation valves could result in a reduction of the overall saf ety of the CRBRP by reducing the decay heat removal reliabil ity. Inclusion of Isolation valves presents additional f ailure modes which could cause loss of heat rmoval in the affected loop.

The chil led water lines penetrating containment shall have one rmote manual isolation valve Iocated outsIde the containment as ciose as practical to the containment. Since this closed system is not connected to the containment atmosphere nor does it contain radioactive materials, f ailures in the system outside of containment cannot result in release.

The penetrations associated with the Primary Sodium Removal and Cecontamination System shall be isolated to meet the requirements of CRBRP General Design Criteria 45 and 47.

O O

6.2-11 Amend 62 Nov. 1981

6.2.4.3 Design Evaluation The containment isolation features of the design of lines penetra-ting containment provides the necessary assurance that the Gntainment system will provide the barrier to release or psad of rar!ioactive gas or particulate matter. Two basic types of containment isolation are provided: closure of lines directly connected to the containment atmosphere; and closure of lines of closed systems.

Automctic isolation of the lines connected to the containment atmosphere is provided by the design. The initiation features described in Secticn 7.3 and the isolation valves for the contaireent ventilation lines, air lines, and vacuum breakers respond to the nsuMd events to prevent releases in excess of the guidelines of 10CFR100 for sodium fires or assumed increases in pressure at.d activity levels.

Earlier in the design both temperature and pressure were considered for initiation of the CIS; however, both were eliminated.

Calculations show that substantial (for detectionpurposes) pressure and/or .

temperature rises are delayed many hours af ter the initiation of accidents l studies, as shown in Figures 6.2-2 & 6.2-3. Further, the maxincn cressure is in the range of 2 psig and the maximum temperature is 250 F. In addition the direct measurement of the limiting parameter, which is radioactivity in this case, is the most desirable. It should be noted that containment isolation can be initiated by either a high radiation signal from the head access area radiation monitors, or a high radiation signal from the containment exhaust radiation monitors. Two isolation 25 valves, with independent actuating trains, are provided. One valve is inside containment; and ene outside as close as practical. This redundancy assures proper isolation assuming single internal random failures of the equipment. Periodic on-line testing capabilities are included. The valves and associated actuators are located in area; which are protected from tornado generated missiles and which are designed to withstand tne seismic forces.

For the lines connected to the reactor coolant boundary, the two valves, either manually or automatically actuated as appropriate, provide the necessary protection. The different classes of the lines are separately discussed below.

l The argon and nitrogen supply line valves provide a double barrier which is automatically activated on loss of the ex-containment boundary.

The valves and associated actuators are located in protected areas and are testable. Remote and local manual initiations are provided.

Amend. 25 i n 6.2-13 Aug. 1976 U

,.- ---.-.-..,---._~---.--..--,,--.,,,,--,,_.y.m

. m., , - - - , , , - , _ . . . _ , - - , . - . . _ , _ _ . , - - - .---,-.cm , , , ~~__.w- w

The nffrogen exhaust line to CAPS has two automatically initiated valves. The valves provide two barriers following closure. The valves and associated acttators are located in protected areas and are testable.

i For the remainder of the penetrations, two valves are provided as barriers to l release. Manual initiaF on will ce adequate to prevent releases exceeding the guicelIne values.

I i

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O 6.2-13a Amend. 62 Nov. 1981 m._____...___.. .- _ ____. _ _. _ _ _ . _ _ _ _ _ _ __ _ _ _ . _ . _ _ _ _ _ _ _ . ~ _ _ _

[

- t J TABLE 6.2-5. Lines Penetrating Containment c 3n a %E %S 3 .

. EES ha a 3  : Be% B 2d

+ 0 . c #5 E Sc Bob c sc 8%-

e 2 +2 2 E8 8%3 +32 2*8 m2 %2 .233 3 tt & "  %; "; Let %t .Eu  %% 7% h".'

Et ta Ka 8 a& 08; ;EE K3a 435 S3 83 EE 48 Penetration E d :2 8 @$ 3 #G 332 f#8 C#$ s uiE i# J!# GE s$

I DecontamI nation Ranote Waste Water Aut& Auto-Return 9.2 2 Gate 3" CIS Closed Closed mat;c Open matic Manual <4 B IHTS Piping loop No. 2 Inlet 5.4 0 N/A 24" N/A N/A N/A N/A N/A N/A N/A N/A N/A lHTS Piping Loop No. 2 Outlet 5.4 0 N/A 24" N/A N/A N/A N/A N/A N/A N/A N/A N/A IHTS Piping

• Loop No. 3 N/A N/A 5.4 0 N/A 24" N/A N/A N/A w/A N/A M/A N/A

. Inlet N

E N

IHTS Piping L Loop No. 3 Outlet 5.4 0 N/A 24" N/A N/A N/A N/A N/A N/A N/A N/A N/A IHTS Piping Loop No. 1 Inte 5.4 0 N/A 24" N/A N/A N/A N/A N/A N/A N/A N/A N/A iHTS Piping Loop No. 1

, Outl et 5.4 0 N/A 24" N/A N/A N/A N/A N/A N/A N/A N/A N/A 1

< m

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8

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to Ev-Cont.

Stor. Tank) 9.3 1 Globe 4" N/A N/A Ci cce d Mareal Closer Manual C/ A <30 C Sedita Trensfer Lfno (EVS Filt & Drain) 9.3 1 Globo 3" N/A N/A Cl ose d Manual Jicsed Manual N/A <30 C l NaK ChRS from Fall in (Neot e R r e.ot e 6" N/A Place Cl ose d Manual Cicsed Manual Parm . <30 H Coi.t .,i nm, n t 9.3 1 G icAo haK CHRS To Fall in Th roto Rtret e Cc nta i nnen t 9.3 1 Globe 6" N/A Pleco C l ose d Manual Ci cced Manual Manual <30 H cn m TUPS t Cold Recote Aute- Fmcte j 61l f$cx 9.5 2 Globo 1-1/2" CIS Closed Closed Manual Open ratic M:nual <10 F l c' .

N cv Pmot e Auto- D ur-o t e CAPS Inlet Closed Cicsod Manual Open ratic Manua! <10 F 61lI Heade' 9.5 2 G!cbo 3" Cl!

P(rote A u t o- Pm.c t o RAPS to Recycio 61 1l.; , Argon Vesse! 9.5 2 Gi ct'o 1-1/2" CIS Cicsnd Cl osed Manual Open ratic Manual <10 F l

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TABLE 6.2-5A O

SUMMARY

OF CONTAINMENT ISOL ATION VALVING CATEGORIES AND APPLICABLE GDC

1. Lines connected directly to the containment atmesphere (Items 1-3) are provided with two isolation valves (one inside and one outside), which are automatically closed when necessary. This is in conf ormance with CRBRP GDC 47.

2. Lines connected to the reactor coolant boundary, primary cover gas spaces, or Inerted cell atmospheres (Items 8-13) are provided with two isolation valves (one h. side and one outside) which have automatic or remote manual closure activation. This is in conf ormance with 'RBRP GDC 46.

3. Lines of closed systems which penetrate the containment and which are neither connected to the reactor cocian > boundary nor to the containaent atmosphere (items 6, 7,14,15,18, and 19) are provided with a single valve located outside of and close to the containment as practical. This is in conformance with CRBRP GDC 48. I The intermediate Heat Transport System lines are not provided with Isolation valves. Justification of this position is included in Section 6.2.4.1.

4 Lines 4, 5,16 and 17 are provided witn two isolation valves, one outside and one inside.

5. The design of the Containment isolation System will provide the capability of remote-manually closing all the containment isolation valves from the Control Roan.

6. Indication of valve position status of all the containment isolation valves will be provided in the Control Room.

O 6.2-28 Amend. 62 Nov. 1981

TABLE 6.2-6 TABLE OF BYPASS LEAK PATHS Type Penetration Service Leakaae (lbs/ day)

[I Decontamination Waste hater Return .0024 Argon Gasline RAPS .0016 Aracn Gasline RAPS .00245 RAPS Bypass to CAPS .0008 Post Accident Monitoring Sample .0008 Post Accident Monitoring Sample .0008 Post Accident Monitoring Sample .0008 Post Accident Monitoring Sample .0003 Post Accident Monitoring Sample .0008 Post Accident Monitoring Sample .0008 Instrument Air .0016

Breathing Air .0016 Service Air .0016 Decontamination Water Supply .0024 Arcon Suppl 3 Line .0016 Argon Recycle Line .0016 Argon Cover Gasline (Sampling) .0008 N Supply Line .0016 2

Floor Drain Sump Discharge .0048 Ventilation Air Exhaust .0038 Ventilation Air Supply .0038 Containment Purge Line .0038 Containment Purge Line .0038 l

Containment Vent Line .0028 Containment Vent Line .0028 Normal Chill Water .0064 Normal Chill Water .0064 Normal Chill Water .0032 Normal Chill Water .0032 Normal Chill Water .0032 '

Herial Chili iL ter 0132 Nnr a', Chill 'later '932 Amend. 30 6.2-?86 Nov. 1976

7.1.2.2 Indeoendence of Redundant Safety Related Systems To assure that independence of redundant saf ety related equipment is preserved, the following specif ic physical separation criteria are liriposed f or saf ety related instrumentation.

o All Interrack PPS wiring shall be run in conduits (or equivalent) with wiring for redundant channels run in separate conduits. Only PPS wiring shall be included in these conduits. Primary RSS wiring shall not be run in the same conduit as secondary RSS wiring. Wiring f or the CIS may be run in conduits containing either primary RSS wiring or conduits containing secondary shutdown system wiring, but never i ni erm i xed. Expanded criteria for physical separation of the CIS are given in Section 7.3.2.2.

o Wiring for other safety related systems may be run in conduits I containing either primary RSS wiring or conduits containing secondary RSS wiring, but never intermixed, provided that no degradation of the separation between primary and secondary RSS results, o Wiring for redundant channels shall be brought through separate containment penetrations with only PPS wiring brought through these penetrations. Primary RSS wiring shall not be brought through the same penetration as secondary RSS wiring. Wiring for the CIS and other saf ety related systems will be brought through the same penetration as the RSS wiring with which it is routed.

o instrumentation equipment associated with redundant channels shall be mounted in separate racks (or completely, metallically enclosed ccopar tments ) . Only PPS channel instrumentation shalI be mounted in l these racks. Primary RSS equipment shall not be located in the same rack as Secondary RSS equipment.

l o The physical separation between conduits, penetrations, or racks containing redundant Instrument channels shalI be spectfled on an Individual case basis to meet the requirements of Regulatory Guide 1.75. This separation shall provide assurance that credible single events do not simultaneously degrade redundant channels or redundant shutdown systems.

o The wiring f rom a PPS buf fered output which is used f or a non-PPS purpose may be included in the same rack as PPS equipment. The PPS wiring shall be physically separatea f rcm the non-PPS wiring. The amount of separation shal1 meet the requiranents of IEEE 384-1974.

o Electrical power f or redundant PPS equipment shall be supplied f rce separate sources such that f ailure of a single power source O

7.1-3 Amend. 62 Nov. 1981

does not cause f ailure of more than one redundant channel . The power sources and associated wiring shal I be separated, as specif led in Section 8.

The criteria fe cable tray fill, cable derating, cable routing in congested or hcstile areas, fire detection and protection in cable areas, and cable markings are defined in Section 8. Separation of redundant safety related equipment within the control boards is described in Section 7.9.

7.1.2.3 Physicai Identification of Safety Related Eculoment The Plant Protection Systan equipment will be identified distinctively as being in the protection systern. This identification will distinguish between redundent oortions of the protection system such that quallfled personnel can distinguisn whether the equipment is safety related and, it so, which channel.

Color coding, cabinet and wire labeling and other techniques as appropriate will be used.

7.1.2.4 Conformance to Regulatorv Guides 1.11 " Instrument Lines Penetrating Primary Reactor Containment" and 1.63 " Electric Penetration Assem-olles in Containment Structures for Watercooled Nuclear Power Plants" There are no instrument lines as def ined in Regulatory Guide 1.11 which penetrate primary reactor containment. All electric penetration assemblies in the containment vessel will be designed, constructed and Installed in accordance with Regulatory Guide 1.63 and IEEE Standard 317-1972.

7.1.2.5 Conformance to IEEE Standard 323-1974 "lEEE Standard for Qualifsing Class IE Eculoment for Nuclear Power Generating Stations" i All Class IE equipment will be qualified to confirm the adequacy of the equipment design under normal, abnormal, and postulated accident conditions for the perf ormance Class IE f unctions. This will be accomplished through a disciplined progran discussed in Reference 13 of PSAR Section 1.6, "CRBRP 61 Requiranents for Environmental Qualification of Class 1E Equipment."

l 7.1.2.6 Conformance to IEEE Standard 336-1971 " installation. Insoection and Iesting Recuirements for Instrumentation and Electric Eculoment During the Construction of Nuclear Power Generating Stations" )

l The installation, insoection and testing of the instrumentation, electrical and electronic equipment during construction will conform to the requiranents of IEEE Standard 336-1971. The quality assurance progran for the safety related Instrumentation and control equipment wilI conform to the requiranents of j Regulatory Guide 1.30. Ref er to Chapter 17 for a description of the quality assurance progran.

l O

! 7.1-4 Amend. 61 l

Sept. 1981

1. e Table 7.2-2 e

)

PPS Design Basis Fault Events fault Events Prirrary_Shutdewn Sys tem Secondarv Shutdown System

l. Ant ic i patgtLfau.l tg III A. Reactivity Disturbances Positive Ramps 5 /sec and Steps 10 Startup Flux-Delayed Flux or StarTup Nuclear Flux- Pressure 5-40% Power Flux-Delayed Flux or Meditled Nuclear Rate or Flux- Pressure Flux-Total Flow 40-100% Power Flux- Pressure Flux-Total Flow y Full Power High Flux Flux-Total Flow Negative Ramps and Steps Flux-Delayed Flux Moditsed Nuclear RaTo e

B. Sodium Flow Disturbances Coastdown of a Single Primary or Primary-intermediate Primary-intermediate Intermediate Pump Speed Ratio Flow Ratio Loss of 1 HTS Loop Pump Electrics Primary-intermodlate Flow Ratio Loss of 3 HTS Loops Pump Electrics Flux-Total Flow t

i l 5ea i O

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9 e e TABLE 7.2-2 (Continued)

Primary Reactor Shutdown System Secondary Reactor Snutdown System I

57 F uit Events C. Steam Side Disturbances Evaporator Module Isolation Valve IHX Primary Outlet Evaporator Outlet Na Closure Temperature Temperature Superheater Module Isolation Valve Steam-Feedwater Flow Evaporator Outlet Na Closure Mismatch Temperature L

Water Side Isolation and Dump IHX Primary Outlet Evaporator Outlet Na of Single Evaporator Temperature Temperature  !

Water Side Isolation and Dump Steam-Feedwater Flow Evaporator Outlet Na '

of Single Superheater llismatch Temperature Water Side Isolation and Dump of Steam-Feedwater Flow Evaporator Outlet Na i

[ Both Evaporators and Superheater Mismatch Temperature 4

o Loss of Normal Feedwater Steam-Feedwater Flow Steam Drum Level Mismatch ]i Turbine Trip with Reactor Trip Steam-Feedwater Flow Steam Drum Level  !

47l Mismatch (Loss of Main Condenser or t Similar Problem)

Inadvertent Opening of Evaporator Steam-Feedwater Flow Steam Drum level Outlet Safety Valve Mismatch Inadvertent Opening of Superheater Steam-Feedwater Flow Steam Drum Level Outlet Safety Valve Mismatch Inadvertent Opening of Evaporator IHX Primary Outlet Evaporator Outlet Na SE Temperature Temperature 5[

g Inlet Dump Valve t

80 i

~

_ _ . _ _ . - . _ _ _ . .. _ ..-____-- - - - - - - . ._. _ _ _ . _ _.-.__ _ - . _ _ - - _ . . . _ - _ _ . _ l

i j Table 7.2-2 (Continued)

Lidflyenh EcJmarX_Shuldewn_ System Sec h ary_,Shutdcwn Systen II. L'n ! Ikely_fAulb A. Reactivity Disturbances (2)

Positive Ramps 35 /sec and Steps 60 l Startup Flux-Delayed Flux or Startup tiuclear j Flux- Pressure

, 5-40% Power Flux-Delayed Flux or Modified Nuclear Rate os Flux- Pressure Flux-Total flow l

40-100% Power Flux-Pressure Flux-Total Ficw j Full Power High Flux Flux-Total Flow i

1 B. Scdlum Ficw Disturbances I

j Primary Pump Seizur' Pr itrar y-i nt ermed i ate Pr ic.ary-i nterrediate Flow N Speed Ratlo Ratto N

Interrrediate Pump Seizure Primary-intermediate Pr irrary-intermediate Flow Speed Ratio Ratio a

Loss of 2 HTS Loops Purr.p Electrics Pritrary-intermediate Flow Ratio

,i C. Steam Side Disturb nces Steam Line Dreak(3) Steam-Feedwater Flow Evaporator Outlet Na Mismatch Terrporat ure Recirculation Line BreakI3) Stocm-feedwater Flow Steam Drum Level j Mismatch i <-

O (G D Feedwater Line Dreak(3) Steam-Feedwater Flow Steam Drum Level

• 5.

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Mismatch

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3 8 9 9 Table 7.2-2 (Continued)

Fault Even11 fr_} maryJ hut dagILS_ystem Secondaryj hutdown System Failure of Steam Dump System Steam-FeedwaTer Flow Steam Drum Level Mismatch i

l Sodlum Water Reaction in Steam Steam-Feedwater Flow Sodlum-Water Reaction Generator Mismatch 111. Extremely Unlike.l.y A. Reactivity Disturbances 1

I Positive Ramps 52.0/sec l

I Startup Flux-Delayed Flux StarTup Nuclear  ;

5-40% Power Flux-Dolayed Flux or Moditted Nuclear Rate or Flux- Pressure Flux-Total Flow 40-100% Power Flux- Pressure Flux-Total Flow j u

m Full Power High Flux Flux-Total Flow l i

(1) The maximum anticipated reactivity fault results from a single failure of the control system with a maximum insertion rate of approximately 4.1 cents per second.

(2) The maximum unlikely reactivity faults result from multiple control system failures leading to withdrawl of six rods aT nor mal speed or one rod at the maximum mechanical speed. -

! (3) The PPS is required to terminate the results of these extremely unlikely events within the umbrella transionT spectried as f l emergency for the design of the major components.  !

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TABLE 7.2-3 ESSENTIAL PERFORMANCE REQUIREMENTS FOR PPS INSTRUMEf1TATION Accuracy Response Time 57l Plant Parameter (b of span) (msec)

Neutron Flux Primary tl.0 <10 Secondary -1.0 10 Reactor Inlet Plenun Pressure 2.0 <l50 Sodium HTS Pump Speeds t2.0 :20 Sodium HTS Flow 10.0 <500 Reactor Vessel Sodium Level 15.0 <500 i

Undervoltage Relay 13.0 <200 Steam Flow t 2. 0 < 500 ,

Feedwater Flow 12.5 - 50 0 Evaporator Outlet Sodium Temperature 22.0 <5000 57 Steam Drum Level +l.0 <1000 .

l IHX Primary Outlet Temperature +2.0 c5000 9/

t Amend. 57 Nov. 1980 7.2-23

Head Access Area Radiation The Head Access Area Radiation Subsystem initiates closure of the f'^ containment isolation valves in the event of large radiation releases in the

\ head access area. Three radiation sensors are located in the head access area to provide early initiation and closure of the isolation vnives to assure that releases from design basis events do not excead the guideline of 10CFR100.

7.3.1.2.2 Essential Performance Requirements To implement the required isolation function within the specified limits, the CIS must meet the functional requirements specified below:

The closure time requirement for the inlet and exhaust isolatico valves is 4 seconds with a thrce second or less detection time in the heating and ventilating system. A 10 second transport time frca sensing point to the valve exists (see Section 15.1.1). The 3 seconds includes 43l sensor time respcnse, comparator and logic time delays. 30 The CIS is designed to meet these requirements for _the 57 environmentaT conditions described in Section 7.2.1.

7.3.2 Analysis The design of the CIS provides the necessary functional performance and design features to meet the requirements of the appropriate standards O specified in 7.1.2 as descri bed below.

7.3.2.1 Functional Performance The analyses in Sections 15.5 and 15.6 shows the results of the postulated fault conditions. These analyses assumed a closed containment where the events occurred with the containment hatch closed. For the limit-ing event, primary drain tank fire during maintenance, scoping analyses have been performed to determine the required closure time of the containment isolation valves. For the primary drain tank fire, closure within 20 minutes is adequate. Further, analyses to determine the required closure time under postulated accident conditions have been performed and are discussed in Section 15.1.1. These analyses are used to determine the available design margin. The results of this assumed condition do not axceed the guideline values of 10CFR100 if the main exhaust and inlet valves are closed within 4 seconds assuming the normal air transport time from the detector to the valve is 10 seconds or more, a '14,000 cfm normal ventilation rate.~~

57 i

I

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l (s- Amend. 57 l

N v. 0 7.3-3

Since the automatic Containment isolation System is designed to isolate within the above time response requirements, al I of the design basis conditions are terminated within the necessary limits f or the present design concept.

7.3.2.2 Design Features The Containment isolation Systam instrumentation, control s and actuators are designed to meet the requirements of IEEE-279-1971 and RDT Standard C16-IT, Dec. 1969. Tne analyses of compliance with these are sunmarized below.

Single Failure No single f al!ure within the CIS nor ranovel from service of any component or channel will prevent protective action when required. There are three independent instrument channel t f or each necessary measurement, four independent 2/3 logics, and two inJependent actuatcrs provided (as shown in Fi gur e 7.3-1 ) .

Byoasses No bypasses are provided.

Multiole Setoolnts Multiple setpoints are not required.

Ccmoletion of Protective Action The CIS is designed so that, once initiated, protective action at the system level must go to completion. Return to normal operation requires manual reset by the operator.

Manual Initiation Tne Cls includes means f or manual initiation of containment isolation at the system level. No single f ail ure wil l prevent manual initiation of the action.

Control and Protection Interaction There are no shared components between the control system and the CIS.

The prev isions f or access, inf ormation read-out, annunciation of trips, and periodic testing are as specif ied f or the Reactor Shutdown Systcm in Section 7.2.2.

' Physical Seoeration The f ollowing criteria assure physical separation f or the CIS.

There wil l be at least one containmeni penetration f or cach of the three Primary PPS Instrument channel conduits and each of the three Secondary PPS

, instrument channel conduits which exit containment. / I l requi rements f or i

separation of PPS wiring through conduits wil l also apply to separation of PPS j wiring through containment penetrations.

l 7.3-4 Amend. 62 Nov. 1981

1 J

O There are three categories of CIS cabling: cables betwren the radiation  ;

monitoring sensors and logic panels; cabling between the logic panels and the

. power breakers; and cabling from the breakers to the valve actuators.

Wiring for the three CIS instrument channels will be routed exclusively with the three Secondary PPS instrument channels.

CIS logic train actuation wiring will be routed through two separated and independent conduits. A conduit will contain only wiring from a single CIS

logic train. No intermixing of CIS logic trains within a conduit will be l

permitted. CIS logic train I wiring will be routed f rom CIS logic panel 1 to CIS breaker 1. CIS logic train 2 wiring will be routed from CIS logic panel 2 4

to CIS breaker 2.

l All of the inside containment isolation valve actuation wiring (both manual and automatic) will be routed through at least one separated and independent l conduit f rom CIS breaker 1 through a separate and independent containment isolation valve actuation containment penetration. inside containment isolation valve actuation wiring will be routed through separate and independent conduits from the inside of the containment Isolation valve actuation containment penetration to the Individual containment isolation valves. No other wiring will be routed through the conduit and containment penetration containing inside containment isolation valve actuation wiring.

( All of the outside containment isolation valve actuation wiring (both manual and automatic) w!ll be r outed through at least one separated and independent conduit from CIS breaker 2 to the Individual outside containment isolation valves. No other wiring will be routed through the conduit containing cutside j containment isolation valve actuation wiring.

i i

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7.3-4 a i

Amend. 62 '

Nov. 1981 i

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

1 O (2) Overflow vessel sodium level

• ( 3) Reactor overflow sodium ini t temperature (4) EVST sodium pump control signal (each pump)

( 5) EVST NaK pump control signal (each pump)

(6) Primary makeup pump control signal (each pump) 46 50 The following EVST cooling and DHRS process variables are moni-tered using a single sensor and redundant cabling and panels.

(1) EVST sodium flowrate (each loop)

(2) EVST NaK flov ate (each loop)

(3) Primary overflow makup sodium flowrate (each pump outlet) 46l ' (4) EVST airblast heat exchanger fan speed setting (each loop)

(5) EVST NaK expansion tank level (each loop)

(6) EVST sodium inlet temperature (each loop)

(7) EVST sodium and NaK remotely operated valve position i indicators (each loop)

(8) EVST airblast heat exchanger damper position indicator (each loop) 1 (9) DHRS heat exchanger bypass valve position indicator 46l (10) DHRS NaK expansion tank level C. Annunciation and Data Handling The following EVST cooling and DHRS process variables are con-

~

nected to the plant annunciator system to alert the plant operator of off-normal conditions:

i (1) Low sodium and NaK pump gas cooling flow rate (each pump) 46 (2) High EVST sodium, EVST NaK and primary makeup pump stator temperatures (each loop) l 26 7.6-3b Amend. 50 June 1979

{

I

(3) Low EVST sodium, EVST NaK and primary makeup flowrate (each pump loop)

(4) High and low EVST sodium inlet temperature (each loop)

(5) High and low EVST NaK expansion tank level (each loop)

(6) High and low EVST sodium level (7) High and low EVST sodium temperature (8) Low sodium valve temperatures (9) High and low DHRS expansion tank level Key process variables that are connec+ed to annunc)Stors are also connected to the plant data handling and display ofstem.

D. Other Features Remotely operated valves in EVST cooling and DHRS circuits incorporate either "falI safe" or "fafI in place" features and are provided with direct manual (reach rods on sodium valves) override capability in event of l&C or gas supply failures. DHRS valves are provided with acc.mulators to provide 1/2 hour startup capability for a period of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> af ter ha gas supply is lost.

Type IE power is supplied to the equipment and instrumentation required to provide the safety related f unctions of EVST cooling and DHRS as shown in Figure 7.6-13. This assures independence of off-site power.

Functional testing of all portion of DHRS that are not used during the course of normal operations will be tested on an annual basis during reactor refueling.

Equipment required to provide power to EVST and DHRS pumps, airblast heat exchangers and the monitoring instrumentation in the control panels shall be designed and tested to Seismic Category I requirem;nts.

Type IE temperature sensors are provided at the EVST sodium and NaK pump field control panels to alert (anrunciate) the operator of loss of cell cooling so that EVST cooling service can be transf erred to the standby circui t.

7.6.3.1.4 Initiating Circuits Reactor decay heat removal through DHRS is initiated ' rom the Cor, trol Room panel as described in Sections 9.1.3 and 9.3.2.

7.6.3.1.5 Bvoass and Interlocks When the DHRS is activated, automatic control of the EVST airblast heat exchangers is bypassed so that control of the complete system is manual. AlI valves in this circuit are also operated on a d' rect or remote manual basis.

l The flow in the primary sodium overflow makeup 9

7.6-3c Amend. 62 Nov. 1981

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1 Amend. 62 9.3-21 Nov. 1981 I

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i 9.4 PIPING AND EQUIPMENT ELECTRICAL HEATING f 9.4.1 Design Basis The piping and Equipment Electrical Heating System provides the electrical heaters, electrical heater mounting hardware, heater power con-trollers and the related temperature measuring and controlling instrumenta-tion and equipment required to heat the following sodium containing process i systems and components:

Reactor Enclosure

! Reactor Refueling (Storage Tank)

Reactor Heat Transport (Primary and Intermediate) Systens 47l Steam Generation System (Dump Tanks and Sodium Water Reaction Product Tanks)

Auxiliary Liquid Metal System Inert Gas Receiving and Processing System Sodium Impurity Monitoring System This heat is required to preheat these sodium process systems prior to fill, to prevent sodium freezing when syster. heat sources such as reactor decay heat and pumping heat become insufficient, and to maintain pre-established temperature di#farances in the system.

To perform the dry heat-up function, the tlectrical Heating System shall be capable of preheating the sodium process systems from ambient tem-perature (70 F) to any temperature between ambient and a maximum of approximate 1;, 450 F before the system is filled with sodium, at a rate j determined by the particular sodium process system requirements.

l The Electrical Heating System shall also be capable of providing the applicable heatup rate for the particular system or components when filled with sodium, and of holding process system temperatures when filled with sodium. Heat provided by this system can be used to melt frozen sodium in piping or components. Freezing of sodium in major systems or components is considered unlikely and is an abnormal event. Melting of frozen sodium is not safety related.

t The heater physical mounting arrangement and the electrical protec-tion of the heater circuitry shall be designed to preclude damage to the

) components being heated.

i Heaters and the associated mounting hardware that are applied to

! components which are safety related shall be designed not to impair the ability of these components to perform their safety function during or after

, a design basis event. Those safety related components which require heaters are listed in Table 9.4-1.

Amend. 47 l

Nov. 1978 i

9.4-1

9.4.2 Svstem Descriotion The electrical heating and control system provides power to the tubular heaters or mineral Insulated (MI) heating cable mounted on the piping and/or components of the systems indicated in Section 9.4.1.

The heat rates required by different components are controlled by using therinocouples to monitor olping and component temperatures and to adjust the power supplled to the heaters, by means of 3 mode proportional temperature controllers and solid state relays.

Tubular heaters apply heat via a spiral wound nickel-chromium alloy resistance wire insulated from its containing metal tubular sheath by tightly racked Magnesia (Mg0) powder. Several inches on each end of each heater are unheated having a heavy electrical conductor to the electrical termination. I n cer ta i n cases, i.e., selected piping smaller than 10 inches 0.D., the heat is appiled by mineral insulted heating cable that consist of a metal sheath drawn down over a M 0 Insulated single heating element of nickel-chromium-Iron wiro.

G The heaters will be stood off from the sodium containing metal boundary for the safety-related piping and equipment which is Iisted in Table 9.4-1. The heater sheath is not electrically insulated frora the metal boundary. For systems and components not listed in Table 9.4-1, the heaters may be either stoodof f or applied directly in contact with the sodium containing metal boundary. Groundfeult Interrupt (GFI) circuits prevent metal damage in the uniIkely event of heater arcing.

Shorting of a heater element to the heater sheath and the possible subsequent arcing to the component is further prevented by the insertion of ground current detection and interrupting devices in the heater circuits. These devices automatically open the heater circuit if a short is Indicated by the ground current exceeding a pre-set value.

Heater power is controlled by a pre-programmed, direct digital control system.

Three individual systems are provided; one for the Reactor Containment Building, one for the Steam Generator Building and one for the Reactor Service Building.

These three subsystems can be controlled by a control terminal in the main control room, or each subsystem can be controlled at the central processing center in each building, in addition, the operator can take manual control of the Individual control zone at the local panels.

l The control system is reprogrammable, on-line or off-line, permitting the creation or deletion of complex control strategies such as cascading or sequenc!ng of systems wIthout the necessity of wiring changes.

The controller compares a temperature control setting (or ramp rate in the heat-up mode) as set by the plant operator, to the actual temperature, and generates an error signal. A trigger unit converts the error signal into a corresponding output which controls the AC currer.t to the heaters.

The digital control system, in addition to controlling the heaters, has an inherent feature to scan and alarm all the monitoring thermocouples.

i 9.4-2 Amend. 62 Nov. 1981

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9.10 COMPRESSED GAS SYSTEM V The Compressed Gas Systems discussed in this section are those which supply Instrument air, service and breathing air, hydrogen for generator cooling, and carbon dioxide for generator purging. Instrument and Service Air Systems are depicted in Figure 9.10-1. The Compressed Air System Safety Class lil Instrument Air Supply is shown in Figure 9.10-2. 'A final drawing of this system will be provided upor completion of desip. The Hydrogen System is shown in Figure 9.10-3, and carbon dioxide system in Figure 9.10-4.

9.10.1 Service Air and Instrument Air Svstems 9.10.1.1 Design Basis The Service Air System is designed to provide clean, oil - free, compressed air which will be used to:

1) Provide air necessary for various maintenance functions.

2) Provide breathable air to required stations.

3) Provide air for the Instrument Air System.

To fulflil these requirements, the system components and piping are designed, fabricated and. inspected in accordance with applicable codes as follows:

o Air receiver tanks, fliter bodies, drying chambers, moisture

() separators, intercoolers/aftercoolers meet ASME Boller and Pressure Vessel Code, Section Vill, Division 1.

o Piping (except containment penetration piping and Isolation valves, instrument air piping and accumulators serving safety related components) meets ANSI B31.1.

o Containment penetration piping and Isolation valves meet ASME Section iIi, Class 2 and Seismic Category 1.

o Instrument air piping and accumulators serving safety related components meet ASME Section Ill, Class 3, Seismic Category 1.

o in addition, service and Instrument air equipment piping and components meet WARD-D-0037 (Appendix 3.7-A), Seismic Design Criteria for Clinch River Breeder Reactor Plant".

Environmental design requirements will moet those for saf ety related equipment discussed in Section 3.11.

O V

9.10-1 Amend. 62 Nov. 1981

9.10.1.2 Eystem Descriotion The Instrument and Service Air System equipment is located in the Turbine Generator Building.

Compressed air for these systems is provided by three 100 percent capacity, oil free compressors artanged in parallel and supplying air to a ccomon discharge header and twin air receivers. The system shall be maintained at a pressure of 165 1 15 psig by the use of a pressure control valve in the distribution header. Each compressor includes: Inteke fi:ter-sliencer, intercooler, aftercooler, moisture separator, and automatic control equipment.

A distribution header, downstream of the air receivers, supplies air to service air anc breathing air stations, and to the Instrument air system.

Service air piping to the instrument air system passes through one of two full capacity prefilters, one of twc full capacity air dryers and filtered by one of two fui t capacity afterfilters. The instrument air is then distributed to various stat Mns located throughout the plant as required.

For those locations where loss of instrument air could result in the loss of safety function of another component, local air accumulators, with dual automatic Isolation valves, are provided. The basis for the size of these air accumulators is provided in the PSAR sections which describe the safety systems supplied by the Instrument air system. The items generally considered in sizing these air accumulators include:

1) the duration through which operation is required,

2) air losses,

3) minimum pressure needed to operate the valve or instrument, and

4) adequate mergins.

Air dryers are installed in the Instrument Air System which dry the Instrument air to a -40 F dewpoir.+. Profilters and afterfilters are located in the inlet and outlet p! ping of the instrument air dryers, and remove all particles I approximately 7.5 microns or larger ;n size, and 98% (by weight) of all particles 3 microns or larger in size. Periodic maintenance checks of all filters, coupled with redundancy in design of flow pat. s, ensure continuous systen re l iab (I !ty (see Figure 9.10-1 ) .

l Automatic back pressure valves are used to isolate the Containment Subsystem of the Plant Protectic.7 System as described in Section 9.10.1.3. Two of these valves are located at +he Instrument Air Penetration and two valves are located at the Service Air' Penetration. One valve of each pair is within the Reactor Containment Building and one valve of the remaining pair is outside the Containment Building. These back pressure control valves can also be closed by manual switches on 359 Main Control Panel.

9.10.1.3 Safety Evaluation l

Under normal plant use, one compressor is in operation; a second compressor is on autcmatic standby for large load swings and the third compressor serves as Amend. 62 9.10-2 Nov. 1981

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Amend. 62 Nov. 1981

9.11 Communications System The Clinch River Breeder Reactor Plant is provided w! 5 the following comunication systems providing ef f ective and diversified means of communication between plant personnel in all vital areas during the full spectrum of accident or incident conditions under maximum potential noise levels:

a. Public Address intra-Plant Communications (PA-IC)

b. Private Automatic Exchange (PAX)

c. Microwave Communications

d. Powerline Carrier Communications (PLC)

e. Maintenance Communications Jacking (MCJ)

f. VHF Radio Station

g. Portable Radio

h. Manual Telephone Switchboard I. Offsite Law Enforcement Radio J. Security Intercom

k. Security Portable Radio System

l. Supervisory Control and Data Acquisitions (SCADA) Remote Terminal Unit (RTU) 9.11.1 Design Bases

a. The Public Address Intra-Plant Communications System (PA-lC) shall:

1. Provide primary communications throughout the plant.

2. Insure availability by being powered from the instrument AC bus which is capable of receiving power from offsite and normal AC power supply or the Non-Class IE station batteries. The PA-IC system equipment shalI be mounted on seismically qualifled supports when located in Seismic Category I structures. The power and sound circuits are arranged so that any disruption of the system in the non-seismic Category I areas does not affect the operation of the system in the seismic Category 1 areas.

3. Provide fire, high radiation, and evacuation alarms throughout the plant to the plant personnel by means of manually actuated multitone generator signals broadcast through the page channel.

3 Amend. 62 9.11-1 Nov. 1981

I

b. The Private Automatic Exchange (PAX) shall provide intra plant communications in all plant areas and inter p! ant communications between the plant and key locations within the TVA system. Office areas shall be supplemented with a key-telephone systen.

c. The Microwave Communications System shall be the primary inter-plant communications system between CRRRP and other TVA facilities, through the existing TVA microwave communications netwcrk.

d. The Powerline Carrier Communications System shall be an alternate inter plant communications system to the Microwave Communications System. 29

! e. The Maintenance Communications Jacking System (MCJ) shall:

1. Provide sound powered communications between the Control Room, local instrument panels in all plant areas for supporting maintenance and instrument calibration activities, and pre-selected stations for fixed emergency communications. 48

2. Provide sound powered communications between the Plant Control System Steam Generator Building remote shutdown panel and all "7

local panels required for the support of remote plant shutdown.

f. The VHF Radio Station shall provide direct communication with the '

l TVA Emergency Staff Operations Office in Chattanooga, Tenn.

g. The Fortable Radio Communications System shall provide a " Total Area Coverage (TAC)" portable two-way radio communication system in all areas of the plant, both inside and outside the buildings for traffic control, maintenance operations, fire control and general lI",8 communications.

h. The Manual Telephone Switchboard shall provide an inter plant communication terminal in the Control Room. Call director type telephones shall be provided for the shift engineer anu unit operator as an extension of the switchboard- '

i. The Offsite Law Enfo;rement Radio 53 m.m shall provide an independent voice channel to the lacal offsite L enfo' cement agency.

l j. The Security Intercom System shall provide communications bot een remote security stations and shal; prov He tone paging as required 47 for use t;y plant secrity personnel.

l l

9.11-2 Amend. 48 Feb. 1979 O

1

k. The Security Portable Radio System shall provide a " Total Area Coverage (TAC): portable two-way radio communication system for use by O plant security personnel.

l. SCADA Remote Terminal Unit shall provide the capability of controlling and monitoring the 161 KV Generating and Reserve Switchyards at CRBRP f rom the TVA Area Dispatch Control Center ( ADCC) at Volunteer.

9.11.2 _Descriotion 9.11.2 1 Public Address intra-Plant Communication System (PA-lC)

This system is powered from the 120 volt single phase Instrument AC bus, backed up by a DC battery / inverter system. The Instrument AC power supply is capable of supplying power to the PA-IC System for a period of two (2) hours, when all plant and offsite AC power sources are lost.

The system consists of handsets, amplifiers, loudspeakers, multitone generators, and provides one paging channel and four party line channels. The system has a future capabliity of being expanded to six (6) party l inos.

The equipment is designed to operate in all plant environments which are warm or cold, humid, dirty, and where constant vibration may be encountered. The handsets, amplifiers and loudspeakers are arranged and located to cover all accessible plant areas.

s The pageline of the PAIC system is used to broadcast alarm signals generated l by a multitone generator, over the entire plant area. The multitone generator produces the following distinct alarm signals.

a. Building Evacuation

b. Radiation Emergency

c. Fire Alerm initiation of the alarm signals is performed manually at the Control Room Operator's Desk. In the event of changing alarm conditions, the higher priority alarm signal will override the lower priority signal.

9.11.2.2 Private Automatic Exchange (PAX)

The PAX System is a network of dial-type telephone handsets located throughout the plant administrative areas and in the plant operating areas which are also serviced by the PA-IC System. The system includes an automatic switching machine capable of handling 200 telephone lines and telephone handsets.

9.11-3 Amend. 62 Nov. 1981

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

The PAX system is connected to the following plant communi-

! cations systems through tie lines and interfacing circuits:

1) Microwave Communications System

2) The page channel of the PA-IC System

3) The Powerline Carrier Communications System l

The PAX system incorporates a dial-in executive right-of-way l feature for the plant unit operator, the Shift Engineer, the Plant 47lManagerandforsecurity. This system includes a key telephone system for administrative inter-office telephone traffic.

The local commercial telephone service is available to the Plant Manager, Shift Engineer and other key locations in the plant. The commercial telephone system is not connected to any of the communications systems within the plant and is not a part of the Communications System.

I i 9.11.2.3 Microwave Communications t

l The Microwave Communications System is the primary TVA operated, l inter plast communications link between CRBRP and other TVA generation l and transmission facilities. The CRBRP microwave radio station operates i l on the 7125-8400 MHZ Government band frequency range for receiving and

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transmitting signals.

The Microwave Communications System provides independent communication channels for the following:

1) The generated power and other related information for unit control.

2) Voice communication between CRBRP and other TVA facilities.

3) Supervisory Control and Data Acquisition (SCADA) for Power Transmission Terminal.

9 11.2.4 Powerline Carrier Communications The Powerline Carrier Communications system is an inter plant communications system linking the CRBRP with other TVA facilities via the 161 KV transmission lines. The system uses low voltage, high frequency signals superimposed on the 161 KV transmission lines, to transmit voice I and plant operation data. 29 9.11-4 Amend. 58 Tiov. 1980

This system is an alternate means of data and voice communications between CRBRP and other TVA generation and transmission facilities and TVA Control Conters.

9.11.2.5 Maintenance Communications Jacking Svstem (MCJ) -

The system consists of sound powered headset / microphones and Jack stations. ,

Each headset / microphone contains a transmitter / receiver and need be only plugged into a Jack station for operation.

The purpose of this system is to facilitate the testing and calibration of equipment instrumentation and to provide for a fixed communications syst;m for effective response to an emergency. The MCJ system may also be used for the support of remote plant shutdown. Jack stations are arranged and located where required throughout the plant. All Jack station loops are connected to the Control Room. All Nuclear Island Jack station loops are also connected to a patch panel located on the remote plant shutdown panel. The user wears a headset / microphone assembly, plugs the cable into either a Jack station or a panel rock mounted Jack and thereby has hands-free communications.

9.11.2.6 VHF Radio Station The VHF Radio Station is provided to transmit emergency voice communications between the CRBRP Control Room and the TVA Power Production Emergency staff operations office. The CRBRP VHF Radio station transmits at 163.175 megahertz and receives at 170.075 megahertz. The radio will be f requency checked in accordance with FCC regulations and be given frequent operating checks.

9.11.2.7 Portable Radio System This system consists of a number of selective call portable radios (walkie-talkies) with paper and voice actuated microphone that transmit a low power signal. The system maintains satellite receivers that relay the signal to the base station and its comparator. The comparator selects the strongest signal received from a satellite receiver (voting systerr.) and then wire transmits the amplified signal to the base station which in turn retransmits the amplified signal.

The portable units have the capability to communicate among themselves on an alternate frequency.

Fixed repeaters which permit use of portable radio communication units are protected from exposure to fire damage by fire rated cabinets.

l l

l t

l O

9.11-5 i

Amend. 58 Nov. 1980 i

9.11.2.8 Manual Telerhone Switchboard This system consists of a dispatch switchboard to handle the communications requirements of the Control Foom. The switchboard is connected to the O'

following plant communications systems through tie lines and interf acing circuits:

1) Microwave Communications System

2) Powerline Carrier Communicaticas System

3) The PAX switching equipment Call cirector type telephones (answer only) connected to the switchboard are located at the shift engineer and the unit operator positions.

9.11.2.9 Offsite Law Enforcement Radio The system provides a dedicated voice channel with tone remote control console for communication between plant security personnel and the offsite law enforcement egency. The radio will be frequency checked in accordance with FCC regulations and be given frequent cperating checks.

9.11.2.10 Security I nter.com The Security Intercom System shall provide communications between CRT terminals and security personnel at all remote multiplexing panels and card readers. The system shall allow hands free operation and shall provide for tcne paging as required on one er mcre selected speakers.

9.11.2.11 Security Portable Radio System

) The Security Portable Radio System provides a " Total Area Coverage (TAC)"

portable two-way radio communications capability for use by plant security perscnnel only.

The system operates at frequencies separate from the plant Portable Radio System.

O 9.11-6 Amend. 62 Nov. 1981

l 9.11.2.12 SCADA Ramote Terminal Unit

) The SCDA Remote Terminal Unit will be located in the CRBRP Switchyard Relay House.

The Remote Terminal Unit will be linked to the TVA Volunteer Area Dispatch and Control Center (ADCC) via a 4 wire dedicated channel. The cummunication interface shall be provided by two data modems (normal and Alternate). The normal modem will interf ace with the microwave system and the alternate modem with powerline carrier system.

9.11.2.13 Evaluation of Communication Systems Separate and diverse means are provided to permit communication with and between plant personnel throughout the CRBRP. Under normal operating j

conditions, all systems are available, in addition, the PAX, PA-IC, VHF Radio and Portable Radio Systems are designed to function continuously on back-up DC battery power supplied in the event of loss of all offsite and onsite AC power sources. As part of the PA-lO system, two redundant signal generators are located at separate locations in the Control Building to provide standby capacity.

The Microwave Communications System provides voice and data channels between CRBRP and TVA Power System Control Center, the Volunteer Area Dispatch and Control Center and the Chattanooga Power Building.

The Powerline Carrier Communication System provides voice and data O communication channels between CRBRP and the TVA Power System Control Center, and between CRBRP and the TVA Volunteer ADCC.

Both the Microwave Communications and Powerline Carrier Communication Systems are powered from the normal plant 120 VAC supply. Each system is backed up by its own battery - battery charger supply and is capable of operating at least

. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> without chargers.

A fixed, two-way VHF radio communications system is provided for a direct -

voice link between the plant and TVA's Emergency Staff Operations Office in Chattanooga, Tennessee through existing TVA owned, VHF radio communication l facilities.

l l The Maintenance Connunications Jacking System is a sound powered system and I

does not depend on a power source for operation.

l l

l i

I O 9.11-7 Amend. 62 Nov. 1981

_ .___ _ . _ _ _ _ . _ . - . _ _ ~ _ , _ _ _ . __, _ -_ __ - . _ . _ _ - _ .

9.15 EOUIPMENT AND FLOOR DRAINAGE SYSTEM 9.15.1 Design Bases The plant Equipment and Floor Drainage System (EFDS) is designed to collect the drainage from all plant equipment such as pumps, tanks, coolers, etc., as well as the floor drainage.

Under normal operating conditions the floor drains in the plant serve for house keeping purposes. However, the EFDS is sized to accomodate the maximum l postulated discharge and flooding limits water eventaccumulation such as a pIDe on therfloor upture, to notank morerupture, than 3 1/2or sprinkler inches. All safety related equipment is mounted on pads at least 4 inches high.

9.15.2 Svstem Descriotion Separate EFDS sumps are provided for radioactive, potentially radioactive and non-radioactive areas of the plant. Each sump contains two vertical sump pumps with one pump serving as a f ull capacity spare.

Equipment and floor drains in areas that do not have the potential of becoming radioactive are collected and discharged into the waste water disposal system.

Floor drains carrying radioactive fluids are routed to the radioactive liquid l waste treatment systen, sump Drains containing potentially radioactive fluids are routed or pumped to a main collection sump in the RSB Radwaste Area and O { monitored for level of radioactive contamination. If the sump influent is contaminated, it is pumped to the radioactive IIquid waste sump where it can l then be processed through the radwaste treatment system, if the sump influent is not radioactively contaminated, then it is pumped to the waste water disposal system.

Treated waier and other process water treatment wastes which do not have the potential to be radioactively contaminated, are routed to seperate sumps for transport to the waste water treatment system.

Where there is a potential for oil spills, the drainage is routeo to oil Interceptors prior to discharge into the waste water disposal system. Oil spills are not allowed to drain in areas that contain radioactively

ontaminated equipment or fluids, in this case, the oil spill is contaminated with curbs and dikes and removed manually. Oil routed to the oil Intercep+or is collected in a waste oil tank and removed f rom the site for subsequent disposal.

All floor drains which are located in areas where sodium and water are present or in areas adjacent to cells containing sodium, are provided with water leak detectors. These leak detectors ar e provided to detect and identif y the location of any water leak.

1 l

(:i) ,

9.15-1 Amend. 62 Nov. 1981

1 61 9.15.3 Safety Evaluation The plant equipment and floor drainage system is designed so that it is not reasonably possible for any radioactive drainage in these systems to be discharged out of the plant without undergoing the required treatment cr processing.

61l Evaluations of radiological considerations f or normal operation and postulated sp!!Is and accidents are presented in Sections 11.2.5 and 15.0 respectively.

The plant Equipment and Flocr Drainage Systems is not safety related except for the p'pirg and valves required for containment isolation (Secticn 6.2.4).

EFDS piping within areas containing safety related equipe, ant is supported with 61 Seismic Category I supports.

9.15.4 Tests and Insoections EFDS pipes embedded in concrete are leak tested. All EFDS piping is tested fcr leaks after i nsta l l ati on. All leaking pipes or joints are repaired before 61 the concrcte is placed. All pumps are tested to ensure ihat their perfernances meet the required design ficws and pressures. A check source will to provided with the radiation monitor to ensure its operability.

9.15.5 Instrumentation Aoolication Each sump is provided with automatic cantrols to start and stop and alternate operaticn of the sump pumps. In case the lead pump falls to start, a high level switch automatically starts tt.e standby lag pump. A high-high level switch is provided in each sump and alarms in the control room to Indicate potential sump overflow. Radiation monitors are installed in the radioactive 61 and potentially radioactive RSB-RWA EFDS sumps.

> 0

, Aaend. 61 l 9.15-2 Sept. 1981

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O O O TABLE 15.1.3-3 SYSTEMS ASSUMED OPERABLE TO MITIGATE THE C0ftSEQUENCES FOLLOWIrlG THE OCCURREf4CE OF EACH ACCIDEtiT EVENT l Required Operable Events System Primary Secondary 15.2.1 Anticipated Events

;         15.2.1.1        Control Assembly Withdrawal at            PPS followed in long        Flux-YPressure    Flux-Total Flow Startup                                   term by decay heat          Flux-Delayed Flux removal (1) 15.2.1.2 Control Assembly Withdrawal at                   PPS followed in long        High Flux         Flux-Total, Flow Power                                     term by decay heat          Flux- QPressure removal 15.2.1.3 Seismic Reactivity Insertions-0BE                PPS followed in long        High Flux         Flux-Total Flow term by decay heat          Flux- QPressure removal 15.2.1.4 Small Reactivity Insertions                      PPS followed in long        High Flux         Flux-Total Flow
     $                                                              term by decay heat y                                                              removal
     $    15.2.1.5        Inadvertent Drop of a Single Control      PPS followed in long        Flux-Delayed Flux Modified Nuclear Rod at Full Power                         term by decay heat                            Rate removal 15.2.2          Unlikely Events 15.2.2.1        Loss of Hydraulic Holddown                PPS followed in long        High Flux         Flux-Total Flow term by decay heat          Flux-QPressure removal 15.2.2.2 Sudden Core Radial Movement                      PPS fol bwed in long        High Flux         Flux-Total Flow term by decay heat          Flux-VPressure removal 15.2.2.3 Maloperation of Reactor Plant                    PPS followed in long        High Flux         Flux-Total Flow Controllers                               term by decay heat g,                                                             removal                     Flux-QPressure                     22
     *5

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i i I 4 iABLE 15.1.3-3 (Continued) f4x;uired Operabl e Ly cts __ _ M fr PElrE.y " F4".2 20 15.2.3 Extremely Unlikely Events i i 15.2.3.1 Cold Sodium insertion FPS followed in long Speed Patio flow Ratio I term t'y decay heat 2 r erov a t 4 j 15.2.3.2 Gas Dubblo Passage through fuel, PPS f ol l owed i n long High Flux Flux-Total Flow 1 Radial blar.ket and Control tern by decay heat l Assemblies ronoval 15.2.3.3 Sel smic Reactiv i ty insertion-SSE F'PS f ol lowed i n long High 6 lux Flue-Total F'ow j term by decay teat Flux- Pressure j removal HTS Pump Electrics l l 15.2.3.4 Control Assembly withdrawal at F^ foll med I .i L on g Flux- Pressure F l ux-Tr. t a l Flow Startup-Maximun Mechanical Spoed term by cecay heat clux-Delayed Flux romoval 4 i 15.2.3.5 Control Assembly withdrawal at PPS follewed in long High Flux f!ux-Total Flow i Pow er term by decay heat re-oval w

       .on              15.3.1           Anticipated Events w
,       I l      c                15.3.1.1         Loss of Of f-Site Electric Power                       PPS followec in long           HTS Pump electrics flux-Tctal Flow

• term by decay heat i

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,                       15.3.1.2         Spurious Primary Pump Trip                             PPS fofIowed In iong           HTS Pump Eloctrics Flow Ratio l                                                                                               term by decay heat             Speed Patio l                                                                                               removal 15.3.1.3         Spurious Intermediate Pump Trip                        PPS followed in long           Speed Ratio        Flow Ratio term by decay heat r mov a l
      =                 15.3.1.4         Inadvertent Clowre of One                              PPS followed in long           S team-Feedw ater  Evap. Outl et Tmp.

j Qg Eveporator or Superheater Module term by decay heat rmov al

      .3                                 lsciation Valve

! a w. e 15.3.1.5 Turbine Trip Long term by decay None Required Nmo Reav rd

      $$                                                                                        beat removal (2)                                                          .

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1 c. . , . I ~ .  ;,.- - Figure 9.16-3 RCB RGCS Amend. 62 9 16-12 Nov. 1981

O O O TABLE 15.1.3-3 SYSTEMS ASSUMED OPERABLE TO MITIGATE THE C0flSEQUENCES FOLLOWillG THE OCCURRENCE OF EACH ACCIDENT EVEf47 Required Operable j Eyc,t1 System Primary Secondary 15.2.1 Anticipated Events 15.2.1.1 Control Assembly Withdrawal at PPS followed in long Flux-YPressure Flux-Total Flow Startup term by decay heat Flux-Delayed Flux removal (1) 15.2.1.2 Control Assembly Withdrawal at PPS followed in long High Flux Fl ux-To tal , Fl ow Power term by decay heat Flux- g Pressure removal 15.2.1.3 Seismic Reactivity Insertions-0BE nigh Flux PPS followed in long Flux-Total Flow term by decay heat Flux- QPressure removal 15.2.1.4 Small Reactivity Insertions PPS followed in long High Flux Flux-Total Flow In term by decay heat i 7 removal i $ 15.2.1.5 Inadvertent Drop of a Single Control PPS followed in long Flux-Delayed Flux Modified Nuclear i Rod at Full Power term by decay heat Rate 4 removal 15.2.2 Unlike;y Events 15.2.2.1 Loss of Hydraulic Holddown PPS followed in long High Flux Flax-Total Flow term by decay heat Flux-QPressure removal - 15.2.2.2 Sudden Core Radial flovement PPS followed in long High Flux Flux-Total Flow term by decay heat Flux-VPressure removal 15.2.2.3 Maloperation of Reactor Plant PPS followed in long High Flux Flux-Total Flow Controllers term by decay heat Flux-yPressure 22 g, removal

• S.

8 MRi

IABLE 15.1.3-3 (Continued)

                                                                     .ru;u i r ed C; er at i e LyuM                                                          _ _ :s kie                 -                  PIkat                          % Wart 15.2.3           Extrr ely Unlikely E v erit s 15.2.2.1         Cold Sodlun Insertien                        FPS followed I ri long                    $;eed Patio                   f l ow Pa*lo term ty decay N4a t rm eval 15.2.3.2         Gas Eutbl e Passage throug5 f uel,           f ?S f ol i c= ed i n icng                High Flux                     Flux-Tctal Flc=

Radial ti arlet and Control torn ty cocay toat Assembiles r mov a l 15.2.3.3 Sei smic Reactiv ity insertion-SSE f PS f ollcwed in long High flux Flue-Total Fic= tern by Cecay Feat Flux- F'ressuro r e oval HTS Pu p Electrics 15.2.3.4 Control Assembly withdre.al at PPS follcwed in Long Flun- Pressure Fluv-Tctal Ficw Stcr tup-Maximum Mechanical Speed term ty decay beat Flux-Eelayed Flux rmotal 15.2.3.5 t'on t r o l Asseebly Withdrawal at PFS f ol i cw ed i n l ong High Flux Flux-Tota! Ficw Pew er tern ty decay teat rmonal w cn

-      15.3.1           Anticipated Events w

I o 15.3.1.1 Loss of Of f-Site Elec+r!c Power PPS follcwed in long HTS Pump electrics Flux-Total Flcw term by decay beat rm oval 15.3.1.2 Spurious Pr!rtary Pump Trip PPS followed in long HT! Pump Electrics Flcw Ratio tern ty decay teat Speed Patio r e ovat 15.3.1.3 Spurious intermediate Pump Trip PPS follcwed in Icng Speed Ratio Flcw Patio term by decay heat r m oval nd 15.3.1.4 IraLertent Closure of One PPS f o l l ow ed i n long Stean-Feed.ater Evap. Cutlet T m . tern by CeCay beat QQa Evaporator or Superheater ModJle isolation Valve rmoval ct 5' 15.3.1.5 Turbine Trip Long term ty decay Nor.e R eq u i r e d Nore Rquired 3$ teat rmoval (2) O O O

O O O TABLE 15.1.3-3 (Continued) Required Operable Events Svstom Primary Secondarv 15.3.1.6 Loss of Normal Feedwater PPS followed in long Steam-Feedwater Steam Drum Lovel term by decay heat removal 15.3.1.7 Inadvertent Actuation of the Sodium- PPS f ollowed in long S team-Feedw ater Evap. Outlet Temp. Water Reaction Pressure Relief System term by deca / heat rmoval 15.3.2 Unlikely Events 15.3.2.1 Single Primary Pump Seizure PPS followed in long Speed Ratic Flow Ratio term by decay heat rmov al 15.3.2.2 Single Intermediate Loop Pump Seizure PPS followed in long Speed Ratio Fic4w Ratio term by decay heat rerroval 15.3.2.3 Small Water-to-Sodium Leaks In (3) Steam Generator tubes 5 15.3.2.4 Failure of the Steam Bypass System PPS followed in long Steam-Feedwater Steam Drum Leval s term by decay heat 4 o removal o 15.3.3 Extremely Unlikely Events 15.3.3.1 Steam or Feed Line Pipe Break PPS followed in long S team-Feedw ater Evap. Outlet Tep. term by decay heat removal 15.3.3.2 Loss of Normal Shutdown CooIIng System PPS followed in long S tearr-Feedw ater Steam Drum Level term by decay heat removal 15.3.3.3 Large Sodlurrr-Water Reaction Sodlum water reactic- Steam-Feedw ater Sodlum Water gg <m pressure retlef system Reaction rupture discs ~. 15.3.3.4 Primary Heat Transport System Pipe (3) 'g Leak ~m 15.3.3.5 Intermediate Heat Transport System PPS followed in long lHX Pr! Outlet Pipe Leak riow Ratio tem by decay heat rmoval

(TALE 15.1.3-3 (Continued) Required Operable Events System Primary Secondary 15 . 5. .! Unlikely Events 15.5. .1 Fuel Assembly Dropped within Reactor (3) Vessel during Refueling 15.5.2.2 Damage of Fue' Assembly due to A".tompt (3) to Insert a fuel Assemb'v Into an Occupied Position 15.5.2.3 Single Fuel Assembly Cladding Failure EVTM Seals and Subsequent Fission Gas Release during Refueling 15.5.2.4 Cover Gas Release during Refueling (3) 15.5.2.5 lleaviest Crane Load Impacts Reactor (3)

Closure llead g 15.5.3 Extremely Unlikely Events L 15.5.3.1 Collision of EVTM with Control Rod (3) h Drive Mechanism 15.6 Sodium Spills 15.6.1.1 Primary Sodium In-Containment Containment Iso-Storage Tank Failure during Maintenance lation System 15.6. .) . 2 Failure of the Ex-Vessel Storage Tank (3)

Sodium Cooling System during Refueling 15.6.1.3 Failure of Ex-Containment Primary (3) Sodium Storage Tank 22 1 58 e2 _5: O O G

15.7.2 Unlikely Events 15.7.2.1 Inadvertent Release of Oil Through Pump Seal (PHTS) 15.7.2.1.1 Identification of Causes The primary sodium pump has oil-lubricated bearings and/or seals above the pump tank which contains sodium. The seals will be designed to prevent oil !eaking into the pump tank for all modes of operation. The primary pump concept incorporates a seal lubrication system with a fixed total oil inventory (see Figure 5.3-14a). Oil that leaks through the lower seal will be collected in a lower seal leakage tank and pumped to waste during servicing. Abnomal leakage must be made up by deliberate manual action to open the system and add oil. The lower seal leakage tank has the capacity to hold the total seal oil inventory and thereby precludes any seal leakage from entering the pump tank in the event of an abnormal leak rate. An additional and last barrier preventing seal leakage from entering the pump tank sodium is provided in the pump design by a shaft oil slinger and reservoir located below the nomal seal rubbing faces. Any oil overflowing the lower seal leakage collection tank or running down the pump shaft is collected in a reservoir which has a capacity in excess of the total oil inventory. The primary pump O' concept, therefore, would require a combination of independent failures to occur coupled with a deliberate manual addition of oil to the system before oil could enter the pump tank. Although the release of oil from the primary pump seal to the primary sodium is considered an extremely low probability event, the results of such an event have been evaluated. Two potential effects have been identified:

1. Plugging Effects

2. Reactivity Effects 15.7.2.1.2 Analysis ot Effects and Consequences If it is postulated that the oil were to be released to react with the primary sodium, the following analysis is presented.

The oil above the seal would flow down the pump shaft and vaporize, or react with the sodium in the pump tank. The reaction of oil and sodium will result in the r~icase of hydrogen and carbon. The carbon compounds will either fladt on the sodium, dissolve (on the order of one ppb), or sink to the bottom of the pump tank. These are small particles which are easily fractured. O Amend. 62 15.7-9  ; Nov. 1981 ; l i

Tha release of these particles from the pump tank to the primary loap will depend upon the manner in which the pump is ope ra ting. If the pump is shutdown, the solids will stay in the pump tank. If the pump continues to operate after a seal failure, the reaction products would eventually go into the primary loop. In the present pump concept tne pump tank will contain approximately 800 gal. of sodium, and will be changing at 700 gpm due to flow from tne IHX return (200 gpm) and bearing return flow (500 gpm). Plugging Effect Three different conditions were evaluated as follows: A. To calculate the maximum plugging temperature in the pump discharge, the following conservative assumptions were made:

1. Pump tank temperature is 1000*F.

2. The pump tank vents to cover gas system through the pump standpipe bubbler. Maximum gas pressure is 12 in. W.G. plus equivalent static head of sodium @l000'F for elevation between nonnal RV sodium level and normal level in the pump tank. This assumes no pump draw down. P res su re is 97 in. W.G.

3. Oil leaks into the pump tank at a rate just sufficient to saturate the pump tank sodium volune of 800 gallons with H2 at the temperature and pressure above. This results in a concentration of 121 ppm of H2 in the pump tank sodium.

4 The pump tank mixture is drawn into the pump and mixed with primary sodium at the ratio of 700 gpm/34000 gpm (IHX and bearing return flow vs pump discharge flow).

5. The resultant pump discharge contains 2.5 ppm of disolved H2 and the plugging temperature is 460 F.

C. To calculate the maximum plugging temperature in the core and the remainder of the system the following conservative assumptions were made:

1. Assume that leakage continues as in the previous condition until the entire 6 gallon inventory of oil in the seal system has leaked into the tank which is at 1000 F. )g 1

15.7-9a O Amend. 19 Play 1976 a ._ _

5.0 INSTRUCTIONS, PROCEDURES AND DRAWINGS b) 5.1 OWNER IMPLEMENTATION The Owner has established and implemented a practice of prescribing in documentary fonn the required quality of plant structures, systems, and components and necessary activities to assure attainment of requisite quality through work activities. This practice includes specifying division of work responsibilities and the project-wide practices to be implemented in execution of those responsibilities. Through this practice, l10 the following documents have been prepared: e Plant Design Guidelines e Management Policies and Requirements e Contract Statements of Work e Environmental and Safety Analysis Reports e Policies, Procedures and Instructions e Reports e Records The Owner has prepared his procedures and instructions in accordance with procedures that prescribe the fonnat to be followed and the identification system to be used. These procedures cover all activities of management, engineering and design control, document review and control, procurement, surveillance activities, audits, and records management. These procedures prescribe methods for performing quality-related activities I in conformance with the requirements of 10CFR50, Appendix B. The Owner procedures are organized under a Management Procedures System which is administered by a procedures coordinator from within the Owner organiza-ti on. The procedures coordinator is assigned the function of controlling the issuance of procedures to assure coordination and consistency in format, content, etc. The procedures system itself is organized along divisional lines (Engineer-ing, Procurement, Construction, Quality Assurance, Public Safety, Operations, 45l Project Control, Administrative Services, and others) which give the responsible 27 managers the responsibilities for: e Assuring that policies of a continuing nature are incorporated in the Management Procedures System (MPS). e Incorporating applicable laws, standards such as 10CFR50, Appendix B, Executive Orders, decisions and directives of the Project Steering Committee (PSC) into the procedures to the extent necessary to show the requirements placed upon the Owner. 10 17A-23 Amend. 52 Oct. 1979

l o Determining the coverage and content of management directives necessary to carry out their assigned functions, assuring the accuracy and currency of the procedures, and arranging for the cancellation of those that become obsolete. o Approving procedures for which they are responsible. Obtaining review and comment by other organizational units when appropriate. o Submitting to the Procedures Coordinator:

a. Draft Procedures for Review of Format l b. Final Procedures for Director Approval o Determining, with concurrence of General Counsel, what portions of procedures, if any, shall be communicated to the contractors.

Furnishing to the Procedures Coordinator the names of contractor personnel to whom such material together with any appropriate supplementary explanation or instructions should be distributed. The Procedures Coordinator assures that style, format, content, terms, titles and numbering sequence of all procedures conform to the requirements of the Management Frocedures System. The Chief, Administrative Services is the prime control officer for procedures and as such: o Effects the printing and distribution of the final approved procedures and subsequent revisions. o Maintains a master file of all current approved CRBRP Project Office procedures and a reference file of previously issued procedures and their revision. o Prepares and maintains an index of procedures. l Organization Unit Managers are responsible for writing and implementing the l procedures necessary for their division. General Administration procedures j cover policies and procedures which apply to all employees. The Project Director approves for issuance all CRBRP Project Office procedures. The individual division procedures are approved by the responsible Division Manager and recommended to the CRBRP Project Director for final approval . 1 Each new procedure or revision of existing procedure is prepared using the Management Procedures System numbering code and format. O 17A-24 Amend. 62 Nov. 1981

8.0 IDENTIFICATION AND CONTROL OF MATERIALS. PARTS AND COMPONENTS 8.1 OWNER IMPLEMENTATION The Owner delegates execution responsibility for identif ication and control of materials, parts and components to the other major project participants ~ through contracts. 8.2 REOUIREMENTS OF OTHER PARTICIPANTS Each project participant, who has an assigned responsibility for materials, parts and components including partially fabricated subassemblies are required, by contract, to establish and implement identification and control practices. Each participant's identification requirements are to be determined during the initial planning stages and their practice will assure: ' l that identification of the item is maintained, on or attached 1o the item or on records traceable to the item as required throughout fabrication, erection, installation, and use of the item; the item (s) can be traced to the appropriate documentation such as drawings, specifications, purchase orders, manufacturing and Inspection documents, deviation reports, and physical and chemical mill test reports; that the method and location of identification does not affect the function or quality of the item being Identitled; and thaT the correct identification of items is accompilshed and verified prior to tne release for fabrication, assembly, shipping and installation. Th,ese practices will be designed to preclude the use of incorrect or def ectivo-msterials, " parts and canponents. O- The Owner monitors major participant identification.and control of materials, ' . parts and components practices and periodically audits the participants' practices to assure proper implementation and adcquacy.

                                                                                                                          /

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17A-33 u Amend. 62 Nov. 1981 *

                                                               ----  .Y   ,e        . - - . - - . - - - - __ _.                 my__

9.0 CONTROL OF SPECIAL PROCESSES lll 9.1 DWNER IMPLEMENTATICfi The Owner delegates execution respersibility for control of 10 special processts to the other raajor project participants through contracts. 9.2 REQUIREMENTS OF OTHEF PARTICIPAN7S i Project participants, who are assigned responsibility for activities where special processes are involved, are required, by j

        .       contract, to establish and implement practices to assure adequate                                                                   10 performance and control of special processes such as welding, heat treating, nondestructive examination, and cleaning.                                                      These practices will include the following elements:

e Qualification of procedures, equipment, and personnel for performance of special processes in accordance with applicable coces, standards, specifications, or supplementary requirements, e Special processes are performed by lualified personnel and l acccmplished with written process sheets, shop procedures, l chacklists, travelers, or equivalent with recorded evidence of verification. a f;uali'icatin, records of procedarcs, equipment, and personnel assoLidfed with special processes are established, filed, and kept current. The Owner n.onitors major participant special process control practices 10 ana Jertadicalli r.dits .ttie participants practices to assure proper i::plementa tiot and acequacy l l i m-r- - 1 l l l 1 1 17A-34 O Amend. 10 Dec. 1975

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15.0 NONCONFORMING MMERI ALS. PARTS OR COMPONENTS 15.1 OWNER IMPLEMENTATION

           ~
                             ..Th'e Ow ner has establisned and imp!emented practices for control, review and                                                                               '

disposition of nonconforming materials, parts or components. These practices are designed to assure that measures are established to control materials, parts, or components which do not conform to requirements in order to provent their, inadvertent use or Installation. The nonconformance control practice includes the -f ollowing elements: o Establish Disposition Responsibility o Docunentetion and Reporting o Review, Evaluation and Disposition All reports of nonconformances af fecting saf ety or utility of items that are proposed to be dispositioned "use as is," "use as repaired" or "use as modified" are forwarded to the Owner for approval. These reports are part of the documentation required at the plant site. Errors or deficiencies reported to, or discovered by the Owner which could

                            . adversely affect safety-relatea structures, systems or components identified in design documents subsequent to their formal release or issue are evaluated by Engineering, Public. Safety, an- Quality Assurance for consideration as a p                        reportable deficiency under paragrap 50.55(e) of 10CFH50. If it is concluded

( . s that the error or deficiency comes under this paragraph, the deficiency together with the proposed corrective action is reported to the Nuclear Regulatory Commission according to regulations. Defects or noncompliance in the plant or basic component supplied to the plant that are reported to, or discovered by the Owner which could adversely affect safety related functions of the plant are evaluated by Engineering, Public Safety, and Quality Assurance for consideration as a reportable deficiency under 10CFR21. If it is concluoed that tho' defects or noncompliance is reportable under part 21, the deficiency is, reported to the Nuclear Regulatory Commission according to

                    ,         regulations.

The deficiency, whether reportable or not, is further evaluated against the

                    ~         procedural requirements thet should have prevented the occurrence. When the

' procedural system is deficient, the affected organization is required to take whatever steps are necessary to achieve appropriate corrective action to the l ysystem to preclude recurrence of the deficiency. The deficiency is reported wIthin the project vle on unusual occurrence report as described in Section 16. _ The' 0wner Quality Asrsurance organization also participates in and monitors the execution, of the nonconfermance control practices and periodically audits or arranges for independeni audit of the control practices to assure r implementation and adecuacy. x 17A-43 Amend. 62 Nov. 1981

      \
        ,-                      _ . . _ . - .     ,.   ._m_.__.--_.l _ ,. . . , - . , , _ , , , . . _ _ . . . _ . . _ . _  ,.__y , m.. _,.__,,,,c       . _ . , . . . _ . . _ _ . _ _ , _

15.2 Rf.QLIIREMENTS OF OTHER PARTICIPANTS Each participant, who is assigned responsibility for procurement, manufacturing, or construction of items of the CRBRP, is required by contract to establish and Irplement a practice for the control of nonconforming materials, parts or ccenponents. These nonconformance control pracilces will I.q:lude the following elements: o The identification, documentation, segregation where practicable, review, disposition, and notif ication of af f ected organizations of ronconforming materials, parts, components, or services is undertaken. O l e 17A-43 a Amend. 62 Nov. 1981 l

I e Documentation identifies the nonconforming item; describes O the nonconformance, the disposition of the nonconformance, and the inspection requirements; and includes signature approval of the disposition. e Provisions are established identifying those individuals or groups delegated the responsibility and authority to approve the dispositioning of nonconforming items. e Nonconforming items are segregated where practicable, from acceptable !

   )7 items and identified as discrepant until properly dispositioned.        l t

e Acceptability of rework or repair of materials, parta, components, systems, and structures is verified by reinspecting the item as originally inspected or by a method which is at least equal to the original inspection method; inspection, rework, and repair  ! procedures are documented. e Nonconformance reports dispositioned "use as is" or "use as repaired" or "use as modified" are made part of the inspection records and forwarded with the hardware to the Owner. e Nonconformance reports are periodically analyzed to show quality 10 trends, and the results are forwarded to management. These practices will assure tnat nonconforming items are reviewed and accepted, rejected, repaired or reworked in accordance with documented pro-O cedures. They will include measures which control further processing, delivery or installation pending proper disposition of the deficiency. The Owner monitors the major participants nonconformance control practices and periodically audits the major participant's nonconformance [0

                                                                                                '1 control practices to assure implementation and adequacy.

16.0 CORRECTIVE ACTION 16.1 OWNER IMPLEMENTATION 10 The Owner has established and implemented a system for corrective action wherein conditions adverse to quality such as failures, nonconformances, malfunctions, deficiencies, deviations and defective material and equipment 27 l that to theare required Owner fornonconformance through reliable and safeandoperation of the plant unusual occurrence are reported reporting procedures. 10 Quality assurance activities found deficient by Owner reviews and audits of the participants are also reported. The corrective action system includes the following elements: 10 ( Amend. 45 July 1973 17A-44

5.0 INSTRUCTION

S. PROCEDURE

S AND DRAWINGS The BOP Supplier has prepared his procedures and instructions in accordance with procedures that prescribe the format to be followed and the identification system to be used. These procedures cover all activities of managernent, document review and control, procurement, surveillance activities, audits, and records m=Jnagement. These procedures prescribe methods f or performing quality-relaied activities in conformance with the applicable requirements of 10 CFR 50, Appendix B. The BOP Suppller procedures are organized under a Management Procedures System which is administrated by a procedures coordinator from within the BOP Suppller organization. The procedures coordination is assigned the function of controlIing the issuance of procedures to assure coordination and consistency in f ormat, content, etc. The procedure system itself is organized along divisional lines (Engineeri ng, Procurement, Construction, Qual ity Assurance, Publ ic Saf ety, Operations, Project Control, Administrative Services and others) which give the responsible managers the responsibilities f or: 1 o Assuring that policies of a continuing nature are incorporated in the Management Procedures System, (MPS). , o incorporating applicable Iaws, standards such as 10 CFR 50, Appendix B, Executive Orders, decisions and directives of the Project Steering Committee (PSC) into the procedures to the extent necessary to show the requirements placed upon the 00P Suppl f er. o Determining the coverage and content of management directives

,                                                    necessary to carry out their assigned f unctions, assuring the accuracy and currency of the procedures and arranging f or the cancellation of those that become obsolete, o    Approving procedures for which they are responsible. Obtaining review and comment by other organizational units when appropriate, i

o Submitting to the Procedures Coordinator

a. Draft procedures for review of format.

l b. Final procedures for Director Approval. o Determining, with concurrence of General Counsel, what portions of procedures, if any, shall be communicated to the contractors. Furnishing to the Procedures Coordinator the names of contractor personneI to whom such material together wIth any appropriate

supplementary explanaticn or instructions should be districutea.

I t i l O 17C-15 Amend. 62 Nov. 1981

The Procedures Coordinator assures that style, format, cortent, terms, titles and numbering sequence of al! procedures conform to the requirements of tne Management Procedures System. The Chief, Administrative Services 13 the prime control officer for BOP Suppller Procedures and as such: o Ef f ects the printing and distribution of the final approved procedures and subsequent revisions, o Malntal ns a master f II e of alI current approved CRBRP Project Of f ice procedures and a reference file of previously issued procedures and their revision. o Prepares and maintains an index of procecures. Organizational Unit Managers are responsible f or writing and implanenting the procedures necessary for their division. General Administration procedures cover policies and procedures which apply to all unployees. The Project Director approves for issuance all CRBRP Project Office procedures. The individual division procedures are approved by the responsible Division Manager and recommended to the CRBRP Project Director f or f inal approval . Each new procedure or revision of existing procedure is prepared using the Management Procedures system numberI ng code and f ormat. Each division establ ishes steps f or the review of draf t procedures within the divisions. If a procedure applies to more than one division the other divisions af f ected receive the draf t procedure for review. A draft is sent to the Procedures CoordInato who reviews It f or format, styIe, and numbering sequence. The final procedure or revision of existing procedure is approved by the i appropriate division manager responsible for that particular subdivision of I procedures, is approved by the CRBRP Project Director, and released for im p l anentat i or . Distribution of each procedure or revision of existing procedure is listed and

 !Iled with the procedure copy in the procedure master file. The register l

shcws which revision is current. The BOP Supplier practice f or documenting, in written f orm, the requirements f or and results of actlvIties af fectIng quality is, itself, executed in I accordance with document control procedures identified under Section 6.0, l Document Control. l l l O 17C-16 Amend. 62 Nov. 1981

O 8,0 IDENTIFICATION AND CONTROL OF MATERIALS. PARTS AND COMPONENTS U The responsIbIIity for execution of Identification and controf of materials, , parts and components, is delegated to BOP equipment suppliers and the CRBRP site receiving organization by contract. Each supplier, who has an assigned responsibility for materials, parts and components including partially f abricated subassemblies is required, by contract, to estabiish and impianent identification and control practices. Each supplier's identification requirements are to be determined during the initial planning stages and his practice will assure that identification of the item is maintained, on or attached to the item or on records traceable to l the item as required throughout f abrication, erection, installation, and use of the item; the item (s) can be traced to the appropriate documentation such as drawings, specifications, purchase orders, manufacturing and Inspection documents, deviation reports, and physical and chemical milI test reports; that the method and location of Identification does not af fect the f unction of quality or the item being identified; and that the correct identification of Items is accompiished and verif ied prior to the reiease for f abricatIon, assembly, shipping and installation. These practices will be designed to preclude the use of incorrect or def ective materials, parts and components. The BOP Supplier monitors supplier identification and control of materials, parts and components practices and periodically audits the suppliers' practices to assure proper Implernentation and adequacy. O l

O i

i 17C-23 Amend. 62 Nov. 1981

9.0 CONTROL OF SPECIAL PROCESSES The responsibility for execution of control of special processes during manufacturing of BOP equipment, is delegated to equipment suppliers by contract. I Project suppliers, who are assigned responsibility for activities where special processes are involved, are required, by contract, to establish and implement practices to assure adequate performance and control of special processes such as welding, heat treating, nondestructive examination, and cleaning. These practices will include the following elements: e Qualification of procedures, equipment, and personnel for performance of special processes in accordance with applicable codes, I standards, specifications, or supplementary requirements. i e Special processes are performed by qualified personnel and accomplished with written process sheets, shop procedures, check lists, travelers, or equivalent with recorded evidence of verification. e Qualification records of procedures, equipment, and personnel associated with special processes are established, filed, and kept current. g The BOP Supplier monitors supplier special process control practices and periodically audits the supoliers' practices to assure proper implemen-tation and adequacy. l 10 l Amend. 10 17C-24 Dec. 1975

15.0 NONCONFORMING MATERIALS. PARTS OR COMPONENTS The BOP Suppller has established and implemented practices for control, review and disposition of nonconforming materials, parts or components. These practices are designed to assure that measures are established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. The nonconformance control practice includes the following elements: i o Establish disposition responsibility o Documentation and reporting o Review, evaluation and disposition All reports of nonconformances affecting safety or utility of items that are proposed to be dispositioned "use as is", "use as repaired", or "use as modified" are forwarded to the BOP Supplier for approval. These reports are part of the documentation required at the plant site. Errors or deficiencies reported to or discovered by the purchaser which could adversely affect safety-related structures, systems or components identified in design documents subsequent to their formal release or issue are evaluated by Engineering, Safety and Quality Assurance for consideration as a reportable deficiency under Paragraph 50.55 (e) of 10 CFR 50. If it is concluded that , the error or deficiency comes under this paragraph, the deficiency together O with the proposed corrective action is reported to the Nuclear Regulatory Ccanission according to regulations. Defects or noncompliance in the plant or In the basic component supplied to the plant that are reported to, or l i discovered by the purchaser which could adversely affect safety related functions of the plant are evaluated by Engineering, Public Safety, and Quality Assurance for consideration as a reportable deficiency under 10CFR21. If it is concluded that the defects or noncompliance is reportable under part i 21, the deficiency is reported to the Nuclear Regulatory Commission according

)

to regulations. The deficiency, whether reportable or not, is further evaluated against the procedural requirements that should have prevented the occurrence. When the l procedural system is deficient, the affected organization is required to take

!         whatever steps are necessary to achieve appropriate corrective action to the system to preclude recurrence of the deficiency. The deficiency is reported wIthin the project via an unusual occurrence report as described in Section 16.

1 Each supplier, who is assigned responsibility for procurement, manufacturing, or construction of items of the BOP, is required by contract to establish and implement a practice for the control of nonconforming materials, parts or components. These nonconformance control practices will include the following elements: j o The Identification, documentation, segregation, review, disposition, ~ and notification to affected organizations of nonconforming materials, parts, components, or services is undertaken. 17C-31 Amend. 62

,                                                                                                     Nov. 1981

e Documentation identifies the nonconforming item; describes the nonconformance, the disposition of the nonconformance, and the inspection requirements; and includes signature approval of the disposition. e Provisions are established identifying those individuals or groups delegated the responsibility and authority to approve the dispositioning of nonconforming items. i e Nonconforming items are segregated when practicable, from 27 17l acceptable items, controlled and identified as discrepant until properly dispositioned. e Acceptability of rework or repair of materials, parts, com-ponents, systems, and structures is verified by reinspecting the item as originally inspected or by a method which is at 27 least equal to the original inspection method; inspection, rework, and repair procedures are documented. e Nonconformance reports dispositioned "use as is," "use as repaired," or "use as modified" are made part of the in-spection records and fomarded with the hardware to the Owner. e Nonconformance reports are periodically analyzed to shcw quality trends, and the results are fomarded to management. These practices will assure that nonconforming items are reviewed and accepted, rejected, repaired or reworked in accordance with documented procedures. 4dThey will include measures which control further processing, delivery or in-stallation pending proper disposition of the deficiency. The B0P Supplier Quality Assurance organization also participates in and monitors the execution of the nonconformance control practices and periodically audits or arranges for independent audit of the control practices to assure implementation and adequacy. 10 The responsibility to control nonconforming material, parts and components related to BOP equipment manufacturing and resulting subcontracts, is delegated to equipment suppliers by contract. 10 Amend. 45 17C-32

CL INCH RIVER BREEDER REACTOR PL ANT O A DESCRIPTION OF THE CONSTRUCTOR QUALITY ASSURANCE PROGRAM TABLE OF CONTENTS PAGE NO. 0.0 INTRODUQI1QN 17F-1 0.1 ORGANIZATION 17F-1 0.1.1 Organizational Arrangement 17F-1 0.1.2 Responsibil ity and Authority 17F-2 0.1.3 Qualification Requirements of the Project 17F-3 QA Manager 0.2 PROGRAM 17F-3 FIGURES 17F-1 Company Organization f or Quality Assurance 17F-4 17F-2 Quality Assurance Department Organization 17F-5 17F-3 Quality Assurance Interrelationships 17F-6 17F-4 Scope of Constructor Quality Assurance Program 17 F-7 Participation 17F-5 Major Elements of the Constructor Program 17F-8 O 17F-! Amend. 62 Nov. 1981

CL INCH RIVER BREEDER REACTOR PLANT DESCRIPTION OF THE CONSTRUCTOR O' QUALITY ASSURANCE PROGRAM

0.0 INTRODUCTION

Stone & Webster Engineering Corporation (S&W) is the constructor f or the Clinch River Breeder Reactor Plant (CRBRP) Project, in this capacity, S&W is responsible f or management and perf ormance of those tasks associated with the overall construction ef f ort. This includes the responsibility to plan, implement, and manage the Constructor portion of the CRBRP overall quality assurance program. This program will be applied to activities within S&W's contractual scope of work that af f ect saf ety related structures, systems, and components (as def ined in Section 3.2 and 7.1). The S&W CRBRP Project Quality Assurance Program is based on the Stone & l Webster Topical Report, SWSQAP 1-74A, " Standard Nuclear Quality Assurance Progr am. " Although some organizational elements and responsibilities have l been shif ted for this project, all requirements contained in SWSQAP 1-74A which are applicable to S&W's scope of work will be implemented. By accomplishing this, the S&W Quality Assurance (QA) Program for CRBRP complies with the applicable requirements of 10CFR50, Appendix B and RDT F2-2. The correlation of 10CFR50, Appendix B and RDT F2-2 is shown in Figure 17.1-3. The changes that have been made in the quality assurance organizational l structure from that shown in SWSQAP 1-74A have been made to respond to project \ conditions where S&W does not have responsibility for engineering or design, as well as requirements of the owner f or the establishment of a project quality assurance organization. As a result, the responsibil ities f or l implementing some requirements contained in SWSQAP 1-74A have been shif ted within the organizational elements of the QA Department. These changes in organization and responsibility are described in the following paragraphs. 0 .1 ORGANIZATION 0.1.1 Organizational Arrangement The S&W management organization, including quality assurance management, for the CRBRP Project is shown in Figure 17F-1. The QA Depariment organization is shown in Figure 17F-2. Figure 17F-3 shows the Project Quality Assurance Organization. The changes in the quality assurance organizational structure f rom that presented in SWSQAP 1-74A f or the Constructor program of the CRBRP Project are: A. Because S&W has no engineeri ng or design responsibil ity f or CRBRP, these organizational elements are not represented. O 17F-1 Amend. 62 Nov. 1981

B. The position of CRBRP Project QA Manager has been created. The CRBRP Project QA Manager, who will be located at the project site, has l overall authority and responsibility for quality assurance functions, both administrative and operational, on the project. The Project QA Manager receives quality assurance direction from, and reports to, the QA Department Manager in S&W Headquarters. The QA Department Manager reports to the Vice President, QA who reports to the S&W Company President. CRBRP project policy is received through the interface shown in Figure 17F-1 with the Project Managers and/cr the Senior Project Manager. , 1 l C. The position of QA Program Administrator will not be established for the CRBRP Project. The Project QA Manager and the project QA Staff wfil perf orm those f unctions normal ly essigned to the coordirator. D. Field Quality Control Division personnel at the site and Procurement Quality Assu"ance personnel in the S&W District Offices will receive project direction from the Project QA Manager. Corporate policy, l corporate administration, and corporate resource support will remain with the present Peadquarters divisions. E. The Project QA Manager's staff will be established to perform both quality assurance engineering and quality assurance management t functions.

r. In addition to the audit function of the Cost and Auditing Division of the QA Departnent in S&W Headquarters, which will retain

, responsibility for audits of the overall S&W CRBRP Project QA Program, an audit function within the Project Quality Verification group will also report to the Project QA Manager. 0.1.2 Responsibilltv and Autho1]fy The shift of quality assurance responsibilities to support the revised organizational structure is as follows: A. Because S&W has no engineering and design responsibilities, these functions will be performed Ly others designated by the owner as described in other sections of this PSAR. B. The Project QA Manager is responsible for performing the quality assurance program management and administrative functions for the project quality assurance organization, as delegated by the QA Department Manager, in addition, the Project QA Manager is responsible for those tasks normally assigned to a QA Program Administrator. The QA Department Manager, located in Boston Headquarters, will provide quality assurance policy and guidance, the interface with corporate management and access to other QA Headquarters divisions. In this position, the Project QA Manager has the organizational freedom and authority to identify quality problems, initiate, recommend, or provide solutions through designated channels O l 17F-2 Amend. 62 l Nov. 1981

1 l 'I l and verify implementation of corrective action. The Project QA ,

Manager is also responsible for establishing necessary interfaces with

other project participantr. on both informal and formal basis as j designated by the owner. ,

i f i l l I  ! i l i l i l 1 i l i , i I i i I , O I i i i i I .I i i i l

• i i

i i l i 17F-2a Amend. 62 Nov. 1981 f. ) 1

C. The Project QA Staf f will be responsible for and will execute many of the Project QA Managers' assigned tasks f or the Constructor program. l D. In addition to audits perf ormed by the Headquarters QA Cost and Auditing Division of the overall quality assurance program, the CRBRP Project QA Manager will maintain a staf f of quallfled auditors to conduct required audits of subcontrac+ ors, the S&W CRBRP FQC organization, and others as requested or directed by the owner. 0.1.3 Oualification Recuirements of the Project GA Manager The qualification requirements of the Project QA Manager will be equivalent to those described in SWSQAP 1-74A for the Chief Engineer, Quality Systems Division. 0.2 PROGRAM The Constructor Quality Assurance Program is a major portion of the overali Project Quality Assurance Program. The scope of the program and the type of project participation covered by the program are shown in Figure 17F-4. The major elanents of the Constructor Program are shown in Figure 17F-5. S&W has been assigned execution responsibi;Ity for the f ull scope of the Constructor program except the area of preoperational testing and start up ectivities. Responsibility for execution of activities related to those areas has been retained by the owner. O 17F-3 Amend. 62 Nov. 1981

I 1 i 0FFICE OF l THE C11 AIRMAN AND PRESIDENT 1 4f N10R VICE PRES 40ENT PflOJECT WAN AGER DENif DintCIOR QU Alli y PROJECT l CON 8f L Cil0H HANAGIR5 MOR.QA 11EADQUARTERS

                                                                                                                                    ~

O A KillD G E l 3fN10R Slif **"""~~~-h============ Caane reostCt COnsinuctios 4___ f 04 WANAGER RIPMstNiallVE

~                                                          CONTRACT M                                                       ADMINISTRATOR E                         ENGINEERING       _       _

LI Att0N CONSTRUCTION ADMlHISTRATOR 1 I I I I 8 '" " surfalt Itt Of NT SAfggy Suff RINif HDE NT surer NTENDINI of CON 31RUCil0H gurtRVISOR of COSI L CONSIRUCIl0H OF C0ft51RUCil0N " 8tRytCES SClif 01, LING EN".NEERING s- ,- ,, ,- , .. ,, L S[ND , PROJECT ORGANIZATION FOR QUALITY ASSURANCE 8 a lif CLINCil RIVER BREEDER REACTOR PLANT PROJECT

     ---- COMMuMICATION & LI AISON DE 6 MSm MGMEREG CORPOMTION

=> O 3 5@ a. G' $$ Figure 17F-1 O O O

                                                    =

2- , - ~ ,.

          %                                                   '\_,                                                  ,,

VICE PRESIDENT AND MANAGER OF QUALITY ASSURANCE I i ASSISTANT ASSISTANT MANAGER MANAGER QUALITY QUALITY ASSURANCE ASSURANCE I I QUALITY QUALITY CRBRP FIELD QUALITY PROCUREMENT :10NDESTRUCTIV1 QUALITY ASSURANCE ASSURANCE PROJECT QA CONTROL QUALITY TEST SYSTDiS PROGRAM COST AND MANAGER DIVISION ASSURANCE DIVISION DIVISION ADMINIS-AUDITING DIV. JRATORS J IviSTnti C T on NOTE: PROJECT DIRECTION IS PROVIDED BY Tile QUALITY ASSURANCE DEPARTMENT ORGANIZATION CRBHP PROJECT QA MANAGER TO OTilER CLINCil RIVER BREEDER REACTOR PLANT PROJECT 5E ORGANIZATIONAL UNITS OF Ti!E QA DEPART- STONE 6 WEBSTER ENGINEERING' CORPORATION

  .$              MENT WilICll ARE PERFORMING ACTIVITIES P             IN SUPPORT OF Tile PROJECT.
  $0 Figure 17F-2

VICE PRESIDENT MANAGER OF QUALITY ASSURANCE PROJECT QUALITY ASSURANCE MANAGER C 7

• QUALITY QUALITY QUALITY ASSURANCE ENGINEERING VERIFICATION SUPPORT SERVICES

                                         *PROCURDIENT

• FIELD QUALITY ASSURANCI- QUALITY CONTROL-DISTRICT OFFICES SITE ANOTE: PROCllREMENT QUALITY ASSURANCE AND FIELD QU/.LITY PROJECT QUALITY ASSURANCE ORGANIZATION CONTROL PERFORM VERIFICATION ACTIVITIES AS AN CLINCH RIVER BREEDER REACTOR PLANT PROJECT 2 INTEGRAL PART OF Tile PROJECT QUALITY ASSURANCE STONE & WEBSTER ENGINEERING CORPORATION j ORGANIZATION AND 41ECEIVE PROJLCT DIRECTION FROM

" i TiiE PROJECT QA MANAGER. CORPORATE ADMINISTRATION, G* CORPORATE POLICY, AND CORPORATE RESOURCE SUPPORT $* ARE RECEIVED FROM TilEIR PARENT DIVISIONS IN Figure 17F-3 BOSTON llEADQUARTERS. e O #

i l i i.

till l I

i t L I i I I l THE CLINCH RIVER BREEDER REACTOR PLANT i PRELIMINARY SAFETY ANALYSIS REPORT i i

CHAPTER 17.0 - QUALITY ASSURANCE APPENDIX l  ;

l A DESCRIPTION OF THE GE-ARSD-RM l QUALITY ASSURANCE PROGRAM i i ^ 1 . l r I i l 1 i GENERAL ELECTRIC COMPANY l ADVANCED REACTOR SYSTEMS DEPARTMENT j SUNNYVALE, CALIFORNIA t i i l 5 i I i lO Amend. 52 Oct. 1979 i i I

CLINCH RIVER BREEDER REACTOR PLANT

   )

A DESCRIPTION OF THE GE-ARSD-RM s./ QUALITY ASSURANCE PROGRAM TABLE OF CONTENTS PAGE

0.0 INTRODUCTION

............................................ 171-1 1.0 ORGANIZATION............... ............................ 171-2 2.0 OUALITY ASSURANCE PR0 GRAM............................... 171-9 3.0 DESIGN CONTR0L.......................................... 171-12 4.0 PROCUREMENT DOCUMENT CONTR0L............................ 171-1/ 5.0 INSTRUCTION

S. PROCEDURE

S AND DRAWINGS................... 171-19 6.0 DOCUMENT CONTR0L........................................ 171-21 7.0 CONTROL OF PURCHASED MATERIALS. EOUIPMENT AND SERVICES., 171-23 8.0 IDENTIFICATION AND CONTROL OF MATERIALS. PARTS AND \ COMPONENTS.............................................. 171-26 9.0 CONTROL OF SPECIAL PROCESSES............................ 171-27 10.0 INSPECTl0N.............................................. 171-28 11.0 TEST CONTR0L............................................ 171-29 12.0 CONTROL OF MEASURING AND TEST E0UIPMENT................. 171-30 13.0 HANDLING. STORAGE AND SHIPPING.......................... 171-31 14.0 INSPECTION. TEST AND OPERATING STATUS................... 171-32 15.0 NON -CONFORM ING MATER I AL S. PARTS OR COMPONENTS. . . . . . . . . . . 171-33 16.0 CORRECTIVE ACTl0N....................................... 171-35 17.0 OUALITY ASSURANCE REC 0RDS............................... 171-36 18.0 AUDITS.................................................. 171-38 FIGURES 171-1 Organization of Quality Assurance Program O Participation........................................ 171-40 171-1 Amend. 52 Oct. 1979

TABLE OF CONTENTS (Continued) PAGE O 171-2 GE-ARSD-Quality Program Management Organization...... 17'-41 171-3 GE-ARSD Product Assurance Organization............... 171-42 171-4 Major elements of the GE-ARSD-RM QA Program.......... 171-43 lablE. 171-1 GE-A:;5D Quality Assurance Program Index Versus Req u i remen t s of 10CF R50, Appen d i x "B" . . . . . . . . . . . . . . . . 171-44 Attachments 171-1 GE-ARSD Quality /ssurance Document Descriptions...... 171-54 171-2 Schedule for issuing Unreleased Procedures........... 171-69 O O 171-11 Amend. 52 Oct. 1979

CL INCH RIVER BREEDER REACTOR PLANT A DESCRIPTION OF THE GE-ARSD-RM QUAllTY ASSURANCE PROGRAM TABLE OF CONTENTS s PAGE

0.0 INTRODUCTION

............................................ 171-1 1.0 ORGANIZATION............................................ 171-2 2.0 OUALITY ASSURANCE PROGRAM............................... 171-9 3.0 DESIGN CONTR0L.......................................... 171-12 4.0 PROCUREMENT DOCUMENT CONTR0L............................ 171-1/ 5.0 INSTRUCTION

S. PROCEDURE

S AND DRAWINGS................... 171-19 6.0 DOCUMENT CONTR0L........................................ 171-21 7.0 CONTROL OF PURCHASED MATERIALS. EOUIPMENT AND SERVICES.. 171-23 8.0 IDENTIFICATION AND CONTROL OF MATERIALS. PARTS AND COMP 0NENTS.............................................. 171-26 9.0 CONTROL OF SPECIAL PROCESSES............................ 171-27 10.0 INSPECTION.............................................. 171-28 11.0 TEST CONTR0L............................................ 171-29 12.0 CONTROL OF MEASURING AND TEST E0UlPMENT................. 171-30 13.0 HANDLING. STORAGE AND SHIPPING.......................... 171-31 14.0 INSPECTION. TEST AND OPERATING STATUS................... 171-32 15.0 NON-CONFORMING MATERIALS. PARTS OR COMPONENTS........... 171-33 i 16.0 CORRECTIVE ACT10N....................................... 171-35 1

17.0 OUALITY ASSURANCE REC 0RDS............................... 171-36 18.0 AUDITS.................................................. 171-38 i FIGURES

! 171-1 Organization of Quality Assuranco Program Participation........................................ 171-40 ) 171-I Amend. 52 Oct. 1979 i _.. - . - ~ . _ , . _ _ . - _ _ - - _ _ _ _ _ . -

TABLE OF CONTENTS (Continued) PAGE O i 171-2 GE- ARSD-Qual i ty Program Management Organ i z at ion. . . . . . 171-41 171-3 GE-ARSD Product Assurance Organization............... 171-42 17l-4 Major el ements of t he GE- ARSD-RM Q A Program. . . . . . . . . . 171-43 l l 1 Dbl &1 l l 171-1 GE-ARSD Quality Assurance Program Index Versus Req u i rements of 10CFR50, Append i x "B". . . . . . . . . . . . . . . . 171-44 Lita bmen1B 171-1 GE-ARSD Quality Assurance Document Descriptions...... 171-54 l 171-2 Schedule for Issuing Unreleased Procedures........... 171-69 O l l l l l 171-11 Amend. 52 Oct. 1979

APPENDIX l CLINCH RIVER BREEDER REACTOR PLANT A DESCRIPTION OF THE CE-ARSD-RM QUALlTY ASSURANCE PROGRAM 0.0 iNTRODJCTION A Quality Assurance Program has been established by the General Electric (GE) Company's Advanced Reactor Systems Department (ARSD) to assure conformance to contractual requirements for the Clinch River Breeder Reactor Plant (CRBRP). The contra:tual requirements invoke Quality Assurance Program Requirements by means of the U.S. Department of Energy Standard RDT F2-2 through Amendment 3. GE-ARSD is a major participant on the CRBRP in the role of a Reactor Manufacturer (RM), having been delegated by Westinghouse ARD-LRM, the execution of responsibility for the design, procurement and manufacture of NSSS systems, components and services, including a portion of the Project's Quality Assurance Program as shown on Figure 171-1. The Quality Assurance Program established to meet contractual requirements contains those elements necessary to comply with the requirements of Code of x Federal Regulations, Part 50, Appendix "B", " Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants" (10CFR50, Appendix "B") for GE-ARSD scope of work. This appendix describes how the Quality Assurance Program meets the applicable criteria of 10CFR50, Appendix "B". The practices described herein will be applied to the planning, design, procurement, and manufacture of those systems, components, and structures defined in Sections 3.2 and 7.1 of the PSAR that are assigned to GE-ARSD and covered in the Contract Scope of Work. I ( l J 171-1 Amend. 52 Oct. 1979

1.0 ORGANIZATION 1.1 GE-AMD PRODUCT ASSURANCE ORGANIZATLON GE's organizational structure for performing quality-related activities associated with management, design engineering, procurement, and manufacture of C S systems, components and services, and the responsibilities and authrs itles of key positions within the GE organization, are described in Sect ion 1.4 c ' the PSAR. GE's organization chart is shown on Figure 1.4-8.

 're Management organization structure having responsibility for GE's quality
 ;;rogram is presented in Figure 171-2.       Their responsibilities are described below:

The Department General Manager of GE-ARSD reports directly to the Vice-President and General Manager, Energy Systems and Technology Division and has overalI responsIbiiIty fcr the GE-ARSD Quality Program. These responsibilities include: (a) Report promptly significant quality problems to the Division Vice-President and General Manager, Energy Systems and Technology Division, for communication through appropriate channe's to the Sector Executive. in addition, when these problems involve significant legal and/or reputation risks, communicate the problems to Corporate Legal Operation and/or Corporate Public Relaticos Operations (Corporate Relations Staff). (b) Establish and issue supporting procedures and take other actions for fulfilling the requirements of the Corporate Product Quality Policy. (c) Not If y the Staf f Executive, Product Qual ity, Technical Resources Staf f (Corporate Technology Staff) of any situation which entails field replacement or rework of a substantial nature. (d) Provide for an annual independent review and assessment of the ef f ectiveness of the Department Qual ity Assurance Program, including its compliance with appropriate contractual requirements (eg., RDT and ANSI). The Department General Manager has responsibility and authority to issue and i implement a Department Product Quality Pclicy. Authority to deviate therefrom

ls reserved to the General Manager. He has delegated the execution of the j following responsibilities

(1) The Section Managers, Clinch River Project, Technology and Special Projects, Design Engineering and Development Engineering have the responsibility and authority to: (a) Define the quality requirements and standards for the products associated with the project (s) assigned by the Department General Manager to the Section and insure that all customer needs and reasonable expectations are recognized and defined consistent with Department and Company policy. (b) Approve the quality program to be employed on Section products. O 171-2 Amend. 62 Nov. 1981

(c) In conjunction with Product Assurance and Services Section, assure O that Section work is accomplished in compliance with the applicable quality program and that each product meets the established quality requirements. (d) In conjunction with Product Assurance and Services Section, insure that supplier's work is accomplished in compliance with the applicable quality program for those projects assigned to the Section. (2) The Section Manager, Appiications Engineering and Planning has the responsibility and authority to: (a) Convey the customer's quality requirements and standards as defined by contract, for each product to each responsible section. (b) Assure that the quality of product service rendered after sale and delivery meets reasonable expectettons of the customer. (3) The Section Manager, GE-ARSD Financial Cperations has the responsibility and authority to provide sound financial information, consistent with quality objectives. (4) The GE-ARSD Legal Counsel has the responsibility and authority to: (a) Keep Department Management apprised of current laws and regulations applicable to the achievement and maintenance of product quality for products of the Department. (b) Assist Department management in formulating product quality requirements in light of laws and regulations applicable to Department products and advise management regarding potentially significant legal and/or reputation risks involved in identified product quality problems. (5) The Section Manager, Procurement has the responsibility and authority to: (a) Convey to the suppller, the Department's quality requirements and

standards for each procured product.

(b) In conjunction with Product Assurance surveillance, assure that supplier work is accomplished in compliance with contract imposed j quality requirements. l (c) Assure that administrative specifications applied to the supplier meet l all quality needs and requirements by obtaining appropriate reviews l and approvals prior to incorporating the specificat,an into a l contract. i i (6) The Manager, GE-ARSD Employee Relations, has the responsibility and i authority to: i j (a) Insure the creation and maintenance of a Department environment in which employees at all levels will have an attitude of striving for excellence in the performance of their work. ! (b) Conduct suitable programs'to ensure employee motivation. (c) Insure the employment of professional employees who are cualified for l the high technology work in the Department. i 71-3 ! Amend. 52 Oct. 1979

The organization of GE-ARSD's Product Assurance and Services Section is presented in Figute 171-3 and is described below: (7) The Manager, Product Assurance and Services, has the responsibility and authority to: (a) Assure that each Department product delivered to a customer meets the specifled quality requirements. (b) identify quality problems in the multifunctional Quality Assurance Program Activities and ensure their satisfactory resolution. (c) Insure that all new work proposals conform to applicable codes and standards, and that appropriate quality requirements are defined. (d) Insure that significant risk to the Company quality reputation are detected, reduced to an acceptable level, ana/or after notifying the Project Manager communicated to the Department General Manager for his information and/or action. (e) Design and maintain a Quality Assurance Program for each Department Product. (f) Obtain customer based Quality Measurement Data. (g) Furnish the Staff Executive, Product Quality Technical Resources Staff (Corporate Technology Staff) wIth the Department Product Quality Policy and revisions as issued. (h) Prevent shipment of unsatisfactory products. (i) Insure, by appropriate verification activities, that Department work at the supplier's plants is accomplished in compliance with the applicable quality program. The Manager of Product Assurance and Services reports at the same organizational level as the highest line managers having direct responsibility for performing quality-reIated activitles, and has the organizational freedom and authority to identify quality problems and ensure satisfactory resolution, and to control further processing, delivery, installation or operation of an item having a deficiency or unsatisfactory condition until proper disposition has occurred. The Manager, Product Assurance Clinch River Project reports to the Manager, Product Assurance and Services with reporting lines to the Manager, Clinch River Project, the Department General Manager and the Vice President and General Manager of the Energy Systems and Technology Division. The Manager, Product Assurar ae, Clinch River Project has the organizational freedom and authority to Identify quaIity probiems and ensure satisfactory resolution and to control further processing, delivery, installation or opcrations of an item having a deficiency or unsatisfactory condition until proper disposition has occurred. The authority to stop work is retained by the Department General Manager. The lManagerofProductAssuranceandServicesandtheManager,ProductAssurance Clinch River Project are authorized to prevent shipment of unsatisfactory products. O 171-4 Amend. 62 Nov. 1981

The Manager cf Product Assurance and Services and the Manager, Product 7 [3 $ Assurance-CRP are responsibie for immedIate notification to affected management, of conditions that in their opinion, require the stopping of engineering or manufacturing work in-process. Affected management,.Upon notification by the Manager of Product Assurance and Services of conditions that the Manager of Product Assurance and Services determines' require the - stopping of work, shall Irunediately evaluate such conditio'ns and inform the Manager of Product Assurance and Services of the actions to be taken. . Ths' affected manager has the option to stop work' activities in his area ~of ~ responsIbiIity untii conditions are corrected to the satisfaction of'the Manager of Product Assurance and Services and the Manager Product Assurance- s CRP. If the affected Manager elects not to stop work, and the Manager of . Product Assurance and Services determines that tha planned' corrective actions are Insufficient to warrant the continuance of work, the Manoger of Product < Assurance and Services has the authority to require that the affected man,ager immediately justify his actions to the Department General Manager. The Department General Manager will decide the actions to be taken and informs the , affected Manager, and the Manager of Product Assurance and' Services, end the' Manager, Product Assurance-CRP. - The Department General Manager is continually involved in appraising the ' status of the quality program and the accomplishments of the Product Assurance and Services organization. This is accomplished by means of Monthly Quality'

                                                                                  ~

Status Reports and Management Review Meetings. , A Monthly Quality Status Report containing quality assurance progress and accomplishments, current problems, non-conformances and failures wit 5 their b analysis and corrective action status, quality trend data, and results of program audits and management reviews, is prepared and transmitted to the LRM, the Department General Manager, and the Vice President and General Manager of Energy Systems and Technology Division. The Manager of Product Assurance and lServicesandtheManager,ProductAssurance-CRPmoetwiththeDepartment General Manager on a regular basis to discuss the status of the Department's quality program and accomplishments of the Product Assurance and Services organization. f The GE-ARSD General Manager will also have an independent review and l assessment of the effectiveness of the GE-ARSD Quality Assurance Program I performed annually. This review to determine the adequacy of the QA program and its compliance with RDT F2-2 and 10CFR50, Appendix "B" will be preplanned, documented, and will identify necessary corrective actions. These corrective actions will be communicated to the responsible organization and trackeo to assure l completeness. l 1 b

 \s!                                                                                  -

17l-5 Amend. 62-Nov. 1981

The responsibility for assuring compliance with the quality assurance requirements of CRdRP is delegated by the Manager of Product Assurance and Services to the Manager, Product Assurance - Clinch River Project. The Manager, Product Assurance - Clinch River Project has the responsibility for

     - devising, developing and ensuring .the ef f ective execution of the Project's quality assurance program for the GE scope of the CRBRP project. In
     ' performing this function he identifies requisite quality assurance activities, approves er concurs in plans und procedures developed for implementation, recommends cetegation of responsluility To Icwer tier organizations for implementat ic n with appropriate approval in the requirements and implementatica documentation of the lower tier organizations, plans and performs contro: and ' verification activities to ensu e ef f ective program execution, and manages and directs the activities of assigned staff members to accomplish the quality assurance program management activities.

In performing these activities he is supported by the Manager, Quality Engineering end Verification-Equipment Projects and the Manager, Quality

      ; Engineering & Verification-Plant Syster.s.      They have fer their scope of work ihe responsibility of devising, developing and ensuring ef f ective execution of the Project's quclity assurance program.        These programs include the design,
 ,     manufacture, test, and delivery of major components. They plan and perform surveillance of.!nternal procurement control, plan and perform surveillance of verders activities including source inspection of product, perform follow up en audits of the supp!!er quality assuFance activities, and initiate, coordinate, and cause closure of product nonconformances and audit findings.

A principal quality engineer prcvides the same control and verification ac+ivity for Project equipment manuf actured by GE-ARSD. The Manager, Quality Assu 's respo6sible ta the Manger, Product Assurance and Services for impleme of quality assurance activities in support of all products produced diro ,1y by !RSD. On the Clinch River Project, this will include quality planning, quality engineering, and quality control activities in response to quality assurance requirements defined by the project organization. Quality Assurance provides verification of ARSD engineering, procurement, fabrication, and test activities carried out in ! support of Project equipment or camponents supplied directly by ARSD. This will include supplier source surveillance and receiving inspection of caterials, as well as in-prccess Inspection of ARSD manufacturing and test j operations. Quality Assurence also has responsibility for tne maintenance of i in-house quality systemc, including the issue and control of ASME code manuals as specified for certification of "N", "NPT" and "U" stamp requirements. in fulfilling these responsibilities, the Manager, Quality Assurance has a staff consisting of a Quality Engineering function, a Quality Control function, and a Quality Systems function. l The Manager, Product Assurance Audits, is responsible to the Manager, Product Assurance and Services'for conducting an internal and external audit program i consistent with the CRBRP Quality Assurance Program requirements. In response l ltorequestsfromtheManager,ProductAssurance,ClinchRiverProject-(i.e. I task definitions and schedules), he develops audit plans and schedules to meet Clinch River Project needs. He is responsible for conducting an i r.terna l l audit program on CRBRP Engineering and Procurement activities, including the preparation and issuance of audit schedules, 17!-6 i Amend. 62 Nov. 1981

       ,            the assembly of a qualified audit team, the preparation of corrective action requests and the follow-up action required to assure effective corrective actions are implemented.              He is also responsible for conducting an external audit program of scoplier activities including the preparation and issuance of audit schedules, assembiing a qualIfled audit team, preparing findings and soliciting correcilve action through contractual channels and the follow-up action required to assure ef f ective corrective actions are implemented.

Managotent Review Meetings of the CRBRP are held regularly by the Manager, l Clinch River Project. The CRP staff and the Manager, Product Assurance-Clinch River Project present the status of the Project for their areas of respons!bility at these meetings to the GE-ARSD staf f. The Manager, Product Assurance-Clinch River Project discusses the status of the Quality Assurance Program at these meetings. In addition to these project review meetings, management review meetings are held periodically by the Manager of Product Assurance and services with the Manager, Freduct Assurance - Clinch River Project to evaluate the overall CRBRP Quality Status. I The collection and maintenance of Product Assurance generated quality records l Is the respontibility of the Manager, Product Assurance - Clinch River Project, the Manager, Quality Assurance, the Manager, Product Assurance Audits, and the Manager, Management Systems. 1.2 PRODUCT ASSURANCE AND SERVICES MANAGER'S OUALIFICATIONS The Manager, Product Assurance and Services and Managers, Product Assurcnce CRP, Product Assurance Audits, Quality Assurance Quality Engineering and VerifIcatlon, and Qualiity Engineering sh!i have the foilowIng minimum qualifications: Education - Shall be a graduate of a four-year accredited Engineering College or Uniwsity with a degree in Engineering or Sclerice. O 171-7 Amend. 62 Nov. 1981 _ . _ , , _ . . .m. __ - -,,._.y. _ . , _ . . . . , . _ . , . . _ , . _ _ _ . . . _ . _ , _ . - - _ _, . _ , . _ . _ _ - . _ . .

Lcerlence - General: Shall have a tr.inimum of 10 years experience in quality O engineering, or manufacturing, associated with *1uclear facilities or other high iechnology product arcar. Specialty: Shall possess a broad knowledge and understanding af Industry and Government Codes, Standards and Regulations defining quality assurance requirements und practices. Shall have a working knowledge and understanding of quality assurance methods and their applicatlon. Managerial: Shall be experienced in crganizing, directing and administering an overalI program or actlvity, or a major portion of en overalI program having broad scope and application. Shall have a perience in the supervision of personnel and planning and management of other resources needed to conduct an extensi/c quality assurance program. The Manager, Quality Control, shall have the following minircun qualifications: Educatien - BS Degree, industrial Engineering preferred. However, suitable industrial experience is acceptable. Experience - General: See Product Assurance Managers' Qualifications above. Spectalty: See Product Assurance Managers' QualIfIcatlens above. Managerial: See Product Assurance Managers' Qualifications above. The Manager, Quality Systems shalI have the folIowing minimum qualifications: Education - BS Degree, Engineering or Science preferred. However, suitab l e industrial experience is acceptable. Exoerlence - General: Shall have a minimum of ten years experience in quality, engineering and/or quality systems involving high rigor quality hardware. Specialty: Shall have a wcrking knowledge of applicable Industry and Government Codes, Standards and Regulations defining quality assurance program requirements. Shall have specific knowledge of quality assurance methods as applied to design, manufacturing, assembly, test and supplier control activities. Managerial: Shall have experience and training in managerrent planning, organizing, integrating and measurement of quality systens or '1 similar functions. O 17l-8 Amend. 52 Oct. 1979

N I 2.0 OUALITY ASSURANCE PROGRAM j GE-ARSD's Product Qual ity Pol icy states:

                          "It is the policy of the Department that:

a) All products of fered be consistent with the public interest and applicable laws and regulations, b) All products of fered be such as to validate buyers' selection of the price-perf ormance canbination involved and thereby contribute positively to the Company's product quality reputation. c) in every market segment served, the Department pursues the objective of achieving a reputation f or product quality that either equals or exceeds that of any competitor serving the same segment." This policy is implemented by the Department General Manager within the Department by assignment of prine and contributing responsibility for l Management of the Quality Assurance Program as described in Section 1.0 of this Appendix. The GE-ARSD CRBRP QA Program is documented in the CRBRP Quality Assurance Program index (QAPI). The QAPI is a document which provides a listing of those GE-ARSD Procedures, instructions, or other Documents, to be utilized in Implementing the specif ic quality assurance requirements f or the program, and O includes an organizaticaal chart that identifies key personnel and their f unctilonal responsibliities and authorities for cualitity assurance activities. The QAPI is issued with Impiomentation direction by the Clinch River Project Section Manager to those f unctions and/or individuals having the responsibility for implementation of quality-related activities. The Q API i s reviewed and revised as necessary to ref lect audit corrective actions and procedure changes. The QAPl is approved by the Manager, Product Asstrance - Cilnch River Project and is subject to review and approval by the LRM. The initial issue and all revisions of the QAPI are distributed both internal ly and externally under controlled distribution with acknowledgement receipt r equ i red. Table 17i-1 shows a matrix correlating the Quality Assurance Program implementing documents to the criteria requiroments of 10CFR50, Appendix "B". Attachment 171-1 provides a synopsis of the GE-ARSD QA Program Documents. The measures cescribing how GE-ARSD meets the criteria of 10CFR50, Appendix "B" are contained in the other sections of this appendix. The Department policy requires that written procedures will be established and issued to f ul flil the requirements of +he Corporate Product Quality Policy. Authority to deviate f rom the Quality Policy is reserved to the Department General Manager. Prime responsibility for implementation verif ication is assigned to the Manager of Product Assurance and Services to monitor and measure the perf ormance of the management and implementation areas to ensure iheir perf ormance is in compliance with the program. The GE-ARSD CRBRP Quality Assurance (QA) Program is established to comply with the contrac1ual requirements of RDT Standard F2-2 which is applicable to 171-9 Amend. 62 Nov. 1981

GE- ARSD's scope o' work on CRBRP. This program, established to meet the contract requirements, also contains those program elements necessary to comply with the criteria of 10CFR50, Appendix "B" as applicable to GE-ARSD's scope of work. The major elements of the GE-ARSD CRBRP QA Program are shown in Figure 171-4 and are applied to the saf ety-related structures, systems, end components as listed in Sections 3.2 and 7.1 of the PSAR within GE's scope of work. Each of the program elements as shown in Figure 171-4 is executed by GE-ARSD. GE- ARSD delegates the execution of these elteents, when appropriate, to suppliers, however, GE- ARSD retains responsibility for adequate implementation and performance by their suppilers. l l O l l l l O 171-9a Amend. 62 i Nov. 1981 L- .__ - _ _ _ . , _ _ _ _ - - ._ _. _ _ _ . __

1 1 l GE-ARSD establishes personnel requirements through the use of position guides. O Quellflod personnel with formal training and/or experience are assigned to the established positions. A periodic formal performance evaluation of each incumbent by his next higher level of management is documented. A Department-wide personnel training program is utilized to insure a knowledge and understanding of Department polIcles, instructions, and procedures. Corporate and Nuclear Energy Business Group job-related training programs are also utilized. For NDE and Control of Special Processes, personnel are trained and quallfled in accordance with applicable codes, standards and procedures. GE-ARSD has established training and Indoctrination programs to assure that

 ;                                 personnel performing quality-related activities understand the requirements and applicability of QA Prograrn requirements. New employees are required to attend an orientation session which includes a familiarization of the technical requirements of CRBRP and engineering and quality program practices.

Training sess!ons are also conducted to provide instruction in the Policies and Procedures to te employed to project / program personnel, quality assurance engineers, designers, procurement, and manufacturing personnel. Personnel who perform inspections, tests, and non-dectructive examinations are required to be trained and qualified for each area of specialty before they can perform the activity. Training is conducted in accordance with appropriate procedures. O l i i l l l

O f

( 171-10 Amend. 62 Nov. 1981 1 m.,---m---,._,-. , y v n, - m yy y , __,mm. . . , _ _ . ,,_ _. . _ ,,_ , _ _ _ ,

t 52 l Inspections and tests performed at GE-ARSD are conducted in accordance with O prescribed instructions prepared for the specific item under consideration. The preparation of these instructions, including the criteria considered, is 52 l discussed in Section 10.0 of this appendix. GE-ARSD does assure for all acceptance tested and deliverable CRBRP equipment, that quality-related activities are performed with specified equipment and under suitable environmental conditions, and that pre-requisites have been satisfied prior to inspection and test. 52 l GE-ARSD procurement documents implement pertinent requirements of 10CFR50, Appendix "B" by identifying the QA Codes and Standards, or portions thereof, contractually imposed on GE-ARSD and other contractual requirements which must be complied with by the Supplier and described in the Supplier's QA programs 52 to insure that the contract requirements imposed on GE-ARSD properly passed on to the Suppller and are traceable, in some instances, depending on the item being processed, ex1racts from the appiIcable Codes and Standards and other contractual requirements are included verbatim in the procurement documents. 52 l GE-ARSD reviews subcontractor and suppiler QA Programs against the GE-ARSD contractually imposed portions of appilcable Codes and Standards and any additional QA requirements included in the contractual document. The principle objectives of the QA program and key functions and elements which it contains are not expected to change over the duration of this project. However, circumstances may make advisable changes in the organization, or in the implementing detalis, and such changes will be made in accordance w ith normal management practices. This PSAR QA description will be O reviewed annually to assure that all required changes have been documented, and it will be updated as necessary. i i O i 171-11 Amend. 52 Oct. 1979

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

3.0 DESIGN CONTRDL 52lCRBRPdesignactivitiesatGE-ARSDareperformedwithinClinchRiverProject, Development Engineering, and Design Engineering Sections. Control, approval, and management of the design activities is exercised by Clinch R!ver Project Sectlon. Assignment of design responsibility is made by the Clinch River Project, Development Engineering, and Design Engineering Section Managers to appropriate subsection managers, who in turn assign the des!gn responsibility to a functional or technical discip!Ine unit manager. The unit manager makes individual assignments to a responsible engineer for perfccming the effort. The unit manager assigned responsibility is supported by the other functional end technical units as required. The centractual design basis is defined to the responsible design 52l organizational units by treans of a Project Master Plan and Work Breakdown Structure Dictionary issued by the Clinch River Project Section which provides the delineation of the program work scope, identification of each item to be delivered, and assignment of responsibility. Interim direction to accomplish tasks or other work efforts not specifically identified is accomplished by means of the issuance of Project Directives. System Design Descriptiens (SDD's) provide the principal means of design 52 l definition and control for each CRBPP system for which GE-ARSD has re spon s i b i l i ty . The SDD is a comprehensive technical document utilized to: define and integrate the various technical, operational, and safety considerations involved, identif y interf aces; and serve as a comon technical basis for project activities. The SDD's are responsive to the Overall Plant Design Description (OPDD) requirements and provide a technical reference and control for detailed component specifications, operational test procedures, l and safety analysis. The Equipment Specifications ("E" Spec's) similarly ! provide the control for cceponents, and define the design, fabrication, i quality assurance, testing, handling and shipping requirements. The "E" Specs l are responsive to the SDD requirements. The SDD's and "E" Specs are prepared and issued as specifications. Where required, principal design data are submitted to the LRM f or approval prior to project use. Design definition and control is established through a technical documentation structure such that the approved document set, when issued, provides assucance that applicable design criteria, code, standards, practices, and requirements for materials, structures, components, systems, facilities and processes, are defined and correctly translated into specifications, drawings, and work l procedures and instructions. Each level in the design definition structure I provides requirements that constrain and control the next lower tier of ! documents. During the conduct of the design ef fort, the responsible eng!neer conducts a study to determine what codes and standards, and their specific requirements, are applicable to the system or component under consideration, and the l l O 171-12 l Amend. 52 ! Oct. 1979

adequacy of such codes and standards. When existing codes and standards are judged to be inadequate, he initiates appropriate action to improve the code O or standard. The applicability of the codes and stande.-ds is specified in the appropriate design documentation. An element of each document review and approval, is the consideration and application of the code or standard to the design. 9 O 171-12a Amend. 52 Oct. 1979

Measures have been established to assure: that appropriate quality requirments, codes and standards are specified in design documents; that parts, materials, and processes are selected on the basis of experience and qualification 1ar 'he intended serv ice; that design interf aces with in GE-ARSD, and with other Project participants, are def ined and controlled; that designs are verif ied by appropriate methods and personnel; that design def inition documentation is approved and issued by appropriate organizations; and that design changes and deviations are subject to the same measures as applied to the original design. The responsible engineer def ines the quality standards that are appropriate f er his design, and documents those standards in the specif ications, drawings, or other technical documents. These documents are reviewed in accordance with established procedures, and changes or deviations f rom the quality standards are identifled and controlled. Changes to quality standards are handled in the same manr.er as any other technical changes to the af fected documents. Deviations are classed as changes if they apply to the f ull set (or lot) of production Itons. If deviations apply only to a specif ic item of herdware, they are identif ied and controlled within the Product Assurance and Services Section. Each deviation is documented as a non-conformance item and is resolved with participation of the appropriate technical, management, and supporti ng organizations. The responsible engineer identifies and includes in appropriate planning, those engineering studies necessary to substantiate the adequacy of the design f or its intended application. Engineering analysis studies f ocus on those aspects and considerations !nvolved in the intended services; such as, operation, maintenance, in-serv ice inspection, rel iabil ity, avail abil ity and safety. The studies are based on analysis of reactor physics, structural stress, thermal, hydraulic, environmental, and other et fects, and other appl icable technical co nsi der ati on s. Engineering studies are documented in engineering memoranda and f ormal reports and are included among the engineering work records. The results of the engineering studies and analyses produce criteria to be satistled by the design. These criteria are defined and documented within l design documentation. The design criteria include perf ormance objectives, operating conditions, reguiatcry requirments, saf ety and avalIability, materials, f abrications, construction, testing, operation, maintenance, and qual ity requi rcments. Design documentation incl udes speci f ications, drawings, and instructions, as necessary, to def ine the speci f ic requirments f or detail design, materials, f abrication, construction, instal lation, testi ng, inspection, maintenance, cleaning, packaging, shipping, storage, operation, and quality assurance. Design documentation is rev iewed by competent special ists in many dif ferent f ielas to assura design producibility, manuf acturability, inspectabil ity, and testat:l l ty i r accordance with the established techn: cal requirments f or the ded gn. O 171-13 Amend. 62 Nov. 1981

Acceptance criteria for inspection and test are developed by the responsible O engineer during the design phase and are identif ied in the design def inition documentation. The acceptance criteria provide the basis f or acceptance or rejection of each designated quality characteristic and, when appropriate, specify the points at which compliance will be accomplished and verified. l O O 171-13 a Amend. 62 Nov. 1981

Identification and control of design Interfaces is accomp!!shed by the responsible engineer and documented by means of SDD's, "E" Spec's and interface Control Drawings (ICD's). The fundamental control document for functional Interf ace data is the SDD which identifies the system interfaces including referencing supporting control documents, i.e., ICD's, and together with ICD's, compictely defir,es requirements for every interf ace within a system. The "E" Spec is similarly utilized to define the functional interf aces for ccinponents. l ICD's are cocuments that identify the physical interface characteristicsICD's necessary to ensure ccrrpatibility between mating pieces of equipment. are distributed to, and used by project participants for assuring compatibIIity of systems and/or components. Interface requirements are transmitted to interfacing organizations and concurrence is obtained prior to issue. Proposed changes are coordinated with interf acing organizations prior to implementation. For the CRBRP Project, the Manager, Project integration and Centrol has been delegated the responsibility for coordinating the control cf interf aces within 62-ARSD's scope of responsibility, and for providing iIalson between GE-ARSD and other project participants concerning interfaces. Prior to the issuance of SDD's and "E" Specs, Controlled information Data Transmittals (CinDT's) are used as a control trethod for transmitting and receiving working level design data. Distributicn and control of design documents and their changes is described in Section 6.0 of this Appendix. The Managers of the responsible design organizations wIthin GE-/RSD are respcnsible for determining the degree and extent of design verification necessary to assure the adequacy of finished designs. Each discrete design or design aspect, or change of design for, or having an effect upon a system, equipment, component cr service, is, as appropriate, verified for design adequacy by a design review, independent calculation, or qualification testing program. The facters utilized in determining the extent of design verification are the relative importance of the design action to safety and reilability, the uniqueness of the design, and the complexity of the design. The Manager of the responsible design organization designates an independent reviewer or review team to perform the independent calculation or perform the design review. GE-ARSD procedures require that inspection and test specificatloas be a subject for design review, and they are included on the design review checklist. Reviewers are technically competent individuals who have no direct resronsibility for the design under consideration, but may be f rom the same organization (Section, Subsection or Unit). Design reviews and independent calculations, and their subsequent documentation, are planned, conducted and documented in accordance with written procedures. Design reviews are shcwn on Project plans and schedules which are subject to LRM approval. The responsible manager is required to resolve the review team findings. When qualification testing is utilized as the verification method, the test program includes requirements for testing under the most adverse design conditions. O 171-14 Amend. 62 Nov. 1981

1 All pertinent operating modes are considered in determining these adverse design conditions. The control measures applied to qualification test are described in Section 11.0 of this Appendix. i GE-ARSD drawings, specifications and instructions are reviewed, approved and Issued in accordance.with established procedures. An Engineering Review Memorandum (ERM) is utilized to record the review and comments of Individuals having competence to determine whether the document under consideration adequately conforms to applicable contract and standards requirements within his assigned areas of review. This review is conducted prior to approval for issue. The ERM is used to designate and record areas of review, reviewers, document status, configuration control level, comments and signatures. Review j areas include interf ace compatibility, productibility code compliance, i materials engineering, verification of design adequacy, calculation verification, System Design Description (SDD) conformance, quality requirements, and safety (and licensing) conformance, and others as appropriate. The responsible engineer with the concurrence of his manager, determines the areas of review requ! red for the particular document. The review always includes a review of quality requirements by Product Assurance engineers to evaluate quality characterfstics and requirements, and verify ! that the document contains or references the appropriate requirements to

achieve and verify required quality. The Manager - Licensing and Reactor i Systems identifies those documents requiring his review to evaluate requirements essential to safety and/or licensing.

The responsible engineer evaluates and resolves the comments received during the review process. The responsible engineer reviews the final document with O the prior reviewers when there is a significant change or a significant comment is not included in the final document. The Manager assigned responsibility for preparation of the document is responsible for verifying that technical requirements and objectives have been met, necessary reviews completed, and comments resolved prior to approving the document for issue.

AlI changes to design data which have been baselIned (l.e., approved by LRM or equivalent) are initiated by an Engineering Change Proposal (ECP) which is the vehicle for identification of the change, documentation of the impact of the change, and for obtaining approval of the change. The Engineering Change Notice (ECN), is the vehicle for making precise changes to specific design documents and implements the ECP. The review, approval, and release is accomplished in the same manner as the original design document. Discovery of a design deficiency or error in issued, or released, design documents, that if undetected would have adversely af f ected saf ety-related l structures,systemsorcomponents,aredocumentedandcorrected,asdiscussed in Section 16.0 of this Appendix. O q 171-15 Amend. 62 9 Nov. 1981 i

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

Parts, materials and processes are selected by the responsible engineer on the basis of proven experience or qualification for the intended service. For CRBRP, a common set of material data, extrapolation, and Interpretations contained in the Nuclear Systems Materials Handbook are the basis for design definition. How ever, for items designed to the ASME Boller and Pressure Vessel Code, the Code data takes precedence. Application reviews are conducted for essential parts, materials, or processes, and include, as applicable, (1) evaluation, analysis and tradeoff studies, (2) coordinat ing and interf acing with other participants, (3) document reviews, and (4) design reviews, t Design documents including changes, and review and approval and verification records, are collected, stored, and maintained as described in Section 17.0 of l 41 this Appendlx. i l l l l l till i l { l l l i l l l 171-16 Amend. 52 Oct. 1979

4.0 PROCUREMENT DOCUMENT CONTROL C4 Procurement actions are initiated by a Material Request (MR). The MR Identif ies the material, equipment, and services to be purchased, references the applicable technical documents and quality requirements, and contains approval signatures. Procurement actions are initiated by the responsible engineer for procurement of plant hardware, or by a procurement specialist for the Fabrication Shop materials. Preparation of technical documents (i.e., drawings and specifications) for use in the procurement of material, equipment, and services is the responsibility of the responsible engineering manager having technical cognizance over the material, equipment, and services to be purchased as discussed in Section 3.0 of this Appendix, included (or referenced) in the technical documents are design basis technical requirements, including the applicable regulatory requirements, components and material identification requirements, drawings, specifications, codes, standards, test and inspection requirements and special process instructions, as appropriate. The requestor prepares the MR in accordance with documented procedures assuring that the recorded Information is accurate and legible; ref erenced documents are included with the procurement package; end product identif ication is recorded, and vendor and sub-vendor quality requirements are specified. The Quality Assurance Engineer reviews the procurement package to verif y that QA Program requirements are correctly stated, inspectable and controllable, ON that there are adequate acceptance and rejection criteria. In addition, by audit, the QA organization verifles that procurement documents have been prepared, reviewed and approved in accordance with the Department procedures. The MR is required, as a minimum, to be approved by the requestor's manager, an authorized Product Assurance representative, and an authorized financial representative. l The MR, coupleted and approved, provides direction to Procurement to prepare a document (Purchase Order, Subcontract, etc.) which becomes the " contract" lbetweenthesupplierandGE-ARSD. Procurement documents contain the requirements as approved on the MR. .O l l t 171-17 Amend. 62 l Nov. 1981 l

l Procurement documents are reviewed by the Quality Assurance Engineer to verif y that the MR requirements were correctly included as specifled by the MR. Material Request Changes are reviewed in accordance with the same procedures and are subject to the same review and approval requirements as the original documents and provides direction for change to the procurement document. MR's and procurement documents for spare or replacement parts, of safety-related structures, systems and components, are subject to identical controls as applicable to those used for original eaufpment. GE-ARSD procedures require that prior to contract award, each contending vendor of safety-related, complex and high cost materials, equipment and services, be evaluated by Product Assurance (and other Department functions as appropriate) to determine the capability of the supplier to meet the procurement qual ity requirements. An approved supplier list is maintained l current and used by Procurement to determine that each suppller is an acceptabIe source prior to piacement of the purchase order. A check to verify approved vendor status is also made when the Quality Assurance Engineer reviews the purchase order. Procurement documents are collected, stored, and maintained as described in Section 17.0 of this Appendix. O 1 l l 171-18 l Amend. 62 Nov. 1981

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

5.0 INSTRUCTION

S. PROCEDURE

S AND DRAWINGS O The GE Policy Procedures system is structured to conform with the criteria for 10CFR50, Appendix "B" as described herein. The system includes Department Policles and Instructions (BR's), Engineering Procedures (ENG's) and , Fabrication and Test Procedures. Also where appropriate, functional routines are used by a manager to control his or her area of responsibility. The Department PolIcles and Instructions are prepared, maintained and published by the Manager, Management Systems. A Policy and instruction Board, which is chaired by the Manager, Management Systems, is comprised of representatives of each Section including Product Assurance and Services. The Board reviews and concurs in the Department Policies and Instructions, with final approval l reserved to the Department General Manager. Amendments are treated in a lIke manner. Deoartment Policies and Instructions (BR's) delineate the sequence of actions to be accomplished in the preparation, review, approval, and control of instructions, procedures, and drawings, and they provide overall direction for the conduct of Department business. Interim authorization is provided therein for the preparation of other derivative documents, such as, Engineering Procedures and Fabrication and Test Procedures, which cover specific fields of interest in more detailed form. The responsibility for the preparation, maintenance and publication of the Department PolIcles and Instructions is assigned to the Manager, Management Systems. All new, revised or amended procedures are approved by the General Manager following Interral evaluation. O Engineering Procedures (ENG's) provide authority to do engineering work in a disciplined manner. They give direction in critical areas such as, design reviews, control of engineering changes, design verification, use o+ codes and standards, and other aspects of engineering work. Engineering Pr ocedures are approved by selected representatives of Sections involved in d%Ign, development, quality assurance, and program control activities. The responsibility for the preparation, maintenance and publication of these documents is assigned to the Manager, Management Systems. I Fabrication and Test Procedures (FTP's) delineate requirements for Fabrication Operations personnel in the planning and performance of f abrication, i processing, assembly, test, and shipping of manufacturing items and the l operation of the Fabrication Operations f acilities. The responsibility for the preparation, maintenance, publication and approval of Fabrication and Test l Procedures is assigned to the Manager, Fabrication Operations. Drafting Routines, included within the Drafting Manual, implement Department l Instructions and Engineering Procedures to the extent of prescribing specific ! practices af fecting engineering documents, e.g., drawings, associated lists and specifications. The responsibility for the preparation, maintenance and t publication of these documents is assigned to the Manager, Drafting and l Documentation. I lO l 17l-19 Amend. 62 Nov. 1981 _ _ _ _ _ _ _ _ _ _ _ ~ . _ _ _ . _ _ . _ _ _ .__.

Manufacturing Process Instructions (MPI's) are prepared to define control instructions and parameters for complex or repetitive f abrication processes. MPI's are approved by responsible managers in the fabrication and product assurar.ce organizations prior to issue and are referenced on the Work Order Record (WOR). Qualltv Control Instructions (QCI's) are prepared that delineate quantitative and qualitative acceptance criteria. The responsibiIIty for the preparatlon, maintenance, publication and approval of Quality Control Instructions is assigned to the Manager, Product Assurance and Services. The Department instructions and Engineering Procedures provide quantitative and qualitative acceptance criteria to verify that important activitles have been satisfactorily accomplished. Department Instructions assign the responsibility to the Quality Assurance Engineer for reviewing drawings, specifications, and fabrication documents to ensure the inclusion of quantitative and qualitative acceptance criteria prior to release for fabrication. Instructions for the fabrication of items manufactured by ARSD are in Work Order Records (WORs). WORs are prepared for each manufactured item and, in conjunction with Manufacturing Process Instructions (MPIs) and Quality Control Instructions (QCis), define fundamental processes (such as, welding, standardized inspections, and non-destructive examination) to be used in fabricating any item. The WOR delineates the sequence of fabrication, references applicable technical documents, identifies inspection and customer or Authorized inspector hold points or witness points. Each operation is required to be signed and dated to provide evidence that the operation has been ccepleted by an authorized individual. l 6.0 DOCUMENT CONTROL i GE-ARSD produces large variety of documents in support of its work on CRP. The fofIowing types of documents are considered to be eritIceI to the definition, documentation, and control of the safety related aspects of hardware and therefore, have strictly controlled review, approval, and issuance requirements:

a. Design documents (e.g., drawings, specifications, stress reports, saf ety, reliability and other analyses) including documents related to definition and verification of computer codes that are used to substantiate designs.

l i b. Purchase Requests. l

c. Instructions and procedures for such activities as fabrication, m difIcatlon, instalIatlon, test, and inspectlon.

55 l l O l l 171-20 Amend. 55 l June 1980

d. As-butIt drawings.

e. Quality Assurance and Quality Control Manuals.

f. Topical reports that substantiate designs.

g. PSAR/FSAR.

h. Nonconformance reports.

I. Preoperational test specifications. Procedures for the review, approval, and issuance of these documents and changes thereto are established and described to assure technicai adequacy and inclusion of appropriate quality requirements prior to implementation. The QA organization, or an Individual other than the person who generated the document but qualified in quality assurance, reviews and concurs with these documents with regard to QA-related aspects. Approval changes are included in ins 1 ructions, procedures, drawings, and other documents prior to implementation of the change. Procedures to assure that obsolete or superseded documents are removed and replaced by applicable revisions in work areas in a timely manner. Master lists to identify the current revision of instructions, procedures, p specifications, drawings, and procurement cocuments. These Iists are updated V periodically and distributed to predetermined responsible personnel. An Engineering Review Memorandum (ERM) is utilized to record the review and comments of Individuals having technical competence to determine whether the drawings, spectfIcations, and test procedures, under consideratlon, adequately conform to applicable standards within his assigned area of review as discussed in Section 3.0 of this Appendix. Engineering Change Proposals (ECPs) and Engineering Change Notices (ECNs) are utilized in accordance with established procedures to initiate and approval changes to issued drawings and specifications. The ECP is a document that describes a proposed change to issued engineering documentation, and which when approved, authorizes the implementation of the change to the documents. The ECN is a document that records in detail, the changes to an issued engineering document and which initiates implementation 55 i by the Drafting and Documentation Unit. The issued O 171-21 Amend. 55 June 1980

ECN is delivered to the affected organization, e.g. Manufacturing or Procurement, for immediate change in the technical requirements. Revision of the affected document is concurrently accomplished by the Drafting and Documentation Unit. Procedures require that all changes excepting those of a clerical type be reviewed and approved, as a minimum, by the same organization that reviewed and approved the original document. Change authority and approval is achieved by means of ECPs which contain dete!!ed descriptions of each document to be changed. Follow!ng the ECP approval, the change implementation utilizing the t0N is directly accomplished in the affected documents, which are then released in revised form to the users. Records are maintained on the approval status of changes in-process and this information is readily available. Standard distribution lists are prepared and maintained to provide for internal and external distribution with required quantitles for issued drawings and specifications. lDocumentdistributionandcontrol is effected by two functional groups - A Communication Document Control (CDC) and a Technical Document Center (TDC). The CDC maintains files on correspondence, both internal and external. The TDC files and controls technical documents; such as, specifications, drawings, procedures, etc. Distribution lists are maintained to provide revised or changed documents to responsible personnel. Additionally, a library facility maintains files on general information and technical reports of a general nature. l l l l l l 1 l l l l l l 171-21a Amend. 62 Nov. 1981 l

An Engineering Drawing List is prepared and maintained as a complete and current list of drawing and specifications, including supplier drawings which are applicable, and is distributed as a minimum to Responsible Managers. The Responsible Managers disseminate the information to assure that proper and current documents are utilized. Documents controlled in accordance with 1he above procedures include: design specifications and design, manufacturing, construction and installation drawings. The review, approval and issue controls for the CPBRP Quality Assurance l Program index (QAPI) was previously discussed in Section 2.0 of this appendix. The GE-ARSD procedures and instructions referenced in the QAPI are reviewed, approved, and Issued in accordance with established procedures as discussed in Section 5.0 of this appendix. Section 4.0 of this appendix describes procurement document control. GE-ARSD receives controiled distribution copies of the Iicensing and principal design documents, i.e. SDD's issued by the Project. These copies are j received in the Technical Decument Center (TDC), where they are logged and then distributed to Responsible Managers. The Responsible Manager receiving the document is required to acknowledge receipt in writing to the TDC. The TDC acknowledges receipt to the issuing organization. Revisions to these documents tre received and acknowledged in the same manner. Manufacturing, inspectlon, and testing instructions are inciuded or referenced O on the shop traveler as discussed in Section 5.0 of this appendix. The Work Order Record (WOR) specifles the applicable revision of the referenced MPI or QCl. All revisions to WOR's are controlled and require approval by the same organization that originally approved the WOR's. Revisions to MPI's or QCI's require a corresponding revision to the WOR prior to implementation. The Work Order Record will normally remain with the item being fabricated until the final inspections are completed to assure that the proper revision of instructions are utilized. For speciel cases where the item is sent outside of GE-ARSD for vendor operations (e.g., special NDE), the item is identified with its Work Order Record Number, but the Work Order Record itself stays with l l the Manager, Fabrication Operations or his delegate, until the item is returned to GE-ARSD and they can be reunited. Document control, relative to source and receiving inspection of purchased

      'tems, is described in Section 7.0 of this appendix.

O 17I-22 Amend. 62 Nov. 1981 l

7.0 CONTROL OF PURCHASED MATERIAL. EQUIPMENT. AND SERVICES Precedures and practices are established and documented to provide assurance that purchased material, equipment, and services, whether purchased directly or through subcontractors, conform to the procurement document requirements. These measures include provisions, as appropriate, for the following: o source evaluation and selection, o review and coproval of supplier work piens and procedures, o appropriate objective evidence of quality furnished by the contractor, o inspection, surveillance, and audit at the source in accordance with written procedures during design, manufacture, inspection and test to verif y compilance wIth quality requirements, o examination of items upon delivery in accordance with written proceduros, and o review and acceptance of quality records required to be delivered with the item to the plant site. The measures described here are applied to all safety-related structures, systems and components not fabricated within GE-ARSD's faelllTies. In the event another General Electric Company Division or Department is awarded a contract to be a suppller to GE-ARSD, the measures described here are applicabIe. Source evaluation anc approval activities are initiated during review of the Material Request (MR) by the Quality Assurance Engineer as discussed in Section 4.0 of this Appendix. Instructions are entered on the MR, defining to

        +he purchasing organization the approval level to which a supplier must be quellfled. Approved Supplier Listings contain Information such as the product or service for which the suppller has been approved, and the quality level (including ASME Code) for which they are approved suppliers. The method of suppilcr evaluatior. may include past performance, supplier Quality Assurance Pr ogra.: cianual review, and/or on-site survey.

Whenever Purchasing pr $ oses to use a supplier that is not listed on the l approved supplier lisi ,>r is listed with a lower approval level, a request for 55l evaltation is initiated by Purchasing to Quality Assurance, who is responsible for performing the appropriate evaluation of the supplier's Quality Assurance Program. When appropriate, qualifled Engineering, Manufacturing, Procuremeni, etc., personnel participate in the supplier evaluation to ensure that a supplier possesses specific capabilities. Results of supplier evaluations are documented and maintained on file at GE-ARSD. O 171-23 Amend. 55 June 1980

The Quality Assurance engineer following revlew and approval of the procurement packege, discussed in Section 4.0 of this Appendix, directs the 55l Plans (RIP's). preparation The SIP or RIPof Source inspection specifies Plans (SIP's) the characteristics and/or to be Receiving verified by Inspectio inspection or test and the point at which the inspection wIII be performed. 55l The SIP or RIP is approved by Quality Assurance Management. Tha SIP or RIP includes provision for verification sign-off of each requirement specified. Supplier-generated documents, including design and manufacturing drawings and manutecturing and Inspection plans for complex equipment, are evaluated to 55l determir.e those characteristics, and supporting documer.tation, to be verified by in-process inspection or witness by GE-ARSD. Notification of these inspection and witness points is made to the supplier. Source inspection is performed at these points (prior to release for shipment) upon notification by the supplier. Source inspection may not be required when all quality characteristics to be verified are accessible for inspection upon receipt, or the quality of the items car. be verified by review of test reports, or by performance of acceptance test. The SIP's or RIP's define where specific characteristics are to be verified or inspected. Source and Receiving inspection Plans define applicable inspection criteria, quantitative and qualitative. Receiving inspection of supplier-furnished material, equipment and services is performed in accordance with predetermined RIP's pricr to use o. Installation to verify that:

1) The material, component, or equipment is properly identif ied and corresponds w Ith receiving documentatlon.

2) Required Inspection records, or certificates of conformance which attest to the acceptance of material, components, and equipment are available and acceptable. These records are supplied to the LRM or site in accordance with contract requirements.

3) Items accepted and released are identifled as to their inspection status prior to forwarding them to a controlled storage area, or releasing them for installation or further work.

1

4) Nonconforming items are segregated, controlled and clearly identifled until proper disposition is made as discussed in l Section 15.0 of this Appendix.

The results of source surveillance, inspection and receiving inspection are recorded on the SIP's or RIP's and provide appropriate objective evidence along w ith the certif ications of conformance statements of the quality of l material, equipment and services f urnished by the suppller. 55l Source surveillance and Inspecticn is performed or controlled by Product Assurance and Services personnel. For major procurements, resident representatives may be assigned to the supplier's f acility. Receiving inspection is performed by QC inspectors. O 171-24 Amend. 55 June 1980

,_.s     Personnel performing inspection, test and surveillance functions are qualified as described in Section 10.0 of this Appendix.

( A Vendor Case Record routine is specified as part of the quality requirements and is utilized by the supplier to report nonconformances to procurement requirements dispositioned " accept as is" or " repair," as described in Section 15.0 of this Appendix. Suppliers' certificates of conformance are verified as follows: (a) Material overchecks are periodically performed on selected materials to verify the accuracy of the supplier's certificates of conformance. ( (b) NDE supp!!ers' inspection and test report results are verified by comparing results obtained with test specification requirements and/or by verifying their NDE results (e.g. by reading of X-ray flims). For processes such as welding, suppliers' weld qualification samples may be independently verified when appropriate to achieve the required level of q ua l I ty. 52l GE-ARSD plans and performs, as appropriate, in-process surveIIIance ad inspections, final source inspections and/or receiving inspections to verify conformance wIth specifled requirements. 52l Quality audits are scheduled and performed by Product Assurance Audits to O verif y supplier ccrnpliance with specified quality requirements. These audits (b are performed as described in Section 18.0 of this Appendix. Requests for corrective action are submitted through Purchasing to the Suppliers for areas determined to be deficient. 52lThecontrolsforspareandreplacementpartsprocuredbyGE-ARSDareidentical to those described in this section of this Appendix. O 17l-25 Amend. 52 Oct. 1979

8.0 IDENTIFICATION AND CONTROL OF MATERIALS. PARTS AND COMPONENTS Procedures are established to define engineering requirements for identification of materials, parts, and components. The identification requirements include such items as model or part number, marking method, and nameplate location. The LRM has provided GE-ARSD the CRP equipment identification numbering structure for use in establishing identification requirements in specifications and drawings. The identification requirements for materials, parts and components are established by the responsible engineer during the initial design. Identification of materials, parts and components is maintained as required throughout manufacturing, fab"! cation, assembly, test and shipment to the customer. Identification and control of materials, parts and components fabricated by GE-ARSD are specified by procedures. These procedures provide the means for assuring that only correct and acceptable liems are used during manufacture. Physical identification is specified and utilized to the maximum extent practical. Identification is made either on the item or on reccrds traceable to the item. Where identification marking of the item is used, the marking is applied in a manner that will no* affect the item's function or quality. Raw materials are identified by tagging or other specified means. Identification includes Purchase Order Number, item number, applicable 55l specification number / drawing number, material type, heat, etc. Materials are introduced lato the fabrication process utilizing the Work Order Record / shop traveler which provides traceability and material identification requirements throughout fabrication, assembly, testing and shipping. The QC Inspector verifles that specified materials are being used end are correctly identified with the Work Order Record number. As fabrication of parts, subassembiles and assemblies progresses, re-identification is accomplished as prescribed by the shop planning. The QC Inspector verifles such re-identification operation. Noncenforming materials, parts and components are identified and segregated as described in Section 15.0 of this Appendix. Pequirements f or identification of purchased items are identifled in procurement documents in Section 4.0 of this Appendix. Purchased materials, parts and components are verified when received to assure that the identification conforms to the requirements specified by the Purchase Order, drawing or specification as described in Section 7.0 of this Appendix and is traceable to certifications. Only those materials, parts or components that conform to their identification requirements are accepted and released. Records that provide traceability of materials, parts or components quality history are accumulated and maintained as described in Section 17.0 of this Appendix. 171-26 Amend. 55 June 1930 k

9.0 CONTRCL OF SPECIAL PROCESSES Special process requirements are spe:Ifled as part of the technical documentation as described in Section 3.0 of this Appendix. Special processes performed at GE-ARSD, such as welding, heat treating, nondestructive examination and others are accomplished under controlled conditions. Personnel performing special processes are qualified (in accordance with written procedures) and certified as required by codes or other contractual documents. Special process operations are Identified on the Work Order Record (WOR) and specify the specific procedure or instruction to be utilized. Nondestructive examination procedures are prepared or approved by a Level ill examiner to define nondestructive examination requirements. The procedures define the methods, equipment, acceptance standard and final examination record requirements. A nondestructive examination report is completed for each examination performed which documents the technique used, the examination results and identification of the certified examiner, who performed the inspection by signature, level and date. In addition, the certifled examiner signs and dates the WOR. NDE personnel are qualifled and certifled in accordance with ASNT, SNT-TC-1A. Personnel are periodically recertified. Records that substantiate nondestructive examination personnel qualifications are maintained by the Level 111 Examiner. O Welding procedure specifications are prepared by a welding engineer and are J maintained in a GE-ARSD welding manual. Welding procedure specifications contain the methods and the essential and non-essential verlables. Welding procedures and personnel are qualIfled in accordance wIth appiIcable codes, standards, or other contract requirements. The welding procedure specifications, the record of welding procedure qualification tests, welder performance qualification tests, and the certified test reports are maintained as quality records. Heat treatment operations, including post weld heat treatments, specified on the Work Order Record when subcontracted to heat treat suppilers are controlled as described in Section 4.0 of this Append!x. Heat treatment records in the form of time temperature charts, or data with furnace identification, are obtained for each operaiion and are maintained as quality records. Cleaning and otner special processes. when used are documented in written procedures prepared by Manufacturing Engineering. Control of special processes for purchased items is specified on procurement documents as described in Section 4.0 of this Appendix and verified as described in Section 7.0 of this Appendix. O 171-27 Amend. 62 Nov. 1981

10.0 lNSPEC110N Inspectico and acceptance testing is perfcrmed by Quality Assurance personnel who are organizationally independent from organizations responsible for inspection and testing are 55 l performance of the activity being inspected. performed to procedures Qualificationbyistrained and qua based on demonstration of proficiency, skill, knowledge or experience, it is the respensibility of Quality Assurance management to assure that personnel under their direction receive appropriate training, maintain qualifications current and receive remedial training when quality trends indicate the need for retraining. Inspection planning, defining the characteristics to be Inspected, drawing acceptance and rejection criteria and requirements for recording of inspection results are documented or referenced as a part of the Work Order Record as described in Section 5.0 of this appendix. For ASME code items, the WOR is reviewed with the Authorized inspector for his designation of mandatory hold and check points. All inspection operations are signed and dated as verification of completion. inspection and NDE reports of the results are referenced on the WOR along with the Nonconforming item Reccrd number when applIcab!e. Personnel who perform nondestructive examinations are qualified and certified as described in Section 9.0 of this appendix. Whenever direct measurement is not possible, process surveillance is utilized as specliled en the WOR. Modification, repair, rework and replacement operations are specified on Work Order Records, and such operatlens are completed and Inspected the same as the original item. Inspection results are reviewed by Quality Assurance personnel prior to release of a product for shipment to provide assurance that inspection requirements have been satisfied and that proper records have been prepared, inspection of purchased items la described in Section 7.0 of this Appendix. O 171-28 Amend. 55 June 1980

11.0 TEET CONTROL Test programs are identified and documented in accordance with established procedures. The test programs cover all required tests, including, as appropriate, prototype / qualification tests, hydrostatic tests of pressure' boundary components, in-process tests of manufactured items, and" proof.and acceptance tests prior to installation. Technical documentation in the form of specifications and crawings define the requirements for in-process tests, acceptance tests, and verification. tests of performance characteristics. . . Operating and Test Procedures are prepared by Engineering and issued as -1 specifications to define the operational requirements and test procedures for , conduct of prototype /qualifIcatlon, proof and acceptance testing. These ' procedures: (a) define the test objectives, (b) describe the test item and Interf aces with the test f acility, (c) describe unique handling requirements and equipment, (d) describe the test details, including acceptance and rejection criteria contained in applicable design documents, pnd enyironmental limitations, (e) Itemize the parameters to be measured, accuracy;roquirements, instrumentation response characteristics, range of v'erlables to by measured, error analysis requirements, (f) provide Instructions for-data. col't ection, reduction, analysis, and retention, (g) describe f acility and f acility equipment requirements, and (h) Quality Assurance inspection and check points. Test items falling to meet specified performance requirements are documented on a Nonconforming item Record and dispositioned by a Material Review Board ., consisting of representatives f rom Engineering and Quality Assurance, and for Os rework and repair dispositions a manufacturing engineer. _ The Material Review Board is responsible for cetermining that items repaired ' - ' and replaced are tested in accordance with the original or acceptable alternative design and test requirements. ', , Responsibility for acceptance of prototype / qualification test results is assigned to the responsible engineering manager. Tests required to be performed during the manufacturing process aredI'entified on the Work Order Record, the prime controlling document for all manuf acturing ' operations, including testing. Quality Control Instructl,ons (QCis) or 55 lManufacturingProcessInstructions(MPIs) are referenccd on the Work Order Record and provide detailed instructions for in-process-and fin.al acceptance testing of items, including (a) Instructions for test method, equisment and instrumentation, and (b) test prerequisites, such as calibrated Instrumentation, equipment, personnel qualification, condition and status of item to be tested, environmental conditions, acceptance criteria, and documentation requirements, for acceptable and nonconforming items. The Work Order Record contains provisions for designating and sign-off of HOLD and CHECK POINTS by the Authorized Inspector or customer representative. Responsibility for review and acceptance of in-process and final acceptance test results is assigned to the delegated Product Assurance Manager. 17l-29 Amend. 55 June 1980

i2.0 f0NIPOL OF MEASUR!NG AND, TEST EQUIPMERI Toc!s, gauges and insfruments used in measuren.ent, inspecticn, and monitoring b 'er product ecceptance are cailbrated and periodically recalibrated and cor. trolled in accordance 'wlih establ ished procedures. Procedures require that ca!'bratiorc star dards have an uncertanty (error) requirenent of no more than 1/10 to 1(4 of N.e allowable tole-ance el the equipment being calibrated, ual(ss limited by Sta 0-of-the-Art. Measurement and test equipment is calibrated utilizing strndards whose calibration is traceable to the U. S.

           ?,ational Rureau of Standarcs, receptea values of natural physical constants, or der ived '6y rat ion type of se; f-calibrat ion techniques.

Responr ib il i ty for calibratio9 ct test jnstruments is assigned to the GE-ARSD l Develrpyent Test and FabrLcatic,n Dper,at : ens. Responsibil ity for cal ibrat ion of toc 1s and gauges is subg6ctr ucted. included witnin this responsibility is performance of Initial and periodic calibration, establishment of calibration f re:;uency, cal ibration standards, cal ibration procedures, records and notif icat ion of Product Assurance when equipnent is f ound to bo out of c;libration, and identification of calibretion status. The assurancn that tools, gauges, and instruments used for product acceptance are of the proper' ~ range, types and accuracy to verif y conformance to design requirements is inc. responsibility of Quality Assurance. Surveillance is maintainec of tools, gauges, and Instruments being utilized by inspection and manufacturing to assure that they.are within current calibration. If a tool, gauge or , ins *rument is found to be beyond iis calibration due date, it is removed from use by tre Quality Centrol Inspector af fixing a Quality Hold Tag. When a tool, gouge, or ihstrument is found to be out-of-calibration during s per 'ocic recal ibrat ion,' Qual.i ty Control is imp ediately notified by the source pprforming the calibration and' informed of the r.ature aad extent of the error, pi investigation is conducted- to determir:ci he rnaterials, parts cr cctnponents t a;fected by the error reported. Based upon the investigation results, re-inspeciicns are made to verify the validity of the previous readings ob te. Ine d . lbacc'ptable e materials, parts, and components are submitted to the Materlat Rei:< w Board on a Nonconforming liem Record for disposition as described in Seption 15.0 of this AppengIx.

                                  /

GE-ARSD ut'lIzis procecures for control of measuring and test equipment which conformjnith tnu requirements of RDT F3-2T, Calibration Program Requirements, and tye.fteasuring and Test Equipment Calibration and Control Requirenients of RDT-r2-2. These procedures require the use of identification labels to

        ' indicate rext calibration due date.

GE-ARSD's majcr subcoatractors and suppliers which have RDT Standard F2-2 imposed, are rcquired to meet the applicable requirements of RDT Standard F3-2T, Cel'bration Program Requirements, and the Measuring and Test Equipment Calibration and Control Requirements of RDT Standard F2-2. All other suppliers 6re regulred to meet o'her calibration standards established by government or industry. Vender surveys and audits verify that these requirements era being complied with. O 171-30 Amend. 62 Nov. 1981

13.0 HANDLING. STORAGE. AND SHIPPING Procedures and Instructions are estcblIshed to provide control of handling, storage, cleaning, packaging, preserving, shipping release and shipping of material and equipment as necessary to prevent damage, deterioration or loss during nianuf acture and shipment. When necessary for a particular item, special coverings, special equipment, or special environmental conditions; such as, Inert gas atmosphere, specific moisture content levels, and temperature levels are specified by Engineering and verified by Product Assurance. As required, special markings or instructions are used to icentify, maintain and preserve a shipment, including Indication of the presence of special environment or the need for special control. Technical documents, such as specifications, specify the requirements applicable to special handling, preservation, storage, cleaning, packing and l shipping. The shop traveler is prepared by Manufacturing Engineering to provide instructions for performing these activities. For both GE-ARSD manuf actured items and for cirect-to-site procured equipment, Product Assurance verifles that identified quality requirements have been completed prior to product shipment release. These verification activities are discussed in Sections 7.0 and 10.0 of this appendix. Necessary Instructions or guidance for Site inspection, handling, preservation, storage and special controls are prepared and are delivered to the Site prior to or at the time of component shipment. GE-ARSD engineering personnel establish the requirements for special handling preservation, storage, cleaning, packaging and shipping. Personnel accomplishing these activities are hired to fill job descriptions which include job qualification requirements. Performing personnel utilize documented procedures to accomplish these activities. For any unique task of this type for which a procedure deps not exist, an Instruction or procedure is prepared and proper training is accompiIshed prior to use. O 171-31 Amend. 62 Nov. 1981

14.0 INSPECTION. TEST AND OPERATING STATUS The Werk Order Record is the prime controlling document for all in-process O 52 l manuf acturing inspection and test operations at GE-ARSD. The Work Order Record delineates the sequence of operations to be perf ormed and is used to document completion of machining, assembly, welding, inspection, examination, testing, preparation for shipment, and other operations. The Work Order Record accompanies the item throughout the manufacturing process. It identifies the item at all stages of manufacturing and provides the status of manufacturing inspection, examination and testing status. As each sequential step is accomplished, verification of completion is made by the performer, affixing his signature, date and pay number to the Work Order Record. Application of welding stamp is made directly on the part or on weld maps, referenced on the Work Order Record. By-passing cf inspection, test, and other critical operations can only be accomplished by revision of the Work Order Record and approval of the Quality Control Engineer. At completion of the manu f actu -Ing cycle, the Quality Control Engineer verifles that all sign-of f s required on the Work Order Record are ccenpleted. During the manufacturing process, any item found to be nonconforming is segregated from conforming items and Identified with a Quality Hold tag. The control of nonconforming items is discussed in Section 15.0 of this appendix. 52 l GE-ARSD imposes requirements on subcontractors and suppllers for identirication of witness and hold points in design, fabrication and test 52 cycles for GE-ARSD participation and specifier that audits will be conducted by GE-ARSD. Suppllers are required to submit for approval a program schedule including manufacturing and test activities. GE-ARSD then identifies witness and hold points within the schedule. 9 17l-32 Amend. 52 Oct. 1979

15.0 NONCONFORMIMG MATERI ALS. PARTS OR COMPONENTS O GE-ARSD controir materials, parts, components, and services which do not conform to specified requirements in accordance with documented procedures in order to prevent their inadvertent use, further processing, or shipment. Procedures provide for iden'Ification, documentation, notification, segregation, evaluation, and disposition of nonconforming items. Nenconformances are identified during inspection ope.ations performed by GE-ARSD or by supplier notification. Nonconformances are documented by GE-ARSD Quality Control personnel on a Nonconforming item Record (NIR) or by the suppiler on a Vendor Case Record (VCR). These oocuments describe the item identification, reference the applicable drawing, specification, or standard, and describe the details of the nonconformance. GE-ARSD f abricated items found to be nonconforming during manuf acturing inspection operations are removed from the production flow, segregated in a quarantine area (size permitting), identified by attachment of a Quality Hold Tag prepared by the QC Inspector and recorded on a Nonconformance Record (reverse side of the Work Order Record). Nonconf ormances which cannot be reworked to the drawing or specification using Standard Shop Practices and Procedures are further documented on an NIR for disposition by the M.R.B. Purchased items found to be nonconforming at Receiving inspection are identified with a Quality Hold Tag prepared by the QC Inspector. The NIR serial number is recorded on the Quality Hold Tag and Receiving inspection [T Plan to provide cross reference identification to the document submitted for O MRB disposition. A Material Review Board is established to evaluate and determine disposition of NIR's for items that cannot be reworked and for VCR's. The Board consists of authorized members representing Engineering and Product Assurance. The 55 l Material Review Board's disposition and directions are recorded on the NIR or VCR. When the nonconformance is associated with en ARSD manuf actured item and the item disposition involves rework or repair operations, the concurrence of a Manufacturing Engineer is also required. Material Review Board dispositions of " Repair" or " Accept as is" for end items are submitted to the LRM for approval, as required by contract, and on ASME Code materials, parts or components are also submitted to the Authorized Inspector for his review and concurrence. Items dispositioned as rework or repair are reprocessed and re-inspected using the Wcrk Order Record to assure that the reworked items conform to drawing or specification requirements and repair items conform to the repair procedure requirements. 171-33 Amend. 55 June 1980

Items dispositioned as reject are quarantined and ultimately disposed of either by scrapping or returning to the supplier. Manufactured items, when scrapped, are marked or deformea to preclude their use. To provide assurance jthatallactionsarecomplete, including C/A as described in Section 16.0, on each Noncenforming item Record prccessed, a final review and close-out of the document is performed by signature of the Quality Assurance PRB Representative. Copies of NIR's and VCR's are made part of the inspection records which support the qual ity of the material, part or component. Coples of NIR's and VCR's are also forwarded to the customer as part of the records package as required to fulfill code, specification or contractual requirements. NIR's and VCR's are analyzed ic determine quality trends and categorized by responsible function, fault code and fault category. As adverse trends are identif ied, corrective action is initiated. Quality Trend Reports are generated and forwarded to affected managemer.t for Information and action. O l l l l l l l l 1 O 171-34 f Amend. 62 i Nov. 1981 (

16.0 CORRECTIVE ACTION l Conditions that are adverse to quality (such as nonconformance, f ailures, malfunctions, deficiencies, deviations, defective material, defective equipment and other anomalies) identified both internally and externally are processed for corrective action in accordance with established GE requirements. Corrective actions are initiated following the determination of a condition that is adverse to quality to preclude its recurrence. GE identifies, documents and implements these corrective actions in accordance with documented procedures. Corrective action requirements result from evaluations of purchased and fabricated item nonconformances as cescribed in Section 15.0, design errors and deficiencies as described in Section 3.0, ar.d audit results as described in Section 18.0 of this Appendix. The Unsatisfactory CondlTion Record (UCR), Corrective Action Request (CAR) and l suppl iers Correc. Ive Action Request (SCAR) are used to document requirements for corrective action and follow-up. UCRs are initiated by GE-ARSD personnel who become awcra of e problem outside of any formal audit activity. C/RE can be initiated by Quality Assurance personnel and oihers acting in an audit capacity as a result of internal audit activities. Deviations 1(antified from external eudits require that corrective action by the Supplier be identified and submitted to GE. Request for corrective action for defiencies discovered by source and receiving inspection on supplier products are made using the SCAR, when appropriate. O l UCRs and CARS are directed to responsible organizations describing the adverse conditions and requesting corrective action to alleviate the problem. Responsible managers at Suppliers are required to respond to the request within a predeterm!ned time period. The recipient completes and returns the document to the issuing organization for review and acceptance with follow-up reviews performed to assure that actions con.mitted are implemented and effective. Upon determination that the action taken has been implemented and is ef f ective to preclude recurrence, the CAR cr UCR is closed out. All CARS or UCRs isst.ed by Product Assurance Audits which have overdue response or action dates are listed on a Monthly Status Report, distributed to Management, and reported at appropriate Management Review Meetings. l CARS and UCRs may identif y problems that require reporting by other reporting documents; such as, Failure and Unusual Occurrences Reports. These reports assure that appropriate GE-ARSD Management and customers receive notification l of unusual occurrences requiring technical evaluation or corrective action or that impact safety. O 171-35 Amend. 62 Nov. 1981

17.0 QUALITf RECORDS GE-ARSD collects, stores, and male.tains records necessary to provide documentary evidence of the quality of items and services in accordance with Procedures and Practice. The Manager, CRP is responsible for establishing a Quality Records Management System and its implementation through the Department's procedural system. The Quality Records Management System is responsive to applicable requirements and regulations and provides: A. for the identification, declaration and indexing of records to be preserved as Quality Records for the CRBRP Project. B. a method of positive identification and correlation of each record with the item or activity to which it applies C. a method of timely transfer of Quality Records to the Owner. The Manager, Technical Operations Support is responsible for estabiIshing a Records Management Organization. Activities to support the impIementetion of the Quality Records Management System wIII include: A. Coordination of a Record identification System B. Coordination of record files / storage as maintained by the responsible organizations and establishment and operation of werking and reference files within a controlled access area. C. Maintenance of a listing, as furnished by the CRP Section, of all indices and/or logs of quality records. D. Coordination of the transfer of records to the customer (owner) as required. Assure preservation and safekeeping of the records. E. Accepted Quality Records will be held pending authorization for transfer to the Owner. O 171-36 Amend. 62 Nov. 1981

s GE-ARSD records include inspections, test, audits, and material analyses; moniioring of work performance; qualification of personnel, procedures, and k equipment; and other documentation; such as, drawings, specifications, procurement documents, calibration procedures and reports; non-conformance reports; and corrective action reports. GE-ARSD inspection and test records for acceptance and delivery for CRBRP contain: A. A description of the type of observation. B. Evidence of compietIng and verifying a manufacturing, inspection, or test operation. C. The date and results of the inspection or test. D. Infcrmation related to non-conformance. E. Inspector or data recorder identification. F. A statement as to the acceptabilliy of the results. The requirements and responsibilities for recc.d transmittals, retention, and maintenance subsequent to completion of work, are consistent with procurement documents and applicable codes and standards. 0 55 O 17l-37 Amend. 55 June 1980

18.0 AUDITS GE-ARSD conducts a comprehensive audit program to verify that the QA Program requirements have been developed, documented, and ef f ectively implemented and complied with. The audit program is applicable in those quality program activities conducted directly by GE-ARSD as well as those delegated to suppliers. Responsibility for conducting the audit program is established in Section 1.0 of this Appendix. Internal and external audit plans are issued annually and updated on a quarterly basis. The audit plans delineate the activities, organizations, processes or products to be audited; the wcrk scope to be audited, i.e., project management, planning, design, procurement, manufacturing, records, training, etc.; and the planned schedules for conducting the audits. GE-ARSD internal and external audits, as well as vendor surveys include an objective evaluation of:

a. Work areas, activities, processes, and items, and the review of documents and records,

b. Quality-related practices, procedures, and Instructions and the effectiveness of implementation.

In developing the audit plans, the frequency, scope and activities to be audited are based on importance and status of the activity. All applicable functional areas are audited at least once during the contract for all systems or major cceponents. The audit team leader is selected on tne basis of his experience, training, demcnstrated capability to perf orm audits and f ami!!arity with the methods, procedures and codes and standards applicable to the audit scope. The members of the audit team are selected on the basis of their familiarity with the methods, procedures, and codes and standards applicable to the audit scope. No member of the audit team can have had any j direct responsibility for the activity being audited. 1 Prior to conducting audits, the lead auditor prepares a specific audit plan identifying the subject of the audit, purpose and scope of audit, audit dates, audit team and audit criteria; and an audit checklist which identifies 1he functional areas to be audited, the audit contact, the reference documents, and characteristics to be evaluated. The specific audit plan is approved by l the Manager, Product Assurance Audits. The audit checklist is utilized by the audit team to record their findings and cbserva+1ons. At the conclusion of the audit, an exit review is held with the management of the functional areas audited to review the findings and corrective action l requirements. Corrective Action Requests (CAR's) are prepared to document each significant discrepancy found and are directed to appropriate management as discussed in Section 16.0 of this Appendix. An audit report is prepared to document the audit results. Internal audit reports are distributed to the audited Managers; Manager, Product Assurance and Services Section; Manager, l Quality Assurance Support; M< nager, Product Assurance Clinch River Project; Manager, Clinch River Project Section; and other responsible Section and Subsection Managers. 171-38 Amend. 62 Nov. 1981

p. l External audit reports with all findings are distributed to the appropriate internal management and are submitted to the purchasing organization for i distribution to the supplier along with a request for corrective action.

Information concerning audit results is provided to the LRM. When results from a number of audits or other surveillance activities indicate that a generic adverse trend is developing, a Quality Trend Report is prepared and distributed to appropriate management. Follow-up audits are conducted as appropriate to verify the effectiveness of corrective action performed. Audits confirm compliance with established procedures fcr interface control. These audits cover interfaces with the LRM, other RM's and with subcontractors and suppliers. Activity audits, both internal and external, verif y the implementation and effectiveness of Indoctrination and training programs. A complete file of records to support each audit conducted is maintained. O 171-39 Amend. 62 Nov. 1981

O OWNER I LE AD RE ACTOR MANUFACTURER I GE-ARSD R E ACTOR M ANUF ACTUR ER R I SYSTEMS SERVICES COMPONENTS I SUBSERVICES

        =      SERVICES                                SERVICES       -
                                                                      =
        "                                              M ATERI ALS M ATERI ALS
        . COMPONENTS I

I I l M ATERI ALS SERVICES l Figure 171-1. Organization of Quality Assurance l Program Participation l l l l Amead. 52 O Oct. 1979 1/T-40 1

                     \                                                                       \

V 1 1 i ) 'l I GE l' ENERGY SYSTEMS & TE CHNOLOGY

DIVISION l

d 5 ADVANCED RE ACTOR SYSTEMS 3 DEPARTMENT 1 I 1 1 w N j 7 A ARSD FINANCIAL CLINCH RIVER APPLICATIONS PRODUCT ARSD N ENGINEERING ASSURANCE ARSD-LEGAL EMPLOYEE W SECTION & SERVICFS j & PLANNING REL ATIONS i i 4 7 TECHNOLOGY & DESIGN DEVELOPMENT

.;                           PROCUREMENT         SPECI AL PROJECTS                                       ENGINEERING           ENGINEERING

] July 1980 1 4 8 "" ' 2 O Figure 171-2, GE-ARSD Quality Program Management Organization

.          <m

. ::3 J

w .CL i CD G WN l j

ENERGY SYSTEMS

                                                                     - - - ~ - - -

I TECHNOLOGY '"] OlVISION I l I I

          !                                         ADVANCED L

l ____.__ REACTOR SYSTEMS DEPARTMENT _ _ __ ] I I I -._._ _. ._ l l PRODUCT ASSURANCE

                                                        &           _ _ _ _ _           __J SERVICES I

l l PRODUCT ASSURANCE QUALITY ASSURANCE CLINCH RIVER TECHNICAL & SPECIAL AUDITS PROJECT PROJECTS

• QUALITY ENGINEERING QUALITY MANACEMENT

              ~                                                  ~"
                   & VERIFICATION              ENGINEERING                              SYSTEMS
                   -PL ANT SYSTEMS O U ALITY ENGINEERING QUALITY
                   & VERIFICATION                               -                           &           -
                      -EQUIPMENT                 CONTROL COCUMENTATION F ROJECTS SCRS                   QUALITY
             ~
                                                                -                     OPER ATIO E      -

SURVEILLANCE SYSTEMS SUPPO RT

• OCES NOT MA KE DIRECT CONTRIBUTION TO CLINCH RIVER ACTIVITIES (StIPPORTS BASE PhCGRAMS AND OTHER ACRK)

Figure 171-3. GE-ARSD Product Assurance Organization 9 Amend. 62 171-42 Nov. 1981

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

8 9 9 i k

                                                                                                                                                                                                                                                                                                                                            ]
 }                                                                                                                                                                                                                                                                                                                                          '

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I t

PROGR AM MANAGEME NT CORRECTIVE ACTION E NGINE E RING % t C', } ORG ANIZA TION DOCUME N T ATION AUDtTS AND REVIEWS l QUAltTY ASSUR ANCE PROGRAM 1 Planning 1. Responsetnlity and Authority 1. Pol cies and Pnn edures 1. Quatiev Audas

2. Traenmg and indoctemation 2. Quahty Recorets 2. Management Revievvs UNUSU AL OCCURRENCE REPORTING ,

2, Quality Asso..nce Program imien  :

3. Personnel Qualibcat.on 3 Quality Status Repo,es f

4 h M ANUF ACTuntNG, F ARHIC ATION AND ASSEMBL Y  ; PROC UM E UF N T i l'ESIGN AND DE VE LOPME NT Procurement Plannmg Planmnq Design Ptanmnq ' l Ps ocurement R equirements inspection arwt Test Plan Des >75 Detmetion and Control

                                                     " ' ' " '"*                                                                                                 Ptor urement Document Revieves                                            Matereat taentehcation and Cort u3                                                               e

2. Cortes. Standaros and Practices Control of Processes

3. Engmeenne Stud.es Evaluation and Select.on of Procurement Sources 1 F abrication and Assembly Processes i

4. Parts. Matenals and Processes i

5. Design Desenptions t , cen,,al Requirements 2. Process Quahfication 6 Specifications. Drawmgs and 2. Acceptable Source List 3. Nondestructive E mammation Instructions 4 Cleenmg

3. Pre Avverd Evaluation

7. Identification 4. Interchange of Source inmorteori and Tests

) B. Acceptance Cnteria Capatmhty Information I 1 General Requirements I w 9. Interf ace Control Control of Confessration 1 y 2. Procedures Document Revieve and Control 3 g j 4 Inspect *on Status Irwiscation t

1. Document Revieves 2. As Built Veratication '

4 S. Certification w 2. Document Control woment Cawation and Stan&s

3. Engineenng Diavumg L ests Document Control 60" 8"'** *"C ' ""d '"'P*C I'0" Rwem Enuipment Cabbration and Standards Receiving inspec tion ,

D e m nt 1. E quipment E veiustion

1. annmq and inwm 2. Control of inspection Measurmg and Test Failure Reportmg and Corrective

                                                                                                                                                                       'O'""'*"'*""                                                                     Equipment                                                                          6 I                                      Ac rvon                                                                                                                        3 Dispos 4tioning of Received items Conteof of Nonconformmg iterns                                                  4 Discrepancy Equipment Control of Receeved items                                                 Statistical Quahty Control and Analysis                                                        [

Control of Nonconformmg Items Corrective Action j O> Handimg. Preser vation. Packa.png. Stora.y and Shippmg O3 et (D 1. Handhng

       .            5                                                                                                                                                                                                                             2. Preservation. Pack aging and Storage                                                  j C3-                                                                                                                                                                                                                              3 Shipping                                                                               '

e mn D tQ Figure 171-4. Major Elements of the GE-ARSD-RM QA Program l 6

TABLE 171-1 GE-ARSD Quality Assurance Procedures Index versus Reauirements of 10 CFR 50, Appendix "B" Sheet 1 of 7 ICCFR50 App B Criteria TItIe impi m entIng Document

• l Organization Organization L ist - ARSD Product Quality Quality Assurance Program index 03-002 (Section 1.2 "Organizatl~

and Responsibilities") ll Quality Assurance Quality Assurance Program Index 03-002 Program [ lil Design Control Engineering HOLDS

  -                                                        Project Baseline Definition and Documentation i                                                        CRBRP Quality Verification Planning A                                                        CRDRP Working Level Design Data CRBRP Saf ety Categories functional Classification Design Criteria Cost Account Planning ConfIguratton Management Design Definition Documentation Structure Engineering Work Records Design Data Hold Design Data Hold - (RBRP Project Engineering Change Notice Engineering Change Notice - CRBRP Project CRBRP Conf iguration Management System CRDRP Conf Iguration Control Board CRBRP Engineeri ng Change Proposal CRBRP Walver Requests Supplier Documentation Design Reviews - Internal Design Reviews - Custcmer Oriented Technical Documentation Technical Documentation - Drawings oB

<g

             'See Attachment 171-1 for description of implmenti ng Documents, cL dR O                                                        O'                                                           O - - - - -

TABLE 171-1 GE-ARSD Quality Assurance Procedures Index versus Requirements of 10 CFR 50, Appendix "B" Sheet 2 of 7 10CFR50 App B Title implementing Document

• Crlteria Technical Documentation - Specifications and RDT Standards Ill (Cont.)

Engineering Memoranda Standard Distribution RDT Standards Review, Apprcnal and issue Drawings, Specifications and Standards Engineering Changes Preliminary Engineering Change (PEC) Engineering Configuration Definition System (ECDS)

 ~

Technical Documentation j IV Procurement Document Control Configuration Management e Engineering Release Instruction

 $                                                       Material Request Material Request - CRBRP Project Supplier Documentation Interdivisional Work Orders Request f or Proposal or Quotation Procurement Contract Preparation Purchasing Action Approval Purchasing Action Approval - CRBRP Project Purchase Review Board Letter Order V               Instructions,                   acceptance Criterla Procedures and                  Receiving, in-Process and Final inspection Drawings                        Fabrication Engineering Manufacturing System Noncoded items Fabrication Part List and Material Ordering Sheet QA Manual for Comp!!ance with ASE Boller and Pressure Vessel Code, Section lli Divlslon 1, Sections 4.0 and 11.0
          *See Attachment 171-1 for description of implementing Documents.

<g 2

TABLE 171-1 GE-ARSD Quality Assurance Procedures Index versus Requirements of 10 CFR 50, Appendix "B" Sheet 3 of 7 10CFRSO App B Criteria Title impimenting Documenta V (Cont.) Quality Control Instruction and Product Assurance Instruction Technical Documentation Technical Documentation - Drawings Technical Documentation - Specifications and RDT Standards Work Order Record VI Document Control Reports Configuration Management Control of Ccunputer Codes Engineering Work Records [ Engineering Change Notice 7 Engineering Change Notice - CRBRP Project n Engineering Release Instruction m CRBRP Working Level Design Data CRBRP Conf Iguration Management System CRBRP Configuration Control Board CRBRP Engineering Change Proposal CRBHP Walver Requests Suppller Documentation Technical Documentation Technical Documentation - Drawings Technical Documentation - Specifications and RDT Starderds Engineering Memoranda Standard Distribution Review, Approval and issue - Drawings, Specifications and Standards Engineering Changes A-E or E-C Design Change Engineering Configuration Definition System (ECDS)

           *See Attachment 171-1 for descripiloa of Implementing Documents, z>

O B . g 04 ~~ O O O

T i 4 I TABLE 171-1

GE-ARSD Quality Assurance Procedures Iridex versus j Requirements of 10 CFR 50, Appendix "B" i Sheet 4 of 7

( 10CFR50 App B ] Criterla Title implementing Document

• Vll Control of Purchased Procurement l Supplier QuallfIcation 4 Materials, Equipment i and Services Business Managed Procurement CompetItIvv Procurament i

Proposal Evaluation Plan

]                                                                                          Proposal Evaluation l

Cost / Price Analysis Supplier Negotiations }; Cost or Pricing Data Purchase Action Approval N Purchase Action Approval - CRBRP Project

• Purchase Review Board I b Contract Award Letter Order l

I Source inspection Configuration Management , Receivinc, in-Process and Final Inspection t i Supplier Contract Administration Material Request 1 Material Request - CRBRP Project Supplier Shipment i 1 plier Shipment - CRBRP Project

 '                                                                                              elving Purchased Material Vili             Identification and              Configuration Management

] Control of Materials Final Inspection - Acceptance Stamps Fabrication Engineering Manufacturing System: Noncoded items l QA Manual for Compliance with ASE Boller and Pressure Vessel I Code, Section 111, Division 1, Sections 3.5, 4.4, 4.5 and 5.3 Handling, Preservation, Packaging, Storage and Shipping

'                                           'See Attachment 171-1 for description of implementing Documents.

sF

         .g 1

1

+        w .O-O

~~

i

~

i

TABLE 17I-1 GE-ARSD Quality Assurance Procedures Index versus Requirements of 10 CFR 50, Appendix "B" Shee. 5 of 7 10CFR50 App B Criteria TItie impiementIng Document

• Vtil (Cont.) Material identification and Control identification System Receiving Purchased Material Supplier Contract Administration Technical Documentation Technical Documentation - Drawings Technical Documentation - Spect f Ications IX Control of Special Receiving, in-Process and Final Inspection

~ Processes Final Inspection - Acceptance Stamps y Fabrication Engineering Manufacturing System: Noncoded items QA Manual for Compliance with ASME Boiler and Pressure Vessel g Code, Section ill, Division 1, Sections 4.5 and 4.6 o3 Technical Documentation Material Request Material Request - CRBRP X Inspection Fabrication Engineering Manuf acturing System: Noncoded items Fabrication Parts List and Material Ordering Sheet QA Manual for Compliance with ASE Boller and Pressure Vessel Code, Section Ill, Division 1, Sections 3.4, 3.5 and 4.5 Quality Control Instructions and Product Assurance Instruction Work Order Record XI Test Control Design Verification Development and Qualif ication Testing Development Review Engineering work Records QA Manual for Cornpliance with ASK Boller and Pressure Vessel Code, Section li t, Division 1, Section 4.8

                                                                               *see Attachment 171-1 for description of implevnenting Documents.

5$ 5eo. co m wN O O O

(

                                                               \                                                                           ,

TABLE 171-1 i GE-ARSD Quality Assurance Procedures Index versus Requirements of 10 CFR 50, Appendix "B"

                                        ,                  Sheet 6 of 7 i

10CFR50 App B Criteria Title leptementing Document

• XI (Cont.) Quality Control Instructions and Product Assurance Instructions Technical Documentation - Drawings Technical Documentation - Specifications and RDT Standards '

Test Authorization and Record work Order Record Xil Control of Measuring Measuring and Test Equipment Calibration and Control and Test Equipment QA Manual for Compliance with ASE Boller and Pressure Vessel Code Section ill, Division 1, Section 7 [ Xlli Handling, Storage and Fabrication Engineering Manufacturing System: Noncoded items y Shipping QA Manual for Compliance with ASE Boller and Pressure Vessel a Code, Section li t, Division 1, Sections 4.6 and 4.7

• Handling, Preservation, Packaging, Storage and Shipping Shipping - CRBRP Project Suppller Shipment Supplier Shipment - CRBRP Project Quality Audit Work Order Record XIV Inspection, Test and Final Inspection - Acceptance Stamps Operating Status Fabrication Engineering Manuf acturing System: Noncoded items QA Manual for Compliance with ASE Boller and Pressure Vessel Code, Section 111, Division 1, Sections 4.5 and 11.2 "

Work Order Record XV Non-Conforming Receiving, in-?rocess and Final Inspection Materials, Parts or Febrication Engineering Manufacturing System: Noncoded items Components Failure Reporting and Management QA Manual for Compilance with ASE Boiler and Pressure Vessel Code, Section lil, Division 1, Section 11.0 l zg 'See Attachment 171-1 for description of implementing Documents.

    $a i  c SR h

TABLE 171-1 GE-ARSD Quality Assurance Procedures Index versus Recuirements of 10 CFR 50, Appendix "B" Sheet 7 of 7 10CFR50 App B Criteria Title impimenting Document

• XV (Cont.) Supplier Identified Anomaly Anomaly System Source inspection Receiving Purchased Materials O Supplier ideatified Anomaly Ti XVI Corrective Action Unusual Occurrence Reporting and Management y Unusual Occurrence Reporting and Management - CRBRP Project Quality Audits

?                                                                Annmaly System

<c O C Unsatisf actory Condition Anomalies 7 c.n Source inspection

• O Receiving, in-Process and Final inspection Fabrication Engineering Manufacturing System: Noncoded Items y QA Manual for Ccnpliance with ASE Boiler and Pressure Vessel 8

Code, Section li t, Division 1, Section 11.0 Failure Reporting and Management v XVil Quality Assurance Record Control Records Shipping Shipping - CRBRP Project Fabrication Engineering Manufacturing System: Noncoded items QA Manual for Compliance with ASK Bolle end Pressure Vessel Code, Section li t, Olvision 1, Sectim 9.0 XVili Audits Quality Audits

                 *See Attachment 171-1 for descr iption of Implementing Documents.

O >3 E I$a.

       ~~

9 O e

ATTACHMENT 171-1 GE-ARSD OUALITY ASSURANCE DOCUMENI_ DESCRIPTIONS

1. Anomalv Svstem This Instruction provides a summary of the departments system for documenting, controlling, reporting and discharging corrective actions for those conditions adverse to the quality of a department product.

2. Business Managed Procuremeqi This Instruction defines the objectives, application, Individual responsibilities and general approach to operation of a procurement team.

3. Comoetitive Procurement This instruction defines adequate price competition, single source, sole source, and requires pursuit of competition.

4. Configuration Management This policy establishes the guidelines for configuration management as a discipline that provides a system to certify that the product conforms exactly and on a continuing basis to its related documentation package through each of its revisions and changes.

O 5. Contract Award This Instruction defines the requirements for executing the contractual document.

6. Control of Comouter Codes This Instruction prov! des basic information for the control of computer codes used for design purposes.

l I a i O l 171-54 Amend. 62 Nov. 1981 l

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

7. Cost Account Planning This Instruction describes the requirements for the content, preparation and responsibilities for Cost Account Planning.

8. Cost or Pricino Data a

This instruction provides the means for compliance with Truth in Negotiations Act, and its principals as stated in federal regulations.

9. Cost / Price Analvsis I

This instruction defines essential terms used in cost / price analysis, provides for the performance of cost / price analysis and establishes responsibility. ,

10. CRBRP Configuration Control Board This Instruction describes the procedure and responsibilities in establishing a Configuration Control Board for the Clinch River Breeder Reactor Plant Project, defines membership and scope of activities.

11. CRBPP Configuration Management Svstem This Instruction identifies the overall requirements and organizational responsibilities for Configuration Management and related activities for the CIinch River Greeder Reactor Plant Project. These requirements O include those actions and responsibilities necessary to define and Implement systems and procedural Instructions to control the activities related to Configuration Identification, Configuration Control, Configuration Accounting and Configuration inspection / Verification.

12. CRBRP Design Data and Basellne This instruction contains the approval process and the required approval l

authority for a baselined design document to be released for Clinch River Breeder Reactor Piant Project use. i

13. CPERP Engineering Change Prooosal This instruction provides the requirements for preparing, formatting, i

reviewing and approving an Engineering Change Proposal for the Clinch ! River Breeder Reactor Piant Project.

14. CPERP Oual Itv Records

, This instruction implements the Quality Records Management Plan and l responsibilities for the Clinch River Breeder Reactor Plant Project. l

15. CRBRP Safety Categories

This Instruction oefines the system of classiffeation of equipment for the Cllnch River Breeder Reactor Plant Project according to its risk l potential during the various phases of plant operation.

171-55 Amend. 62 Nov. 1981

    ---,-...,c. ,,_,..,,.,.m     .- - -     ,   . - , - . , , y m,,,   m -.._,-.-m,.r_--          - . - ~ - - , . . .       ,.m... ,, .._,_..,.e...m,--..-->..,             , ~ , . . . - - . . , - -

16. CRBRP Walver Recuests This instruction establishes a method for walving a requirement on a specified cceponent(s) when unusual circumstances are encountered on the Clinch River Breeder Reactor Plant Project.

17. CPBRP Work!ng Level Design Data This instruction establishes a method for preparing, controlling and releasing the transfer of nonbaselined, working level design data that does not impact cost or schedule within the Clinch River Breeder Reactor P1 ant Project.

18. Desien Criteria This document establishes the requirements for documenting the cr!toria address areas such as performance objectives, reliability and safety requirements, applicable codes and standards, etc.

19. Deslan Data Hold This document provides instructions for identifying and monitoring Design Data Holds on Department drawings, specifIcatlens and standards.

20. Design Data Hold - CRBRP Project This addendum states additional or modified instructions for identification and monitoring of Design Data Holds for the Clinch River Breeder Reactor Plant Project.

21. Desian Definition Documentation Structure This instruction defines the hierarchical series of documents that provides control of and direction to a design as it evolves from the l initial general contract requirements to the specific characteristics and requirements necessary for fabrication and use.

l 22. Deslan Reviews l This document provides guidelines and responsibilities for formally l revtewing designs at component and systems level.

23. Design Verification - CPERP Project l

This document states the additional or modifled requirements and l responsibilities for preparation of CRBRP design verification plans.

24. Design Verification l

l l This document defines the requirements and responsibilities for l l preparation of design verification plans and the formal verification of l l design requirements. 1 171-56 Amend. 62 i Nov. 1981 l

f

25. Develooment and OualifIcation Testing N This document establishes the requirements for development and qualification testing, as necessary 1o demonstraie, evaluate and substantiate the fulfillment of the design objectives.

26. Develooment Review This document specifies the requirement for conducting and documenting management reviews for the purpose of integrating, guiding and evaluating criteria, testing and the application of development results.

1 27. Drawings and Soecifications This instruction establishes the responsibility and methods for documenting specific design definitions.

28. Drawings and Soecifications - CRBRP Project This instruction states additional or modified design definition for the Clinch River Breeder Reactor Plant Project.

l

29. Engineering Changes This instruction describes the requirements and responsibilities for initiating and approving engineering changes to all Department drawings, specifications and standards.

30. Engineering Change Notice This Instruction establishes the Engineering Change Notice as the vehicle which defines and authorizes changes to all Issued Department engineering documents, i.e., drawings, specifications and standards.

! 31. Engineering Change Notico (ECN) CRERP Project This addendum details the Clinch River Breeder Reactor Plant Project, procedures and organizational responsibilities for the preparation,

review, approval and processing of ECN's.

32. Engineering Configuration Definition System This document establishes the Engineering Configuration Definition System as the way to maintain a complete and current listing of all engineering drawings, specifications and standards, including the supplier's which

, are applicable to each project. 4 i I O 171-57 Amend. 62 Nov. 1981

33. Ingineering Release Instruction l

The purpose of this document is to establish the Engineering Release l Instruction as the vehicle by which Engineering authorizes and releases I engineering document for their intended project application. 34 Engineering Work Records This instruction defines the guidelines and responsibilities for the generation and retention of Engineering Work Records.

35. Fabrication Engineering Manufacturing System-Non-Coded items This document authorizes the ASME Code Quality Assurance Manual for non-coded-Items and providas modification instructions for use on non-coded items.

36. Febrication Parts List and Material Ordering Sheet The purpose of ihis Document is to provide guidance in the preparation and use of the Fabrication Parts List and the Material Ordering Sheet.

37. Faiiure RescrtIng and Management This document establishes the requirements for reporting, documenting, and correcting items which fall to function as planned after final acceptance.

38. Functional Classification This procedure provides for a gradation of quality requirements, criteria f or the proper selection of quality requirements and cctnmunicates the selection within the department classification.

39. Handling. Preservation. Packaging. Storage and Shioping l This document establishes the requirements for the handling, preserv?ng, packaging, storing and shipping of raw, in-process and completed parts and assembIles.

O 171-58 Amend. 62 Nov. 1981

_ _ = _ _ _ _ _ . . _ _ _ __ _ . _ _ _ _ _ _ _ _ _ _ _ _ , l 4 j ' 40. Interdivisional Work Orders O This instruction establishes the requirements for conducting procurement activity with other General Electric Divisions or Departments. l i

41. Letter Order This instruction defines policy and procedures to govern issuance and definitization of undefinitized orders or subcontracts which are to be documented on a purchase order or a subcontract form.

42. Manacament Holds This Instruction implements procedure for the Identification, reporting and management of holds imposed during design, development, procurement, f abrication, construction, installation, and operation activities. A hold is to flag or control an item for which information must be ,

r available before the item can be released for further processing. 4

43. Material Identification and Control i This document establishes the methods and control utilized in the manufacturing area to insure suitable material identification throughout processing.

44. Material Reauest O This instruction provides for the preparation and processing of a procurement action except for specialized items which control is already established.

45. Material Recuest - CRBRP Profect

> This addendum describes additional instructions applicable to Material Requests under Clinch River Breeder Reactor Plant Project's contracts.

46. Measuring and Test Eautoment Calibration and Control This document establishes the requirements for the calibration, adjustment, maintenance and control of all measuring and test equipment used to accept department products.

f 47. Organization List GE-ARSD i This instruction defines the function, title and name of Management l

                                             !                             personnel for GE-ARSD.

l 1 i I \ lO 171-59 Amend. 62 i Nov. 1981

48. Preliminarv Engineering Change (PEC)

The purpose of this document is to permit the preparation of PEC's f or expediting the immediate implanentation of on-the-spot-changes to developmental hardware and test facilities. PEC's are not applicable to prototypes, production work or products intended f or custmer shipment.

49. Preliminarv Engineering Change - CRBRP Project This instruction states additional or modif ied use of a PEC on the Clinch River Breeder Reactor Plant Project.

50. Procurement This document def ines the requirments f or performing procurment activities such as the purchase of material, equipment, and services required by the department.

51. Procurement Contract Summary and Contract Preoaration This instruction def ines a method f or summarizing a procurment action and preparing the required contractual document.

52. Product OualItv This document establishes a quality policy for the department. It def ines the broad structure of the quality systs to be used in implementing the policy; identi f ies prime and contributing responsibilities, relationships, and checkpoints for managerial reviews and actions; identi f ies relationsh ips with custcmers, vendors and government agency representatives and i denti f ies supporting document systuns for measuring product quality.

53. Project Baseline Definition and Documentation This instruction establishes a Project Baseline, including the Work Breakdown Structure, the Project Master Plan, Project Budget Baseline and Project Schedule Baseline.

54 Project Baseline Definition and Documentation - CRBRP Project This instruction adds or modif ies information f or baseline def inition and documentation on the Clinch River Breeder Reactor Plant Project.

55. Prooosal Evaluation This Instruction def ines the requirments f or opening proposals, saf eguarding proposal inf ormation and eval uation of proposal information.

, 56. Prooosal Evaluation Plan

This instruction def ines the requirment f or preparing and establishing a proposal evaluation.

17I-60 Amend. 62 Nov. 1981 _- j

                                                                                                                    \

57. Enocurement Actfor.Acorovat Thit instruction provides the requirement f or obtaining review and approval of a procurenent action.

1 l i < i 1 i 1 e i i I i i I l

i. O l

171-60 a Amend. 62 Nov. 1981 l l

58. Procurer.ent Action Acoroval - CRBRP Project This addendum describes the instructions f or purchase actions approved under the Clinch River Breeder Reactor Plant Project contracts.

I

59. turchase Review Board This instruction describes the activities and responsibilities of the Purchase Review Board in the review and control of critical ltan pr ocur ements.

60. Qualltv Assurance Manual for Ccepliance with ASME Boller and Pressure Vessel Code - Sfetion lil Division l This manual cescribes the QA Prog am and documents the controlled manuf acturing systen utilized by GE-ARSD to comply with the ASME B&PV Code, Section 111 Division I requirenents.

61. Qualitv Assurance Program Incex - CPERP Project ThIs document cescribes the Qual Ity Assurance Progran and Identi f les procecures which the departrrent shal l implanent to assure conf ormance to contr actual require-ents f or the assigned scope of work on Cilnch River Breecer Reactor Plant Project.

62. Qualitv Audits This instruction def ines the procedure and responsibility for establishing and implementing Internal and External Audit requirunents.

63. Quality Control Instruction (OCl) and product Assurance Instructions (PAI)

This instruction authorizes preparing: Issuing and applying of QCI's and pal's incl uding def Ining thei r f ormat and subject matter.

64. Cualitv Verifitaflon Pianning This instruction establishes the criteria and responsibility for Quality Verification Plans, incl uding test and Inspection plans.

65. PDT Standards This document establishes the criteria, procedures and responsibilities f or the use and development of RDT Standards of US D')E f unded prograns.

66. Receiving Purchased Material This document establishes the requiranents associated with the f Icw of purchased materials f rom the time of department receipt to its in-plant destination or return.

171-61 Amend. 62 flov. 1981

I

67. Receivino. In-Process and Final Insoection This instruction states the requirements for the inspection of purchased items received at a Department site, itms during the fabrication process and Itms prior to their ultimate use.

I l O i l l O 171-61 a Amend. 62 Nov. 1981

68. Receiving Insoection Deficiency Tnis Instruction provides a method f or review and disposition of a Receiving Inspection Def Iciency.

69. Record Control This document establishes the departments records system and identi f les the required quality assurance records, the record custodian, the r ecor d location, end retention period.

70. Recorts This instruction provides requirements f or the preparation, production review, approval, and release of reports and the conf ormation therein.

71. Recuest fer Procosal or Ouotation This instruction establishes methods f or obtaining proposals and qt.otations f rom suppliers.

72. Review. Aeproval and Issue. Drawings. Soecificat: ens and Standards This document establishes the requirements f or conducting reviews of technical documentation prior to initial i s s ue . Add i t i ona l ly , it speci f ies the requi rements f or resol ving comments made by technical personnel representing various disciplines (i.e. saf ety and licensing, QA, etc.) and the approval level required f or i s s ua nce. The requi rements f or documenti ng the above sequence are al so i denti f ied.

73. Shioning This document establishes the requirements which must be met prior to shipping an item ir.cl uding Product Assurance si gn-of f.

74 jihloning - CPEPP Pro iect This instruction provides direction f or preparation of shipping plan and related activi'y to Clinch River Breeder Reactor Plant Froject.

75. Source insoection This Instruction provices the instructions necessary to perf orm In-process qual ity hol d points and/or f inal inspections at supplier's facility.

76. Standard Distribution This procedure establishes the use of pre-es+=blished listing f or the Icentif ication of distribution f or issued t. ings, specif ications or reports, i

171-62 Amend. 62 flov. 1981

77. Sucoller Contract Administration d This instruction provides uniform routines for administering purchase contracts fran point of execution untii beginning of termination or cl osecut.

78. SucoIIer Documentation This instruction provides the requirments for establishing supplier document submittals and the processing, controlling and reviewing and/or approving of suppller document transmittal packages.

l O O 171-62a Amend. 62 Nov. 1981 l

79. Sucoller identiffed Ancealv This document establishes the system to be folicwed for the cocumentation and processing cf unplanned events which occur at a suppller's f aciiIty during the fulfillment of a contract with GE-ARSD. The Vendor Case Recorc is the form utilized.

80. Suceller Negotiations This instruction describes methods to be used in negotiating contractual documents and changes and defines documents which do not need to be negotiated.

81 Sueoller Evaluation This instruction provides syste i for evaluating and approving potential suppllers of materials, cceponents and services to be purchased.

82. SuoniIer ShIrment This instruction establishes requirements fcr effecting shipment frce, a suppller.

63. Sueoller 3hioment - CPERP Project This addendum prcvides additional requirements for effecting shipment f rec supplier directly to the Clinch River Breeder Reactor Plant site.

64 . Testing Anceal les

 '      Tals Instruction defines responsIbIIitles and describes the actIvitles for documenting, controlling, reporting and resolving any anomaly detected during testing.

O 17l-63 Amend. 62 Nov. 1981 i

85. Test Authorization and Record (TAR) n'

\-- This procedure describes the preparation and use of the Test Authorization and Record. A signed off tar authorizes initiation of a test.

86. Unsatisfactorv Condition Anomalles This instruction states the requirements for reporting and processing any unsatisfactory condition.

87. Unusua! Occurrence Reoorting and Management This document, in response to RDT Standard F1-3T, establishes the requirements for reporting, documenting and correcting the type of unplannod events known as " Unusual Occurrences". These events have by definition an impact on safety and/or a significant programmatic effect.

88. Unusual Occurrt nce Reoorting and Management - CRBRP Project This addendum describes additional instructions pertinent to and including certain events to Nuclear Regulatory Commission for the Clinch River Breeder Reactor Plant Project.

89. Work Order Record

/) \~ / This document estab!!shes the requirements for preparation, review and approvel of the Work Order Record. O 171-64 (Next Page is 171-69) Amend. 62 Nov. 1981

Attachment 171 -2 Schidule for issuing Unreleased Procedures item Title Schedule O I O 17l-69 Amend. 52 Oct. 1979  : i

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11. Juality SOP A-01 ISG Policles and Procedures '

Assurance 50P M-10 Progrm Managreent 50P Q-10 ESG Quality Assurance Pr oyrm SOP Q-16 SiatIty Assurance (QA) - Pr c, m Suppori f unct Ions t 50P Q-12 Quality Assuranco Progrm AuJ 's SOP y-18 ESG Quality Recor ds PMD No. 16 Quality A.surance Managrenent Reviews IFO No.11 CRBRP Document liold Status System i FMD No. 20 CRORP Training and Indoctr ination

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                                                                                                          >                                                                 Pr oc ur evnen t (w uments and Cm ponnnts                                                       CS*H 5.'                                                               Pr o< ur men t Qual i t y Ver I i f r a t ir.n Instr oc t Ions (c or : I nued )                                                    C S?M 7.3, 7.4                                                         Pr oc urmont Vnr i f icat ion ( r.ou r r o are! Per ol v i riq V er i f i c a t ion )

CS3M H.2 I dent i f f ra t ion and Con t r ol of Ptat er f al s an<1 Items CS3M 8.4 Materlal Chocklists CS5M 9 Contr ol of Const r uc t i on Pr or < < so s CS3M H.! Weldi ng and Br ailnq Matr r i al s CS!" 1 *- Noncon t cu m ing M iter f al s and ltoms Pf4 P+ 4 Centrol Stations I P U M- 5 -6 Froductive Matosfal (x >n t r o l l l IX. Control of E Pi 5 - 21 Mat er f a l s a,d Pr 4.r. *" <r C< n t r o l Syston l Spm l a l Pr m es!.es Q AOP N 5.02 [SG Sportal loollng l ~ Q f,OP P3.01 Mawf m f or ing Pr oduction Or dor ( Stop teaveler<> j N CS?M 9 Contr ol c' t Con st r uc t i on Pr oc es so =, I QAOP N6.01 Qu si l f f(at ion of Wol di ng Fr m edur er, a ro j w o l <1i ng Por son no t - j S. Q ACP t.6.02 Qualification and CortIfIcatIon of Nondns t r u< t I v o W CS3M 2.4 E x nm i na t i c,n Fer r.onon i i. CS?M 5.11 Cl eani ng Proc edur es Fictre 17J-4. ')uali ty Assurance Procedure Index vs. Requirer:ents of 10 CFR 50, Appendix B (Sheet 8 of 13) 52?

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j ESG Impiment ing Document cv Prm,edur e Appendix B - - - - - l Criterion Number TItIe i ._ _ . _ . _ _ _ _ _ _ _ _ _ _ . . _ . . . . . . . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ . . . IX. Control of QAOP N6.03 Special CS3M 5.9 Nondestructive Examination Prm.edures Processes QA0P N6.05 Qualification of Special Processns (continued) CS3M 5.4 Welding and Brming Procedures and Per sonnel CS3M 9.3 Control of Welding Operations CS3M 5.5 Heat-Treating Procedures CS3M 7.8 Subcontracted fur nace Brailng Services CS3M T.9 Subcontracted Heat Treat Sor wicas CS3M 7.10 Subcontracted Nondestructive Examination Se* vicns hit H-3-15 Qualif ication of Walders, Wnlding Oper ator s, and Welding Pr ocedures X. Inspection 50P K-90 Receiv ing and t r:spection of inccaing Mater f al and Equipment QAOP N1.21 Quality Assurance P1ans QAOP N1.22 Quality Assurance Acceptance Procedures OAOP N4.02 Procurment Quality veri f ication Instructions g QAOP N4.04 Receiving inspection N QAOP N4.03 Procurment Quality Assurance - Source inspection /',urvellianca p QAOP N4.04 Procurment Quality Assurance - Receiving inspaction CRBRP Receiv f rig inspection Overchack Requirrmants 4:. QAOP N4.02A CD QAOP N5.01 Manuf acturIng Production Order ( Shop Ir av ei er s) Figure 17J-4. Quality Assurance Procedure Index vs. Requirements of 10 CFR 50, Appendix B (Sheet 9 of 13) l

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                                                       '2-4                                             Control Stations u F3-10                                               Packaging and Shipping XIV.             Inspection, SOP F-00                                               Receiving and Inspection of locoming Mater t al and Equ'pment Test and Oparating    SOP K-84                                               Warehousing of Direct-Chargei Purchated Materials t$y Status                                                                    Traf f ic and warehousing 50P P-46                                               Handl ing and S tor age of Project Cr i t ical Hm dw are 50P Q-18                                                ESG Quality Records QAOP N1.21                                             Quality Assurance Plans QAOP N3.02                                              ESG special tooling QA0P N4.04                                             Procuroment Quality Assurance - Recniving inspection y                                     QAOP N5.01                                             Manufacturing Production O. der (Shop Treselw s)

G CS?M 9 Control of Construction Processes E QAOP N6.04 Weld Material Control 03 QAOP N7.00 Product Acceptance Tests QAOP N7.01 Pressure Testing QAOP N9.00 CS3M 14.2,14.3,14.4 Issuance, Use, and Contr ol of Stamps QAOP N9.02 Serialization of Ha dware QAOP N10.00 Nonconf ctn.Ing Materials and Items Figure 17J-4. Quality Assurance Procedure Index vs. Reaciraments of 10 CFR 50, Appendix B (Sheet 11 of 13)

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fcend. 62 17J-49 NOV. 1981

r_.__= - - i l l O O O I ) h ESG Impimenting Document << Pr oc edur e l Appendlx 0 - - - - - - - - - - - - - - - - - - - - - - - - ~ ~ - - - i l Criterion Nuriber TItie i _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ . _ _ [ 1 I ( XYl. Corrective 50P K-90 Receiving and Inspection of inccaing Material and Fquirent ! Action l SOP Q-14 Corrective Action Systen l SOP M-80 Unusual Occurrence Repor ts - RDT Progrms fHD No. 48 Unusual Occurrence Repor ting  ! EMP 5-19 Fallure Reports EMP 5-20 i nci dent Repor t s i QAOP N4.03 Procurment Quality Assuranca - Source inspec t ion /survei l l ance QAOP N1.04 Procurment Quality Assurance - Receiving inspection t QAOP N10.00 Nonconf orming Materials and items I , QAOP N14.00 Corrective Action I l CS3M 16 Corrective Action SOP Q-20 Reports to the Nuclear Regulatory Commission (NRC) Concerning Def ects and Noncompilances FND 18 CRBRP Qual ity Records Managment Syetem l C XV1I. Quality SOP Q-18 ESG Quality Records $ p Assurance EMP 3-1 Engineering Documentation Process . m Recor ds 04P 2.126 Case Fiie DocumeniatIon O QAOP N13.02 Quality Assurance Data Packages ' CS3M 17, 7.11 Procurment Quality Assurance Rncords CS3M Appendix A Contracting for the Fabrication of a Code i tem n in N-Cer ti t Icate ' Holder Retalning Ovnrail Responsibility by Cer a fication and  ; Stamping [ XVill. Audits 50P Q-12 Quality Assurance Progre Audits j QAOP N1.04 Quality Assurance Audits  ! CS3M 18 Audits I

                                                                                                                                                                                                                                                   )

Figure 17J-4. Quality Assurance Procedure Index vs. Requirements of 10 CFR 50, Appendix B t l (Sheet 13 of 13) 6 i

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a t we l e I CD Ch MN  ! t I I l t 6 i i

OUALITY ASSURANCE MANUAL PROCEDURE DESCRIPTl0113 STANDARD OPERATING POLICIES (SOF M SCP A ESG Policy and Procedures This S0P defines the types of ESG administrative policies and procedures authorized, and establishes minimum format and distribution requirements fer such pol icies and procedures. It identifies the F'gh , level of management, corporate or otherwise, f responsible fo il ishing qual ity pol icies, goals, and objectives. A clear path of cc. ... cation between Quality Assurance organization and corpcrate manage nent is defined. Positions and groups responsible for defining both content and changes to the Quality Assurance Program and manuals are Identif ied, in addition to the management level responsible for the approval of the Quality Assurance Program and manuals. Provisions are established for controlling and distribution of Quality Assurance manuals and revisions. SOF J Prepara lon and Processing of the Purchase Requisition This SOP establishes methods and policies applicable to the preparation and processing of the Perchase Requisitions (Form N25-R-2). The requisition is used for authorizing procurement, through Purchasing, of materials, equipment, and services from suppliers. l Procedures are established that delineate the sequence of actions to be l accocplished in preparation, review, approval, and control of the Purchase Requisition. lSOPK-90-Receivingof incomina Material and Equipment Receiving inspection of supplier-furnished material and equipment is per f ccmed in accordance with the following. The material is properly identified and corresponds with receiving documentation. Inspection is perfccmed and judged acceptable, in accordance with predetermined instructions, pricr to use. Items accepted and released are identified as to their inspection status, pricr to release. Nonconforming items are segregated, contralled, and identif ied until proper disposition is made. 17J-51 Amend. 62 Nov. 1981

S0P K Warehousing of Direct-Charged Purchased Materials by Traffic and Warehousing Methods are specif ied to identif y and control materials. Verification of correct identification of material, prior to release, is required. Material shall t,e protected against loss, damage, and deterioration f rom environmental conditions. lS0PP-46-HandlingandStorageofProjectCritical':ardware Special handling, preservation, storage, packaging, and shipping requirements are specified and performed by qualified personnel under predetermined ir,structions. S0P K Shipping Special packaging and shipping requirements are specified and accomplished by quallfled individuals, in accordance with predetermined instructions. Procedures are prepared in accordance with design and specification requirements which control the packaging and shipping of materials, components, and systems to preclude damage, loss, and deterioration. SOP K Material Handling Equipment Special handling requirements are specified and accomplished by quellfled Individuals, in accordance with predetermined Instrucilons. Procedures are prepared in accordance with design and specification requirements L which control the handling of materials, components, and systems, to prevent damage. , S0P Q Calibration of Measuring Instruments and Equipment Procedures describe the calibration technique and frequency, maintenance, and control for all measuring instruments and test equipment which are - used for obtaining data, where traceable calibrations are required. Measuring end test equipment is identified and the calibration test data l Is identified with the associated equipment. Measurement and test l equipment are calibrated at specified intervals, based on the conaltions I affecting the measurement. When measuring and test equipment is found to ( be out of calibration, any items measured with this equipment are withheld i until the accuracy of the results is evaluated. The complete status of all items under the calibration is recorded and maintained. Reference and transfer standards are traceable to national standards, if national standards do not exist, the basis for calibration is documented. I i O 17J-52 l Amend. 62 I Nov. 1981

SOP L Laboratory ar.d Engineering Notebooks It is the policy af the company to record all scientific and laboratory research and development activities in laboratory and engineering notebooks to be used by scientific and engineering personnel, primarily to record results of scientific studies and lab work, whether conpany or customer oriented. Innovations, inventions, discoveries, and improvements will be recorded for the purpose of fulfilling contractual obligations and protecting 4g company interests.

52) S0P K Procurement and Control of Supplier Data

? Procedures are established for preparation, review, and control I of instructions, procedures, drawings and changes thereto. These dccunents and changes thereto are procedurally controlled to assure adequacy. Provisions are established, identifying the personnel responsible for these activities. Changes are reviewed by the same organizations that performed the original review, unless delegated by the applicant to qualified responsible organizations. Approved changes are promptly included in the appropriate documents. SCP M Prcyram Management This SCP sets forth principles and guidelines for the managements of Energy Systems Group Business Programs. The Guidelines include 52l organizational framework, program management processes, performance l monitorir.g, and reporting systems. l 55 l SCP N Configuration Summary Reports l This SCP establishes the policies, methods, and responsibilities for the preparatiori, issuance, and use of Configuration Summary Reports. The primary purpose of this report is to aid the Manufacturing, Quality Assurance, and Engineering functions in determir.ing configuraticn and effectivity reglirements for product hardware. 52l 50P Q Corrective Action System Evaluation of nonconformances an. determination of the need for corrective action follow established procedures. Vampt corrective action is initiated, fol'owing the determination c. nonconformance to procedural or technical requirements. Adverse conditiens signi-ficant to quality, their causes, and corrective actions, are reported to the appropriate levels of management. Amend. 55 17J-53 June 19R0

w .. wwn. m .- w u w +. w .- u - -ea~~.e-ma.-_.w-- . i l i i !O I l l CRBRP PROGRAfi MANAGEMENT DIRECTIVES (PMD's) { t PMD CRBRP Correspondence Control I t This procedure delineates the method for identifying, controlling, and  ! accounting for all incoming and outgoing correspondence, and for capturing I 55152l commitments on the Commitment Status Report system. j l PMD CRBRP Document Hold Status System [ 1

45) t This procedure applies to holds and TBD': on all released (for 52l project use) Principal Design Data for wh:ch ESG is responsible.

The current status of each Hold and TBD in these documents which i impacts Level 2 or Level 3 activities is maintained in the Document Hold system as described in this directive.  ; PMD-12. - Quality Assurance Review and Approval of Engineering I I Requirements Documents f f I j l This directive establishes the requirement and procedure for formal l l 52l l review and approval by Quality Assurance personnel of ESG-generated l i 1) drawings, 2) specifications, 3) specification amendments, t I

4) Engineering's Change Proposals, 5) System Design Descriptions (SDD),

40 and 6) Engineering Orders. { PMD CRBRP Licensing Administrator This directive defines the responsibilities of the ESG CRBRP Licensing i i i Administratcr for implementing and controlling licensing criteria in accordance with Section 9.0 of the Management Palicies and Requirements l 52 (MPR). i PMD Schedule Cevelopment and Control  ! This directive delineates the method for development, processing, i , 52l approval, maintenance and change control of the ESG :;chedule j hierarchy which defines the CRBRP effort within the requirements i of ESG Program Management System. . This directive defines both the vertical integration of schedules i for CRBRP from the contractual interface to the detailed work package structure and the horizontal breakout over the '.ime of the j various schedular levels and documents. i I 9 17J-56 Amend. 55 June 1980 f s , t s

FMD Quality Assurance Management Reviews This procedure implements a Quality Program requirement for periodic quality assurance management review meetings to assess CRBRP Project quality accceplishments, discuss program qualliy audits, and resolve management problems affecting quality. PMD CRBRP Quality Records Management System This procedure implements the qualItf records req;irements of the CRBRP Management Policles and Requirements Document, Section 11.0, " Project Records Management", for ESGRM activities. PMD CRBRP SLS dreparation and Revision This procedure defines the methods for preparation and maintenance of CRBRP System nesign Descriptions. O O 17J-57 Amend. 62 Nov. 1981

PMD Materials and Processes for CRBRP This directive is estabilshed to ensure that all CRPRP design work will be based upon one common set of materials data as well as on consistent extrapolations and interpretations of these data. PMD Baselining of Documents This procedure gives the method for defining documentation as part of the CRBRP base lIne. PMD Review of Supplier Data This directive establishes spectfic requirements for the revlew of suppller data and augments the general requirements of 50P K-78. PMD Unusual Occurrence Reporting The purpose of this procedure is to provide for DOE Unusual Occurrence Reporting and for identification of those occurrences which require special consideration as deficiencies reportable under 10CFR50.55(e) and 10CFR21. PMD SHRS Reliability Program This directive defines the requirements of the reliability program at ESG on CRBRP. C PMD instructions for Required Documentation and Procedures for Shipment of Components to CRBRP Site or Other Designated Areas This directive describes the required documentation and the suomittal sequence to be followed prior to and during shipment of components and equipment to the CRBRP Constructor, Stone and Webster Engineering Company. PMD Acceptance Test Requirements and Specifications This directive defines the requirements for systems acceptance testing specifications which are to be prepared by Al-ESG. i PMD Storage, Maintenance, and inspection of Material Parts and Components. This directive describes the requirements and responsibilities for storage, maintenance, and inspection of material, parts, and components for CRBRP that are under the cognizance of ESG. l l l 17J-60 l Amend. 62 Nov. 1981

ENGINEERING MANAGEMENT PROCEDURES (EMPs) EMP 1 Pref ace to the Engineering Management Procedures Manual This procedure describes the scope of the Engineering Management Procedures (EMP) Manual. EMo 2 Engineering Studies This procedure establishes the requir(nent f or conducting stuules to establish that the cesign meets the design criteria, is based upon proven practices or analysis, and is adequate f or the intended serv:ce. It cescribes the method for preparing, releasing, and controlling Engineering Studies. EMP 2 Design and Acceptt;nce Criteria This procedure delineates the need for design and acceptance criteria to be defined and published in the appropriate design basis documents. EMP 3 Engineering Documentation Process This procedure describes the scope of the procedures which control the preparation, release, and control of speci f ications, draw ings, and reports by Engineering. EMP 3 Engineering Release System This procedure pro, ides instructions f or the preparation, numbering, l release, and control of drawings for the Engineering Release System, and provices guicel ines f or appl ication of the standard release. EMP 3-5.1 - Code Release System This procedure supplements EMP 3-5 and provides specific Instructions for ! the release and control of engineering drawings in compliance with the ASME l Code, t 1 l EMP 3-5.2 - Standard Release System l This procedure supplements EMP 3-5 and provides specif ic Instructions f or the release and control of engineering drawings in support of deliverable end items. EMP 3-5.3 - Experimental Release System This procedure supplements EMP 3-5 and provides specif ic instructions f or the release and control of experimental draw:ngs. O 17J-61 Amend. 62 l Nov. 1981

m n- --- *-- A - - EMP 3 Engineering Change Control This procedure defines the method for requesting, evaluating, approving,

 !                  and executing engineeri ng changes.

1 EMP 3 Interface Control This procedure establishes the criteria for Interf ace definition and the methods for describing and controlling the interf ace in appropriate documentation drawings and specifications. EMP 3 Control of Engineering Documents This procedure describes the methods f or control of drawing originals and prints, released by both the Standard or Limited Release Systems. EMP 3 Engineering Orders - Preparation Instructions This procedure describes the preparation and use of an Engineering Order to

.l                  release drawings or specifications, and defines requirements.

EMP 3 Preparation and Control of Supporting Documents This procedure establishes the types of supporting documents and defines the requirements for their preparation, release, and change. EMP 3 Component Traceabil This procedure describes the elonents and respon_ibility for establishing item traceability, i O 17J-62 Amend. 62 Nov. 1981 1.

EMP 3 Engi neer ! - Requirements f or Serial ization This procedure si~ ccnditions under wh!ch Engineering requires serialization o' 2ngcaen+s or parts f or traceability purposes. EMP 3 Mater , 5e a,titutions This procedure establishes the method and conditions un 1r which substitute materials, in .' lace of those called f or on engineering drawings, may be used, tMP 3 Request f or Document Change This procedure describes the f ormal treans f or requesting a change to a released drawing or specification and the approval and processing of that request. i EMP 3 Engineeri ng Managment System f or Speci f ications l This procedure def ines the method f or the preparation and control of Engi neer i ng speci f Ications. l EMP 3 Welcment Checkl ist This procedure provides the checklist to be completed f or critical weldmenTs, and the system f or its imp!omentation. EMP 3 Engineering Release Plan of Action This procedure gives the f ormat and requirernents f or a plan describing the means of preparaticn and release and approval of program documents, i l l l l l 17J-o3 Amend. 62 Nov. 1981

EMP 3 Documentation, Release, and Control of Scientific and Technical Computer Programs This procedure describes the documentation formats for scioatific and technical (S&T) computer progrms used and/or produced within the Research and Engineering Department. Those S&T programs that are developed outside of ESG shall also be documented to the same extent specified by this EMP alIowing f or vendor documentetion f ormats. EMF 4 Test Procedures This proceduro gives the f ormat f or preparation of Test Procedures. EMP 4 Test Reports This procedure gives the f ormat f or preparation of Test Reports. EMP 5 Design Reviews This procedure establishes the requirements for independent design reviews, and the means of their scheduling, conduct, and reporting. EMP 5 Checking of Engineering Drawings This procedura establishes the responsibilities f or checking of all engineering drawings. EMP 5-A - Fail ure Re,mrts Fall L,re Reports are to be used when a component or system under test has f allea or deviated from expected conditions on all ESG programs as defined in Paragraph 3.1. EMP 5 Incident Reports incident Reports are to De used when an incident or failure occurs in a tect other than on the component being te6ed on all ESG programs as defined in Paragraph 2.1. EMP 5 Materials and Processes Control System This procedure establishes the policy and responsibilities f or control of materials and processes. O u 17J-64 Amend. 62 Nov. 1981 l

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

l I EMP S Application of Standards This procedure provides guidance and direction for the application of codes and standards. It categorizes various types of standards and establishes responsibilities for their collection and application. i l 35 l CORPORATE ATO AI MATERI AL PROCEDURES (CMP's/AIMP's) l 45 I a0l AIMP 1.1.1 - Procurement Policy This procedure describes the procurement policy of Rockwell International, and supplements i'. tc cover procurement reflecting DOE requirements. 45 l 33 l 521 CMP 3.121 - Source Selection This procedure defines Rockwell International's practice concerning selection cf procurement sources and making commitments. 52l CMP 2.14 - Changes to Purchase Orders and Other Directions to Suppliers This procedure establishes standards for accomplishing changes to purchase crders and ef fecting other directiori to suppliers. 40 l CMP 2.126 - Case File Documentation This procedure establishes the documentation required to be i accumulated in procurement case files. 52l AIMP 3.109.1 - Procurement from Approved Suppliers This procedum requir>.; procurements to Code requirements, to ensure that Qualiti :4ssurance-approved suppliers are obtained. 40 Amend. 55 June 1980 h 17J-65

j QUALITY ASSURANCE MANUALS

• PROCEDURES O QA0P N1.00 - Preface to Quality Assurance Manual 40 The preface to each Quality Assurance Manual delineates the pure?se and authority of the manual.

I QA0P N1.01 - Quality Assurance Department Functions This document outlines the 'metions of the individual grNps .aithin the Quality Assurance Depar- 4 t. QA0P N1.03 - Vision Requirements for Quality Assurance Personnel This procedure establishes vision standards for Quality Assurance Department personnel and defines responsibilities for administering an eye examination program. 40 b2l0A0PN1.04,CS3M18-QualityAssuranceAudits 1 These procedures outline the Quality Assurance responsibilities for implementing and maintaining an audit program to determine the over-5 21 all effectiveness of the ESG and supplier quality programs and to identify areas where corrective prevention action is required. QA0P N1.21 - Quality Assurance Plans This procedure defines Quality Assurance Department responsibilities for participating in the preparation of Quality Assurance Program Plans or Quality Assurance Program Indexes and for preparing Quality Assurance Functional Plans. l QA0P N1.22 - Quality Assurance Acceptance Procedures This procedure defines requirements and responsibilities of the Quality Assurance Department for the preparation, release, and control of Quality Assurance Acceptance Procedures (QAP's). i QA0P N1.23 - Quality Status Reports This procedure establishes Quality Assurar.ce Department requirements and responsibilities for preparation of periodic Quality Assurance Program Status Reports and for submittal of the reports to Energy 52 Systems Group customers. i l

• Energy Systems Group Quality Assurance Department Procedures (QA0P) l E'2 Energy Systems Group ASME Code Section III Manual (CS3M) i

. O                4 l                                                                                                   Amend. 52

! 17J-66 Oct.1979 l

QAOP N2.03 - Document Control This proc 9 dure provides direction f or the control of engineering and shop draw ings, including custaner drawings applicable to products to be fabricated in the ESG Manufacturing Shops. The purpose of such control is to assure the f abricailon, proces si ng, inspection, and testing of products to the proper drawings. QAOP N3.00, CS3M 12 - Calibration of Measuring and Test Equipment These procedures def ine requiranents f or cal ibration control of tools, gauges, instrum?nts, and test equipment used by Manuf acturing and Quality Assurance to measure products (material s, parts, componen ?s, and appurtenances) or to control processes related to the product. QAOP N3.02 - ESG Special Tooling This procedere def ines the requirenents and responsibilities f or control of tooling used by Manuf acturing and Quality Assurance Depart ents in product f abrIcation. QA0P N4.00, QAl N4.00A, CS3M 4 - Procuranent Documents These procedures def ine requirenents and responsibilities f or preparation, rev iew, and approval of procuranent documents associated with the purchase of rnaterial s, parts, and services. QAOP N4.01, CS3M 7.2 - Approved Procurenen ? Sources These procedures def Ine Quality Assurance Department requiranents f or eval uation and approva' of procurenent sources (suppliers) of material, parts, and services used in ESG products. QA0P N4.02, (.S3M 5.3 - Procurenent Quality Verif ication Instructions fhese procedur es def Ine Qual Ity Assurance Department equirenents and responsibilities f or pr ecaring inspection !nstructions applicable to procured itans and services. QAOP N4.03 - Procurement Quality Assurance - Source inspection / Surveil lance This procedure def ines Qual ity Assurc...ce Department requi r ments and respons!.* ? lities f or quality verification of ,vocured Itans and services at a supplier's f;.cility. 0/OP N4.04 - Procurenent Quality Assurance - Receiving inspection These procedures def ine Quality Assurance Department requirenents and responsibilities f or inspecting ar.d testing incoming procured itens and services. 17J-67 Amend. 62 Nov. 1981

i f 1 QAl N4.02A - 01BRP Overcheck Requirenents This Instructica def ines requironents f or the overcheck ot raw  ! j materials procured for the CRBRP Project tcr ESG "make" Itans and f or materIat ! verif ication f or ESG " buy" Itans. l 1 1 i i  ; ( l' i I l i \ i l i i I i . l l l l i i i e  ;

                                                                        ". 7 J-67 a Amend. 62   >

Hov. 1981

QAOP N5.01 - Manuf acturing Production 0-der (Shop Travellers) This procedure def ines the requirements and responsibilities f or the preparation . 1 utilization of t.*s Manuf acturing Production Order (MPO). CS3M 9 - Control of Construction Processes These procedures def Ine the guidelines used to authorize and control the process, f abrication, instal lati on, inspection, examination, and testing of components, parts, and appurtenances. QAOP N6.01, CS3M 5.4 - Welding Procedures and Personnel These procedures establish requirements and responsibilities f or qualifying welding and brazing procedure specifications and welding and brazing personnel (welding, welding operators, brazers, and brazing operators) employeo in f abrication of Code items. QAOP N6.02, CS3M 2.4 - Qualif ication and Certi f ication of Nondestructive Examination Personnel These procedures establish requirements and responsibilities f or the training, examina1 Ion, quali fication, and certi fication of Energy Systems Group personnel engaged in the following nondesTructivo examination processes: p'g Radiographic Magnetic Particle Liquid Peietrant Eddy Current s ,/ U ltr asoni c Leak Detection , l l l O 17J-68 Amend. 62 Nov. 1981

l QAOP N6.03, CS3M 5.9 - Nondestructive Examination Procedures These procedures establish requirements and assign responsibilities for preparing and controlling nondestructive examination (NDE) procedures used for determining compilance of products to requirements of applicable codes and standards. QAOP N6.04 - Weld Material Control This procadere defines requirements and responsibilities for issuanco and control of we. l ding mater ial s (el ectrodes, rods, spool s, and f l ux) . QAOP N6.05 - Qualification of Special Processes This procedure defines requirements and responsibilities for qualification of special processes used during fabrication or inspection of products at Energy Systems Group. QACP N7.00 - Product Acceptance 7ests This procedure defines requirements and responsibilities of Quality Assurance Department personnel in performing acceptance tests, or witnessing acceptance tests performed by others on parts, material, subassemblies, assemblies, subsystems, and systems (items) that require acceptance by Quality Assurance. l QAOP hl.01 - Pressure Testing This procedure defines the requirements and responsibilities for perfcrming hydrostatic or pneumatit tests of ESG-fabricated ASME Code or other products. QAOP N7.02 - Qualification ana Certification of Visual and Dimensional Inspection Personnel This procedure defines requirements and responsibilities to provide a mandatory program of training, examination, and certitication for personnel performing dimensional inspection. The Drogram will provide perlediu updating to acccamodate changes in requirements and maintain the level o' knowledge necessary +o perform dimensional inspection assignments. QAOP N8.00 - Statistical Quality Control Program This procedure establishes Quality Assurance Department requirements and responsibilities for implementing and maintaining a Statistical Quality Control Program. O 17J-69 Amend. 62 Nov. 1981

i r i h 52l QAOP N9.00, CS3M 14.2, 14.3, 14.4 - Issuance, Use, and Control of Stamps  ! i These procedure- afine the requirements and responsibilities for the I l issuance, apph u. ion, and control of stamps used for markings that identify personnel ~arforming examination, inspection, test, welding, l j and brazing operations. t l QA0P N9.02 .. .alization of Hardware j t This procedure defines Manufacturing and Quality Assurance Department j l requirements associated with the serialization of parts and assei?blies  ! that are fabricated or procured by Manufacturing. 55 l

          ' QA0P N10.00, CS3M 15 - Nonconforming Materials and Itens These procedures define requirements and responsibilities for                       i control and disposition of nonconforming materials and items in the product manufacturing / procurement processes.

i i 53l QAI N10.000 - CRBRP Hardware Nonconformance Processing  ; i < This instruction supplements Procedure N10.000 by providing specific l ! details for CRBRP nonconformance items in accordance with LRM and 52 Owner requirements. l I' l QA0P N12.00 - Packaging and Shipping Inspection l i j This procedure defines Quality Assurance Department responsibilities  ! ' for inspecting and packaging and the preparation for shipment of  ;

ESG products. It applies to products requiring Quality Assurance j l acceptance that are shipped from ESG, to an ESG -anstruction site, i l

52 to an ESG customer, or to an ESG supplier. <

                                                                                                     \

i  ! l QAOP N13.02 - Quality Assurance Data Packages ! This procedure provides fonnat requirements for the preparation of , ! Quality Assurance Data Packages for transmittal to the customer. i Contractual requirements take precedence over this procedure, in l case of conflict. [ i  ! } i i ' l i Amend. 55 l Juna 1980 2 17J-70 1 4 l

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1  ! I l

QAOP N14.00, CS3M 16 - Corrective Action f or Nonconf ormance Products These procedures establish requircaents f or taking action to correct O conditions causing nonconf orming material, parts, and components. Its purpose is to provide increased assurance that ESG products will meet design, conf iguration, and pert crmance requircnents. lCS3M2.3-TrainingandIndoctrinatio.1 This procedure defines requirenents and responsibilities f or training and Indoctrination of personnel perfccming activities affecting quality or Code ccepliance, as necessery, to assure that suitable proficiency is achieved and maintained. CS3M 3, 6 - Design Documentation Control These procedures establish the requirm.ents and responsibilities as an Owner's Agent, tnd for the control of design activities and documents associated with items being constructed in accordance with the requirements of the Code. lCS3M7.3,7.4-Procurcce..tVerification(SourceandReceivingVerification) These precedures define requirments for source and receiving Inspection, ext.mination, and test of procured materials, parts, and services. CS3M 6-.2 - Identification and Control of Materials and items These ,nr:cedures define requircrents and responsibilities for implementing and maintcining ;rerial checklists required by the Code. CS3M 6.3 and Appendix A - Procurement as an Engineering Organization This procedure defines the requirccents and responsibilities when Energy Systems Group, as an Engineering Organization, assumes responsibilities, in addition to the design, for purchasing and contracting activities, it establishes quality assurance interf aces with the manuf acturer, material manufacturer, material supplier, or installer involved in the activities of the Engineering Organization. CS3M 13, 5.7, 5.8 - Handling Preservation, Packaging, Storage, and Shipment These procedures establish measures for handling, preservation, packaging, storage, and shipping to prevent damage to Code items. CS3M 8.,4 - Material Checktists This procedure defint. - ~JlrCcents and responsibilities for impICcenting and maintaining materias checklists required by the Code. O 17J-71 Amend. 62 Nov. 1981

CS3M 8.3 - Wel ding and Brazing Material s Ti:ese procedures def ine requirenents and responsibilities f or control of Code + ding and braz ing material s (electrodes, fil ler wire, fl uxes, gases, anu wel d insert material s) used in f abrication and assembly of Code items. l CS3M 9.3 - Control of Welding Operations These procedures def ine requirenents and responsibilities f or controlling production welding and brazing operations on Code Itans. l CS3M 5.5 - Heat-Treating Procedures These procedures def ine requirenents for controlling heat treating processes performed by Energy Systans Group. It is applicable to heat-treating processes other than weld preheat and interpass tanperature, which are controlled in accordance with methods specif led in qualifled wel d procedure speci f ica tions. CS3M 7.8 - Subcontracted Furnace Brazing Services This procedure def ines requirenents and responsibilities f or control of subcontracted f urnace brazing services. CS3M 7.9 - Subcontracted Heat Treat Services This procedure def ines requirenents and responsibilities f or control of subcontracted heat treat services. US3M 7.10 - Subcontracted Nondestructive Exanination Services This procedure def ines requirenents and responsibilities f or control of rubcontracted nondestructive examination operations performed on Code ma+erials and itens. CS3M 10, 11, 5.10, 14.5 - in-Process and Final Exan ination, Tests, and inspections. These procedures def ine requirenents and responsibilities f or examinations and tests of Code itens, during fabrication and upon completion of f abrication to assure their compliance with Code requirenents. l CS3M 2.5 - Authorized Nuclear Inspectcr This procedure def ines Energy Systans Group requiranents and responsibilities f or assisting the Authorized Inspectcr in perfccming his duties, in accordance with Code requirenents. O d 17J-72 Amend. 62 Nov. 1981

CS3M 17, 7.11 - Procuranent Quality Assurance Records This procedure def ines requirenents and responsibilities f or accumulating records generated during design and/or f abrication of Code itens at Energy Systans Group, transmitting records to the owner or custaner, and retention of records by Energy Systans Group. CS3M 5.11 - Cleaning Procedures This procedure def ines requiranents and respons'$llities f or preparing and controliIng cleaning procedures. O O l 17J-72a Amend. 62 Nov. 1981

t MANUFACTURING MANUAL PROCEDURES (mms) M-2 Control Stations This procedure defines the activities of the Control Stations under the jurisdiction of Manuf acturing. The f unction of a Control Station is to , ensure an even f low of materials and components through the Manuf acturing areas, in accordance with an approved schedule and instructions on the Manuf acturing Production Order (MFO). l M-3 Productive Material Control This procedure provides direction f or the control of material whose f abrication or Initiation of procuranent action is under the jurisdiction of Manutacturing, bh3 Packt.ging and Shipping This procedure delineates the responsibilities f or, and the methods of assuring, proper packaging of components, materials, assemblies, and tooling when special in-plant handling containers are required, or when the items are prepared f or stcrage or shipment to the custaner. M-3 Numbering and Control of Manuf acturing Material Processing Procedures (MPP) This procedure establishes the methods and responsibilities f or the numbering and control of Manuf acturing Material Process Procedures (MPPs). M-3 Qual ; f Ication of Wel ders, Wel ding Operators, and Wel ding Procedures This document def ines the procedure and responsibility for initiating, controlling, and maintaining welder and welding operatcr performance qualifications, in accordance with the applicable codes, standards, and spect f Icati ons. l O 17J-73 Amend. 62 Nov. 1981

l I i i 'e i i i i i t ! t I  ; I l i 'l i l i 11  : 4 43 This page intentionally blank. l  ; i 'l l l l I I i. l t

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j i l i l I l I, i. t 1 t g_17 Amend. 43 ' I Jan. 1978 l $ 1 1 ] - i

   ..n

B.1.3.5 E-5 Loss of Primary Pumo Pony Motor with Failure of the Check Valve to Shut The initial conditions for this event are the scae as for tne pony motor failure upset condition (U-8) except Toat the check valve in the affected loop remains fully open. Flow will reverse in the affected loop while intermediate sodium flow remains constant. Thermal driving i.ead in the affected loop is not considered. This condition will cause an abnormal temperature distribution in the IHX, a rapid decrease in the hot leg temperature in the af fected loop and increasing core temperature due to the bypassed flow. B.1.3.6 E-6 Design Basis Sodium / Water Reartion Eteni The event begins with a small water / steam leak, from a steam generator tube, which escapes operator action, raises the intermediate heat transport system pressure close to the SWRPRS set point and causes localized overheating of another tube to the point of pressure rupture. The rupture crea is equivalent to a double ended guillotine tube fallere and produces a water / steam Jet which causes localized overheating of a second tube to the point of pressure rupture after one second. The second rupture also equals one EDEG tube failure. The scenario repeats to the point of a third and final pressure rupture equalling one EDEG tube failure after anotter second. The first equivalent double ended guillotine tube failure causes rupture disc actuation, followed by automatic isolation and blowdown of the water / steam side of both the evaporators and the superheater in the affected loop. In addition the reactor, the sodium pumps and the turbine are tripped. For the unaffected loops, the event is similar to a reactor trip from full power. Decay heat removal is maintained through the two remaining loops. The event is classified as faulted for the affected Steam Generator unit, for the interconnecting piping between the affected steam generator unit and the associated rupture disc, for the super heater-to-evaporators sodium piping, and for the injected reaction products separation tank (s). For the rest of the loop, the occurrence is classified as an Emergency event. O Amend. 62 B-18 flov. 1981

A shell-side hydreuilc model test is providing Information on the potential [' for tube or tube sheet vibration in addition to shell-side flow distribution. The potentia! for thermally damaging the steam generator tubes as a result of departure from nucleate boiling (DNB) is also being experimentally investigated within the steam generator development programs by exposing tubes to severe DNB condit ions. Descriptions of the steam generator prototype test and "f ew tube" test are provided in Section C.6.3.2A. The shutdown heat removal function requires the inteocity and operation of the steam piping, main steam line valves, turbine bypass valves, steam generator modules and steam drums. All of these components include levels of redundancy during shutdown heat removal. Light Water Reactor and conventional steam plant experience and data as well as acceptance tests will be used to assure that adequate SHRS reliability can be established without special testing directed toward these components. C.4.2.4 Direct Heat Removal Service (DHRS) Areas identiflee for Reliab!!Itv Emohasis The DHRS incorporates two primary coolant heat transfer loops. An inner loop transfers heat from the core to the outlet plenum via circulation in the PHTS. Heat rejection from the outlet plenum is accomplished via injection of cold sodium into the outlet plenum via the makeup nozzle and extraction of hot sodium via the overflow nozzle. An essential element for the successful operation of this system is the ef fective heat transfer between sodium (~.} circulating in the two loops. This heat transfer takes place by means of \./ mixing of sodium from the two circulation loops in the outlet pienum. The eff ectiveness of this mixing mechanism has been demonstrated in the 1/21 scale water tests performed at ARD. Further confirmation will be obtained from the 1/4 scale water tests to be conducted in the Integral Reactor Flow Model at HEDL. The DHRS uses the components of the primary sodium service system and the EVST cooling system. The DHRS introduces only the overflow heat exchanger and additional valves. The Integrity of primary piping and other elements of the DHRS coolant bcundary will be supported by the materials testing programs identified for the primary coolant boundary in Section C.6.1.2. The performance of DHRS will be supported by informetion from flow testing of the reactor vessel outlet plenum described in Section C.6.1.2A. Other testing has not yet been identified for DHRS with the possible exception of manufacturer acceptance testing of the overflow heat exchanger and the air blast heat exchangers. The latter component is similar to the FFTF air blast heat exchangers and its reliability will be supported by testing done for the FFTF components. A C.4-15 Amend. 62 Nov. 1981

C.4.2.5 Interfacing Systems Areas Identified for Reliability Emphasis The SHRS has the capability of functioning in the natural circula-tion wde in the primary, internediate and steam / water loops. The requirement for ;1actrical power is that the battery and instrument air supplies be avail-able + ' caerate control instrumentation in the SGAHRS oi DHRS. These components are of cor .entional design, and generic reliability data are available to support their reliability. O Amend. 36 C.4-16 March 1977

A. Sie g Generator Mcdule There are substantial materials properties and weld development programs which su?oe:t the development of a reliable heat transfer surf ace for the steam l generator module. The following two tests also precide Information applicable to the steam generator module reliability. 11 Shell-Side Hydraulle Model Test The shell-side flow and tube vibration tesi were comple%J In June 1976. This test included a full diameter 757 tube bundle in a shell-s!de water flow test. Flowratesupto32,00ggpmwereused. Local tube cross flow Reynolds numbers were as high as 2 x ;0 . The tett included simulated (alr/ water) two phase steam / water flow to determine the significance of tube side vibration excitation. Data have been evaluated for application to both prototype and plant unit design. The evaluations indicate adequacy of the design with respect to tallure modes resulting from vibration or shell-side flow distribution.

2) DNB Corrosion and Heat Transfer Effects Tests Two test programs have been conducted in support of the CRBRP evaporator design in the areas of heat transfer and corrosion effects. Tests at Argonne National Laboratory (ANL) over the period of January - December 1976 with a 42 ft (12.8 m) long single tube sodium heated test section produced data to characterize Departure from Nucleate Bolling (DNB), DNB-associated thermal oscillations and post-DNB heat transfer. Subsequent tests conducted in the GE d DNB Effects Test Loop during 1976 produced heat transfer data complimentary to the ANL tests and demonstrated that operation with DNB in the evaporator would not produce excessive corrosion under worst case CRBRP thermal oscilIations and water chemistry conditions. Following the heat transfer tests, an endurance test was performed during the March - November 1976 period in which the test section was exposed to 4181 hours of steaming time of which 2820 hours were wIth DNB held within a 24-inch (0.61 m) test zone. Post-test examination showed no localized or accelerated corrosion and based on the uniform corrosion found, a long term lif e of about 30 years would be predicted for the CRBRP evaporators.

O C.6-13 Amend. 62 Nov. 1981

O B. Steam Generator Leak Detection Systg The steam generator leak detection system development program includes development of instrumentation to (1) detect hydrogen in sodium and cover gas and (2) detect oxygen in sodium. Programs are in place to develop the detection elements. The detection levels and decision logic for use in the systen will also be established through these tests. In one of these programs, leak detection eierents will be integrated into leak detection systems which will be tested on the "few-tube" test described earlier. O Amend. 47 Nov. 1978 C.6-13a

i l l l  ! ! I l 1 j I O i l i AMENDMENT 62 t LIST OF RESPONSES TO NRC QUESTIONS L i There are no new NRC Questions in Amendment 62. i i 1 I 1 O i l I i 9 h h f i Qi , i t i i l I l

{ L i I I a j Question 001.2  ! ! i I Is Appendix D considered part of the Reference Design Submittal?

Response

l With the deletion of Appendix D in Amendment 24 this question is no - longer applicable.

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i  ; I i 1 l 1 t t t 1 i L I t i l 1 l I i i' f l  ! i 4 l l i 4  ! I 1 I l

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l l i i l 4

                                                                                          )

Q001.2-1 Amend. 62 Nov. 1981

O Cuestion 001.73 (5.2.1.3) Please provide details of the potential sodium leak paths under CCA conditions; describe what provisions exist for deflection of sc dium. Mcw much would be predicted to escape?

Response

Under SMBCB loading conditions, an upward pressure pulse would cause the large rotating plug, intermediate rotating plug, and small rotating plug to impact sequentially against the margin shear rings. This loadirg would make the head / riser assembly deflect with maximum deflection occurring at the small rotating pl ,. The riser structures would maintain their ccncentricity at the upper er : and provide passage for sodium through the riser er'eluc . Letic;e c iths to the head access area could develep or.;y in*co; tre :+aring assemblies. The leakage requirements to be met by the head / riser assemblies are given in Section 5.3.2 of CRSRP-3, Volume 1 (Reference 10a of PSAR Section 1.6). O V t l i l

 /                                                                0001.73-1 i

Amend. 62 flov. 1981 l

1 Question 001.74(5.EljQ ! leow will the guard vessel be periodically tested to verify its in:.agrity and availability? Will it be periodically inspected? I j Response: There is no planned periodic testing of the reactor guard vessel. However,l the inner surface of the guard vessel is accessible for in-service inspection by the same television camera equipment that is used for

 ,   in-service inspection of the external surface of the reactor vessel.

l This is described in more detail in Section 5.2.4.5 of the PSAR. l The guard vessel will be inspected periodically. I ) ) i I I I O Q001.74-1 Amend. 62 Nov. 1981

1 i i l Ouest' ion 001.82 (5.2.4.4) What criteria will be used as a basis of decision as to whether or not backup seals are required for the CDA?

Response

The design of the Riser Assemblies incorporates a backup :eal (margin seal) on the assumption that the inflated elastemer dynamic seals may not have the capability of containing the mass of liquid sodium propelled upward during HCCA loadings. I The margin seals are designed to stop sodic.n flow out of the riser i assemblies and bearing races. The seals would maintain an elastomer-to-metal contact against the races and risers, closing the leakage paths.

                       .The seals are self-energized so as to provf &a sealing under no pressure difference. A pressure difference across t:.a seals causes them to seal more tightly with a force prcportional to the pressure differe ce.

The details of the leakage requiremer.ts on the seals are given in Section 5.3.2 of CRSRP-3, Volume 1 (Reference 10a of PSAR Secticn 1.6). I l l 0001.82-1 i Amend. 62 Nov. 1981

i i 1  ! i 1 Questien 001.92 (5.J.l.1-f) . l i } Does the 300 MJ sec reference design leading for a CDA establish limiting j values for any of the heat transport system. f Rescense: I l The 300 M4 sec loading referred to in the question is absolete. The SMBCS i .i leading requirements are identified in Section 5.2 of CRBRP-3, Volume 1 l

(Reference 10a of PSAR Section 1.6). i Sased en !Se preliminary analyses perfor ed to date, these SM503 leadings de not establish lin

ting conditions on any com;cnent of the heat transport i system.  ;

I i j ' 1 4 I i i i f,O i I l i I i 0001.92-1 Amend. 62 flov. 1981

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{ l I i I i j i 4  : Question 001.102 (5.3.2.3.2) l l In the event of an IHX leak, describe thc/ procedure for identifying f

and plugging a defective tube. -

) I Response: l l i l At present, the details for locating and plugging a leaking tube have not l been developed. ' 'owever, plans are ur derway to develop details for both. ' l This data will be provided when it be,omes available. i  : r R R e i i 4 i I t

<                                                                                                                                  1 d

L j , } l i . t l t J l t I  !

                                           \    Q001.102-1                                       Amend. 62

, Nov. 1981 1

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Question 001.111 (5.3.3.1.6) Describe how overpressure loads from a CDA will affect the primary sodium pumps. Will seal failure occur? Will the pump integrity be maintained? Response : The HCDA is not part of the design basis for the CRBRP Reference Design as explained in Sections 1.1 and 15.1. As indicated in Section 5.3.1.1, Item F under " Transients", SMBDB loadings must be accommodated by the primary coolant boundary. SMBDB design requirements appropriate t- the pumps appear in CRBRP 3, Volume 1 (Reference 10a, Section 1.6). The SMBDB loadincs on the primary pump of 550-575 psi will l result in high primary pressure loads at the pump tank nozzles--the most critical area in the pump in terms of stress. These pressure loadings in the primary pump will cause stresses in the discharge nozzle to approach the yield point. In the pump tank suction nozzle, the yield stress may be exceeded, but the resulting stress will be within the allowable stress for ASME Section III faulted conditions. The pump procurement specification requires that the primary pump tank be capable , of sustaining one occurrence of SMBDB loadings at the end of plant l life without loss of ability to contain the sodium. It is expected that the SMBDB loadings will not adversely effect the l impeller /dif fuser. Shaft seal failure is not expected to occur under SMBDB loadings. l O' The pressure pulse at the sodium argon interface in the tank will be attenuated as it passes through the annular space between the shaf t and the shield plug and as it passes through the labyrinth bushing below the shaft seal. Above the bushing, there are two paths open to the l cover gas; one is a low impedance path through the leakage oil reservoir to the RAPS system, and the second is through the seal oil cavity. If the pressure pulse which reaches the shaft seal exceeds the seal oil cavity pressure of 14.5 psig, it may produce a pressure unbalance which would cause the seal surfaces to separate, thereby allowing some gas to escape. As soon as the unbalance pressure is dissipated, the seal springs would cause the saaling surfaces to reseat, and the seal would then con-tinue to function properly. Because of the low pulse pressure in the gas and the low impedance path to the RAPS system, opening of the shaft seal faces is not expected to occur. O Q001.111-1 Amend. 62 Nov. 1981 3

Questien 001.147 (5.3-28) Clarify what is meant by the statement that the " piping supports will be designed to fail if the 1 cads are well beyond nomal operating conditions". Rescense: If the SMSDS loadings were to cause rapid and extreme pipe motion the seismic restraint could cause excessive pipe leads. This situation is avoided by use of shear pins in the snubber shafts which are set to fail cnly under such a severe event. 9 I I l l 1 1 l i Q001.147-1 6 faend. 62 lbv. 1981

l l Question 001.151 (5.3-34) l Provide the schedule for submitting the results of the evaluation identified in 5.3.3.1.2. l Response: The structural evaluation of PHTS pressure-containing components listed f { in the referenced section will be contained in their final design reports. l j The current schedule for completion of final design reports is given  ! below. Each of these reports will be prepared by the component vendor  ! with the exception of the PHTS piping and the flowmeter. At the flowmeter i location,the sodium pressure is contained by the piping. Design and j analysis of the PHTS piping is being done by Westinghouse. The final  ; . PHTS piping design report will include analysis of all associated parts I ! under Westi.1ghouse design cognizance such as thermowells and connections to pressure sensors. l l , i l I PHTS Component Final Design Report Due . ( Coolant Pump November 1982 l Intermediate Heat Exchanger presently available ! Cold Leg Check Valve presently available Piping October 1983 Reactor Vessel April 1982 l l r l l l t 1 I r i t i i [ Amend. 62 Q001.151-1 Nov. 1981 , l

Ouestion 001.209 (15.1.1.;.14 ) The PEAR states that the Overflow Heat Removal System (OHRS) is not re-quired to function following the third level dynamic-loadings. It appears that the third level margins are simply mechanical and provide ro level of assurance that the reactor would be maintained in any safe configuration following a third level event. Explain what means of cooling would be available to the reactor core af ter a third level event.

Response

The Direct Heat Removal Service, DHRS (formerly OHRS) has been included in the design to provide a diverse means of heat removal. It provides capability to protect the plant in the highly unlikely event that heat removal through all of the steam generators is impossible as a result of a sequence of failures. For that highly unlikely situation, the DHRS is ' capable of removing the full decay heat assuming no other heat removal from the time of reactor trip. The SMBDB margins are provided as additional assurance that the health and safety of the public is protected even in the event cf the occurrence of unforseen circumstances. These margins are not associated with a specific event. The specification of SNBDB requirements provides capa-bility for a spectrum of highly improbable circumstances and involves . judgment based on test and calculational results together with a C.A knowledge of system behavior. The dynamic loading requirements within the reactor vessel and in other parts of the Primary Heat Transport System assure that the system will remain intact. Since the system elevations have been specified to be consistent with natural circulation, this capability would exist following the dynamic loads. The heat removed through the normal cooling systems would be a function of the availability of flow paths through the reactor. In the postulated event of failure of heat removal through the PHTS following a hypothetical core disruptive accident, the heat is removed through the TMBDB process described in CRBRP-3, Volume 2 (Reference 10b of PSAR Section 1.6). l l l l l Q001.209-1 Amend. 62 V Nov. 1981 l

J Question 001.214 (15.5.2.3.2) Provide an analysis of the release of all volatile fission products, assuming failure of the elastomer seals and loss of EVTM cooling.

Response

This question appears to be motivated by a concern that there might be a common mode failure due to the loss of EVTM cooling that could result in the release of all volatile fission products and failure of the seals. Such a common mode failure is not considered to be a credible accident, as discussed below. The EVTM cold wall is designed to remove decay heat from 20 Kw spent fuel a ssemblies. As discussed in detail in PSAR Section 9.1.4.3, neat from the cold wall is removed by forced air flow, provided by an air blower. In the event of a complete loss of forced air cooling, the cooling mode is autonatically

 ;           and inherently changed from forced to natural convection. This change is effected by fail safe controls, not by operator action. The maximum fuel cladding temperature of a 20 kw spent fuel assembly in a gCCP in the natural air convection cooling mode was calculated to about 1500 F. At this cladding temperature, only randon fuel rods would be expected to release fission gas into the EVTil. The accident discussed in Section 15.5.2.3 conservatively, assumed fission gas release from all 217 fuel rods in the EVTrl.

All steady-state seal temperatures in the EVTM during natural convection are below 200F, which is less than the upper limit of 35CPF considered as detrimental to the integrity of elastomeric seals. This is achieved by protecting seals from direct heat radiation, placing them in areas benefiting from convective air cooling, and providing a geometry with high thermal resistance and heat capacity between the heat source and the seals. It is, therefore, concluded that loss of forced air cooling cannot lead to a common mode failure resulting in the sudden release of volatile fission

,            products from the EVTM.

A potential mechanism of fission gas leaving the EVTM is by s',ow diffusion through the seals. The radiological consequences of such an event were analyzed in Section 15.5.2.3.2 and resulted in dose rates less than , , the limiting values. As described in Section 9.1.4.3 of the PSAR, 4 all surfaces in the EVTM to be sealed against radioactive gas are provided with at least two seals in series with pressurized gas between them. The i seals will be leak tested before reactor refueling operations are initiated. As discussed in the response to Question 310.22, leak testing of the EVTM will be required by a technical specification. This specification will assure that accidental releases of fission gas from the EVTM are below the liniting values. O Q001.214-1 Amend. 62 Nov. 1981

Question 001.258 (15.7.3.4.1) Provide an analysis of the consequences of a release of cover gas to the HAA.

Response

The Head Access Area (HAA) atmosphere communicates freely with the upper RCB atmosphere. To provide a conservative, upper bound estimate of the potential consequences of postulated cover gas releases to the HAA, an instantaneous release of the entire primary system cover gas inventory to the RCB is evaluated. Such a release is considered hypothetical; its consequences are evaluated to demonstrate that even for this limiting case release, no danger to the health and safety of the public exists. It is assumed that the entire primary system cover gas inventory (Reactor Cover Gas, Overflow Vessel Cover Gas, and PHTS Pump Cover Gas) is in-stantly released to the RCB. The cover gas activity used for the analysis is based on continuous plant operation with 1% failed fuel - the design basis failed fuel fraction. Fellowing such a postulated release, the automatic containment isolation system, described in Section 6.2 and 7.3 of the PSAR, would be activated and containment isolation effected; the potential consequences of this event would be limited to direct gamma shine exposure from the radioactive cover gas released to containment and to leakage of the cover gas activity through the low leakage RCB. k The design leak rate of the RCB is 0.1% Vol/ Day at 10 psig. For the postulated event considered, no mechanism exists to pressurize containment. However, for conservative analysis, a constant 1 psig containment overpres-sum was assumed. This 1 psig overpressure is a conservative allowance for building heatup, following containment isolation and possible barometric va ria tions . Based on a square root pressure-leakage relationship, contain-ment leakage at 1 psig is 0.032% Vol/ Day, or on a fractional basis, 3.7 X 10-9/sec. Table Q001.258-1 itemizes the isotopic primary system cover gas inventory, based on continuous plant operation with 1% failed fuel; for this analysis this radioactive inventory is assumed instantly released to the RCB. Column 2 of the table lists the activity per isotope released to the envir-onment during the first 2 hours following the postulated event. Column 3 l lists the total activity per isotope released to the environment. These ! environmental releases were determined considering radioactive decay during holdup in the RCB and continuous leakage from the RCB at 0.032% Vol/ Day. Table Q001.3-2 summarizes the potential site boundary and low population l zone doses res ilting from this postulated event. As the table indicates, a i large margin (. reater than a factor of 10 6) exists between the pctential l

doses and the DCFR100 guideline values. It is therefore concluded, that l even for this Lypothetical case cover gas release, no danger to the health and safety of

he public exists.

t l Amend. 62 Nov. 1901 Q001.258-1 l i

1 r i I Table Q001.258-1 Radioactivit/ Release Following Hypothetical Cover Gas Release to RCB (Curies) Primary Cover

• Environmental Release Isotope Gas Inventory 0-2 Hrs. Total Xel31m 26.2 6.89-4 *
• 1.16-1 Xel33m 816 5.57-1 8.41-1 Xel33 14,900 3.90-1 35.2 Xel35m 2,340 1.13-2 1.13-2 Xel35 56,900 1.39 9.98 Xel38 3,710 1.98-2 19.93 Kr83m 1,410 2.64-2 5.09-2 Kr85m 3,930 8.88-2 3.29-1 Kr85 0.49 1.29-5 4.61-3 Kr87 3,600 5.83-2 8.89-2 Kr88 6,840 1.42-1 3.64-1 i Ar39 35.2 9.28-4 3.33-1 Ar41 27.0 4.99-4 9.39-4 I;e23 1.41+6 2.79-1 2.79-1 H3 8.82-3 2.32-7 8.30-5 Based on continuous operation with 1% failed fuel. Incl udes Reactor, Overflow Vessel and PHTS Pumps Cover Gas.

                    *

• 6.89-4 = 6.89 x 10-4 l

e Amend. 62 Nov. 1981 Q001.258 2 l

   ,-t ....-we.--.-       ww--w-,,y. gm wy  ,-,n3   m.., . , ,          .y--. .,,.          mm, ._y _, ._. _ _-         _m . _ _ - _ _ _ _ . _ - * - - - - _   _m.

l 1 i l l i

O i

Table Q001.258-2 i Potential Off-Site Doses Following Hypothetical i Instantaneous Cover Gas Release to RCB l

\

j Dose (Rem)*

• Site Boundary Low Population Zone (2-hours) (30-days) l (0.42 miles)

(2.5 miles) l t

                                                                                                              -4 l'                                             Total Whole Body

• 6.03 x 10-4 3.26 x 10 i

                                                                                                              -11 2.17 x 10 -11 1

.' Thyroid 6.22 x 10 l _c -6

Lung 1.34 x 10 - 6.87 x 10 i

Bone 0 0 t ! lh i 10 CFR 100 Whole Body 25 25 i Thyroid 300 300 i i !

• Includes gamma cloud and innalation doses. .

! *

• Atmospheric dispersion based on 95th percentile;E/Q's per i

Amendment 38 to Chapter 2 of the PSAR. i I l 'l i i O i Q001.258-3 Amend. 62 Nov. 1981 I j .I

Question 001.295 (4.4.2.4.2) Specify the maximum cladding midwall temperatures in the fuel assemblies and in the radial blanket assemblies on which the orificing scheme has , been based. '

Response

The CRBRP fuel and blanket assemblies orificing is discussed and explained in detail in Section 4.4.2.5. Ratner tnan specifying a maximum cladding temperature, the flow in each assembly is orificed to simultaneously satisfy various constraints, such as attainment of lifetime /burnup objectives, satisfaction of transient limitations, and assurance that the assemblies l exit temperatures and temperature gradients result. in an acceptable themal environment for the upper internals structure. All the above constraints are quantitatively translated in terms of equivalent limiting temperatures (which are individual characteristics of each assembly) and the flow necessary to satisfy the most restrictive constraint (the lowest equivalent temperature) is determined. Assemblies are grouped togheter in orificing zones (a maximum of eight discriminators in fuel plus inner blanket is allowed) and the total flow allocation to fuel and blanket assemblies must not exceed 94% to account for cooling requirements of other reactor components. Section 4.4.2.5.1 discusses the orificing philosophy, approach and constraints; Section 4.4.2.5.2 presents the method adopted in calculation of the O- equivalent limiting temperatures, while results are reported in Section 4.4.2.5.3. O 0001.295-1 Amend. 62 Nov. 1981 e --r--,----.--- ----.:,t yv w-e...e--.---, yn,, ..,---r, ,,---- .--------,,-..,-n---y,,,,,,,---.,---,..c-v.-,.-,r<--,.w,, - . - - - - - - - - ,-- -----+-- -- w- --+--e-- .

l l i Question 001.313 (15.2.1.3) i Provide the results of the analysis'for the SSE and OBE fer the secondary-control rods.

Response

Analysis of a step reactivity insertion postulated to occur as a result of an SSE and terminated by the secondary control rods when tripped by the secondary portion of the Plant Protection System is provided in Section 15.1.4 of the PSAR. This event was selected as the " umbrella event" for analysis reflecting the latest design infomation. The consequences of the event are shown to be within applicable limits. I i The consequences of an OBE terminated by the secondary control rods would l be less severe than those for the SSE. Detailed analysis results for the l OBE will be provided concurrently with an overall update of PSAR Section 15.2 which will be provided prior to NRC issuance of the Construction Permit. I l O O Amend. 62 Q001.313-1 Nov. 1981

t Question 001.343 (15.4) Consider a completely blocked, fueled subassembly at full power. Analyze tne possibility of subsequent subassembly to subassembly propagation up to and including whole-core involvement. Specify the nature and extent of initial fuel or ' cladding plugging above and belcw the core, fuel dispersal and ficw regimes and liquid-liquid heat transfer. Include the possibility of failure propagation via the inlet pienum as well as via the S/A duct walls. Perform a siminar analysis but with scram occurring. Consider maximum values of decay heat and minimum values of shutdown pony-motor ficw in adjacent subassemblies. The analysis should include the transient effects of initial pump costdewn and changes in decay heat. Resconse: Fuel assembly blockage as a potential initiator of an HCDA is addressed in Section 3.3.1.4 of CRSRP-3, Volume 1 (Reference 10a of PSAR Section 1.6). It was concluded that large blockages sufficient to cause coolant boiling are highly improbable. Even if an assembly-to-assembly propagation scenario is hypothesized, the consequences would be enveloped by other failure sequences that involve the whale reactor and have been analyzed in detail. ( Q001.343-1 O Amend. 62 flov. 1981

Ouestion 001.379 (15.0) Provide the number L.d size of the reactor head and vessel support bolts as well as the high strain rate data for the eaterials used in the hold down bolting system. Resconse: As indicated in Section 5.2.2 and CRERP-3, Volume 1 (Reference 10a of PSAR Section 1.6) the reactor vessel support design has been modified, replacing the bolted attachment of the reactor vessel flange to the support ring with an integrally welded attachment. The closure head assembly which consists of the th.ee rotating plugs, interconnecting risers, and attached components, is mounted to the reactor vessel flange via the outer risers. In addition, the plugs and flange are interconnected by shear rings which would transfer upward load-ings directly to the flange, in lieu of through the riser assemblies. The entire reactor vessel / closure head assembly is secured to the reactor support ledge of the reactor cavity by a row of bolts through the support ring. These bolts would experience tension due to upward loadings on the closure head. The design specifies 69 hold down bolts fabricated frcm SA193 Class B7 (120,000 psi yield strength) with a total cross section of 415 square inches. The bolt system remains elastic during the SMBCS leading. Therefore high strain rate data are not required. O i t ( l l l l l QC01.379-1 O l v Amend. 62 Nov. 1981

i I Question 001.380 (15.0) Provide sketches and/or drawings of the details of the reactor inlet and outlet nozzles ' indicating the clearance between the inlet and the outlet piping and associated guard piping.

Response

Figure 5.2-1A specifies the clearance between the bottom 00 of the vessel piping and the bottom ID of the guard vessel to be 2.00 inches minimum. This minimum clearance is required at full pcwer operating conditions. For further details on the inlet and outlet nozzles see Figure 5.2-1B. The reactor vessel inlet pipe will have an 00 of 24.00" and the reactor vessel outlet pipe will have an 00 of 36.00", whereas guard vessel inlet pipe will have an ID of 38.00" and the guard vessel outlet pipe will have an 10 of 50 inches. O QC01.380-1 Amend. 62 Nov. 1981

i I Question 001.390 (9.1.2.1.3) Provide the analysis showing that the EVST top cover can absorb the heaviest drop loads carrit.d above it without changing the lattice spacing of the fuel storage tubes. List the weights of the heaviest objects that may be carried over the EVST.

Response

The upper surface of the EVST closure head assembly consists of a 6.5-inch l thick striker plate, supported at its periphery (see Figure Q001.390-1). j The striker plate fonns part of the RSB operating floor and protects the ' underlying thick steel closure head by absorbing all normal and off-normal structural loads, including accidental impact loads. The distance between the lower surface of the striker plate and the upper surface of the steel closure head is about 10 in. O i The heaviest load normally carried over the EVST is the EVTM floor valve (9 tons). The lift height of the EVTM floor valve is limited to 2 feet. No heavier loads which might be carried over the EVST can be identified. The spent fuel shipping cask, for example, is being transferred only between the cask shaft and a railroad car; both facilities are more than 20 feet distant from the periphery of the EVST. All heavy maintenance equipment is transported by the large j component transporter (LCT) between RSB and RCB (see Section 9.2.1.2.2). The heavy maintenance equipment and the LCT are handled only by the double reeved main hook of the RSB bridge crane, and they are not carried over i I the EVST. Seismic restraints prevent equipment loaded on top of the LCT from toppling onto the RSB operating floor or EVSI during an earthquake. 1 l O Amend. 62 Q001.390-1 Nov. 1981

In spite of these considerations, a hypothetical heavy weight due to maintenance operations has been postulated to drop onto the EVST. The 25-ton load limit of the RSB bridge crane auxil:ary hook was selected as hypothetical weight for this " umbrella" event. A drop height of 2 feet above the EVST striker plate was assumed. This represents the maximum handling height above the operating floor during maintenance operations. Stress calculations were performed to determine the maximum deflection of the striker plate, and the maximum stress due to the postulated impact load. The analysis was based on the following ground rules: (1) The load drops onto the center gf the EVST striker plate with a load impact area of 1 ft (2) The impact load was converted to an equivalent static force, using a spring constant which takes the presence of inspection holes and fuel transfer port holes in the striker plate into consideration. (3) Conventional flat plate formulas for deflection and stress were used. The results of the analysis are as follows: (1) The weight of 25 tons dropping 2 feet is equivalent to a static force of 1.15 x 106 lb acting on the striker plate center. (2) Due to this force, the striker plate would experience a maximum deflection of 3.93 inch into the air space l between striker plate and steel top shield. This amount of deflection is not sufficient for the striker plate to touch the underlying closure head, therefore, the impact load would not be transmitted to the steel closure head. (3) The maximum striker plate stress due to the ampact load is about 13,000 psi, which compares to a minimum yield strength of 35,000 psi for the striker plate material . From the above considerations, it was concluded that the accidental drop of the heaviest object carried over the EVST could not lead to a change in lattice spacing of the fuel storage tubes. QO )1. 390-2 Amend. 62 Nov. 1981

O Question 001.392 (9.1.2.2):

I When new fuel, at ambient temperatures, is inserted in the EVST the fuel subassemblies could be subjected to a thermal shock. Discuss how this is controlled and what stresses occur in the fuel subassembly, and where the most severe loads are and do they approach the yield point of the material?

Response

The purpose of the gas filled (dry) preheat stations in the EVST, nientioned in Sections 9.1.2.2 and 9.1.4.1, is to provide a slower heatup rate for new fuel assemblies than would be achieved by direct immersion into sodium. The preheat stations are thimbles filled with EVST argon cover gas. The gas is in thermal equilibrium with %e sodium surrounding the thimbles. New Fuel assemblies are inserted into the thimbles with a lowering speed l of 2 ft/ min. Heat transfer calculations using the computer code DEAP (Differential l Equation Analyzer Program, described in Appendix A of the PSAR) with conservative assumptions were performed to obtain temperatures gradients in a new fuel assembly during preheating. The new fuel assembly was assumed to have an initial temperature of 700F, and was suddenly subjected to a 4250F temperature step due to full submersion in the argon gas of the preheat thimble. l The results of the analysis indicated that a maximum temperature differene of about 3500F will occur between the center fuel rod and a fuel rod in ] the outer row, located near a corner of the hexagonal fuel assembly duct. The maximum rate of temperature increase will occur in the outer 0 fuel rod and wili be 12.5 F/ min. The maximum temperatur e gradient through the fuel cladding is less than 10 0F/in. in radial direction. This gradient is two orders of magnitude less than a local temperature gradient of 12000F/in. required to induce thermal stresses which would l exceed the material design stress of 40,000 psi. It is concluded that immersion of new fuel assemblies in preheat stations does not impose any severe thermal loads to the cladding. l l O Q001.392-1 Amend. 62 Nov. 1981

Question 001.393 (9.1.2.2.1): Since it is expected that the FHC will be contaminated by alpha, how is the spread of alpha contamination controlled when the cell plugs are removed into the RSB which is also open to the RCB during fuel handling, and to the atmosphere through the H&V system? It is realized that a i floor valve and transfer machine is used but it does not seem likely that 100% control of alpha particles is realistic.

Response

Prior to fuel handling operations which involve transfer of fuel assemblies in and out of the FHC or EVST, the fuel transfer port plugs will be removed. This is accomplished using floor valves and the EVTM. The EVTM couples with its closure valve to a floor valve on top of a fuel transfer port and removes the port plug. During transport to the RSB port plug storage facility, the port plug is within the inerted containment boundry and a drip pan in the EVTM closure valve receives any sodium which might drip from the bottom of a port plug. After the EVTM has traveled to the inerted P,SB plug storage facility, its closure valve is coupled to a floor valve and the plug is deposited into an empty position of the rotatable plug storage facility rack. During the plug transfer process and during the entire time of plug storage, no plug surfaces are ever in contact with the atmosphere of the O RSB/RCB. However, a small annular area underneath the EVTM closure valve and a matching surface on t1e floor valves could contain small amounts of contaminated sodium. The fcilowing administratively controlled, precautionary procedures are presently anticipated whenever the EVTM is decoupled from a facility with potential alpha-contamination, 1.e. , the reactor, EVST, or FHC. These procedures are preliminary, pending further refinement o# the Reactor Refueling System equipment and facility design and their operation. A. General Techniques for Contamination Confinement During Fuel Handling

1. The immediate area of the EVTM r.,0vement path is barricaded or roped off prior to any fuel or plug transfer.

Only a mechanic and health physicist (HP) are present inside the barriers, specifically near the floor valves being serviced by a transfer machine. Both wear appropriate protective clothing and respiratory protection.

2. Before the mechanic can proceed with any hands-on operation involving potentially contaminated surfaces, the HP surveys the direct radiation and radioactive contamination of these surfaces and records the data. If the radiation contamination i level is acceptable, he notifies the mechanic to proceed with the operation.

3. At frequent intervals the EVTM cask volume within the pressure boundary is purged with fresh argon from the floor Amend. 62 Q001.393

___._._________._____-1_ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . . _ - . -

l service stations. The contaninated gas is vented to the same stations. These stations are connected to the plant argon supply and processing systems.

4. Interface purging is performed by the EVTM before opening the closure valve gate to exclude air from the transfer passage. The interface is again purged after closure, prior to separation from the floor valve, to dispose of any contamination gas.

B. Specific Procedure After Machine Decoupling

1. The EVTM is uncoupled fully raising the extender.

2. The mechanic covers the floor valve with a plastic sheet and secures it to the valve. This sheet is not removed until the next time the EVTM will mate again to the same flror valve. He then places a second plastic sheet on top uf it. This will serve as a disposal bag after the next operation is completed.

3. The mechanic cleans the annular area of the closure valve, e.g., by wiping it with alcohol danpening swabs. He then puts the used swabs onto the second sheet on top of the closure valve, and closes it forming a disposal bao around the swabs and any drippage. Later the bag will be transferred to the Radioactive Waste System in a drum or other appropriate container.

4. Finally the mechanic covers the cleaned lower surface of the closure valve with another plastic sheet and secures it to t tne closure valve.

t

5. The EVTM is moved to its next location. There the plastic cover sheet is removed from the closure valve before coupling to the floor valve. The removed cover sheet is transferred to the Radioactive Waste System.

The following three items will all contribute to minimizing the occurrence and/or spread of alpha-contamination during fuel handling and storage: Q001. 393-2 O Amend. 62 Nov. 1981

1  ! l l I 1 i l 1 Question 001.402 (9.5.2.1) 1 1 4 Elaborate on the sodium fire control system in which it is statea that [ 2,500 scfm can be provided to any one of elevcn cells in the intermediate  ! i bay. Are these cells designed to withstand the increase in qas pressure i I i resulting from gas inflow, products of combustion, and the heat added to the mixture?

Response

i ! The nitrogen flooding capability has been eliminated from the fire l protection system. PSAR Section 9.13.2 provides a complete description i of the Sodium Fire Protection Systems and Section 15.6.1.5 provides an evaluation of the plant capability to withstand the effects of a sodium fire in the IHTS cells. l

I i

l t 1 1 1 1 l 1 )! i i l lh Q001.402-1 Amend. 62 Nov. 1981 i l I

                                                                                      ~ . _ _ _ _

4 Question 001.405 (9.6) 1 l Justify the open containment concept for both the RSB and the RCB as being consistent with the low as practicable objective and providing defense in depth for the design basis accident yet to be defined for CRBR. It would seem that the H & V system provided has the prime objective of pro-viding a dilution mechanism for the escaped radioactivity frcm the reactor and defeats the concept of confining radioactivity for treatment, decay, and later disposal. Secondly, the open containment concept becomes an active system rather than a passive system depending upon the closure of dampers and valves to become effective rather than the reverse situation in which no active action has to occur to provide containment. Justify your selection of this ccntainment concept. Resconse: The CRBRP containment is designed and the plant operating philosophy is developea on the basis of an open containment. The design of the contain-ment and the containment ventilation system is provided with sufficient safeguards to prevent accidental release of radioactive materials. The presently identified steady state release of radioactivity during plant operation, identified in PSAR Section 11.3, results in effluents and asso-O ciated doses orders of maanitude lower than the levels of 100FR20. Section 11.3 lists the total annual gaseous effluent release for CRBRP as 710 Ci/yr, compared to the minimum total gaseous effluent release of LWR's studied in W 1258 of 3600 ci/yr. , PSAR Section 11.3 lists the integrated dose to the po miles of the CRBRP site in the year 2010 as 1.7 x 10 pulation c man-rem /yr. The within l50 added cost of a closed containment over the reference design open contain-ment cannot be justified from a cost benefit analysis because of the low ' operational releases from CRBRP. An integral part of the CRBRP design is to allow personnel access during normal plant operation, in order to ensure j equipment operability and to perform routine operations of equipment located within the containment. This equipment is located in containment so that it j is closer to the primary system equipment it serves. (Examples are sodium sampling, inservice inspection, access to I&C cubicles, and large Sodium Component Cleaning Vessel use.) The airborne activity would be too high to allow continuous occupancy during operation if the containment were not 7 l purged (open containment). RCB In order to provide a very low leakage barrier at the primary containment boundary, a seismic Category I, tornado hardened concrete confinement struc-ture is provided around the outside of the inner steel containment vessel with an annular space separating the two structures. Amend. 62 N v. 1981 Q001.405-1

The annular space ba. ween the inner and outer containments will be maintained at a negative press're relative to atmospheric pressure during normal operation and accident condition and is exhausted through high efficiency filters. The filtered exhaust point is chosen to obtain maximum dispersal of the radio-active material prior to reaching the Control Room intake. In addition, the recirculation system for the Control Room atmosphere is increased in capacity to 8500 cfm. The containment atmosphere ventilation system 1s reduced in capacity from 50,000 cfm to about 14,000 cfm in order to minimize the potential release of activity from containment during valve closure time. The containment supply and exhause penetrations will be reduced from 48" to 24" Contain-ment isolation of the HVAC system exhausts will be designed to meet item 4 of the CRBR Design Criteria 47. The containment ventilation / purge system is provided with a time delay duct to preverit the release of radioactive materials during accident conditions. Tha time delay duct is sized for such velocity, that the containment isolation valves will close before the contaminated air reaches the valve zone. Radiation monitors which pro-vide signals for initiating closure of the containment isolation valves are provided at the inlet of the time delay duct and in the HAA. During normal plant operation and all accident conditions, the containment / confinement annulus space is maintained at a minimum 1/4" water gauge r9 ative pressure with respect to the outside atmosphere. During normal pu nt operation, the RCB Operating Floor is maintained under slight negative pressure (<l/8" water gauge). Capability is provided to filter the contain-ment / confinement annulus exhaust through the annulus filter units during normal plant operation and all accident conditions. The tilter system will consist of two 100^; redundant filter-fan units consisting of prefilter, demister, heating coil, HEPA filter bank, absorbent filter bank, after HEPA filter bank and fan components (approximately 14,000 CFM capacity). A tornado missile protected, Seismic Category I enclosure is provided for the RCB annulus filter-fan units, the RCB normal exhaust fans and the annulus pressura maintenance fans. Shielded wall partitions are provided in the HVAC equipment room between the redundant annulus filter-fan-units, RCB exhaust fans, and the annulus pressure maintenance fans. Tornado missile protected, Seismic Category I air intake and discharge openings are provided. RSB The design for the RSB is described in PSAR Section 3.4 and analyzed in Section 15.6. The resulting doses are significantly below appropriate 10CFR100 guidelines values and meet or exceed all of the Design Criteria specified in PSAR Sectic, 3.1. However, modifications to the RSB HVAC system were made to limit air infiltration and to provide recirculation and filtration capabilities during all operating conditions as discussed  ! in Section 9.6-3. Q001.405-2 Amend. 62 O Nov. 1931

Question 001.433 (15.A.3.3.1 Yellow) O In the analysis of the EVCC, the initial temperature is taken as 1075 F, and the elevated temperature due to decay heat is limited to 1200 F. The flow blockage analysis in Figure F 6.4-5 shows however that the sodium temperature in the vessel can rise as high as 1400 F prior to melt-through to the EVCC. Discuss the effect this would have on the partition factors for the fission products dispersed in the sodium, and consequently, on the RC source term (Table 15. A.3-4). Jus ti fy extending the method of partition fractions, as used by Castleman, if the sodium temperature is calculated to rise beyond 1200 F; note that Cs and Rb boil below 1300 F.

Response

In Amendment 24 to the PSAR, the Project withdrew the Parallel Design from further consideration by the NRC staff. This question requests additional information relative to analyses conducted in support of the Parallel Design. Accordingly, the question is no longer directly applicable. The considerations associateo with developing the source term for the TMBDB analyses are discussed in Section 4 of CRBRP-3, Volume 2 (Reference 10b of PSAR Section 1.6). O d O Amend. 62 I Q001.433-1 Nov. 1981

 --r-y   _- _ . , - , , ._-.. ..,--,...--,-mm,_.__,,-_,,,y,_._,,.,,-_,,,%_.w..m.,,.,,,,e.,,            , ..w.__m..,-.,,,.,. ,.,,.v._ _e%,_       . , , - - - ,

O Question 001.500 (F6.4.1.1 ): The particle size distribution for fuel and steel fragments are based on the ANL materials M-series with tests. These tests involved interaction of molten liquid sodium. Hcwever, for some CDA sequences, namely, the transition phase regime of the LOF accident, it is expected that two-phase vapor / liquid core materials will be ejected into the sodium pool. For this case no experimental results have been cited for the particle size distribution. a) Provide the justification for using particle size distributions from liquid / liquid materials interaction tests (ANL M-series) to represent the size distribution resulting from interaction of two-phase fuel materials with sodium, b) Provide estimates of the depth of a thermally stable bed as a function of the particle size distribution. c) Provide a discussion and description of any proposed R&D programs that will yield the necessary information relative to Items (a) and (b)above. Resconse: The information requested is provided in the response to question 001.633.l O Q001.500-1 Amend. 62 Nov. 1981

l Question 001.501 (F6.4.1.1.2) Catton and Dhir (

Reference:

Post-Accident Heat Removal for LMFBRs, UCLA-ENG-7593, dated October,1975) report predictions for stable debris depths, based on theoretical and experimental treatments, that are at variance with the ANL data, generally being much shallower. They found factors such as trapped vapor and wall interference played an important role in fixing the dryout heat flux for a given bed depth. The expressions developea to predict the maximum stable depth were intended to encompass these interfering factors with a conservative choice of an emperical constant. Discuss how this section makes allow-ances for these considerations.

Response

The factors considered by Catton and Dhir in Reference Q001.501-1 are not considered to alter the evaluation presented in CRBRP 3, Vol. 2 (Reference 10b of Section 1.6) for the following reasons: a) Trapped vapor: The particulate beds postulated in the CkBRP are formed by the settling of particles through several feet of sodium. The gases present in the system (fission gases released from the fuel pin at the time of breach or from the fuel at the time of fuel melting (q,/ would separate from the particles due to turbulence) are bouyant and would rise to the sodium surface rather than being entrapped by the falling particles. A discussion of the effects of sodium vapor generation within the debris bed is provided in Reference 0001.501-2. This is a different situation than that considered in the UCLA experiments in which a dry solid material and a liquid are placed in a small container, trapping vapor already present in the system. b) Wall effects: Catton and Dhir's experiment: were conducted in a small (4.7 cm diameter) beaker and wall effects could be expected to have an important influence on their results. The surface area of settled debris in the CRBRP is orders of magnitude greater than the surface area of the beaker, and wall effects would be expected to have little influence on the bed dryout heat fluxes. c) Eed depth comparisons with ANL data 1 Catton and Dhir have developed a correlation for a deep debris bed, i.e. , a height greater than 5 cm, by using a correlation factor which describes a line running through the lower three points of their data. Q901.501-1 i Amend. 62 Nov. 1981

l l t f For the power levels used in the analysis, 2 to 3%, Catton 9 l and Dhir's correlation predicts bed depths of 0.7 to 1.4 cm. . Depths of this magnitude have a heat transfer mechanism different from that of deep beds, primarily because the vapor channels extend the full depth of the bed, and a comparison with ANL data (Reference Q001.501-2) should be made on the basis of a shallow bed correlation. The assessments of debris bed behavior and their bases are provided in Ar,oendix G.1 of CRBRP-3. Volump 7 (Roforonro inh of PSAR Section 1.6). O Re fe rences Q001.501-1. I. Catton and V. Dhir, " Post-Accident Heat Removal for LMFBRs", UCLA-ENG-7593, November 1975. Q001. 501 -2. L. Baker, Jr. , et. al . , " Post-Accident Heat Removal Technology," ANL/ RAS 74-12, July 1974. I O Amend. 62 Q001.501-2 Nov. 1981

l Ouestion 001.502 (F6.4.1.2) i i Expand Section F6.4.1.2 to provide the distribution of debris between the internal'reacter vessel structures and the primary heat transport system for a spectrum of CDA's ranging from highly energetic to benign ones, including scenarios that lead to the transition phase. Resocnse: l The analyses questioned above have been superseded by those currently presented in Section 3.1.1 and Appendix I of CRSRP-3, Volume 2 (Reference 1Cb of PSAR Section 1.6). Q001.502-1 Amend. 62 Nov. 1981 l

i l Ouestien 001.503 (F6.4.1.2) Fer the particle settlement analysis presented in this section, describe

 ,       and discuss the experimental evidence, if any, that supports this analysis.

Resecense: i The infor: nation requested is provided in the response to question 001.688.l I i l i i l 1 i i I Q001.503-1 f Amend. 62 I i flov. 1981

I O Questien C01.504 (F6.4.1.2) Provide the basis (including accident energetics) of the estimate that 58". of the debris is expected to enter the Primary Heat Transport System (PhTS) . In addition, provide information on the consequences of such an occurrence, including consideration of the final destination of the debris in the PHTS and settling and dryout of debris leading to failures of PHTS ccmponents. Rescense: The analyses questioned above have been superseded by those currently presented in Section 3.1.1 and Appendix I of CRSRP-3, Volume 2 (Reference ICb of PSAR Section 1.6). O

  .)                                   q001.SO4-1 v/                                                                         Amend. 62 Nov. 1981

d Question 001.505 (F6.4.1.3) Melt-thrcugh cf the icwer reactor vessel head is predicted to occur in not less than 1600 seconds. However, non-uniformities in the steel melting could result in significantly earlier genetraticcs. The melting irregularities may develop due to the lighter , n.olten . 31 bubbling up through the more dense UO2 at preferred locatient This bubble spacing would be established by a Taylor-type instability. If the bubble release points were to remain spacially staticnary, the heat flux through a thinner molten layer between these points would be higher than the average value predicted by a uniformly advalcing melt front. Since the thermal load en the Ex-Vessel Core Catcher (EVCC) and the External ' Cooling System (ECS) is strongly dependent upon the time required for a melting attack to breach the reactor and guard vessels, provide an assessment of the imcact that earlier arrival of molten material would have on the performance of the EVCC and ECS, Resocnse: The infor ation requested, as applicable to the TM3D8 design, is addressed in the respense to questien 001.633. O QC01.505-1 l O'- Amend. 62 Nov. 1981

i 7 l i l 1 1 ' t l l6  ; Ouestion 001.5C6 (F6.4.1.3.1) 5 }

'                                                                                                                       l Provide complete details on your calculation of time to melt through the        i i                                        lower heads of the reactor and guard vissels. These details should include      l

} the decay heat level and sodium temperature versus time, individual masses i of core debris and steel components being melted, criteria for vessel . l penetration, and all relevant assu..ptions.

Response

The information requested can be found in CRSRP-3, Volume 2 (Reference 10b  ; of Section 1.6), Appendix 8. - 1 l \ i I O G Q001.506-1 Amend. 62 flov. 1981

1 I i I \ l j Ouestien 001.507 (F6.4.1.4)  ! Clarify how debris beds in a dry environment on the thermal baffle can - I i radiate to the cavity EVCC external c oling system [ i l Rescense: 4 l } With the deletion of the Parallel Design in Amendment 24 this question i ! is no longer applicable as the features upon which the questien is based i ] are no longer a part of the design. t I

! i i

! I

i i

) i t i i i  ! i . t i  ; i 1' Q001.507-1 O Amend. 62 Nov. 1981

4 ]. I !1 i 4 O Question 001.508 'F6.4.2) t Provide the basis for not requiring CHRS to be cperational following a CDA. Resconse: The objective of the Direct Heat Removal Service, DHRS, (formerly the OHRS) is specified in Section 5.6.2.1.1 as preventing loss of in-place coolable geometry of the core following reactor shutdown in the event of simultaneous failure of the nornal shutd6wn heat removal systems. A ' hypothetical CCA, by definition, goes beycnd the point of ma.intaing in-place j coolable core secretry. As such, no credit has been taken for either ' j the normal shutdown system or for the DHRS capability in the TMSDB analysis. } i i i ! i r i t l I I l l r I, t I I l l l [ I q001.5081 f Amend. 62 e Nov. 1981 l i. f i i

e O l { Questien 001.509 (F6.4.2) For the case of extensive reactor vesst,. flew blockages esulting in temperatures exceeding 1230*F, ard subsewent failure of the reactor and  ! I guard vessels, provide the basis for t.ssuming that core debris will not be introduced en any other portion of the reactor cavity except the EVCC.

Resconse

l l With the deletion of the Parallel Design in Amendeent 24 this question is no lenger relevent as the EVCC is no longer a part of the design.

i I I i i L l l 1 I 1 I l \ I t Q001.509-1 Amend. 62 ! Nov. 1981 i-L i

s ! I

t i

l l Question 001.510 . 1 , l [ ! Provide the techno il . cation for not considering the heat load j associated with the of radioactive nuclides in the reactor steel i structures dJe to nt Japtures. I i l Response: f i The activation of react .r steel structures is confined primarily to the

stainless steel concentrated in the core, blanket and near-core shielding

} regions. The associated decay power at shutdown from this activation is , i on the order of 112 kw. This is negligible when compared to the fuel  : and blanket decay heat loads from Table C.1-3 of CRBRP-3, Volume 2 l (Reference 10b of PSAR Section 1.6). [ i i 4 i l I I i i I 4 r 1 l l 1 l J l i l I l i l ) l l l j  ! i r itill  : Q001.510-1 Amend. 62 i Nov. 1981 i

O Question 001.511 (F6.4.3.4): Provide a discussion of the consideration that has been given to precluding thermal stress induced cracking in the sacrificial bed because of sudden contact with hotfolten core debris. Resconse: With the deletion of the Parallel Design in knendment 24 this question is no longer relevant as the features upon which th'e question is based are no longer a part of the design. O QC01.511 -1 Amend. 62 Nov. 1981

O Ouestion 001.512 (F6.4.3.5): In view of the uncertainties of solubility rate and heat transfer frem molten fuel pools in sacrificial beds, provide the follcwing information: a) provide estimates of the uncertainties and design margins associated with the final pool configuration and the thermal load distribution to the EVC system. b) In determining the capabilities of the EVCC and ECS, provide a discussion of the consideration that has been given to possible non-uniformities in sodium temperatures and molten pool heat fluxes. c) Because of the potentially significant effect on heat traatfer and penetration into the sacrificial bed, provide a disetnion of the censideration that has been given to solid UO2 and/or stainless steel crust formatien on the troltan fuel in the core catcher. d) Provide a description and discussion of any proposed R&D programs that will yield the appropriate information relative to Items (a), (b) and (c) above. _ Response: With the deletion of the Parallel Design' in Amend cnt 24 'this question is no longer applicable as the features upon which the ' question is based are no longer a part of the design.' l O Q001.512-1 Amend. 62 Nov. 1981 I

Ouestion 001.513 (6.2.6.1.3 Yellow) Provide additional details on how the cargin seals on the closure head plug risers prevent sodium or core debris egress into the sealed head

access area following a CDA.

Resconse: The sealed HAA is not part of the present design. Also, the CCA is now only considered as a hypothetical event beyond the design base. The margin seals must meet the SMPOS sodium and gas leakage requirements given in Section 5.3.2 of CRSRP-3, Volume 1 (Reference 10a of PSAR Section 1.6). Cetails on the margin seals are given in PSAR Section 5.2.4.4. Tests in support of the margin seal design are discussed in Section 5.4.1.5 of CF3RP-3, Volume 1. l

O J

O Q001.513-1 Amend. 62 Nov. 1981

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

Question 001.514 (6.2.6.1.3 Yellew) i Provide additional details on how the structural integrity of the sealed  ! head access area concrete is assured considering the additional heat load [ introduced following a CDA. Resconse: l In Amendment 24 to the PSAR, the Project withdrew the Parallel Design from further censideration by the NRC staff. This question requests additional design informatien en a specific feature of the Parallel Cesign. There-fore, the question is no longer relevant. i F I  ! t i a

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( r 1 i l i r l i l l l r I k l 0001.514-1 9 Amend. 62 Nov. 1981

O Question 001.515 (6.2.7.2 Yellow) Provide the therec-physical properties of the magnesia pewder relative to the initial ir. pact of UO2 on the core debris receptacle. Include in your response a discussion to address the possbility and consequences that the 5 inch thick layer of MgG pcwder may be flushed away by the impingement of a hot, dense liquid stream, thereby diminishing resistance to thermal shock. Resconse: With the deletion of the Parallel Design in Amendment 24 this question is no longer applicable as the features upon which the questien is based are no longer a part of the design. O O Q001.515-1 Amend. 62 Nov. 1981

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

o 4  !

O j Ouestien 001.516 (6.2.7.2 Yellcw) i Figure F6.4-ll shcws significantly greater temperature gradients at the  ;

i centerline of the sacrificial bed than along the side walls. In view of ' i this, provide a discussion of the consideration that has been given in  ! your design to accommodating local hot spot effects, such as varying either cooling coil spacing or NaK ficw rate.  !- 1 . Resconse: 1 J With the deletion of the Parallel Design in Amendment 24 this question is no longer applicable as the features upon which the question is based are j no longer a part of the design. l I I i !O t I , l t t 9 Q001.516-1 Amend. 62 Nov. 1981 l _ = _ _ _ - _ . . - _ - - _ - - - . - - . - - . - - _ __

I i Questien 001.517 (5.2.7.2 Yellew) Provide the technical justificaticn that the 35 feet height for the cooling system annulus is adecuate to maintain structural integrity of the reactor cavity structures and concrete folicwing a CDA. Resconse: With the deletion of the Parallel Design in Amendment 24 this questien is no longer applicable as the features upon which the question is based are no longer a part of the design. l I } 0001.517-1 Amend. 62 Nov. 1981

q

\_f Questien 001.518 (6.2.7.2 Yellow)

Provide a discussion of the censideration that has been given to ccoling the reactor guard vessel support structure, including whether conduction frem hot sodium along the steel structure could result in localized hot spot areas of cencern as to structural adequacy. Resocnse: After penetration, no ecoling of the guard vessel support structure is necessary since the guard vessel serves no function in the TMSDB scenario. The heating of the support structure could be expected to cause sagging of the guard vessel until it ccmes to rest on the outlet nozzles of the reactor vessel.

%)

73 0001.518-1 a q,) Amend. 62 ilov. 1981

1 i l i !O 1 1 Ouestion 001.519 (6.2.7.2 Yellow) Provide additional details on the structural support of the various Ex-Vessel Core Catcher (EVCC) components, including the impact loads that are being j considered. Provide the impact loads that are being used in the design of

the EVCC, including a discussion of the consequences of reactor and guard vessel impact.

Resocnse: I j With the deletion of the Parallel Design in Amendment 24 this question is no longer relevant as the features ucon which the question is based are no longer a part of the design. I i ) < l i I 1 l ' 1 }  ! 1 i i  ! ! I i t f i - ! I I  ; 1  ; I  !

i

} l 1 \ 1 l QC01.519-1 Amend. 62 + Nov. 1981 t

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

1 I< l Ouestion 001.520 (6.2.7.2 Yellew) Provide a discussion of the impurity content and control, especially moisture for the magnesia sacrificial bed, including the consequences of such impurities on materials interactions foliosing introduction of core debris. Provide a description of the instruments, systems and measures to be employed to monitor impurity levels in the magnesia bed. Resocnse: With the deletion of the Parallel Design in Amendment 24 this question is no longer relevant as the features upon which the question is based are no longer a part of the design. i l i ( O i I 1 l i i O 0001.520-1 Amend. 62 Nov. 1981 4

    -r        -,y,,,..-w,-ew.-        ,.,,ewm-w              .- - ve ,. n e.                     e e rm , ..+ w eve .      .vm .      ww.,.,---,.,,-+-~%-we.we-=---

I l i O Ouestion 001.521 (6.2.7.2 Yellow) Provide additional details on the type and thermo-physical properties of the } graphite material that will be used to fill the cooling system annulus. l Resconse: With the deletion of the Parallel Design in Amendment 24 this question is no f l longer relevant as the features upon which the question is based are no longer a part of the design. l l l l l l l l \ e  ! i I l I i I i l l, r l I I !G 0001.521-1 Amend. 62 Nov. 1981 i P m 9

                                                                                                                    . . , ,r . .'y"

        . . . - - -           ~ . -       -      -

i I O Ouestion 001.522 (6.2.7.2 Yellow) l Since significant adverse materials intera;tions have been observed to take I place between sodium and silica-containine, materials, justify the use of i silica-alumina firebrick insulation in the area behind the steel liner. I Resconse: I i With the deletion of the Parallel Design in Amendment 24 this question is no icncer applicable as the features upon which the questien is~ based are no icnger j a part of the design. 1 i l I l ! l O i i I . i i 1 l l [ i i I i O Q001.522-1 Amend. 62 flov. 1931 i I

i i ( 4 i i l i l l 4 i till  ; Question 001.523 (6.2.7.3.1 Yellow) Provide a discussion of the consequences of sodium absorption in the sacrificial  : bed on the molten pool heat transfer, including the impact of sodium vapor generation on the structural integrity of the sacrificial bed and reactor cavity. , Resconse-i With the deletion of the Parallel Design in Amendment 24 ".11s question is no l ) longer applicable as the features upon which the question is based are no f l longer a part of the design. l { l i I l l i l i l \ l I r I l \ l l I l i r k P Q001.523-1 Amend. 62 f Nov. 1981  ! r

                                                                                                           -J

l l !O Question 001.524 (6.2.1.3.1 YeQc d: The assumption appears to have been made that the sodium pool above the molten core debris is at uniform temperature and that any sodium vapor, created by local boiling, will be cendensed, precluding any increase in the i reactor cavity pressure. Natural convection may result in an upswelling i i plume of hot sodium in the center of the sodium pool above the EVCC and hot core debris. The vapor pressure of the hot sodium surface near the center of the pool could influence the extent of overpressure in the reactor cavi ty. Consider this in your evaluation and provide an assessment of the consequences; the capability of thc reactor cavity design to cope with these consequences, including overpressures r:Ust be addressed.

Response

i l The EVCC is not part of the current design. The infonnation requested, as applicable to the TMSCB analysis, is addressed in the response to ques tion 001.617. i i I l l !O \ I l i i i l I i i i l l l g qc01.524-1 Amend. 62 Nov. 1981 l I i i F

=

                                                                                                                                                                                )

l O Question 001.525 (6.2.7.3.1 Yellew): There does not appear to be adequate design margin for the sacrificial bed depth in view of the uncertainties associated with solubility rate . of bed material into the molten core debris, and variations in heat flux distribution over the pool boundary. Justify that conservative solubility i and heat transfer rates were used in calculating final pool size. including a description and discussion of applicable R&D programs. In addition, provide a discussion of the sensitivity of the molten pool termination j point to variatiens in solubility and heat transfer rates. t i Rascense: With the deletion of the Parallel Design in Amendment 24 this question is no longer applicable as the features upon which the question is based are no longer a part of the design. O I r i O q001.525-1 Amend. 62 Nov. 1981

 ,,--,en       ,n. ---. , -- - . _,. -,e.,-    ----.-,--n.     .-..--            . -           , , , , . .  .,,r-                 - - - - - - -, - -

l i , i , t a i Oues tion 001.526 (6.2.7.3.2 Yellcw) Since maintenance is intended to be perfomed on one of the ECS cooling loops during reactor cperation, provide the technical justification to  ; demonstrate that, in the event of a single failure in the operating loop,  ! ! sufficient time is available for restoration of one of the ccoling loops t j to service before either sodium or reactor cavity structural temperatures reach ' j unacceptable levels following a CDA. l 1 l l Resoonse: l ' i

 !           With the deletion of the Parallel Design in Amendment 24 this question is no                                                                                             i longer applicable as the features upon which the question is based are no longer a part of the design.

! i 4 4 l i. 1 I I lO i i i i l { i l l l i I i l i iO i Q001.526-1 l Amend. 62 l Nov. 1981 l I

O Ques tion 001.527 (6.2.7.3.2 Yellew) Confirm whether both ECS cooling systems can be operated simultaneously in the event of unanticipated higher heat loads. Resocnse: With the deletion of the Parallel Design in Amendment 24 this questicn is no longer acplicable as the features upcn which the question is based are no longer , a part of the_ design. l i 4 1 4 { lO O QC01.52 7-1 Amend. 62 Nov. 1981

l f I i alip > j Ouestion 001.528 (6.2.7.3.1 Yellew) 1 l i Provide the estimated ECS NaK ccolant activity levels following a CDA and i an assessment of the radiological consequences of any NaK spills, leakage or other release in the DHX building. - Resconse: With the deleticn of the Parallel Design in Amendment 24 this question is no longer applicable as the features upon which the question is based are no longer l a part of the design. L r i 1 r G Q001.528-1 Amend. 62 flov. 1981 i t t 6

O Questien 001.529 (6.2.7.3.2 Yellow) Provide the sequence of operator actions and correspending ticas to  : manually activate the ECS in the event of a CDA. In addition, provide t a discussion of the necessary instrumentation that will be provided to assist the coerator in detemining the actions to be taken. Address the potential for and consequences of inccmplete or inadequate operator ' action. Respense: In Amendment 24 to the PSAR, the Project withdrew the Parallel Design from further consideration by NRC staff. This question requests additional infomation pertinent to a specific feature of the Parallel Design. Accordingly, the question is no longer directly applicable. However, the . operator actions associated with the TMSDB features are discussed in Section 2.3 of CRSRP-3, Volume 2 (Reference 10b of PSAR Section 1.6). Instrumentation requirements are provided in Section 2.1.2.12 of the  : re ference. O Q001.529-1 Amend. 62

Nov. 1981

Question 001.530 (6.2.6.1.2, 5.2.1.3 Yellow) In veiw of the uncertainties associated with sealed head access (SHAA) head r.argin seal performance during a CDA, provide an upper bound estimate of the quantity of sodium that can be ejected into the SHAA before nitrogen inerting beccmes a necessity. Justi1? on a conservative basis that ejection of the above amount of sodium is precluded by head margin seal design. I Resconse:

 ;                      With the deletion of the Parallel Cesign in Amendment 24 this question is no lenger applicable as the features upon which the question is based are no longer a part of the design. See the response to question 001.513 for a discussion of potential head leakag'...

1, O i i i Q001.530-1 Amend. 62 Nov. 1981

i Question C01.531 (3A.l.3 Yellow) Since the crimary sodium overflow and make-up line penetrations are extensions of the reactor cavity and part of the inner containment boundary, provide the basis for the differences in design pressures as listed in Table 3A.1-3. Resconse: With the deletion of the Parallel Design in Amendment 24 this question is no longer applicable as the inner containment boundary is no longer a part of the design. 1 l j I l l l Qcol.531-1 W i9hf (

O Question 001.532 (9.13.2.2 Yellow) For the EVCC cooling system cells in the Reactor Centainment Building which contain an air atmosphere, provide the technical justification that structural integrity of these cells and independence of cooling loops will not be compromised in the event of a NaK spray fire on the side walls and ceiling, or in the event of a NaK spill onto a catch pan where the nitrogen flooding system fails to operate. Resconse: In Amendment 24 to the PSAR, the Project withdrew the Parallel Design from future consideration by the NRC staff. This question requests additional design information on a specific feature of the Parallel Design. Accordingly, the question is no longer applicable. Q001.532-1 Amend. 62 O flov. 1981

O Question 001.533 (11.3.2.6 Yel10w) Since leakages through the reactor head seals and buffered seals represent the principal constituents of the total off-site radioactive release, provide the basis for not processin Atmosphere Processing System (CAPS)g in order the SHAA atmosphere to satisfy as low as is through the Cell reasonably achievable release criteria.

Response

With the deletion of the Parallel Design in Amendment 24 this question is no lenger applicable as the features upon which the question is based are no longer a part of the design. O QC01.533-1 O' Amend. 62 Nov. 1981

h J Questien 001. 534 (FS.4.3.5, 6.2. 7.2, 8. 3.1.1 ) The capacity (2000 gpm) for each of the two EM pumps in the Ex-Vessel Core Catcher Cooling System appears to be.beyond the current state of the art. Since these pumps will be required to operate for an extended period of tirne for decay heat removal, provide a description and discussion of the experience and/or R&D program that exists for EM pumps of this capacity. Resocnse: With the deletion of the Parallel Design in k.endment 24 this question is no longer applicable as the features upon which the question is based are no longer a part of the design. 0\ U Q001.534-1 Amend. 62 flov. 1981 N

A U Question 001.535 (6.2.7.3.1 Yellow): Jahn and Reineke (

Reference:

Paper NC2.8, International Heat Transfer Conference. Tokyo,1974) have shown that a wide variation exists in the heat transfer coefficient along the submerged surface of a volumetrically heated liquid pool held within cylindrical cavity. This result indicates the possibility of the reactor vessel being initially penetrated by molten 007 at a point on the vessel sidewall near the surface of the U02 pool, and not at the midpoint of the lower dome. Similar possibilities exist for the reactor guard vessel penetration. Provide a discussion of the capability of the EVCC to cope with introduction of core debris in an asymetric manner, including thermal, structural, and criticality considerations, include a discussion of the dependence of molten pool growth on the initial core debris distribution in the sacrificial bed. Furthemore, provide a description and discussion of the R&D program proposed to resolve this question. Resconse: With the deletion of the Parallel Design in Amendment 24, the EVCC is no longer a part of the design. Response to this question, as applicable to tne TMSCB analysis, is provided in the response to question 001.617. O U Q001.535-1 Amend. 62

   )                                                                                                  Nov. 1981 J

O Questior 001.536 (6.2.7.3.1 Yellow) If the sodium pool is deep enough to submerge a portion of the cylindrical guard vessel support skirt, two isolated ullage regions could be created above the sodium pool in the bottom of the reactor cavity. If this occurs the vapor pressure associated with higher temperature sodium near the center of the pool would cause the surface to move downward within the support skirt while raising the level in the annular region outside the skirt. Discuss the effect that the uneven sodium level would have on the reactor cavity and ECS. In addition, confirm whether or not cpenings through the guard vessel support skirt would be provided to eliminate the potential sodium level unevenness.

Response

With the deletion of the Parallel Design in Amendment 24, the ECS is no longer part of the design. The effect of unevenness of sodium pool level within the reactor cavity has little effect on the reactor cavity. The sodium level outside the guard vessel skirt would be raised less than ene foot frca the ccmplete expansion of gases entrapped to fill the area in question to the level of the bottom of the guard vessel. Once the gas has expanded to push the sodium level below the elevation of the ruptured area of the guard vessel, the entrapped gas would escape into the reactor cavity between the reactor vessel and guard vessel. The guard vessel support skirt includes openings as indicated in Section 2.1.2.3 of CRSRP-3, Volume 2 (Peference ICb of PSAR Section 1.6). Q001.536-1 Amend. 62 I!ov. 1981

i O Question 001.558 (6.2.7.2 Part II) i Several 4" vents are provided behind the cavity liner which vent into the RCB in the event of a meltdown. Indicate the rate of pressure rise in the RCB due to such venting and the time it will take to reach 10 psig. Resconse: In the present configuration the Reactor Cavity floor liner is vented to the Reactor Containment Building (RCB) and the wall liner is vented to Cell 105, which is connected to the RCB. In addition, the Reactor Cavity atmosphere is vented to the RCB via a planned vent path actuated by rupture disks which would open shortly after Reactor Vessel penetration due to heat up and pressurization of the Reactor Cavity atmosphere. After a liner section is predicted to fail, the area behind the failed section is assumed to vent into the Reactor Cavity. The TMBOB scenario has been analyzed with these release paths in CRBRP-3, Volume 2 (Reference 10b of PSAR Section 1.6). In the base case, the RCS reaches 10 psig in approximately 20 hours. (A short term pressure transient due to hydrogen burning exceeds 10 psig at approximately 10 hours). O l l 0001.558-1 O Amend. 62 Nov. 1981

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

I Question 001.566 (FII 1.5.2.8.5) Specify what, at this time, would limit the design of the enclosure of the RAA and indicate if there are temperature limits on the seals, concrete, or pressures from the reaction products. Discuss the current limits on the design pressure, in terms of structural acccmmodation for higher pressures. ' Resconse: I With the deletion of the Parallel Design in Amendment 24 this question is no longer relevent as the features upon which the questien is based , are no lenger a part of the design.

I 9

i l l l Amend. 62 9 Q001.566-1 g;0v, 19g1 l I

Question 020.28: . Provide criteria and bases used in determining the size of the liquid argon and liquid nitrogen storage on site. Fresh argon supply rate should be based on the possibility of failure of the Radioactive Argon Proces;ing Subsystem (RAPS) and, consequently, no purified argon return from RAPS to the Primary Recycle Cover Gas Storage Tank. Provide information to demonstrate that argon can be delivered to the site in the event of extreme natural phenomena, such as rain, snow, and resultant floods before depleting onsite stored gases.

Response

There are five liquified gas storage complexes in the IGRP system. Two of these are at the RSB pad, two are at the SGB pad, and one is in the RSB. l The RSB argon supply consists of two 1500 gal, dewars arranged to l deliver gas in sequence, or in parallel. Any dewar can be charged at will. The size of these vessels is determined by the projected consumption and the desired reserve capacity. The normal usage of argon, once the system has been filled and settled in its operation will be modest, and a single dewar will provide a minimum of 30 days of normal service. About half of dewar will be required to re-inert the Fuel Handling Cell. Therefore, the two dewars provide the necessary back-up when this larga cell is being serviced. The RSB nitrogen supply consists of two 6000 gal. dewars arranged to deliver gas in sequence, or in parallel. At the design-value use rate required to supply inerting gas to the RSB and RCB cells, each dewar can provide a 6-day supply. When sodium component cleaning operations are in progress, one dewar can provide a 3 day supply. Vessels provide a minimum of six days of service at the maximun use rate. The 6,000 gal. dewar size was chosen to coincide with the capacity of a standard long haul cryogenic tanker truck. Such a tanker is l expected to be used to provide scheduled recharging service. The currently identified source of supply is located in Huntsville, Ala. , which is approximately 175 miles from the plant site via primary surface roads. Normal transit times can be projected to be less than 5 hours so that the 3 day reserve provides sufficient time to recover from late deliveries due to natural causes (weather, accidents, etc.) The delivery of argon will be on a similar basis. Amend. 62 QO20.28-1

The SGB argon supply consists of two 1500 gal. dewars arranged to deliver gas in sequence, or in parallel. Each dewar is expected to provide normal service for at least 30 days. Maintenance and sodium transfer operations in SGB are not expected to require more argon than can be supplied with adequate reserve by the two dewars. These dewars also provide a back-up supply via a tie-line to the RSB dewar system. l l ,?e normal SGB nitrogen supply system consists of two 3,000 gal. j dewars. For normal operation, the service period of each tank is ' expected to be about 30 days. However, a sodium cleaning operation has been projected for an SGB location. Its needs would require one tank's , capacity in 4 days. Therefore, the two dewars provide an adequate l reserve for futu,e needs. The sodium-water reaction nitrogen supply consists of one 3,000 gal. dewar with a connection to the normal SGB nitrogen supply for emergency use. This supply is provided for use following sodium / water reaction events. The nitrogen is used as the SWRPRS inert cover gas and for holding the pressure on the water side of the steam generators following sodium dump. The steam generator system has only a small normal use-rate, and will be recharged to fill the dewars when the supply tanker arrives for any liquid nitrogen service. The dewar can supply service to one steam generator module for about 36 hours. O QO20-28-2 Amend 62 Nov. 1981

.i ) !O Question 020.40 (9.5.1 Yellow) Provide design criteria and bases used to determine the size of ernergency argon storage facility inside the Dump Heat Exchanger (CHX) building. J

Response

l With the deletion of the Parallel Design in Amendment 24 this question , is no lenger relevent as the features upon which the question is based are no longer a part of the design. 1 t i J l l c

O i

i QO20.40-1 Amend. 62 O ilov. 1981

I O Questien 020.41 (6.2.7.2 Yelicw) Provide a description, preliminary Svout drawings and P&ID showing the heating, ventilating and cooling, and other auxiliary syste:::s required in the CHX building. Discuss the effect of the additional requirements of the DHX building on the capability of the related auxiliary systems included in the reference design. Rescense: In Amendment 24 to the PSAR, the Project withdrew the Parallel Design from further consideration by the NRC staff. This question requests additional design information on a specific feature of the Parallel Design. Accordingly, the question is no lenger relevant. O . l 1 i QO20.41-1 Amend. 62 l s Nov. 1981 I__.___._-... _ _ _ _ _ _ _ _ _ _ _ . _ - . _ _ _ _ . . - . _ _ _ _ _ . _ _ _ - . . _ . . _ _ _ -~ _ -- -

1 i

 )

l i 1

 )                                                                Question 020.42 i

Identify the means proposed to isolate the OHX in the event of fire and . j include the treasures to be used to inert the atmosphere. (Refer to our j previous request 020.20). , Resocnse: 1 With the deletion of the Parallel Design in Amendment 24 this question is no lenger applicable as the DHX is no longer a part of the design.

 !                                                                                                                                                                                    I i

t I  ! i

                                                                                                                                                                                       ?

t l  : j k l l [ QO20.42-1 O Amend. 62 Nov. 1981 l

O_ue s tion 020.43 (9.5 Yellow) It appears that credit is being taken for operation of the inert gas receiving and processing system following the postulated CDA. Speci fically, the capability of the argon gas distribution, radioactive argon processing (RAPS) and cell atmosphere processing (CAPS) subsystems aapear to be assumed to be available. The subsystems are not designed as engineered ' safety features (Table 6.1-1) and postulated single failures (e.g., failure of the cryogenic column) apparently are not considered. The need for these subsystems is not clear. Revise your analyses accordingly or prepose revised systems' designs to reflect their safety-related function, if any. _ Response: The current TlCDB analyses [see CRSRp-3, Volume 2 (Reference 10b of PSAR Section 1.6)] take no credit for operation of the inert gas systems during the postulated scenario. O 4 1 i ( I QO20.43-1 Amend. 62 O Nov. 1981 , - - . . - , _ _ _ _ _ _ _ _ _ - _ _ _ _ ~ _ . _ .

(' Question 040.19 Clarify that lines penetrating containment which are connected to the reactor coolant boundary, primary cover gas space, or inerted cell atmospheres will have the capability of periodically testing the operability and leakage of the containment isolation valves (Section 6.2.4.1.).

Response

Seven inert gas process pipes penetrate the containment building. Two isolation valves are provided at each penetration point as shown in Figure 6.2-10 and discussed in Table 6.2-5. Ten of the isolation valves close automatically when a CIS signal is received. Four valves close on loss of line pressure. Test taps are provided outboard of each isolation valve to enable perfornance of leak testing. Valve position indicators on each automatic closing valve verifies operation of the valve when each valve is exercised by its respective remote valve switch located on the CIS panel. The cover gas sampling line isolation valves will have the same leakage and operability testing capabilities as the process line valves. A typical schenatic for the valve arrangements is shown in Figure Q040.19-1. %.) Amend. 31 Nov.1976 0040.19-1

0 0 C C I l I i l l 1  ; I I I I O U IS -. ZS

                                                                                                                                              ~                                                    '
                                                                                                                                          /     's                                                                           f ' ' N' a

x) - O U T SI D E 1 x& {' INSIDE

                                                                                                                                                                                                                            ;x; N s  /
                                                                   ?N CONTAINMENT PE NE TR ATION
                                                                                                                                                         -3
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                                                                                                                                                     %,.'                                                             ~    /

LJ LJ FIGURE 0040.19-1 Typical Containment Penetration 2, Configuration E5 ii iti P0802 9

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p Question 110.19 (3.6, 3.6.1, 3.6.2) V Your intent to use the J.F. O' Leary Letter of July 12, 1973, for pipe break locations, break sizes and orientations for systems inside and out-side containment is not completely acceptable. State your intent to comply with the latest criteria as specified in NRC Standard Review Plan (SRP) 3.6.2 " Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping" for all systems in which breaks are postulated. Wi til respect to SRP 3.6.2, include the following and provide illustra tions  ; applicable:

1) Provide the details to be used for piping penetrations of containment a reas . Indicate the use of protective assemblies or guard pipes. -

State whether such protective assemblies serve to provide an extension of containment, prevent overpressurization, or provide both functions.

2) Indicate the use of moment limiting restraint's at the extremities or within the penetration assembly.

3) Provide the criteria for the design of the process pipe within the penetration assembly. Include type of material ISeamless or welded),

allowable stress level, and loading combinations. 13 4) Provide the design criteria to be used for flued heads and bellows V expansion joints.

5) Provide the design criteria applicable to any guard pipe which is utilized with the assembly.

Response

CRBRP compliance with the J. F. O' Leary letter was discussed with NRC as documented in Reference Q110.19-1. The response to question 020.4 will further document the CRBRP compliance with the intent of the NRC position. In response to the specific concerns identified, there are no high energy piping penetrations in the RCB; however, the foliowing discussion is provided for the IHTS containment penetrations:

1) Figures Q110.19-1 shows the typical arrangement of the pipe penetra-tion. The boundary of the containment shell is evident from the figure.

2) There are no pipe rupture restraints for the intermediate bay side of the penetration.

iO iV Amend. 26 Aug. N76 Q110.19-1

3) Design loading for the penetration provides ior the maximum forces and mcments whici could be imposed by the pipe. The intermediate pipe is 316 SS in hot leg and 304 SS in the cold leg. There will be a transition weld between the carbon steel containment penetration and the 316 or 304 SS adaptor. The integrity of the assembly will be demonstrated against sodium spray by analysis demonstrating confor-mance to the ASME Section III limits.

4) The design criteria for the flued head is ASME Section III, Class 1.

5) There is no guard pipe in the conceptual design (as evident from the attached figure).

The following discussion is provided for the compressed gas, chilled water and drain system piping penetrations: (1) Figures Q110.19-2 shows the typical arrangement of the pipe pene-trations which serve to provide an extension of the containment. (2) These piping systems are considered moderate energy systems and moment limiting restraints for pipe rupture loadings are not required. However, pipe stops / guides are provided within the penetration as-sembly to limit moment loadings on the closure due to other normal design conditions. (3) Process pipe within the penetration assembly will meet the allowable stress levels anc loading combinations as required by ASME Section III. (4) Bellows expansion joints and closure plates (or flued heads) will meet the design criteria of ASME Section III, Subsection E. (5) Guard pipes will not be used. The detailed configurations of other penetrations have not yet been determined, but will be similar in concept to the IHTS penetrations in that (1) a cylinder is welded to the containment vessel; (2) the process piping passes through the cylinder without guard pipes or piping rupture restraints; (3) the design criteria for the penetration will be at least as conservative as the criteria of ASME Section III, Class 2; (4) no expansion joints (bellows) will be used except as necessary for testing purposes. Reference Q110.19-1: Letter S:L:653, P. S. Van fiort to R. S. Boyd,%mmary of CRBPP/tmC Meeting on Pipe Breaks Outside of Containment," March 3,1976. Amend. 62 Q110.19-2 fiO V - 1981

l < l l 1 9 Question 110.57 (6.2.7.3.2 Yellow) < \

}             The External Cooling System (ECS) must be described in greater detail than        l j             that described in Appendix F. The description should contain the functional      l

, details of the system, loading and operating design criteria, operational l testing methods, and the exolicit codes and standards to be adapted.

]
)             Resocnse:

With the deletion of the Parallel Cesign in Amendment 24 this question

;             is no lenger applicable as the features upcn which the question is based         ,

j are no lenger a part of the design. j 4 i i i" i l 1 1 i 1 + 4 l9

,                                                                                             i I

i l, I l l l i l i ' I l i l l l Q110.57-1 9 Amend. 62 flov. 1981 i

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Question 120.28 (5.2.3.4) Describe methods that will be used to verify the integrity of the core support structure of the vessel during service life.. Response : The integrity of the com support structure during service life will be verified by material surveillance. The CRBRP material surveillance program provides for core support structure material specimens and will include tensile specimens. These specimens will be placed in surveillance locations having a higher flux than in the region of the comoonent whose material pro-perties are being verified. Verification of the expected material behavior in conjunction with analysis is the best known means of verifying the integrity of the core support structure. The material behavior, as determined from the surveillance specimens, will lead the actual component neutron exposure. The core support structure tensile specimens are selected on the basis s of a minimum total residual elongation of ten percent. The first measurable l loss of ductility in austenitic stainless steel occurs at 1 x 1021 n/cm2

 >0.1 MeV with increasina loss of ductility at higher fluence levels. Thus, tensile specimens are selected for all regions of the core support structure whem the end of life fluence is equal to or greater than 1 x 1021 n/cm2    Notch ductility degradation, as well as strength property changes, are also progressive with increasing fluence above 1 x 1021 n/cm2 O                                    Q120.28-1 Amend. 62 Nov. 1981

l Question 120.58: The material presently specified for the Steam Generator Auxiliary Heat Removal System (SGAHRS) is carbon steel. Describe the procedures to be taken to ensure against caustic gouging, stress corrosion cracking, pitting corrosion, incompatibility with insulation and methanation.

Response

Specific procedures for maintaining and monitoring SGAHRS water chemistry are yet to be developed, however, the system has been designed to allow recirculation mixing (if required) to assure that representative samples are taken during sampling. The minimum water purity level specified is 1 as follows: Cation conductivity (at 70 F) <10 micro Mho/cm pH of 9.5 to 10.0 by ammonia addition hydrazine (catalyzed) 150 25 ppm Suspended solids < 5 ppm This water chemistry is a common wet lay up chemistry used in commercial plants which is designed specifically to prevent pitting corrosion. The extremely low corrosion rates associated with this chemistry along with the low operating temperatures preclude methanation. Caustic gouging is not a problem since the SGAHRS contains no caustic. SGAHRS water will contain no sodium hydroxide,only ammonium hydroxide which will not cause stress corrosion cracking in carbon steel. (V) Since the usual effect of increased velocity is an increase in corrosion rate, flow velocities have been limited to the approximate ranges specified below: Type of Service Maximum Velocity - fps Pump Suction 10 Pump Recirculation 70 Steam Drum Feed 20 Saturated Steam-PACC 22 Saturated Steam-Turbine Supply 200 Turbine Exhaust 500 Superheater and Steam Drum Vent 500 The minimum corrosion allowance, including cleaning, in terms cf additional thickness of material, shall be 0.10 inches. This shall be deducted from the available structural material before strength calculations are performed. O kJ Amend. 69 Q120-58-1 Nov. 1931

O Question 120.62 (120.15, 120.41, 120.47, 120.49) Responses to previous questions are not sufficient to conclude that the mechanical properties of welded austenitic stainless steel will not degrade during the life of the plant. The information submitted does not address adequately the long term thermal aging effects. The specific weld filler rod and welding procedures to be used in CRBRP will affect the weld ferrite content, and, after thermal aging, the sigma phase morphology. Sigma phase can degrade the weld joint mechanical properties. It is our position that tests should be initiated prior to plant construction to evaluate the long term thermal aging effects upon the mechanical properties, toughness and crack propagation of welds using materials and procedures spec-ified for CRBRP. Response: (Interim) Qualitative assessment of the limited available data regarding micro-structural and property changes of austenitic stainless steel weldments as related to service temperature and exposure time indicate that the materials utilized in CRBRP have a high likelihood of acceptable mechanical performance. However, the need is recognized to expand the data base, particularly the data on weldments thermally aged for suitable times at temperature and/or N proven methods of extrapolating available short term data. An experimental program to examine the effects of long term thermal aging on welded austenitic stainless steel is included within the base technology program. This program would utilize the specific base metals, weld filler metals, and weld processes to be utilized in CRBRP fabrication. Specimens will be evaluated in the condition prototypic of start of plant life and after various thermal aging times. The duration of thermal exposure would be terminated on a case-by-case basis. Material combinations selected for investigation would include prototypic primary hot and cold leg piping welds as a minimum. The properties to be evaluated would be selected to provide insight on property degradation as related to likely failure modes of the component involved and may include any or all of the following: o microstructural evaluation e tensile properties e creep properties Q120.62-1 Amend. 62 Nov. 1981 _ , - .-_ _ _ _ - _ = .. .- . - -

r crack propagation toughness as determined by the J-integral method l ! The schedule for the experimental program will be dependent upon the fabri-cation schedules of a number of relatad components, for some of which vendors are yet to be selected (such as the PHTS pipino). A detailed schedule for this experimental program therefore cannot be defined at this time. However,flRC will be inforred when such a schedule is available. O l ( l l Amend. 30 flov. 1976 Q120.62-2 t

Question 120.63 (3.2, 5.3.2.1.3) ASME Code Case 1594 is applicable for the examination of elevated tem-perature Section III, Class 1 components only. Provide a listing and technical basis for preservice nondestructive examination requirements that you are specifying for ASME Section III, Class 2 and Class 3 com-por.ents which have not been uograded to Class 1, and where metal tem-perature exceeds those for which allowable stress values are given in Section III.

Response

The Auxiliary Liquid Metal System and the Impurity Monitoring and Analysis System have components designated as ASME Section III, Classes 2 and 3, in which metal temperatures exceed those for which allowable stress values are given in Section III. The following components by system are Section III, Class 2 or 3, and must be designed for elevated temperature service in accordance with an applicable Code Case. EVST NaK air blast heat exchangers ! NaK piping from overflow heat exchangers to the EVST ABHX including valves

        'T                              Primarv Na cold trap economizers (G

Piping between primary Na cold traps and the first isolation valve Primary Plugging temperature indicator (*PTI) and associ9ted piping & valves Primary sodium sampling package (SSP) and associated piuing and valves Intermediate sodium cold trap pumps Intermediate sodium cold trap economizers l Intermediate sodium cold trap piping and valves (applies only to normally flowing circuit-not applicable to drain lines, transfer piping, or piping between cold trap economizer and crystalizer) Intermediate sodium characterization package and associated piping and valves EVS Multipurpose Sampler , Q120-63-1 Amend. 62 flov. 1981

 .- -..    . ..._ .__ _ . _ .~ _ ... _ ___ _ _. ._.-____ _ _ _ .                         _ . _ . _ _ _ _ _ _ _ _ _ _ _ . _ _

10 ASME Section III, Class 2 or 3 comoonents of the Steam Generator System (SGS) and Steam Generator Auxiliary Heat Removal System (SGAHRS) which wili see elevated tenperature service are as follows: a Superheater Outlet Steam Piping e Superheater Outlet Isolation Valve e Superheater Outlet Check Valve o Sucerheater Relief Valve Inlet / Outlet Pipinq e Superheater Relief Valve l l e Reaction Products Separator Tank l e Sodium Rupture Discs to Reaction Products Separater Tank Piping e Rea: tion Products Separator Tar'<s Eoualizer Piping e Superheater Steam Vent Inlet / Outlet Pining e Superheater Steam Vent Valve l e Superheater Steam Vent Isolation Valve l Preservice inspection for all ASME coded plant components is addressed in response to NRC Question 120.66. O Q120.63-2 Anend. 62 Nov. 1931

__._._.__--____m___. .._._ . . _ _ _ . _ . . _ . . . _ _ _ _ . _ . _ . _ _ _ _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ . _ . _ . _ i i 1 Question 130.29 (RSP) (3.7.2.1) i j Your response to Question 130.3 is not complete. An acceptable criterion to determine a sufficient number of modes to assure participation of all significant modes is that the number of modes should be such that  ! inclusion of additional rudes does not result in more than a 10% increase l . in responses. Respont.e : , ! This criterion is discussed in the revised Section 3.7.2.3. l I ! i l i i i i a 1 5 ,O ! I i I I i

                                                                                                                                                        ^

l l l < i L l L l l Q130.29-1 Amend. 62 Nov. 1931 l l r

O Ouestion 130.42 (F6.4.2.5 Yellow) Figure F6.4-13 depicts the temperature for the cavity liner as well as the coolant temperature. Indicate the corresponding concrete temperatures behind the '.iner, the extent of structural damage expected, if any, and provide the resulting margins of safety against loss of required function. Resacnse: In Amencment 24 to the PSAR, the Project withdrew the Paralici Cesign frem further consideration by the NRC staff. This question, bewever, requests analytical inforcation pertinent to the TMSC8 Design. An assessment of the concrete te'nperatures in and the structural evaluation of the reactor cavity folicwing reactor vessel penetration can be found in Section 3.2 of CRBRP-3, Volur.e 2 (Reference ICb of PSAR Section 1.6). O Q130.42-1 ' Amend. 62 Nov. 1981 1

O Ouestion 130.45 (6.2.7.2 Yellcw) Indicate the extent of the sacrificial concrete arcund the reactor cavity circular steel shell. Cescribe the method of analysis used te detemine the extent of the concrete degradation or cracking. Resocnse: In Amendment 24 to the PSAR, the Project withdrew the Parallel Design from further consideration by the NRC staff. This question, however, requests additional information pertinent to the TMBC8 Design with its additional thermai margins as discussed in CRBRP-3, Volume 2 (Reference 10b of PSAR Section 1.6) and will be responded to in that context. There is no sacrificial concrete around the reactor cavity circular steel shell, as discussed in CRBRP-3, Volume 2. However, a layer of insulating (light weight) material is added between the cell liners and the structural concrete as shown in Figure 3-34 of CRSRP-3, Volume 2. The method of analysis used to determine concrete degradation is discussed in Appendix C.3 of CR3RP-3, Volume 2. O l l l l l l Q130.46-1

                                                                     ' Amend. 62 l                                                                      Nov. 1981 l

I

O) (,, Question 130.47 (Accendices E & F) Question 130.37 requested idditional information with regard to the reactor cavity and equipment cell iiner response to sodium spills. Expand the response to this question to enccmpass the idiosyncrasies of the parallel design, i.e., EVCC and setled HAA enclosure cells, rapid pressure building on the RC cell, restriction of hot liner expansion in the EVCC, steam buildup behind the RC liner, etc. Rescense: In Amendment 24 to the PSAR, the Project withdrew the Parallel Design from further consideration by the NRC staff. This question requests additional design information on a specific feature of the Parallel Design Accordingly, the question is no longer applicable. Amend. 62 Q130.47-1 Nov. 1981

i

 >           Question 222.28 (5.6, 7.4)

Provide the design description, design bases, and design criteria for the Steam Generator Auxiliary Heat Removal System (SGAHRS) Instru-mentation discussed in Section 5.6.1.15 and 7.4.

Response

The design description and the design basis of the SGAHRS are presented in PSAR Sections 5.6.1 and 5.6.1.1 respectively. The functional ' requirements for the SGAHRS instrumentation are identified in Section 5.6.1.1.6. SGAHRS Instrumentation and Control Design Criteria IEEE Standard criteria for Class IE engineered safety features systems as identified in Section 7.1 of the PSAR, will be used to demonstrate l the ability of SGAHRS I&C to meet their functional requirements under conditions produced by the Design Basis Events. l It is a requirement for the SGAHRS I&C system that it be designed to accommodate an initiating event with an additional single active failure. Independent power sources will assure continuous SGAHRS I&C operation in O the event of Design Basis Events as specified for the SGAHRS hechanical design. SGAHRS I&C shall provide sufficient surveillance instrumentation in both the

control room and at local panels, to assure availability of operator infor-mation of all critical SGAHRS parameters necessary to operate this system under all postulated plant conditions, and during all SGAHRS operational modes.

I Where immediate operator information is needed, individual annunciators and indicators and the plant data handling and display systems will supply that information to meet the intent of Regulatory Guide 1.97. Instrumentation and Control for the redundant flow paths in each Heat Removal , loop shall be designed such that no single failure will impair its operation. ! Cabinets field wiring and sensor location selection shall meet all applicable l separation criteria, namely IEEE 384-1974 and USNRC Regulatory Guide 1.75. ! Three divisional separation will be adhered to, to obtain maximum availability of SGAHRS I&C consistent with the requirements for the SGAHRS mechanical system design. The intent of BTP APCSB 10-1 as applicable to instrumentation and Control is met by assuring diverse systems powered by different energy sources. l This design includes availability of off-site and standby diesel generator AC l power, as well as three divisional DC battery power. In conjunction with l the Steam Turbine Driven Auxiliary Feedwater Pump, the battery system design O will permit auxiliary feedwater instrumentation and control operation with loss of all AC power. ! Amend. 62 Q222.28-1 Nov. 1981 1

In order to meet GDC 19, SGA: IRS Instrumentation and Control shall be available O from a location remote to the control room. Transfer from the control room to local control shall be accomplished at local SGAHRS I&C panel.s. Shorts, open ground or hot wires in the control room shall not disable SGAHRS renote control from its local panels. Three divisional separate power supplies are used for renote control of SGAHRS I&C. Loss of one division of power will not effect SGAHRS Operation. Instrur entation and Control equipment qualification will be performed as outlined in Chapter 3.11 of the PSAR, ENVIRONMENTAL DESIGN OF ELECTRICAL EQUIPMENT, and the environmental conditions specified for the CRBRP using applicable standards, Regulatory Guides, BTPs, Federai Regulations, etc. , identified in Sections 1.1.3, 3.1 and 7.1 of the PSAR. The method and type of qualification used for SGAHRS I&C will be submitted as part of the F5AR. Seismic qualification will be performed as outlinea in Chapter 3.10 of the PSAR, SEISMIC DESIGN OF CATEGORY I ELECTRICAL EQUIPMENT. I&C equiprnent shall be qualified to OBE and SSE conditions as specified for the CRBRP and identified in Section 3.7 of the PSAR. Plant safety criteria as applicable to the SGAHRS Instrumentation and Control are identified in Section 7.1.2 of the PSAR. The design criteria for the SGAHRS Instrumentation and Control System are com-patible with the Plant Protection System, which initiates SGAHRS I&C by enploy-ing a 2/3/ logic of LOW STEAM DRUM LEVEL or HIGH STEAM TO FEEDWATER FLOW RATIO. l 1 Amend. 62 0:,I Q222.28-2 Nov. 1981

l l 1 Question 222.31 (C.l.3.3)_ Section C.1.3.3 presents an initial rel! ability allocation for the shutdown system (SDS), shutdown heat renal system (SHRS), and for other faults leading to loss of in-place coolable geometry. Provide a table which sub-allocates the SDS, SHRS, and other systems whose faults could lead to loss of inplace coolable ge:.netry down to the major component level such as is shown in Figures C2.2-1 and C2.2-2.

Response

Reliability goals allocations are no longer applied to reactor shutdown and shutdown heat removal systems and components. The major thrust of the CRBRP Quantitative Analysis Program (Appendix C) is to identify the major contributors to plant unrealibility and to evaluate approaches to appropriately minimize the impacts of these contributors. O O Q222.31-1 Amend. 62 Nov. 1981

i

   ,                 Question 222.36 (C.2.01 1

Describe the degree of confomance to IEEE Standard 352-1975, " General Principles for Reliability Analysis of Nuclear Power Generating Station O Protection Systems" for the CRBRP assessment of reliability. j Response: a 1

~

The CRBRP Reliability Program is structured along the guidelines provided in IEEE 352-75 and conforms to the intent of that l r i standard. i The stated objective of IEEE-352-75 is to present the general principles

 '                   that may be used to evaluate the qualitative and quantitative reliability and availabiilty of safety related nuclear power plant systems. Thus, while general principles may be followed, utilization or non-utilization of specific methods described in the standard does not indicate a deqrne of conformance. The standard discusses four basic areas of Reliability evai l ation;

1. Qualitative Analysis Principles

2. Quantitative Analysis Principles t

3. Data Acquisition and Use

! 4. Establishment of Test Intervals l Any well defined reliability program must address each of these topics and the CRBRP progran is not an exception. The in-depth qualitative reliability assessments of the safety systems that are being performed are described in Appendix C of this PSAR. The reliability program has been designed to provide additional assurance beyond the rormal design process that the probability of a loss of core coolable geometry is as low as reasonably achievable. This program addresses the Plant Protection System, and any other equipment whose failure could i prevent or degrade the reactor shutdown or shutdown heat removal functions. Tha ralytical principles and intent of IEEE Standard 352-1975 are in evi-dence in the program's elements, described in Appendix C of the PSAR and I discussed as follows: t i

1. A listing of the equipment critical to the qualitative reliability program objectives is maintained. This safety-related equipment is subjected to Failure Mode and Effects Analysis (FMEA) at the component and system level. The FMEAs are extended to address the effects of faulted interfaces from auxiliary and supporting systems. The FMEAs are used in conjunction with tailored checklists to search out the l

j suse.eptibility of components and system functional redundancies to common causative factors. These qualitative analyses are conducted on a schedule to pennit integration of the program and its findings l into the design development process. The reliability assessment is a i formal part of the design review for . aponents and systems. The qualitative reliability program assessments are documented at the system level and selected major lower equipment levels in Reliability Design Support Documents (RDSDs). Amend. 62 Q222.36-1 Nov. 1981

2. The quantitative analysis principles of the standard are employed in block diagram success path modeling of the reactor shutdown system including the plant protection system. Failure path black diagram modeling of the shutdown heat removal system is employed. The Project uses these tools as aids to the design decision-making process. They provide valuable insights into the identification of the dominant contributors to unreliability.

Sensitivity studies of the dominant contributors are conducted to

                ~

rere intensively evaluate the uncertainties associated with failure rate assignments, operating assumptions, and to examine the benefit of potential design modifications. The overall objective is a balanced design through identification and minimization of the effects of major risk contributors.

3. To accomplish this objective, the modal is quantified with point estimates of failure rates decived from all available data sources.

A failure rate data book is maintained for the SHRS model that documents the derivation of the failure rate of each failure mode assignment in the model . The basis of failure rate assignments in the RSS nodel are an integral section of the assessment document. A statistically meaningful data base for many first-of-a-kind equipment in CRBRP does not exist. For this reason, the Project can establish only limited statistical confidence in the calculated overall RSS or SHRS failure predictions in these assessments. Although this lack of empirical data deters from the Project's statistical confidence in the overall system failure predictions, it does not negate the models' value as a design decision aid. With emphasis on even application of realistic failure rate assignments (not overly conservative, nor overly optimistic) com-parative insights into the overall balance of the RSS and SHRS designs can be gained. These comparative insights can be refined through sensitivity studies that vary the significant modeling parameters to evaluate the system effect of data uncertainties.

4. The last major topic addressed in IEEE 352-75 is that of test interval l defi ni tion. Implications of test interval on CRBR safety system avail-I ability are addressed in C.5.3. Test intervals defined by test l and analysis will be incorporated into the proper plant operating l specifications.

l l l Amend. 62 O Q222.36-2 Nov. 1981 l

t Y Question 222.37 (C.3.0) Describe the qualification procedures and methods as outlined in ! IEEE Std. 323-1974, Section 6.0, for all Class IE equipment associated with the shutdown systems and shutdown heat removal systems. List and j justify any exceptions to IEEE Std. 323-1974. Provide a list of i equipment to be qualified. j Response: Reference 13 of PSAR Section 1.6, "CRBRP Requirements for Environmental , Qualification of IE Equipment", establishes the qualification program for qualifying all Class IE equipment to perfonn its required function under , normal, abnormal, design basis event and post design basis event condi-tions. The entire program is designed to conform to IEEE Std. 323-1974 as clarified by the forward issued by NPEC on July 24, 1975 as IEEE Std. 323A-1975. 1 ) Class IE equipment includes the essential safety related electrical i equipment of the Reactor Shutdown System, the Containment Isolation System, the Steam Generator Auxiliary Heat Removal System, and other i safety related systems. A specific list of Class IE equipment to be qualified is listed in Reference 13 of PSAR Section 1.6. Type testing, operating experience, analysis or combinations thereof will be used to document that Class IE equipment will meet or exceed its performance: requirements throughout the equipment's qualified life. The qualification will be based upon the most severe environment predicted to occur prior to and during those portions of the specific accident transients for which the component is required to perform its safety

function.

Much of the safety related instrumentation in the plant has been l

successfully employed by the nuclear industry and has already been qualified for nuclear power plant service. Where new instrumentation has been developed (e.g., Reactor Vessel Sodium Level), development < test experience will be used for qualification, supplemented by analysis where necessary. FFTF production, acceptance and environmental cycling test results are available for use in qualifying some Class IE equipment (e.g., PPS comparators and logic). To qualify Class IE equipment by type test, IEEE 323-1974 requires that the equipment to be qualified be aged to simulate its end of qualified life condition and then subjected to the qualifying type tests. After development of industry standards for methods to adequately " age" electric and electronic equipment, the Project will incorporate the appropriate requirements into Reference 13 of PSAR Section 1.6. In summary, the CRBRP design will comply with the requirements of IEEE-323-1974. The detailed qualification test result summaries for each item of equipment will be submitted upon completion of ' he c necessary

                  .     . analysis and planning as part of the FSAR.

l

                                                                                                                                                               "  O Q222.37-1                                             Nov. 1981 1

i l Question 222.38 (Q222.37) ] In reference to Acceptance Review Question 222.37, provide: (1) a sche-

dule which identifies the Class IE equipment (includinq dates) to be i qualified, (2) verification that all Class IE items will be qualified,

! and (3) identification of the method of qualification for each item, i.e., type testing, operating experience or qualification by analysis.

Response

(1) Class IE equipment to be qualified is identified in Reference 13 of , j PSAR Section 1.6, "CRBRP Requirements for Environmental Qualification l l of IE Equfpment". Class IE instrumentation and control qualification  ; is scheduled tu start in late 1977 and to be concluded prior to the } j expected issuance of the Operating License Safety Evaluation Report. (2) All Class IE items will be qualified to the CRBRP design basis event environmental and seismic conditions, using applicable indus- j try standards, Regulatory Guides, Federal Regulations, etc. , as  ! discussed in Reference 13 of PSAR Section 1.6. (3) The methods for qualification will comply with the requirements of l IEEE-323-1974 as discussed in Reference 13 of PSAR Section 1.6.  : ' Identification of the method of qualification will be provided in l l the FSAR. A documentation data base will be established which will j i contain environmental qualification data about each item of IE  : ! equipment. This information will be available when the FSAR is l l submitted. l i  ! I t ll  ! I t I l l E i Q222.38-1 Amend. 62 Nov. 1981

i 1 i i i i l l Question 222.40 (C.l.3.3)

                                                                                  -7 In reference to the initial reliability goal allocation of <10         for the Shutdown System, <8 x 10-7 for the Shutdown Heat Removal System, and <10-7 for other systems whose faults lead to loss of in-place coolable geometry, provide your method for incorporating a margin of error to account for errors in prediction, measurement, and test conditions.                                   t

Response

, Reliability goals allocations are no longer applied to reactor shutdown j and shutdown heat removal systems and components. The major thrust of i the CRBRP Quantitative Analysis Program (Appendix C) is to identify l the major contributors to plant unrealibility and to evaluate approaches i to appropriately minimize the impacts of these contributors. . i l I i ) Ih I i  ! J i i i l i l i n l I Q222. 0-1 i lh Amend. 62 riov. 1981 4 1 I

     -_ =_                                         - . - _ _ . _ . _ - _ _                  _ - - - . _ _ _ _ . .__                      . ______.____-

l l I t Question 222.49 (C.3.0) (C.3.2.2.1) In addition to the design verification and reliability test program information contained in Section C.3.2.2.1 of Appendix C, provide the following for a'l , eleven tests: (1) the analytical success criteria for each proposed test item; (2) the expected test results; (3) the risk of failure associated with each test, i.e., high, moderate, or low; (4) the components considered to be critical items; and (5) the impact on the program schedule due to failures of major com-ponents of the SDS and SHRS to meet the required goals.  ; 1

Response

j The section of PSAR Appendix C referred to by this question has been deleted. However, test evaluation is still germaine to the Reliability Program. The 1 l Reliability Program is discussed in Section C.1.3, C.I.3.1, C.1.3.2, C.1.3.3, and i i C.1.3.4 of Appendix C. Our current Reliability Program emphasizes qualitative ' analyses and numerical assessments based upon system functional models and  ! failure parameters gleaned from historical experience and carefully considered  ! i engineering judgments. However, the established and ongoing test programs i ! are continuously monitored from the standpoint of reliability. Although a statistical data base is not being sought, any failures during testing that impact reliability will lead to corrective design actions. Various testing programs are discussed in the following sections of Appendix C: I l C.5.1.1 C.6.1.2 C.6.5.2 C.5.2.2 C.6.2.2 C.6.6.2 i C.S.3.2 C.6.3.2 l C.S.4.2 C.6.4.2 , l l l  : l

                                                                                                                                                        ?

{ I Q222.49-1 Amend. 62 j Nov. 1981 r

                                                                                         ..                _ _ _ _ _            _ - _      _ _ _ -_ )

Question 222.51 (7.1.2.2) (RSP) The physical separation criteria listed in Section 7.1.2.2 provide that wiring for the containment isolaticn and other safety related systems may be run in conduits containing either primary shutdown system wiring, or conduits containing secondary shutdown system wiring. - Furthermore, your criteria provide that this wiring may be brought through the penetrations of either primary shutdown or secondary shutdenn systems, provided that no degradition of the separation between primary and secondary shutdown systems results. These provisions of your separation criteria would not be acceptable because: (1) There would be an inherent degradation of independence between the primary and secondary shutdown systems by potential sneak paths through the containment isolation and other safety related systems; and, (2) The potential interaction between the primary and secondary shutdown systems (de-energize to actua'e) on one hand and the containment isolation and other safety related systems on the other (energize to O actuate) due to a single event such as fire or overheating may put in jeopardy the initiation and/or completion of safety functions. Amend your criteria to alleviate these concerns, or provide an analysis to show that your preser,i criteria will not make your design subject to these concerns. Response : Amended and expanded physical separation criteria for the Containment Isolation System has been included in Sections 7.1.2.2 and 7.3.2.2. The Containment Isolation System is similar to the Primary and Secondary Shutdown Systems in that the signals from the Control Room to the isolation valve operators function as de-energize to actuate. O Amend. 62 Q222.51-1 flov. 1981

Ques tion 222.54 (7.1.2.5)(RSP)

 'le require that your design comply to the recommendations of IEEE Std-323-1974 and Regulatory Guide 1.89. The source term for establishing the radiation environment for the reference design should be accepted prior to the calculation of the applicabl radiation doses.

Supplement the infonnation given in Sections 3.11.1 and 7.1 by submitting in tabular form a listing of all safety related equipment and components (e.g., motors, cables, sensing, and control devices, etc.), located in the primary containment and elsewhere that are required to function during and subsequent to any of the design basis accidents: provide their loco-tion, and for each location, define the worst case design basis environ-mental conditions in terms of temperature, pressum, and humidity, chemical contaninants, and radiation. We will require that the following qualificaticq test pmgram information be provided for all Class IE equipment as part of the FSAR: (1) equipment design specification requirements; (2) test plan; (3) test setup; (4) test procedums; and, (5) acceptability goals and requirements. This information will have to be provided for at least one item in each of the following groups of Class IE equipment: (1) switchgear; (2) motor control centers; (3) valve operators; (4) motors; (5) logic equipment; (6) cable; and, (7) diesel generator control equipment O Amend. 23 Q222.54 -1 June 1976

i

Response

e CRBRP design will comply with the requirements of IEEE-323-1974 and Regulatory Guide 1.89 as discussed in Reference 13 of PSAR Section 1.6. The detailed qualification test programs for each item of equipment will be submitted upon co~pletion of the necessary analysis and planning as part of the FSAR. l l l I I I O l 1 l 1 l l l 1 O Q222.54-2 Anend. 62 flov . 1931 l

Question 222.59 Update Tables 7.1-2 and 7.1-3 to include all Regulatory Guides and IEEE Standards listed in Table 7.1 of the Standard Review Plan.

Response

The applicable Regulatory Guides and IEEE Standards identified in SRP 7.1 are addressed in Section 7.1. The following Regulatory Guides were not addressed in Section 7.1 which applies to safety related instrumentation and control systems for the reasons given. Regulatory Guide 1.6, " Independence Between Redundant Power Sources and their Distribution Systems" is complied with as discussed in Sections 8.3.1.2 and 8.3.2.2 of the PSAR with other discussions on power supplies. Regulatory Guide 1.7 " Control of Combustible Gas following a loss of Coolant Accident" is noted in Section 1.1.3 to be not applicable to CRBRP. Regulatory Guide 1.11 " Instrument Lines Penetrating Containment" is not applicable to CRBRP since there are no instrument lines penetrating con-tainment as discussed in Section 7.1. Regulatory Guide 1.29 " Seismic Design Classification" is discussed in PSAR Section 3.2 " Classification of Components" where it is stated " structures, systems, and components are classified in full conformance with Regulatory O Guide 1.29." A list identifying instrumentation and control equipment classified as Seismic Category I is provided in response to question i 12 4 222.54 Regulatory Guide 1.68 " Test Programs for Water-Cooled Power Reactors" details guidelines for all plant preoperational testing. As noted in PSAR Section 1.1.3 and Chaoter 14.0, Regulatory Guide 1.68 will be censidered in preparing test plans, though specific plans other than those for LWR's in the guide must be developed for CRBRP. Regulatory Guide 1.70 " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants" applies to the entire PSAR and provides a suggested format for LWR SAR's. In general, the CRBRP is consistent with the Draft SFAC for LMFBRs, but the guide is not considered appropriate for inclusion in a list of guides applicable to the design of safety related instrumentation and control. Regulatory Guide 1.78 " Control Room Habitability During Chemical Release" applies to design of LWR control room habitability systems. PSAR Section 6.1 discusses the CRBRP Control Room Habitability System. It is not considered appropriate for a list of guides applicable to safety related instrumentation and control systems. O Q222.59-1 Amend. 24 July 1976

i 1 i

                                                                                                       )

1 I Regulatory Guide 1.89 " Qualification of Class IE Equipment for Nuclear Power Plants" specifies the applicability of IEEE Standard 323-1974 ll which will be applied as discussed in Section 7.1.2.5. The regulatory guide also identifies a source term based on an event which would re-

sult ,' con failure of all the equipment oeing qualified. As indicated in new PSAR Section 7.1.2.12, that guidance will be followed in intent, but not detail.

Regulatory Guide 1.96 " Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants" applies specifically to a boiling water reactor system and, as indicated in Section 1.1.3, will not be applied to CRBRP safety-related instrumen-tation and control systems. l l l O Amend 22 Q222.59- 2 June 1976

Question 222.64 (Q 222.5) Your response to Question Q222.5-1 provided in revised Section 3.10.1 and revised Table 3.2-3 is not complete. We requested that you

                         " Identify specifically and provide a listing of all safety related equipment and structures that will be seismically qualified as Seismic Category I." To the extent possible at this stage of the design, provide a complete response to this request for additional documentation.

Response

A listing of all safety-related electrical equipment which is Seismic Category I is provided in Reference 13 of PSAR Section 1.6. All Seismic Category I mechanical equipment is listed in Table 3.2-2. O - i l i O Q222.64-1 Amend. 62 Nov. 1981

O Question 222.70(7.2.1.2,7.3.2.1) With regard to the environmental qualification requirements of cafety related equipment, you state in Section 7.2.1.2.3 that the environmental extremes are being established and will be supplied at a later date. Provide a specific schedule for the submittal of this information.

Response

A complete discussion of the CRBRP environmental qualification program is provided in Reference 13 of PSAR Section 1.6. O Amend. 62 Q222.70-1 Nov. 1981 O

Question 222.73 (7.4.1.1, RSP) You state in Section 7.4.1.1.4 that " Control interlocks associated with the operation of active components have not been completely defined" for the Steam Generator Auxiliary Heat Removal System (SGAHRS). Specify your time schedule 'or completing the design of this system. It sNuld be noted that your design should provide a diversity of motive and cor.rol power so that both trains of the SGAHRS will not depend on one type o' power. An example of an acceptable auxiliary feedwater system would be to have one train relying on DC and steam and another train relying on AC only, either one of which can provide the required flow. Identify the power sources used to provide motive and control power to SGAHRS.

Response

SGAHRS Instrumentation and Control System design of interlocks associated with the operation of active components is complete. PSAR Section 7.4.1.1.4 includes a functional description of required SGAHRS contro? interlocks. With regard to the request to identify the power searces used to provide motive and control power to SGAHRS, refer to NRC Question 020.11 and its response. Compliance to the applicable portions of BTP APCSB No.10-1 is met with the l following design: In order to meet the requirement of an auxiliary feedwater system to consist of at least two full capacity independent systems, including diverse power sources, SGAHRS is designed with two 50% capacity motor driven pumps, each connected to a separate diesel, and one 100% steam turbine-driven pump. Thus, there is 200% total capacity with half of it electrically powered and half steam powered. There are also two parallel auxiliary feedwater paths to each dram, one from the tu' bine-driven pump through a control valve and one from the headered motor-driven pumps through another control valve. The drum level setpoints are set such that flow normally is supplied from the motor-driven pumps only with the turbine pump recirculating. If the drum level falls below the turbine pump setpoint, the control valve downstream of that pump opens and begins supplying flow. System "A" and System "B" switchgears provide power to the respective System

   " A" and System "B" motor driven SGAHRS pumps. The control power for the System "A" and System "B" switchgears is provided from the System "A" and System "B" battery supply as identified on Tables 8.3-2A and 8.3-28. The control power required for the 100% turbine driven SGAHRS pump is supplied by Battery System "C" as identified in D.C. battery loading table 8.3-2C. This battery system design .1il permit auxiliary feedwater sub-system operation with loss of all AC power.

O Amend. 62 Q222.73-1 Nov. 1981 [

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

Question 222.74 (7.4.2) l() Provide a Functional Control Diagram and a P&ID for the Outlet Steam - Isolation Subsystem (OSIS). Identify the power sources supplying  ! control and motive power to the OSIS. ] i i Response: } The OSIS functional control is discussed in Section 7.4.2 and is j . illustrated as part of PSAR Figure 7.5-6 (Sh. 5). The isolation valve l 1 location in the Steam Generator System is shown on PSAR Figure 5.1-4,  ;

                                     " Steam Generator Schematic Flow Diagram". The power source for this system is Class IE instrument power, both for control and motive power.                                                 j i

i 1  ! l  ! i I i ! l ltill i i i l l - 1 l i l l l I i i l l i l i L I i I t till l! l 0222.74-1 Amend. 62 Nov. 1981

Question 222. 76 (7. 6.1.1, 7.6. 3.1, 7. 5.6.1, Q222.14) l Provide the design bases, design criteria, P&ID's and Functional Control Diagrams for the Treated Water Instrumentation and Control System, the Overflow Heat Removal Service (0 HRS) System, and the Sodium Water Reaction Pressure Relief System (SWRPRS). Provide the same information for the systems listed in Table 7.1-1 under "Other Safety Related Instrumentation and Control".

Response

The design bases for the chilled water systems and service water systems are provided in PSAR Sections 9.7 and 9.9, resoectively., Those bases apply to the instrumentation and control portions of the systems as well as the other portions. The detailed information requested concerning the in-strumentation and control for the Treated Water System, which includes the Emergency Chilled Water and Emergency Plant Service Water, is not available at this time. It will be supplied when available. l The SWRPRS design basis events are identified in section 15.3.3.3, "Large Sodium Water Reaction". Detail on the instrumentation and control functions to meet these design basis events is provided in PSAR Section 7.5.6. The design basis for the SWRPRS, including the instrumentation and control equip-ment, is to protect the plant by tripping the reactor and mitigating the pressure pulse in the IHTS for reactions enveloped by a three tube leak in any steam generator module. ' O PSAR Figures 5.1-4 and 7.5-6 provide P&ID and Functional Control information for the SWRPRS. The design criteria for the SWRPRS I&C equipment is as follows: To monitor the SWRPRS rection products vent line and initiate plant shutdown upon detection of reaction products. Initiate activation of and monitor the SWRPRS equipment required to mitigate the consequences of a large sodium / water reaction. l The SWRPRS I&C equipment necessary to initiate plant shutdown and initiate acti-vation of SWRPRS equipment upon detection of a significant sodium / water reaction shall be designed and qualified in accordance with the appropriate requirements of Section 7.1. This includes the monitoring equipment necessary to detect events requiring plant shutdown and SWRPRS operation. As discussed in PSAR Section 5.6.2, the design basis for the DHRS (OHRS) is to provide a single safety class decay heat removal train, capable of be'ng initiated manually from the Control Room and from a local control s ta ti on. i The bases for the design of the DHRS instrumentation and control is to monitor the operation of the DHRS a~nd provide the control functions as O required to meet the, following DHRS design criteria and requirements. Amend. 62 Q222.76-1 flov. 1981

a. Provide the capability for manual initiation from the Control Room and a local control station.

b. Provide sufficient information to enable the operator to verify that the DHRS is performing its function of removing heat from the reactor,

c. Provide sufficient instrumentation to allow the operator (s) to monitor and control (manually) the temperatures of the sodium and NaK loops to maintain the system within operating limits.

d. The I AC equipment necessary to assure that DHRS will carry out its decay heat renoval function shall be designed and qualif-ied in accordance with the appropriate requirements of Section i

7.1. Sufficient DHRS instrumentation will be provided to enable the operator (s) I to verify that the Auxiliary Liquid Metal Systen valves are aporooriately alianed, pumps and fans operating and heat is being transported to and removed from the ABHX's. Instrumentation and control capability will be provided to allow the operator (s) to monitor and contrcl (manually) the temperatures of the sodium and NaK loops to maintain the system with-in the operating limits. The control for the containment isolation valves ! in ' a NaK subsystem will meet the design bases for the CIS provided in Sect on 7.3.1.2. The design bases for the Heating, Ventilating and Air Conditionining (HVAC) O System, instrurcentation and control, are discussed in revised PSAR Section 7.6.4. The information requested concerning the instrumentation and control for the Recirculating Gas Cooling System will be provided in a future amendment. l l l O Amend. 62 Nov. 1981 Q222.76-2

O Ouestion 222.90 (7.1.2.5 F1-2) Provide the infurraation reouested in our RSP 222.54 and Question 222.70 to reflect environmental conditions applicable to the parallel design. Include in your response an identification of these components required for safety which may be affected by the design basis loading referred to in Section 3.7.2.1.2 (yellow). _Resoonse: In Amendment 24 to the PSAR, the Project withdrew the Parallel Cesign from further consideration by the NRC Staff. This question requests additional design information on specific features of the Parallel Design. There fore , the question is no longer applicable. The environmental conditions associated with operation of the TMBCB features are included in Section 2.1.2 of CRSRP-3, Volume 2 (Reference ICb of PSAR Section 1.5). O l Q222.90-1 Amend. 62 l i' Nov. 1981 l 4 i l

b l t k J Question 222.91 (F5.2.1) t i Provide the information requested in our Question 222.77 regarding the l l accident and post-accident monitoring of plant parameters applicable i to the parallel design. L ]i Resoonse: I l i This question is on design details of Parallel Design features. In ' Amendment 24, the Project withdrew the Parailel Design; hence the question  ;

;                     is no longer directly applicable. However, CRBRP-3, Volume 2 (Reference 10b       l j                      of PSAR Section 1.6) identifies the instrumentation provided for post             !

3 accident monitoring in the case of events which challenge containment j integrity. i s  ! I l I

>                                                                                                       l 1                                                                                                        I

! I J ltill l l I l t i i i  ! i l I l l l l l I l , 1  ! i l I l t t Q222.91-1 Amend. 62 l l 1 Nov. 1981  ! I r l l l l l

1 i a O Question 222.92 (F5.3.7.3.6.2.4.3.7.3.1.1) l j Provide the additional information en the CIS requested in RSP 222.71 j as it applies to the parallel design. Cescribe the head access area i instrumentation, including the sensors,that will be used to isolate the j sealed head access areas frca the main containment area. } Rescense: l In Amendment 2? to the PSAR, the Project withdrew the Parallel Cesign j from further consideration by the NRC staff. This question requests I i additional design information on a specific feature of the Parallel Cesign ,

;                   accordingly, the question is no lenger applicable.                                            I i

1 l )'  ; i 1 b i  ! i I ! i till , I I 4 l t i Q222.52-1 Amend. 62 i-i

+

Nov. 1981 i

l-

, t e i I

4, Ouestion 222.93 (1.2.6 Yellcw) In Section 1.2.6 (yellew) you state that the Plant Control System receives signals from the Ex-Vessel Core Catcher System. Identify all such signals and describe their intended function. Res:ense:

 ,                                                With the deletion of the Parallel Design in Amendment 24 this questien                                                 l is no longer applicable as the features upon which the question is based are no longer a part of the design.

i I I I I l \ i  ! l f f i t i i l i l j Q222.93-1 ' Amend. 62  ; Nov. 1981

tO Question 222.94 1 On.page 3.1-15a you state for "for the design basis core disruptive ' j accidents...the electric power systems provide pcwer as required to assure adequate cooling of the fuel traterials within the containment". Provide specific infonnation as to what power systems support what safety related systems and associated safety functions within the scope of your , i. above statement.  ! \ I j Resconse: l With the deletion of the Parallel Design in Amendment 24 this question ! is no lenger applicable as the features upon which the question is based are no lenger a part of the design. I i l lO \

                                                                                                 \

l i ! l l i i e l l l r Q222.94-1 Amend. 62 I O Nov. 1981 ( t

                                          ...  ..                    .                          J

i l 1 I i l l ! l ! l i > l Question 222.95 (5.2.7, 7.9.1.5. Q222.78) f Provide the infomation requested in our Question 222.78 with respect to l the compliance of your parallel design to the General Cesign Criteria  ; entitled " Control Recm". Res:ense: p { With the deletion of the Parallel Design in Amendment 24 this question is no longer applicable. However, the Control Room habitability requirements for TPSCB cenditiens are provided in Section 2.1.2.15 of CRBRP-3, i

 <       Volutne 2 (Reference ICb of PSAR Section 1.6).

I I i i le I l l 1 l ! L l ? l j Q222.95-1 Amend. 62 Nov. 1981  ; j 4 4 l

i 1 i < i ! i Ouestien 222.97 (8.3-1 A Yellew) i In the diesel-generator lead listing of Table 8.3-1A you list a number  ! cf 4.16 and 13.8 XV leads. For the reference design, the 4.16 KV loads  ! were listed as 13.J KV loads. Explain the reasons for these changes I and describe how you will derive a 4.16 KV supply. Such a supply is  ! not identified in Figure 8.3-1. Identify all non-Class IE loads listed l in Tables 8.3-1 A and 8.3-18. Response: , t I With the deletion of the Parallel Design in Amendment 24 this question  ! is no longer directly applicable. However, the electrical power system I requirements for T.V308 features are identified in Section 2.1.2.13 of CRSRP-3, Volure 2 (~.eference ICb of PSAR Section 1.6).  ; l i i l t , I ' l L O t i I l l l l. I l l l l Q222.97-1 9 Amend. 62 Nov. 1981  ! l l

) 1

,                                                                                     t i                                                                                      l i                                                                                      ,

I > l Ques tion 241.20 (4.2.1.4.2)_ The list of typical inspections under Fabrication Examination should be j . complete. Notably absent are the following: fuel chemical impurity ! analysis, cladding metallurgical state, wire wrap analysis, fuel and tag gas analysis, and fuel densification tests.

Will 100", radiography distinguish between blanket and fuel pellets with the same fuel rod?

For on-site acceptance tests, describe the procedures or tests which will assure that the fuel, blanket, and absorber pellet columns are in their proper location within the cladding.

Response

In Section 4.2.1.5.2 (Tables 4.2-59A and 4.2-598) a reasonably complete but not exhaustive listing of the characteristics of the Quality Conformance inspection plan for fuel assembly components is given. All of the characteristics listed are included in the characteristics of the inspection plan with the exception of fuel densification testing. l Analysis has shown that densification is not a problem in CRBRP (see CRBRP-ARD-0168 provided in response to NRC Question 3, Reference Q241.20-1). It will be possible to distinguish between fuel and axial blanket pellets within a fuel rod using 100% autoradiography. i After manufacture, all fuel and control rods are 100% inspected to insure

that all pellet columns are properly located within the cladding. As i indicated in the response to NRC Question 241.12, the standard analysis techniques utilized to evaluate the pellet column holddown springs insure that these springs will maintain column position during shipping and  !

handling loads up to 6g. CRBR core assembly shipping casks will have l acceleration indicators which will trip when shipping and handling loads  ; exceed the 69 limit. These indicators will be checked on site to insure  ! that the core assemblies have not been subjected to loaos above this  ! limit. A tripped indicator will be cause for rejection; the assembly  ! would then be returned to the fabricator for more detailed evaluation. l l t Reference Q241.20-1: Letter, T. P. Speir to P. S. Van Nort, dtd. Oct. 6, 1975 9 Amend. 62 Q241.20-1 Nov. 1981

Question 241.54 (4.2.1.1.2.2) Provide a discussion which compares the CDF to conventional design limits such as total inelastic strain, generalized stress, and engineering toughness. The discussion should include the methods for normalizing between the CDF and other design limits (e.g., a total engineering toughness limit for component failure - stress times strain - and a cumulative inelastic strain - number of times about the PEL).

Response

In general, the CDF technique described in PSAR Section 15.1.2.1 calculates loss of structural integrity due to two failure modes. When the CDF equals 1, then:

1. The equivalent stress exceeds the material residual strength, and/or

2. The maximum principal stress for a given period of time exceeds the creep strength for the same period of time.

These two limits are directly comparable to time independent and time dependent " generalized stress" limits, respectively, such as those given in the ASME high temperature code cases. However, the CDF is based on a numerical evaluation of test data, and provides a better correlation bet',een environment and performance for a specific material than do the more universal stress limits of the ASME code. More importantly, the CDF correlates the interdependence of the ductile rupture and creep O rupture failure mechanisms, whereas simple generalized stress limits do not. . The inelastic strain criteria as given in Section 15.1.2.2'of the PSAR essentially guard against the same failure modes as the CDF; the difference being that the limits are expressed in terms of strain. In other words, a " strain CDF" could also be defined as follows: CDFs, Acc ,

                                                      'p
                                       ' lim        ' lim -< l.0 where:           Ac = Creep strain increment c

Ac = plastic strain increment p clim = 0.3% strain limit , This " inelastic strain" criteria would also predict loss of integrity,

i .e. , CDFs = 1.0, for either of conditions 1 or 2 above. It should be i

noted that the plastic damage term of the CDF, Lp, accumulates damage for stres., excursions be. yond the proportional elastic limit. This is also true of the plastic strain tenn in the " strain r0F". Q241.54-1 Amend. 62 Nov. 1981 l

The 0.3 strain limit was selected to be conservative, and provides l

generally a lower limit on the failure strain.

The CDF does not include a damage correlation based upon " engineering i toughness", and no attempt has been made to correlate data using this } failure model. However, as discussed in the responses to .lRC Questions 241.48, 241.13, and 001.36, available experimental data show that the fuel rod performance is adequately predicted by those criteria presented in PSAR Section 4.2.1.1.2.2. O l 1

                                                                                         )

l 1 l i I 1 1 i I O Q241.54-2 Amend. 62 fio v . 1981 l

Question 241.91 (4.2.1.3) The discussion of CDF due to steady state operation and the combined s upset events show that at the design level for assembly 6 failure would occur at 540 full power days. Using the same type of analysis and

      ,   assumptions, viz. , maximum uncertainty in properties, 2a plant, etc. ,          i as used for the assembly 6 analysis, how many assemblies would be shown to fail before their goal lifetimes for equilibrium operation?

Response

If the same conservative CRBRP first core material properties, environments, and failure criteria are applied to determine the fuel and blanket rod l'fetimes during equilibrium cycle operation with steady state and worst upset transient loads (as simulated by cycle 3-4 conditions), it is x found that: a) At least 70% of the 156 fuel assemblies achieve the 550 day (2 equilibrium cycle) lifetia.e goal >ith a worst rod cladding hot spot CDF lass than 1.0. Ho ever, only 5% M the rods in a typical fuel assembly bundle exprience cladding temperatures similar to those for the worst rod. Furthermore, a cladding CDF equal to one implies a survival probability of at least 80 to 90 percent for eluilibrium cycle cladding hotspot temperai;res. Thus, it is anticipated that future refinements in orificing, environmental uncertainties, material properties and design criteria fo' 'recluding cladding failure will result in 100% fuel eisembly lifet.me goal achievement for equilibrium cycles. b) All inner blanket assemblies currently achieve the 550 day lifetime goal with all rods having hotspot cladding CDF's less than 1.0. c) Insufficient environmental information is available to predict radial blanket performance over 4 to 5 year equilibrium cycle operation. However, all radial blanket assemblies achieve their first core life-goals (878 full power days for cycles 1 through 4) as noted in PSAR Section 4.2. It should also be noted that the homogeneous CRBRP core assembly designation l numbers indicated in this questicn have been changed tc heterogeneous l core assembly numbering scheme shown in Figure 4.2-10B. Q241 1 1 Amend. 62 O- Nov. 1981

Question 241.93 (4.2.1.3.1.2) This section requires greater specificity concerning the FFTF fuel assembly structura l analyses. Describe or reference these analyses and show how they are related to the CRBR structural analysis. In particular, show how the FFTF analyses led to the selection of the 3 duct locations considered to be limiting areas on page 4.2-47 of the PSAR. For these 3 cases, for whic:: the calculated stresses, margin of safety, and fatigue damage are summarized in Table 4.2-8, please discuss how "it can be concluded that the duct will probably meet the desired design lifetime" when the safety margin for some of the stress categories are as low as 0.06 and 0.11. Discuss in detail the planned additlenal analysis to censider the effects of irradiation damage and sodium exposure, including brittle fracture modes.

Response

Under the revised sub-section numbering in the PSAR, the stress analysin of the core assembly structural components is addressed in Section 4.2.1.'.2.3. This section has been updated to reflect the most current analyses methods, structural criteria and design margins in the CRBRP core assemblies. The concerns raised in Question 241.93 have been resolved by @V using the state-of-the-art analyses procedures which go beyond the scope and the depth v analyses used previously. The results from the new analyses, as described in Section 4.2.1.3.2.3, cWinn the structural adequacy of the CR2RP core assembly structural components. A number of design verification tests are planned for CRBRP and these are summarized in Tables 4.2-19 and 4.2-20 of the PSAR. Q241.93-1 Acend. 62 Nov. 1981

B f

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,I i I Question 310.31 (15.1) All assumptions in Appendix D9 concerning the retention of fission products y in intact fuel, molten fuel, and sodium must be adequately supported. Re-  ; ference to experiments simulating LWR LOCA conditions does not constitute , 3 l adequate justification of arbitrary fission product release fractions. < Provide justification for the conservatism of the fission product release f and transport assumptions using as a data base those experiments which can

!       be shown to be applicable to the LMFBR accident conditions. Note that the     ,

Contamination / Decontamination experiment results are not applied directly i to LWR release fraction computations as they are not adequately represen-tative even of LWR meltdown conditions, i j Response:  ! I l Appendix D has been deleted from the PSAR. Analyses of hypothetical  ! I accidents involving significant molten fuel are discussed in CRBRP-3, f Volume 2 (Reference 10b of PSAR Section 1.6). Section 4 of that document j provides results of a range of radio-nuclide release scenarios that l address the concern reflected in this question. j l i I

l i ,

,O i i l  ! l i l  ! l i l L i i i I Amend. 62 ( O Q310.31-1 Nov. 1981 t l l 1 i

_ _-~ _. l 1 1 Question 310.33 (App. F) P The following events are postulated in the PSAR for the purpose of evalu-ating the reactor cavity (RC) radiation source terms:

l. A melt through of the reactor vessel and guard vessel occurs wherein 100% of the core is deposited in the ex-vessel core catcher (EVCC).

2. The molten fuel in the EVCC is covered with the primary sodium that drains out of the reactor vessel and associated primary sodium piping.

With the exception of the noble gases (all which are assumed to be I released to the sealed head access area), the fission products in the molten fuel are released to and mixed with the primary sodium.

4. The resulting fission product inventory in the RC atmosphere above the sodium is determined on the basis of equilibrium concentration relation-ships.

It is r.ot clear if the above sequence of events results in the most severe RC source term that can be expected with respect to radiological consequences. For example, if reactor vessel and guard vessel melt through can be postulated to occur with only a fraction of the core fuel participating in the nelting, then subsequent drainage of the primary sodium into the RC would leave the remainder of the core in a " dry configuration." Subsequent melting of the

                 " dry" fuel could release fission products directly to the RC atmosphere, 4

bypassing the attenuating effect of the primary sodium. Re-assess your evaluation and provide a response to address this concern.

Response

A radiologicai source term appropriate for the sequence of events leading to a " dry" core, suggested in this question, has been conservatively developed. This response is provided in the cor. text of the thermal margin beyond the de-sign base (TMBDB) design features, since the scenario represented in the question is not within the spectrum of design base accidents. While the analyses referenced herein were performed for the homogeneous core arrange-ment, the conclusions reached are equally valid for the heterogeneous core arrangement. It is also noted that the EVCC and the sealed head access area are not part of the CRBRP design, and thus were not assumed to exist for the analysis. . The principal release mechanism for this " dry" fuel case would be vapor- ) ization of the more volatile fission products as the fuel heats to its melting temperature. An analysis was conducted to determine the rate at which fuel melting would proceed given that the core was lef t in a " dry" configuration. The analysis assumed complete drainage of the reactor vessel at 1000 seconds (the minimum calculated vessel melt-through time), end of equilibrium cycle decay heating rates, and adiabatic conditions in-vessel af ter sodium drainage. Under these conditions, the range of times involved in melting assemblies under their own decay power was determined. The average inner core assenbly would melt in 8 minutes, the average outer core assembly in 10 minutes, the highest power radial blanket assembly in 32 minutes and the lowest power i radial blanket assembly in 90 minutes. The fuel is supported by stainless steel (either the cladding or surface within the vessel on which the fuel Q310.33-1 ! Amend. 62 flov. 1981

t l e O debri s ha s collected). Penetration of the steel would occur at temper-atures in the range of the steel melting temperature (125000 F). The al ternative to this continuous downward slumping is eventual sclid-ifica tion o# the fuel-clad mixture postulated by assuming sufficient mixire witn structural steel such that the mean temperature of the mixture is beicw the steel celting point. In either instance, the time during which " dry" fuel could be releasing materials in the reactor vessel is limited. The fission product release from the fuel during the meltdown phase was treated conserva tively. Based on the compilation of experimental data on meltdom release fractions presented in Reference 0310.33-1, it was conservatively assumed that 100% of the noble gas, halogen, and volatile (Cs, Rb, Te, Sb, Se) fission products and 1% of the solid fission products were released f rom the " dry" molten fuel inmediately. Because of the e<.tremely low vapor pressure of fuel at temperatures in the ranne of the meltina te7erature of steel, the amount of fuel released as a vapor would be insignificant. Thus the source tern would not contain fuel. It was assumed that this source tern immediately enters the Reactor Con-tainment Building (RCB). Except for the noble gases, the airbo ne fission products in the RCB atmosphere would be reduced by plate out and deposition on internal surfaces. The rate of aerosol depletion in the RCB was calcu-Ia ted by tbe HM-3? computer code. During the period from zero to twenty-four hours, reactor containment integrity would be maintained as indicated in Table Q310.33-1 (based on CACECO results). It is noted that the containment conditions w,uld be less severe for this " dry" fuel case than for the " wet" fuel case in which l all fuel and fission products are released to the reactor cavity at the l tine of reactor vessel and guard vessel melt-through. This results since the release of heat producing fission products to the Reactor Containtrent Building reduces the heat in the sodium pool and, consequently, delays the onset of sodium boiling and the rate of sodium vapor generation. Lealane of tne source term from the RCB to the environment would be by way of the A'onulus Recirculation and Filter System. The leakaqe rate of the source term to the annulus was calculated by the CACECO code. Unfiltered hwass leakage was also consi f ered. The potential radiological consequences of off-site exposure resulting from the dry fuel scenario are compared with tne corresponding conseque< ces of the wet fuel case in Table Q310.33-2. At-l mospheric dispersion factors used were the fif tieth percentile x/Q values based on site measured data. Although the doses for the " dry" case are higher than for the wet fuel case, the conclusion that the doses are ac- I centably . low is not changod. References 0310.33-1 WASH-lano, Appendix VII, " Release of Radioactivity in Reactor Accidents" (Draf t August 1974). 0310.33-2 Amend. 62 l Nov. 1981

i i l 'l !9 4 TABLE Q310.33-1  ; ., Comparison of RCB Atmosphere Conditions at 24 Hours for " Wet" Fuel Case and " Dry" Fuel Case l i  ! i

                                             " Wet" Fuel Case                               " Dry" Fuel Case                                                               (
 ;                                                                                                                                                                   l     l Temperature ( F)                    495                               310 k

I Pressure (psig) 16 8 l I j %H 1 0 2 1 4 i

                                                                                                                                                                           )

i I 1 i ! I i i9 i i l l l l l l l 1 i l e Q310. 33- 3 Amend. 62 Nov. 1981

i TABLE Q310.33-2 C0" PARIS 0N OF POTENTIAL OFF-SITE EXPOSURES i for Wet" FueT Case arid' Dry"~ Fuel Case l Doses in REM Organ " Wet" Fuel Case " Dry Fuel Case l Bone 0.027 0.029 2 Hour Lung 0.0029 0.109 E. B. Thyroid 0.0017 5.75 Whole Body

• 0.16 0.26 I

I Bone 0.020 0.028 f i 24 Hcur Lung 0.0024 0.11 LPZ Thyroid 0.0059 5.33 Whole Body

• 0.099 0.16 l
• Includes: Inhal nion, external gamma cloud, and direct ga=a shine l l

1 l l O n110 33-4 A nd.6c? Nav. 1931

O EVS Sodium Cold Trap _ l A conservative assessment of the consequence of a postulated EVS cold trap fire has been conducted by assuming a sodium spray and subsequent pool fire. In the scenario which is analyzed herein, it was assumed that _a sodium leak occurs in the cold trap cell, causing the cold trap cooling system to fail. The plant operators would be alerted by se.crai sodium leakage, temperature, smoke, radiation and pressure detecters, but it is assumed that they fail to isolate the cold trap circuit with the isolation valves which are located in a separate environment and thus not affected by the fire. The leaked sodium burns all the oxygen available in the cell. It should be noted that a leak in this cell is not expected to release the radioactive contents of the cold trap inventory since this inventory consists of condensed solids in the trap. Nevertheless, it is conservatively assumed that the radionuclide concentration in the leaked sodium is that which would exist if all of the cold trap inventory was redissolved in the EVS sodium. In fact, most of the redissolution would take place long after the available oxygen was consumed. For a sodium leak, the resultant cell temperature,and thus the cold trap temperature, would not rise to a value above that of the leaked sodium. Analysis for the Primary Cold Trap shows that it takes tens of hours to redissolve the cold trap contents, even if the cola tr3p temperature rises to 600 0F. Similar analysis for the EVS cold trap has not been done, but similar behavior would occur. In actuality, the EVS cold trap V release would be expected to be slower since its normal inlet sodium temperature is less than 500 F.0 The temperature rises to 6000 F infrequen-tly, only during operations to remove sodium surface impurity buildup in tne EVST. The decay heat generated in the cold trap is not sufficient to cause a significant temperature rise in the cold trap. If the trap were cut off from both its cooling system and the sodium process stream, the trap tem-perature would not rise above its normal operating temperature. The radiological consequences of this event were analyzed as follows:

1. The total inventory of the cold trap is redissolved into the EVS sodium. Tne radioactivity within the 60,000 gal o, EVS sodium includes the normal inventory plus that from the cold trap. It was obtained from the cold trap inventory as given in Table 12.1-24 and the distributions ratio between the cold trap and the EVS sodium as given in Table 12.1-1. The concentration was assumed uniform through-out the volume.

2. Following failure of the EVS cold trap.15,250 gal. of sodium spills l into the cold trap cell. This is the maximum amount that can spill due to the arrangement of the lines from the EVST. With the conser-vative assumptions followed, the radiological consequences are in-l sensitive to the total volume of sodium spilled.

Q310.49-5 Amend. 62 Nov. 1981

All of the oxygen within the colr_ trap cell and the adjoining cells S 3.

which share a common atmosphere reacts with the sodium. The total ]

volume of these cells is 33000 cubic feet. These cells contain a .

l maximum of 2 w/o oxygen in nitrogen.

4. All of the sodium oxide produced is lla20 and all of it is airborne as an aerosol.

l S. All of the aerosol generated is immediately released to the atmosphere without credit for plate-out or settling within the cold trap cell or the Reactor Service Building (RSB), depletion along the leak paths from the cold trap cell 'o the RSB or filtration through the RSB exhaust system.

6. The concentration of radioactivity in the aercsol is the same as that in the spilled sodium. The discussion forwarded in response to Question 310.50 reviewed the partitioning of fission products in the aerosol resulting from burning of the spilled sodium. It was shown that the assumption of equal activity in the pool and aerosol sodium was conservative.

7. Radioactivity decay during the accident is neglected.

! 8. Fallout (cloud depletion) cf radioactive material during downwind transport is conservatively neglected. Using the dose calculational methodology described in Chapter 7 o' the ER and 955 X/Q's, potential offsite doses have been determined. Doses at the Exclusion Boundary and the Low Population Zone are itemized in Table Q310.49-2. l l l l t i Q310.49-6

f. mend. 45 July 1978

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